[Federal Register Volume 69, Number 16 (Monday, January 26, 2004)]
[Rules and Regulations]
[Pages 3698-3814]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-35]
[[Page 3697]]
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Part III
Nuclear Regulatory Commission
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10 CFR Part 71
Compatibility With IAEA Transportation Safety Standards (TS-R-1) and
Other Transportation Safety Amendments; Final Rule
Federal Register / Vol. 69, No. 16 / Monday, January 26, 2004 / Rules
and Regulations
[[Page 3698]]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 71
RIN 3150--AG71
Compatibility With IAEA Transportation Safety Standards (TS-R-1)
and Other Transportation Safety Amendments
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations on packaging and transporting radioactive material. This
rulemaking will make the regulations compatible with the latest version
of the International Atomic Energy Agency (IAEA) standards and codify
other applicable requirements. This final rule also makes changes in
fissile material exemption requirements to address the unintended
economic impact of NRC's emergency final rule entitled ``Fissile
Material Shipments and Exemptions'' (February 10, 1997; 62 FR 5907).
Lastly, this rule addresses a petition for rulemaking submitted by
International Energy Consultants, Inc.
EFFECTIVE DATE: This final rule is effective on October 1, 2004.
Portions of Sec.Sec. 71.19 and 71.20 expire on October 1, 2008.
FOR FURTHER INFORMATION CONTACT: Naiem S. Tanious, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; telephone (301) 415-6103; e-mail
[email protected].
SUPPLEMENTARY INFORMATION:
Contents
I. Background
II. Analysis of Public Comments
III. Discussion
A. TS-R-1 Compatibility Issues
Issue 1: Changing Part 71 to the International System of Units
(SI) Only
Issue 2: Radionuclide Exemption Values
Issue 3: Revision of A1 and A2
Issue 4: Uranium Hexafluoride (UF6) Package Requirements
Issue 5: Introduction of the Criticality Safety Index
Requirements
Issue 6: Type C Packages and Low Dispersible Material
Issue 7: Deep Immersion Test
Issue 8: Grandfathering Previously Approved Packages
Issue 9: Changes to Various Definitions
Issue 10: Crush Test for Fissile Material Package Design
Issue 11: Fissile Material Package Design for Transport by
Aircraft
B. NRC-Initiated Issues
Issue 12: Special Package Authorizations
Issue 13: Expansion of Part 71 Quality Assurance (QA)
Requirements to Certificate of Compliance (CoC) Holders
Issue 14: Adoption of the American Society of Mechanical
Engineers (ASME) Code
Issue 15: Change Authority for Dual-Purpose Package Certificate
Holders
Issue 16: Fissile Material Exemptions and General License
Provisions
Issue 17: Decision on Petition for Rulemaking on Double
Containment of Plutonium (PRM-71-12)
Issue 18: Contamination Limits as Applied to Spent Fuel and
High-Level Waste (HLW) Packages
Issue 19: Modifications of Event Reporting Requirements
IV. Section-By-Section Analysis
V. Criminal Penalties
VI. Issues of Compatibility for Agreement States
VII. Voluntary Consensus Standards
VIII. Environmental Assessment: Finding of No Significant
Environmental Impact
IX. Paperwork Reduction Act Statement
X. Regulatory Analysis
XI. Regulatory Flexibility Act Certification
XII. Backfit Analysis
I. Background
Before developing and publishing a proposed rule, the NRC began an
enhanced public-participation process designed to solicit public input
on the part 71 rulemaking. The NRC issued a part 71 issues paper for
public comment (65 FR 44360; July 17, 2000). The issues paper presented
the NRC's plan to revise part 71 and provided a summary of all changes
being considered, both International Atomic Energy Agency (IAEA)--
related changes and NRC-initiated changes. The NRC received 48 public
comments on the issues paper. The NRC enhanced public participation
process included establishing an interactive Web site and holding three
facilitated public meetings: a ``roundtable'' workshop at NRC
Headquarters, Rockville, MD, on August 10, 2000, and two ``townhall''
meetings--one in Atlanta, GA, on September 20, 2000, and a second in
Oakland, CA, on September 26, 2000. Oral and written comments, received
from the public meetings by mail and through the NRC Web site, in
response to the issues paper were considered in drafting the proposed
rule.
The NRC published the proposed rule in the Federal Register on
April 30, 2002 (67 FR 21390), for a 90-day public comment period. In
addition to approving the publication of the proposed rule, the
Commission also directed the NRC staff to continue the enhanced public
participation process. The NRC staff held two public meetings to
discuss the proposed rule. The first meeting was held in Chicago,
Illinois, on June 4, 2002, and the second was held at the TWFN
Auditorium, NRC Headquarters, on June 24, 2002. In addition, the
Department of Transportation (DOT) staff participated in these
meetings. Transcripts of these meetings were made available for public
review on the NRC Web site. The public comment period closed on July
29, 2002. A total of 192 comments were received. Although many comments
were received after the closing date, all comments were analyzed and
considered in developing this final rule.
Past NRC-IAEA Compatibility Revisions
Recognizing that its international regulations for the safe
transportation of radioactive material should be revised from time to
time to reflect knowledge gained in scientific and technical advances
and accumulated experience, IAEA invited Member States (the U.S. is a
Member State) to submit comments and suggest changes to the regulations
in 1969. As a result of this initiative, the IAEA issued revised
regulations in 1973 (Regulations for the Safe Transport of Radioactive
Material, 1973 edition, Safety Series No. 6). The IAEA also decided to
periodically review its transportation regulations, at intervals of
about 10 years, to ensure that the regulations are kept current. In
1979, a review of IAEA's transportation regulations was initiated that
resulted in the publication of revised regulations in 1985 (Regulations
for the Safe Transport of Radioactive Material, 1985 edition, Safety
Series No. 6).
The NRC also periodically revises its regulations for the safe
transportation of radioactive material to make them compatible with
those of the IAEA. On August 5, 1983 (48 FR 35600), the NRC published a
revision of 10 CFR part 71. That revision, in combination with a
parallel revision of the hazardous materials transportation regulations
of DOT, brought U.S. domestic transport regulations into general accord
with the 1973 edition of IAEA transport regulations. The last revision
to part 71 was published on September 28, 1995 (60 FR 50248), to make
part 71 compatible with the 1985 IAEA Safety Series No. 6. The DOT
published its corresponding revision to title 49 on the same date (60
FR 50291).
The last revision to the IAEA Safety Series 6, Safety Standards
Series ST-1, was published in December 1996, and revised with minor
editorial changes in June 2000, and redesignated as TS-R-1.
Historically, the NRC has coordinated its part 71 revisions with
DOT, because DOT is the U.S. Competent Authority for transportation of
hazardous materials. ``Radioactive Materials'' is a subset of
``Hazardous Materials'' in 49 CFR under DOT authority. Currently,
[[Page 3699]]
DOT and NRC co-regulate transport of nuclear material in the United
States. The NRC is continuing with its coordinating effort with the DOT
in this rulemaking process. Refer to the DOT's corresponding rule for
additional background on the positions presented in this final rule.
Scope of 10 CFR Part 71 Rulemaking
As directed by the Commission, the NRC staff compared TS-R-1 to the
previous version of Safety Series No. 6 to identify changes made in TS-
R-1, and then identified affected sections of part 71. Based on this
comparison, the NRC staff identified 11 areas in part 71 that needed to
be addressed in this rulemaking as a result of the changes to the IAEA
regulations. The NRC staff grouped the part 71 IAEA compatibility
changes into the following issues: (1) Changing part 71 to the
International System of Units (SI) only; (2) radionuclide exemption
values; (3) revision of A1 and A2; (4) uranium
hexafluoride (UF6) package requirements; (5) introduction of
the criticality safety index requirements; (6) type C packages and low
dispersible material; (7) deep immersion test; (8) grandfathering
previously approved packages; (9) changes to various definitions; (10)
crush test for fissile material package design; and (11) fissile
material package design for transport by aircraft.
Eight additional NRC-initiated issues (numbers 12 through 19) were
identified by Commission direction and NRC staff consideration for
incorporation in part 71. These NRC-initiated changes are: (12) Special
package authorizations; (13) expansion of part 71 Quality Assurance
(QA) requirements to Certificate of Compliance (CoC) holders; (14)
adoption of the American Society of Mechanical Engineers (ASME) code;
(15) change authority for Dual-Purpose Package Certificate holders;
(16) fissile material exemptions and general license provisions; (17)
decision on petition for rulemaking on PRM-71-12, Double Containment of
Plutonium; (18) contamination limits as applied to Spent Fuel and High-
Level Waste (HLW) packages; and (19) modifications of event reporting
requirements. The first 18 issues were published for public comment in
an issues paper in the Federal Register on July 17, 2000 (65 FR 44360).
Also, the authority citation for part 71 has been corrected to include
section 234.
This final rule has been coordinated with DOT to ensure that
consistent regulatory standards are maintained between NRC and DOT
radioactive material transportation regulations, and to ensure
coordinated publication of the final rules by both agencies. The DOT
also published its proposed rule regarding adoption of TS-R-1 April 30,
2002 (67 FR 21328).
II. Analysis of Public Comments
As previously stated, the NRC held two facilitated public meetings
in 2002 to discuss and hear public comments on the proposed rule.
(Three other facilitated public meetings were held in 2000 before
drafting the proposed rule.) Each of these meetings was transcribed by
a court reporter. The meeting transcripts and condensed summaries of
the comments made in the meeting are available to the public on the
NRC's interactive rulemaking Web site at http://ruleforum.llnl.gov. and
the Public Document Room (PDR) located at One White Flint North, 11555
Rockville Pike, Room O-1F23, Rockville, MD. The NRC has made copies of
publicly released documents available on the Web site at http://www.nrc.gov/waste/spent-fuel-transp.html.
This section provides a summary of the general comments not
associated with the 19 issues but rather with general topics related to
this rule and the rulemaking process. These are organized under the
following subheadings: Compatibility with IAEA and DOT standards,
Regulatory Analysis (RA) and Environmental Assessment (EA), State
Regulations, Terrorism, Adequacy of NRC Regulations and Rulemaking
Process, Proposed Yucca Mountain Facility, and Miscellaneous (including
comments to DOT). A summary of public comments associated with a
specific issue is included in Section III of this SUPPLEMENTARY
INFORMATION.
Compatibility With IAEA and DOT Standards
Comment. Several commenters generally supported NRC's efforts to be
consistent with IAEA regulations. The particular reasons for this
support varied among commenters but included such issues as approving
of harmonization and encouraging NRC's coordination with DOT. For
example, some commenters stated that harmonization enhances the
industry's ability to import shipments and conduct business in
compliance with both national and international regulations. One
commenter urged the NRC to move swiftly to complete this rulemaking
effort and to remain consistent with DOT regulations. One commenter
stated that uniform international regulations were in the public's best
interest for the safe movement of nuclear materials. Further, this
commenter urged the NRC to accelerate the ``harmonization'' with
international regulations to simplify procedures for companies that
ship nuclear waste both domestically and internationally.
Response. The NRC acknowledges these comments, and the NRC
continues to work to finalize this rule as expeditiously as possible.
As with the issuance of the proposed rule, the NRC will continue to
coordinate closely with the DOT in this effort to ensure consistency
between regulations for the transportation of certain radioactive
materials.
Comment. A commenter supported harmonization but said that adoption
of new or modified requirements into the domestic regulations for
transportation of radioactive materials must be justified in terms of
cost and the need for improved safety and performance. The commenter
added that some of the changes, including the additional technical
complexity of the proposed regulations (e.g., nuclide specific
thresholds), are not warranted based on the history of performance in
the transportation of radioactive materials.
Another commenter noted several areas of incompatibility between
DOT and NRC proposed rules. The commenter also suggested that NRC work
with DOT to agree on a consistent approach in organizing the A1 and A2
values for international shipments in Table A-1. A third commenter
noted that DOT has already issued a proposed rule, HM 232, which
focuses on using the registration program to affect the enhancement and
security of radioactive materials in transport.
Response. NRC's goal is to harmonize our transportation regulations
to be consistent with IAEA and DOT, while ensuring that the
requirements adopted will benefit public health, safety, and the
environment. The NRC has conducted an evaluation of the radionuclide-
specific thresholds (the exemption values), including a regulatory
analysis and an environmental assessment, and concluded that adoption
of these values is warranted, in spite of the technical complexity. NRC
has been working with the DOT. The NRC has completed a regulatory
analysis that supports harmonization in terms of cost and regulatory
efficiency.
Comment. One commenter stated that NRC should use the latest
medical knowledge from independent sources (i.e., not IAEA or
International Commission on Radiological Protection (ICRP) data)
regarding the medical effects of radiation.
Response. The NRC considers a variety of sources of information
[[Page 3700]]
concerning the health effects attributed to exposure to ionizing
radiation. Two primary sources of information are the National Research
Council/National Academy of Sciences (NAS) and the United Nations
Scientific Committee on the Effects of Atomic Radiation (UNSCEAR). Both
groups provide an independent and comprehensive evaluation of the
health risks associated with radiation exposure. The NRC currently is
sponsoring an NAS review of information from molecular, cellular, and
animal studies of radiation, other environmental exposures, and
epidemiologic studies to evaluate and update previous reviews of the
health risks related to exposure to low-level ionizing radiation. These
studies focus on the latest published information available.
Comment. Several commenters questioned the credibility of the IAEA
and the ICRP because these organizations are not publicly accountable.
Three of the commenters further questioned the process of the NRC
simply accepting what the IAEA does, noting that agencies in Europe
have challenged ICRP assumptions. One of these commenters stated that
regulated or potentially regulated bodies should be allowed more
involvement in the IAEA decisionmaking process. Furthermore, the
suggested lack of public involvement led one commenter to express a
general lack of trust for these organizations and question the
credibility of their conclusions. This lack of public involvement was
at issue with another commenter who added that the proposal would only
``make things easier for the transportation and nuclear industries at
the expense of public health.''
Response. The United States is represented at the IAEA for
transportation issues through the DOT acting as Competent Authority
(the official U.S. representative organization). The NRC consults with
DOT on issues related to nuclear material transport. NRC disagrees with
the statement that the NRC simply accepts what the IAEA does. When the
NRC (and the DOT) seeks to amend its regulations to harmonize with
IAEA's, it does so through a deliberate and open process via
rulemaking. The public has been afforded in the past, and will continue
to be afforded, the opportunity to comment on DOT's and NRC's proposed
rulemakings. This effort can result in NRC regulations not matching the
IAEA guidance. Further, the NRC does not ``simply accept'' the IAEA
standards. In many instances, the NRC has chosen to implement
regulations that differ from the IAEA's. Issues 7 and 11 of this final
rule, discussed elsewhere in this SUPPLEMENTARY INFORMATION, are just
two examples of where NRC has differed from the IAEA requirements by
implementing more stringent requirements.
Information on the IAEA and ICRP can be found at their respective
Web sites: www.iaea.org and www.icrp.org. These Web sites provide
background on each organization that should address the concerns about
the credibility of each organization.
Comment. One commenter stated that the burden of proof for
departing from IAEA standards is shifted by the regulators to the
regulated entities. Another commenter suggested that the burden of
proof for rejecting the proposed regulatory changes is being shifted to
citizens and stakeholders.
Response. Both the NRC and DOT are participating members of the
IAEA and have direct input to the development of new transportation
standards. Before DOT or NRC proposes U.S. regulations for
harmonization with IAEA standards, each agency completes a technical
evaluation and makes a determination if each new standard should be
adopted by the U.S. The public involvement process for rulemaking
solicits stakeholders to suggest changes to proposed rule language or
to suggest the rejection of a proposed regulatory change. With
sufficient justification, public comments have resulted in modification
to regulatory text.
Comment. One commenter asked if either NRC standards or IAEA's
could protect the public from ``real world'' problems. The commenter
inquired how NRC accounts for the fact that a cask might burn for
longer than existing standards require it to withstand fire. The
commenter believed that such rationales were particularly relevant in
light of recent incidents, such as the Baltimore Tunnel fire and the
Arkansas River bridge accident.
Response. The NRC notes the questions on how realistic the
transportation standards established by the NRC and the IAEA are. Both
NRC and IAEA standards require that cask designs be able to withstand
hypothetical accident conditions. The conditions bound (or are more
severe than) those conditions that would be expected in the vast
majority of real world accidents and therefore provide protection for
the cask designs. Additionally, the NRC has periodically revisited and
evaluated the effects of actual accidents to look at the forces and the
challenges that would be presented to casks in ``real world''
transportation accidents. For example, in response to the Baltimore
Tunnel fire, the NRC staff has conducted two sets of independent
analyses and has determined that the conditions that existed in the
fire would not have caused a breech of a current spent fuel
transportation cask design had it been located in the tunnel for the
duration of the fire.
Comment. One commenter stated that the timeline by which NRC would
adopt IAEA requirements should be changed. The commenter also stated
that the current 2-year cycle for changes is too frequent.
Response. The timeline for adopting IAEA standards and the cycle
for making changes at the IAEA are beyond the scope of this rulemaking.
Comment. One commenter stated that the proposed rule might allow
weakening of transportation cask safety testing and increase the risk
of the release of radioactive materials during transportation
accidents.
Response. This concern is acknowledged, but the NRC does not
believe that this rule weakens testing standards.
Comment. One commenter stated that all radioactive shipments should
be regulated and labeled so that transportation workers and emergency
responders are aware of the risk.
Response. The comments are acknowledged. DOT regulations include
requirements for labels, markings, and placarding packages and
conveyances of radioactive materials, and training of Hazmat workers.
Existing and proposed regulations for the transportation of radioactive
materials consider the potential risk to workers and emergency
responders of exposure to these materials. The NRC believes the
thresholds for regulation of the transportation of radioactive
materials protect the health and safety of workers and emergency
responders.
Comment. One commenter pointed out that due to the increase in the
number of nuclear shipments, the NRC and DOT must strengthen their
standards to protect the millions of people, thousands of schools, and
hundreds of hospitals residing directly along transportation routes.
Response. The NRC routinely reevaluates the effectiveness of its
regulations to ensure that it is meeting its mission to protect the
public health and safety. In regulating safe and secure transport of
spent nuclear fuel, the NRC has conducted risk studies to consider the
fact that a large number of shipments might be made to a future
geological repository using current generation cask designs. These
studies have confirmed that the current NRC regulations are robust and
protective of the public during transportation of
[[Page 3701]]
spent fuel. Therefore even with an increase in the number of shipments,
these shipments can be made safely in large numbers to a centrally
located storage facility.
Comment. On behalf of the nuclear industry, one commenter said that
harmonization is logical in terms of cost and safety. Harmonized rules
and uniform standards and criteria allow members of the nuclear
industry to know how safe a package is, regardless of where it comes
from. Because many other nations have already adopted many of these
proposed rules, U.S. transporters are already required to meet these
standards in many cases. The commenter also voiced support for
exempting certain domestic shipments from these international
regulations.
Response. Harmonization with TS-R-1 should maintain the safety of
shipments of radioactive materials while eliminating the need to
satisfy two different regulatory requirements (i.e., domestic versus
international shipments). The NRC believes that by clarifying and
simplifying shipping requirements, harmonization will help all who are
involved in the transport of radioactive material to comply
successfully with regulations.
Comment. One commenter stated that there has already been much
deliberation over the proposed regulations. He stated that his
organization and the industry at large have been looking at these
proposed changes for well over 10 years.
Response. The comments are acknowledged.
Comment. One commenter stated that harmonization is a ``value
neutral process'' and isn't necessarily good or bad.
Response. Harmonization can be viewed as a value neutral process,
although the NRC believes that harmonizing domestic and international
regulations generally improves efficiency and safety in the transport
of radioactive material. NRC's proposed changes are based upon the
careful evaluation of specific issues and provisions in TS-R-1. At this
level, the NRC believes that the negative (i.e., costs) or positive
(i.e., benefits) value of a particular change can be assessed
effectively. These costs and benefits have been carefully evaluated in
our decisionmaking process.
Comment. Four commenters opposed harmonizing rules. One commenter
opposed harmonization because it ``appears to be occurring to satisfy
demands of the nuclear industry and affected governmental bodies'' to
facilitate commerce, rather than in the interest of public safety.
Another commenter noted that the primary objective of these changes
should be to protect public health, safety, and the environment.
Another commenter argued that harmonization should not be used as a
justification for violating a country's sovereignty or a State's right
to maintain stringent standards. The commenter said that U.S. rules
were already harmonized before these proposed changes and that the
authors of international regulations should not dictate U.S.
regulations. The fact that other countries have adopted the IAEA
regulations is not sufficient justification for the U.S. to adopt these
regulations. The commenter agreed that some degree of harmonization
makes sense but emphasized that the U.S. needs to maintain control over
its own rules.
Response. The IAEA periodically updates international regulations
for the safe transport of radioactive material in response to advances
in scientific knowledge and technical experience. These changes are
implemented with the purpose of improving public safety, as well as
facilitating commerce. The U.S. has substantial input into the IAEA
development of these periodic revisions through official representation
by the DOT. While the NRC aims to harmonize its regulations closely
with those issued by the IAEA, NRC independently evaluates proposed
changes in the interest of protecting public health, safety, and the
environment. This rule reflects this extensive process; NRC routinely
suggests adoption or partial adoption of certain provisions and
nonadoption of others.
Comment. Two commenters asked if NRC could quantifiably prove that
harmonization is necessary. One asked if NRC's failure to comply with
the IAEA regulations has disrupted commerce or jeopardized public
safety, and whether members of the international community have accused
the U.S. of disrupting commerce by not complying with these
regulations.
Response. DOT and NRC accomplish harmonization by adopting domestic
rules that are compatible with international rules. DOT and NRC rules
may differ from those of IAEA where it is necessary to reflect domestic
practices. However, these differences are kept to a minimum because
regulatory differences can lead to confusion and errors and can result
in unsafe conditions or events. U.S. failure to comply with
international safety regulations could easily result in disruption of
U.S. participation in international radioactive material commerce, with
no commensurate justifiable safety benefit, because other IAEA Member
States are under no obligation to accept shipments that do not comply
with international regulations.
Comment. One commenter wanted to know how the IAEA drafted its
regulations and statistics. The commenter questioned who the IAEA is
and why NRC should accept its statistics. The commenter also asked how
much input the American public has had on these regulations and noted
that Congress and the public have previously rejected IAEA regulations.
Response. The comments concerning the IAEA standards development
process and U.S. citizen input to that process are both beyond the
scope of this rulemaking. However, as noted in the public meetings held
to obtain comments on the proposed rule, DOT is mandated by law to help
formulate international transportation standards, and to ensure that
domestic regulations are consistent with international standards to the
degree deemed appropriate. The law permits DOT the flexibility to
accept or reject certain of the international standards. The NRC/DOT
evaluation of the IAEA standards has resulted in the two parallel sets
of final rule changes. Rejection of an IAEA standard could be based on
technical criteria as well as on public comment on proposed rules. The
IAEA has Member States that develop standards as a collegial body, and
the U.S. is one of those Member States.
Comment. Several commenters urged NRC to improve its scientific
understanding and basis for the proposed rulemaking. Two commenters
suggested that NRC complete the comprehensive assessments of TS-R-1 and
future IAEA standards, the Package Performance Study (PPS), and full-
scale cask tests before proceeding with this rulemaking. A commenter
stressed that ICRP does not represent the full range of scientific
opinion on radiation and health and ignores concepts such as the
bystander effect and synergism of radiation with other environmental
contaminants. This commenter also stated that the exposure models used
to justify certain exposure scenarios are inadequate.
Response. The NRC acknowledges these comments and notes that NRC
participates or monitors the work of major, national and international,
scientific organizations in the fields of health physics and radiation
protection. As such, NRC has access to the latest scientific advances.
Moreover, the NRC has completed an assessment of TS-R-1 as part of the
development of this rule. The PPS is a research project independent of
this rulemaking. Also,
[[Page 3702]]
see the following comment regarding the ICRP.
Comment. Several commenters stated that the IAEA rulemaking process
is not democratic, and their documents are not publicly available and
were developed without public knowledge or input. One commenter
suggested that the public should have had an opportunity to ``comment
on or otherwise participate in the earlier formation of the IAEA
rules.'' Another commenter proposed that the NRC act as an intermediary
between public opinion and IAEA by improving communications with the
public and regulated bodies, providing advanced notice of rulemakings,
and receiving comments on proposed rules.
Response. The NRC acknowledges the comments about the IAEA
rulemaking process, the ICRP representation of scientific opinion, and
the observation on NRC's role as intermediary between the American
public and the IAEA, but each of these comments brings up issues that
are beyond the scope of the proposed rulemaking. Therefore, no changes
were made to this rulemaking. The NRC notes that the IAEA has begun to
discuss ways to foster public participation in its standards
development process.
Comment. Several commenters stated that IAEA and ICRP regulations
should not dictate domestic U.S.-based regulations. Two commenters
stated that IAEA does not necessarily consider the risk-informed,
performance-based standards that are important to rulemaking in the
U.S. The commenters added that the NRC must recognize that while IAEA
standards generally have good technical bases, they are consensus
standards that do not necessarily consider the risk-informed,
performance-based aspects of regulations that we have developed in the
U.S.
Response. The NRC acknowledges the comment about IAEA and ICRP
regulations dictating U.S. based regulations and notes that this
comment is not accurate and is considered to be an opinion. The NRC is
a participating member of both the IAEA and the ICRP, and neither body
dictates to the NRC what regulations or standards must be adopted. As a
participant, the NRC suggests transportation standard changes and as
such, the NRC both proposes and comments on the language of new
standards. This participation permits the NRC to infuse its ideas on
risk-informed regulations, when possible.
Comment. The effort to harmonize regulations was supported by
several commenters. One commenter spoke for Agreement States and
expressed support for harmonizing regulations. Two others explained
that the benefit of harmonization would be consistent national and
international regulations and improved safety, yet U.S. regulators (and
regulations) would retain the legal authority to act when and as
necessary. Another commenter emphasized that given how new information
is found all the time and the IAEA is on a 2-year standards revision
schedule, it does not make sense to hold back harmonizing U.S.
standards with international standards pending the outcome of any
studies.
Response. The NRC believes that its effort to promote regulatory
harmonization will maintain and/or improve safety, increase regulatory
efficiency and effectiveness, as well as reduce unnecessary regulatory
burden. The NRC's aim is to harmonize its regulations with IAEA
regulations by adopting many of the provisions in TS-R-1. However, the
NRC does not propose wholesale adoption of TS-R-1, but only when
adoption provides the best opportunity to maintain and/or improve
public safety, health, and the environment.
Regulatory Analysis (RA) and Environmental Assessment (EA)
Comment. Several commenters found the RA to be deficient in various
aspects. One commenter asserted that updated quantitative data should
be included in the RA that would include the following information: the
number of exempt and nonexempt packages; the number of exempt and
nonexempt shipments; the average number of packages per shipment; and
the detailed information on curie counts by shipment categories. The
commenter noted that all stakeholders are affected by these
deficiencies, notably public information groups and Western States.
Two commenters focused on the RA's cost analysis with one stating
that no changes should be made without a cost analysis and the other
stating that the RA had not adequately considered the cost of the
proposed rule. The second of these commenters stated that specific dose
information, calculations, and information regarding the impact of the
new regulations should have been included in the draft RA and EA. They
found the RA to be deficient because of its failure to recognize likely
impacts of the changes to the double containment of plutonium
regulations, particularly regarding the agreement between the Western
Governors' Association, the individual Western States, and the
Department of Energy (DOE) for a system of additional transportation
safeguards.
Response. Quantitative data was requested throughout the rulemaking
process. These requests were made during the development of the
proposed rule, and a request was again made in the proposed rule. Where
this information was available, it was used in the development of NRC's
proposed positions. To the extent that information was provided, it has
been considered in the development of NRC's final position.
Comment. One commenter asserted that the proposed rule is a major
Federal action, thus deserving of a full Environmental Impact Statement
(EIS). The commenter also stated that an EIS dating from 1977 and a
study dating from 1985 do not suffice as adequate analysis of the
proposed rule's impact, due to changes ``in population, in land use, in
the transportation system, in laws, in issues of national security.''
Response. NRC acknowledges this comment and notes that it has
prepared an EA. Based on the results of the EA, the NRC staff has
concluded that this rule is not a major Federal action requiring an
EIS. As noted in the proposed rule, NRC is interested in receiving
additional data, and to the extent that the data was received, it was
included in the analyses leading up to the final rule.
Comment. One commenter said that the EA and the rulemaking are too
carefully tied together. The commenter said that this fact precludes
NRC from actually finding an environmental impact from the rule.
Response. The draft EA is a study that is required as part of a
rulemaking to ensure that the potential impacts to public health and
safety and the environment are adequately evaluated as part of the
decisionmaking process. As such, the rule and the EA are necessarily
``tied together.''
Comment. Two commenters found the EA to be deficient in various
aspects. One commenter stated that specific dose information,
calculations, and information regarding the impact of the new
regulations should have been included in the draft EA and RA.
A commenter believes that the EA and RA lack the following pieces
of information: the number of exempt and nonexempt packages; the number
of exempt and nonexempt shipments; the average number of packages per
shipment; and the detailed information on curie counts by shipment
categories. One commenter believes that the EA should include
transportation scenarios, updated data rather than 1982 data, and a
quantitative analysis along with a qualitative analysis.
[[Page 3703]]
The NRC was criticized for a portion of the EA (page 43), which
first identifies information necessary to make a risk-informed decision
on the proposed regulation and then discusses the lack of information
in the EA. The commenters noted a discrepancy in NRC's efforts,
particularly the number of NRC staff and resources devoted to this
rulemaking for the past 2 years versus the lack of resources devoted to
updating the 1982 data. They stated that the costs associated with the
Type C package changes were not included in the EA and that process
irradiators are shipping sources equaling about 50 million curies, much
greater than the curie count listed in the proposed rulemaking.
Response. The NRC acknowledge the comments regarding the lack of
information in some portions of the draft RA and EA. The draft EA and
RA were developed based on the best information available to the NRC at
the time. Moreover, NRC solicited in the proposed rule FRN, additional
information on the costs and benefits of the proposed requirements,
including the Type C package changes. All the information received has
been considered in NRC's final decision. The NRC staff notes that the
majority of the proposed changes are such that the specific dose
information and calculations are not required to determine the
appropriateness of adopting or not adopting the change being
considered.
Comment. One commenter expressed concerns about NRC's findings of
``no significant impact'' on radionuclide-specific activity values for
a number of issues. The commenter requested that more detailed
information be provided ``on how many and which radionuclide levels
will rise or fall'' as a result of proposed changes. The commenter also
asked the NRC to define its use of ``significantly'' and to explain how
it determined the level of ``risk.''
Response. Detailed information on the identity of radionuclides
whose specific activity values rise or fall relative to the previous
definition of 70 Bq/g (0.002 [mu]Ci/g) may be determined by inspection
of Table A-2. The context for ``significantly'' is provided in the
background section. NRC has used estimated dose to the public, as
determined through the use of radionuclide transport scenarios, as an
indicator of risk.
State Regulations
Comment. One commenter asked if these new regulations would
threaten a State's right to regulate radioactive materials that NRC has
deregulated. Two commenters stated opposition to the proposed rule due
to their belief that it would lower standards. The first commenter
stated that the proposed rule would override State and local laws that
are stricter than Federal regulations while the second commenter stated
that the proposed rule would reduce environmental protection. Four
commenters added that ``harmonization'' with international law was a
poor and ultimately insufficient justification to weaken U.S.
regulations.
Response. State and local governments do not have authority to set
regulations for the transportation of radioactive materials that are
stricter or more stringent than those of the Federal government. In
accordance with section 274b of the Atomic Energy Act, as amended,
Agreement States programs must be compatible with those of the NRC for
the regulation of certain radioactive materials to assume authority for
the regulations of these materials from the NRC. Because of this, the
Commission developed the ``Policy Statement on Adequacy and
Compatibility of Agreement State Programs'' which became effective on
September 3, 1997 (62 FR 46517). One of the provisions of this Policy
Statement is that an Agreement State should adopt program elements that
apply to activities that have direct and significant effects in
multiple jurisdictions' elements in an essentially identical manner as
those of the NRC (see definition of Compatibility Category B in section
VI of this notice). This is needed to eliminate any conflicts,
duplications, gaps, or other conditions that would jeopardize an
orderly pattern in the regulation of radioactive materials on a
nationwide basis. Those part 71 requirements applicable to materials
regulated by Agreement States are designated as Category B and must be
adopted in an essentially identical manner as those of the NRC because
they apply to activities that have direct and significant effects in
multiple jurisdictions.
Terrorism Concerns
Comment. Six commenters expressed concern with the increased threat
of terrorism and its impact on radioactive material transport. One
commenter suggested that shipping standards be strengthened due to both
an increased threat of terrorist attacks and the decline in rail,
highway, air, and waterway infrastructure. Two commenters stated that
they were concerned that many of the new regulations would make
transported radioactive material more vulnerable to terrorist attacks
and wanted to know how NRC anticipated responding to the threat of
these attacks. Three commenters mentioned that the threat of terrorism
should be taken into account when changing container regulations, with
one commenter highlighting double versus single containment of
plutonium. The final commenter stated that the NRC should reconsider
the scope of the proposed rule due to the ``altered circumstances of
our nation's vulnerability to terrorist attack.'' The commenter also
suggested that the proposed rule be withdrawn and that the NRC
``recalculate the full adverse consequences and the full long-term
financial, health, and environmental costs to the public, the nation,
and the economy of worst case terrorist actions.'' The commenter also
stated that in a time of increased national security threats, the
safety of containerization must be maximized.
Response. As discussed on the NRC's Web site (see www.nrc.gov/what-we-do/safeguards/911/faq.html), most shipments of radioactive materials
involve materials such as pharmaceuticals, ores, low-level radioactive
waste, and consumer products containing radionuclides (e.g., watches,
smoke detectors). A variety of Federal and State government agencies
regulate the shipment of radioactive materials.
High-level nuclear waste materials, such as spent nuclear fuel, are
transported in very heavy, robust containers called ``casks.'' Over the
past 30 years, approximately 1300 shipments of commercially generated
spent fuel have been made throughout the U.S. without any radiological
releases to the environment or harm to the public. Federal regulations
provide for rigorous standards for design and construction of shipment
casks to ensure safe and secure transport of their hazardous contents.
Casks must meet extremely demanding standards to ensure their integrity
in severe accident environments. Therefore, the design of casks would
make any radioactive release extremely unlikely. After September 11,
2001, the NRC issued advisories to licensees to increase security
measures to further protect the transportation of specific types of
radioactive materials, including spent fuel shipments. Additional
measures have been imposed on licensees shipping specific quantities of
radioactive material.
Comment. Another commenter, who lives near a route proposed for
shipping nuclear waste across the country, recommended that NRC
strengthen radioactive transport regulations. One commenter opposed the
adoption of new transport regulations that reduce
[[Page 3704]]
the protection to the public from transporting nuclear wastes.
Response. The NRC believes that the regulations contained in part
71 adequately protect public health and safety. The changes being
adopted will not result in any undue increase in risk to public health,
safety, or the environment.
Comment. Several commenters were concerned that the proposed
regulations may increase vulnerability to terrorist threats using
radioactive materials. A commenter believes that labeling radioactive
materials could aid terrorists by identifying the packages as
radioactive, while another commenter stated that shipments with or
without labels provided potential terrorists with the materials for a
dirty bomb. Another commenter requested that NRC put protective
measures into place at ports and to guard all nuclear shipments with
U.S. military forces. One commenter stated that nuclear shipments
should be transported at off-peak hours while all side roads, tunnels,
bridges, overpasses, railroad crossings, access to exit ramps, etc.,
should be secured before the transport vehicle arrives, and that NRC
should create a ``vehicle-free'' buffer zone ahead and behind the
shipment. This same commenter advocated FBI background checks on all
transporters, drivers, and crew workers involved with nuclear
transport. Two commenters asserted that all new rules should be mindful
to the threat of terrorism, which would be superior to considering
terrorism in separate rules.
Response. The NRC acknowledges these comments and notes that NRC
has taken immediate regulatory actions to address the potential for
terrorist activities; these include issuing orders and advisories to
its spent fuel licensees prior to initiating rulemaking which takes a
longer time, and initiating shipment vulnerability studies. Also, the
NRC will make the necessary rule changes, based on these studies, as
appropriate. Moreover, the NRC staff notes that several of the comments
above were addressed in recent regulations (March and May, 2003), which
were published jointly by the Department of Homeland Security and the
DOT requiring shippers and carriers to submit security plans and
requiring background checks on drivers.
Adequacy of NRC Regulations and Rulemaking Process
Comment. Three commenters believe that the NRC should better
account for low-level radiation. One commenter stated that NRC should
use the latest medical knowledge from independent sources (i.e., not
IAEA or ICRP data) regarding the medical effects of radiation. Another
commenter stated that low-level radiation could cause cell death,
cancer, genetic mutations, leukemia, birth defects, and reproductive,
immune, and endocrine system disorders. This commenter added that long-
term exposure to low levels of ionizing radiation could be more
dangerous than short-term exposure to high levels. Another commenter,
who was similarly concerned with low dose and low dose-rate radiation,
stated that ``arguments of nuclear industry proponents that new
information need not be considered is invalid and since the NRC's legal
mandate is to protect the public's health and safety'' the NRC needs to
consider ``cautionary information that is now available in the peer
reviewed literature.'' The commenter suggested that NRC not focus on
the ``standard man'' but instead focus on the ``most susceptible
portions of the population--ova, embryo, fetus, rapidly growing young
child, elderly, and those with impaired health'' when drafting
regulations. Lastly, the commenter implied that NRC should attempt to
``assess and incorporate impacts of additive exposures to other forms
of life and to ecosystems'' as well as the impacts associated with ``an
individual recipient of the combinations of and synergies among
radiation and other contaminants to which people are exposed.''
Response. As discussed on the NRC's Web site (see http://www.nrc.gov/reading-rm/doc-collections/fact-sheets/bio-effects-radiation.html, radiation may kill cells, induce genetic effects, and
induce cancer at high doses and high dose rates. However, for low
levels of radiation exposure at low dose exposure rates, health effects
are so small they may not be detected. No birth defects or genetic
disorders among the children born to atomic bomb survivors from
Hiroshima and Nagasaki have been observed at low doses of radiation,
i.e., < 25 rad (Chapter 6, ``Other Somatic and Fetal Effects,'' of Beir
V, Health Effects of Exposure to Low Levels of Ionizing Radiation;
National Research Council, 1990). Consequently, few if any similar
effects are expected from exposure to low doses of ionizing radiation.
Moreover, there is no epidemiology data, published in peer reviewed
journals, to support the concern expressed by the commenter that long-
term exposure to low levels of radiation may be more dangerous than
short-term exposures to high levels. Humans have evolved in a world
constantly exposed to low levels of ionizing radiation. The average
radiation exposure in the U.S. from natural sources is 3.0 mSv (300
mrem) per year. Although radiation can have health effects at high
doses and dose rates, for low levels of radiation exposure at low dose
exposure rates, the incidence of biological effects is so small that it
may not be detected. For example, information developed by the Health
Physics Society suggests that the incidence of health effects, if they
exist below 10,000 mrem (100 mSv), is too small to be observed. People
living in areas having high levels of background radiation--above 10
mSv (1,000 mrem) per year, such as Denver, Colorado, have shown no
adverse health effects.
The NRC actively and continually monitors research programs and
reports concerning the health effects of ionizing radiation exposure.
NRC staff monitors the Low Dose and Low Dose Rate Research Program
sponsored by the Department of Energy (DOE). The research project is
designed to better understand the biological responses of molecules,
cells, tissues, organs, and organisms to low doses of radiation. NRC
also is co-funding a review of the Biological Effects of Ionizing
Radiation (BEIR) by the National Research Council. The BEIR committee
will also review and evaluate molecular, cellular, and animal exposure
data and human epidemiologic studies to evaluate the health risks
related to exposure to low-level ionizing radiation. Both groups
provide a comprehensive evaluation of the health risks associated with
radiation exposure.
Finally, existing regulatory guidance suggests that protection of
individuals (humans) is also protective of the environment. IAEA
Technical Report Series No. 332 (Effects of Ionizing Radiation on
Plants and Animals at Levels Implied by Current Radiation Protection
Standards) suggests that, in most cases, the environment is being
protected by protecting humans.
Individuals in occupational or public areas may be exposed to
radiation and chemical exposure which result from materials present in
these areas. The NRC, however, has no regulatory authority over any of
the materials present other than source, byproduct, or special nuclear
material. In many situations, exposures to chemicals and non-NRC
regulated materials are under the purview of the U.S. Environmental
Protection Agency (EPA).
Comment. Seven commenters opposed the proposed rule because of
increased exposure, danger to public health, and increased public
health risk.
Response. The NRC disagrees that the proposed rulemaking will
result in any significant increase in exposure,
[[Page 3705]]
endangerment to public health, or increase in health risk. See earlier
comment responses for further details.
Comment. One commenter stated that U.S. agencies have not
adequately represented public opinion regarding transportation safety.
The commenter was concerned that the number of irradiated fuel and
plutonium shipments in the nation will increase as the proposed
regulations weaken container safety standards.
Response. The DOT and NRC represent the United States before the
IAEA, DOT as the U.S. Competent Authority supported by the NRC. Both
agencies are aware of public opinion regarding transportation safety in
the United States. The NRC disagrees with the comment that U.S.
agencies have not adequately represented public opinion. Additionally,
NRC and DOT prepare their rules in compliance with Administrative
Procedure Act (APA) requirements. The APA requires that public comments
be requested, considered, and addressed before a final rule is adopted
unless there are exigent reasons to bypass the public comment process.
Although the number of irradiated fuel and plutonium shipments in
the future may increase, the number of shipments to be made is
independent of this final rule. Lastly, the comment that the regulation
weakens transportation container safety standards is a statement of
opinion without supporting data or information.
Comment. One commenter suggested that NRC staff needs to address
fully any comments submitted by the public, even when the NRC might
consider these comments beyond the scope of the proposed rule.
Response. Although NRC is careful to address all comments with the
scope of the rulemaking, there are instances when a comment is
sufficiently outside the scope of a proposed action that it need not be
addressed. NRC resources need to be used to address issues related to
the rulemaking for efficiency and effectiveness.
Comment. One commenter stated that the proposed rule did not
specifically incorporate ``issues to improve the protective adequacy of
the regulations'' that were raised by the public during meetings held
in 2000. The commenter stated that ``changes that were adopted in
response to public comments in 2000 must be specified in a revised
Proposed Rule.'' The commenter also asked that further public meetings
be held before DOT and NRC proceed with further revisions of the
transportation regulations.
Response. The current rule stems from NRC's scoping efforts in
2000, and no rule changes were adopted by the Commission at that time.
For this proposed rulemaking, public meetings were held in Chicago, IL,
as well as in Rockville, MD (as previously noted). NRC accepted and
included all comments received, even those received after the July 29,
2002, deadline. For these reasons, the NRC believes its proposed
rulemaking meets the intent of conducting an ``enhanced public
participation process.''
Comment. Eleven commenters requested an extension to the comment
period. One commenter said that the proposed rule is written in a
manner difficult for the public and even watchdog groups to understand.
Because the proposal would affect large portions of the general public
by dramatically changing the standards of radioactive transport, the
commenter urged the NRC to extend the comment period. Two commenters
suggested that the NRC extend the comment period 180 additional days
beyond the July 29, 2002, deadline to allow both the public and the NRC
more time for further consideration. Commenters added that the proposed
rule was not urgent and required further analysis and research.
Finally, one commenter stated that the proposed rule's July 29, 2002,
deadline for receipt of public comments would prevent it from
accounting for the impact of Yucca Mountain. The commenter suggested
that a 1- or 2-month rulemaking extension would be beneficial.
Response. The NRC believes the 90-day public comment period was of
sufficient length, especially in view of the availability of the
proposed rule on the Secretary of the Commission's Web site for over a
year (i.e., the Commission decided to make the proposed rule available
to the public in March 2001, while it was under consideration).
Therefore, the public had the opportunity to comment prior to the
official comment period. Moreover, while not required to do so, the NRC
chose to accept and consider comments received after the July 29, 2002,
deadline. Further, as part of the NRC public participation process, NRC
held two open meetings accessible to the public at which the NRC
answered questions on the proposed rule and accepted comments. As part
of the proposed rule, the NRC solicited additional information from the
public which was considered in the development of the final rule.
Comment. One commenter suggested that the NRC separate the comment
period for the EA and RA from the comment period for the proposed rule.
Response. The commenter's suggestion is noted but is not feasible
to implement because the proposed rule and its supporting RA and EA
must be considered concurrently within the rulemaking proceeding.
Comment. One commenter asked if there is any systematic process by
which the NRC has performed or will perform a cost-benefit analysis of
these proposed regulations.
Response. Whenever the NRC pursues a cost-benefit analysis
(otherwise known as a regulatory analysis), the NRC works diligently to
ensure that monetized, quantitative, and qualitative data are included.
These data are studied to avoid including faulty and/or misleading
data. The draft regulatory analysis in NUREG/CR-6713 has been revised
to take into account the quantitative and qualitative data contained in
the public comments on the proposed rule.
Comment. Two commenters asked for clarification of the proposed
rulemaking's scope in light of the May 10, 2002, letter from Commission
Chairman Richard A. Meserve.
Response. Former Chairman Meserve's May 10, 2002, letter to Senator
Richard Durban provides information on questions posed by the Senator
on transportation of spent fuel and nuclear waste to the proposed
repository at Yucca Mountain, Nevada. The letter provides information
on the NRC's certification process of cask designs, the safety record
of spent fuel casks, and the NRC's authority with respect to
transportation of radioactive materials and its relationship with DOT
and DOE. The issues raised by this letter do not affect the amendments
to part 71.
Comment. One commenter asked if the NRC was aware that, on February
23, 2002, Chicago Mayor Richard M. Daley and 17 other mayors signed a
letter to President Bush that expressed concerns about nuclear waste
transportation. The commenter also made reference to the fire in the
Baltimore tunnel and wondered about safety if the fire had involved
radioactive materials.
Response. The NRC searched its Agency Wide Document Access and
Management System (ADAMS), and no record was found for this letter;
however, the NRC is aware of concerns about spent nuclear fuel
transportation issues that have been voiced by public officials. There
has been significant interest in the Baltimore tunnel fire that
occurred on July 18, 2001, by State and local officials, and the impact
that such a fire might have had on a shipment of
[[Page 3706]]
spent nuclear fuel, had such a shipment been in the tunnel during the
time of the fire. In response to the Baltimore Tunnel fire, the staff
has conducted two sets of independent analyses and has determined that
the conditions that existed in the fire would not have caused a breech
of a spent fuel transportation cask of recent design vintage had it
been located in the tunnel for the duration of the fire.
Comment. One commenter stated that changes in the scientific
community's understanding of radiation injury would affect the risk
assessments and other aspects of the proposed rule. The commenter said
that both the DOE Biological Effects Division's and NASA's study of the
impacts of low dose radiation impacts may require that NRC reconsider
its current standards.
Response. The DOE is funding a 10-year Low Dose Radiation Research
Program to understand the biological responses of molecules, cells,
tissues, organs, and organisms to low doses of radiation. Using
traditional toxicological and epidemiological approaches, scientists
have not been able to demonstrate an increase in disease incidence at
levels of exposure close to background. Using new techniques and
instrumentation to measure biological and genetic changes following low
doses of radiation, it is believed that a better understanding will be
developed concerning how radiation affects cells and molecules and
provide a more complete scientific input for decisions about the
adequacy of current radiation standards. These data are reviewed by
other groups like NAS and UNSCEAR to provide an independent review of
this health effects information. NRC reviews the programs and data
being generated by the DOE and NASA-sponsored research as well as the
reports published by the NAS and UNSCEAR. All of these data sources are
used by the NRC for estimating radiological risk, establishing
protection and safety standards, and regulating radioactive materials.
Comment. Several commenters expressed concern and doubts about the
data used to develop the proposed rule and the information the NRC
provided to support its proposal. One commenter urged NRC to ensure
that the adopted rule represents a risk-informed, performance-based
approach. Two commenters criticized the proposed rule for not
accounting for an expected increase in radioactive shipments. Given
such an increase, one commenter criticized the NRC for using 20-year
old data to justify rule changes that will reduce public safety. This
commenter claimed that the data was out-of-date, inaccurate, not
independently verified, and did not consider the concepts of
radiation's synergistic effects when combined with other toxins.
Another commenter argued that DOT and NRC should use more current data
and future projections including the expected increases in actual
nuclear shipments to estimate the impacts of the rule change. Realistic
scenarios and updated data must be used to project doses and thus
estimate the impacts of the proposed rule's changes, rather than
relying on old data, ICRP, and reliance on computer model scenarios (or
simply stating the lack of data). In addition, DOT and NRC should
include the expected increases in actual nuclear shipments. Another
commenter expressed doubt that the proposed rule's technical benefits
are legitimate and stated that these benefits are not supported in the
draft EA. One commenter stated that the NRC should wait to adopt any
new regulations until there is more information available about the
costs and benefits of such regulations.
Response. The IAEA developed its latest standards through a
cooperative process where experts from member nations proposed and
supported changes to the previous version of the safety standards. The
NRC has provided detail on the justification for the proposed changes
in the statements of consideration for this rulemaking. The commenter
did not provide sufficient detail on which data were of concern for NRC
to further address.
The comment that the NRC is relying on 20-year old data for
justification of its regulations is unfounded. The NRC has completed
risk studies related to the safety of transportation as recently as
2001 and is currently engaged in a research program that will include
the full scale testing of casks, to demonstrate the robust nature of
certified cask designs.
The comments about the quality of data and benefits are considered
to be the opinion of the commenter and were not substantiated. Lastly,
the NRC notes that a cost-benefit analysis has already been conducted
and is reflected in the NRC's RA.
Comment. Four commenters expressed concern that there is inadequate
quantitative data to support the risk-based approach of the proposed
rule and that some of the provisions are based on incorrect or outdated
information. Two commenters were specifically concerned that DOE and
some commercial nuclear facilities are negligent in keeping radiation
exposure and release records. These commenters questioned how NRC data
was gathered and noted that a failure to keep accurate records
constrains NRC's ability to determine whether the proposed
harmonization is economically justifiable. Furthermore, these
commenters added that lack of records undermines the NRC claim that
hundreds of thousands of radioactive material shipments are conducted
safely every year.
Response. See response to the previous comment. Also, the NRC notes
that the commenter's statements regarding DOE and commercial
facilities' negligence is an opinion and was not supported by factual
evidence.
Comment. Three commenters stated that pertinent documents and data
were not readily available or were too difficult to access for the
general public. One commenter requested improved public access to
``sources of codes and IAEA documents that were cited by reference in
the draft'' rule.
Response. The NRC staff worked diligently to ensure that rulemaking
documents, including all supporting documents, were available either
electronically, over the internet, or in hard-copy upon the public's
request in a timely fashion. This includes facilitating public access
to the internet site of the publisher of IAEA documents in the U.S.
Comment. Four commenters stated that the NRC should finish the PPS
and consider its results before finalizing the proposed rulemaking as
well as the rules governing irradiated fuel containers. Another
commenter requested that the PPS be completed and thoroughly analyzed
before this rulemaking is carried out because the current design
requirements for irradiated fuel containers are inadequate and should
be improved.
Response. The NRC believes that shipments of spent fuel in the U.S.
are safe using the current regulations and programs. This belief is
based on the NRC's confidence in the shipping containers that it
certifies, ongoing research in transportation safety, and compliance
with safety regulations and the conditions of certificates that have
resulted in an outstanding transport safety record. Thus, an
established system of regulatory controls protects every U.S. shipment
of spent fuel from commercial reactors. The NRC sponsored PPS is part
of an ongoing confirmatory research program to reassess risks as
shipment technologies change and analytical capabilities improve.
Comment. Three commenters urged the NRC to require more stringent
testing of transport packages in real-world (not computer-modeled)
testing.
[[Page 3707]]
Response. NRC regulations permit certifications through testing,
analyses, comparison to similar approved designs, or combinations of
these methods. A full scale testing is not necessary for the NRC to
achieve confidence that a design satisfies the regulatory tests, as
long as the analyses are based on sound and proven analytic techniques.
Comment. One commenter suggested that the NRC ensure that the
economic value of these regulations is not skewed. That is, the
commenter does not want the needs of one particular industry to shape
the regulations, when the regulations could have a greater impact on a
different industry.
Response. The overall value or impact of the proposed changes
results from the interaction of several influencing factors. It is the
net effect of the influencing factors that governs whether an overall
value or impact would result for several different attributes (i.e.,
different industries or the public). Similarly, a single regulatory
option could affect licensee costs in multiple ways. A value-impact
analysis, such as was undertaken as part of this rulemaking effort,
quantifies these net effects and calculates the overall values and
impacts of each regulatory option. A decision on which regulatory
option is recommended takes into account the overall values and impacts
of the rulemaking.
Comment. One commenter stressed that when the NRC has decision
makers review public comments, the NRC staff should look at primary
documents instead of summary documents. The commenter cited NUREG/CR-
6711 as an example where the regulator runs the risk of having decision
makers read summaries of public comments without understanding the
underlying context and content.
Response. In our decisionmaking process, the NRC did not rely on a
summary document to support the development of the proposed rule. NRC
used primary documents to fully understand the underlying context and
content of the technical information. The summary documents the
commenter refers to were developed to provide the public with a
comprehensive, yet condensed, version of the underlying information.
Further, these underlying documents were also made available to the
public on the NRC Web site during the rulemaking process.
Comment. One commenter asked which countries have already adopted
the proposed guidelines.
Response. The IAEA has conducted a survey that provides the status
(as of July 1, 2003) of each Member State's plans for implementing TS-
R-1. Based on that survey, many States have already implemented the new
requirements of TS-R-1 (e.g., European Commission, Germany, and
Australia). Other States have indicated that they are actively
implementing these requirements and intend to finalize implementation
by the end of 2003. No State indicated that it would not adopt these
standards. This survey is available at http://www-rasanet.iaea.org/downloads/radiation-safety/MSResponsesJuly1 2003.pdf
Comment. One commenter requested clarification on NRC assumptions
for future radioactive materials transportation. Specifically, the
commenter wanted to know whether NRC is assuming the amounts will
increase or remain consistent with past levels.
Response. The NRC's draft RA and EA relied on existing information
to determine the future impacts of the proposed changes. NRC solicited
information on the costs and benefits for each of the proposed changes
as part of the proposed rule. The NRC considered available information
on future radioactive material shipments in its decisionmaking process.
Information that was received as part of the public comment process was
considered in developing NRC's final position. The NRC staff conducted
some sensitivity studies, see for example Comparison of A1
and A2 new and old values in the EA, Table A-1, Appendix A.
Comment. Three commenters opposed weakening regulations that would
reduce the public safety and health through new definitions or accepted
concentration values. One commenter worried that the proposed rule
would weaken regulatory control, allowing increased quantities of
radioactive materials and wastes ``into the lives of individual
citizens without their knowledge or approval,'' thus violating ``the
most fundamental premises of radiation protection.''
Response. The NRC acknowledges the concerns but believes that the
rule continues to protect the public's health and safety in a risk-
informed manner.
Comment. One commenter particularly opposed NRC and DOE studies,
including the EIS to review alternative policies for disposal and
recycling of radioactive metals. The commenter requested that the NRC
maintain stringent controls on all materials being recycled, disposed,
or otherwise reused. Two commenters expressed opposition to the
proposed rule due to a belief that the proposed rule would deregulate
radioactive wastes and materials and allow the deliberate dispersal of
radioactive materials into raw materials and products that are used by
the public and are available on the market.
Response. The NRC acknowledges the commenters' references to DOE
and NRC studies related to the disposal and recycling of radioactive
metals. This rule is not related to the referenced studies.
Comment. One commenter expressed concern that NRC's proposed
regulations could increase the variety of materials that are regulated
as ``radioactive'' for transportation purposes.
Response. The rule does not expand the scope of regulated
radioactive material.
Comment. One commenter expressed concern that the proposed rule
enables commercial and military nuclear industries to ``revive and
expand, thereby generating ever more wastes to be stored, transported
and ultimately * * * sequestered from the biosystem.''
Response. The comment is beyond the scope of this rulemaking.
Proposed Yucca Mountain Facility
Comment. One commenter expressed opposition to sending shipments of
nuclear materials to the proposed Yucca Mountain facility.
Response. Potential shipments to the proposed geologic repository
at Yucca Mountain are beyond the scope of this rulemaking.
Comment. Two commenters raised issues related to the possible
approval of the Yucca Mountain site. One commenter expressed concern
about the safety of dry casks. The commenter asked if the NRC was aware
of the accident at the Point Beach Nuclear Plant in Wisconsin on May
28, 1996, and how similar the dry casks that will ship radionuclides to
Yucca Mountain will be to the casks used at Point Beach. The commenter
noted that once one buries a dry cask, one cannot change it; therefore,
the U.S. will have to be sure that it uses safe casks. The second
commenter urged the NRC to consider the transportation issues
associated with the possible approval of the Yucca Mountain site as the
NRC makes rules pertaining to the packaging and transportation of
radioactive materials.
Response. The Nuclear Waste Policy Act (NWPA) requires DOE to use
casks certified by NRC for transport to Yucca Mountain, if licensed.
Transport casks are generally not the same as storage or disposal
casks. Issues regarding the licensing of the Yucca Mountain site and
the safety of spent fuel storage or disposal casks are beyond the scope
of the proposed rulemaking. The NRC believes compliance with the
[[Page 3708]]
regulations in part 71 provides for safe transport package designs.
Comment. Three commenters expressed belief that increases in future
shipments have not been adequately considered in the rulemaking. The
first commenter stated that these regulations could have important
implications for the shipment of high-level radioactive waste. The
commenter asked if NRC had considered the financial impact of the
opening of the Yucca Mountain facility before proposing the
regulations.
Response. This comment is primarily focused on future shipments to
Yucca Mountain. The Commission has not received any application
relative to the Yucca Mountain site, and a final decision has not been
made on opening the site itself. Any conclusion made now by the NRC on
future shipments would be purely speculative. Moreover, the commenter
did not specify which aspect of the proposed rule would have a
significant bearing on the Yucca Mountain facility.
The NRC did not identify where major impacts would result, none
were identified that would impact spent fuel shipments. Furthermore,
the existing regulations pertaining to spent fuel have been in effect
for a significant time and have resulted in more than 1300 spent fuel
shipments being conducted without any negative impacts to public health
and safety.
Comment. Two commenters asked how NRC factored the possible
approval of the Yucca Mountain repository into our rulemaking. One
commenter urged NRC to seriously consider the likely increase of
radioactive material transportation in Illinois, Michigan, and
Wisconsin that will occur if the Yucca Mountain repository is approved.
The commenter also provided data from DOE's Yucca Mountain EIS on
projected transportation volume through Illinois.
Response. The comments are acknowledged. However, they are beyond
the scope of this rulemaking. As part of the rulemaking process, NRC
solicited information on the costs and benefits, as well as other
pertinent data, on the proposed changes. NRC appreciates the
commenter's submission of data related to projected transportation
volumes of high-level waste. The NRC believes compliance with the
regulations in part 71 provides for safe transport package designs.
Miscellaneous (including comments to DOT)
Comment. One commenter opposed any use of radioactive materials
entirely.
Response. This comment is beyond the scope of the rulemaking. This
rule deals solely with regulations that govern the transportation of
certain types of radioactive materials and does not address issues
related to the use of radioactive materials in commerce.
Comment. One commenter included a comment letter that was
previously submitted in September 2000, discussing all of the issues in
this rulemaking. The letter was resubmitted because the commenter
believes that the NRC did not respond to the comments previously and
might have lost the original comment letter. The commenter also
included several diagrams and an article entitled ``New Developments in
Accident Resistant Shipping Containers for Radioactive Materials'' by
J. A. Sisler. This article discusses the safety tests required for
shipping containers.
Response. The current proposal stems from NRC's scoping meetings
held in August and September 2000, to solicit public comments on the
part 71 Issues Paper. NRC accepted all verbal and written comments
received at the meetings or later in a letter form and considered these
comments in developing the proposed rule.
Comment. One commenter stated that the public's opinion is that
nuclear power and weapons should remain sequestered from the
environment and the public for as long as they remain hazardous.
Response. The comment is beyond the scope of the rulemaking. This
rule deals solely with regulations that govern the transportation of
certain types of radioactive materials and does not address the use of
nuclear power or weapons.
Comment. One commenter expressed a general distrust of business and
urged NRC to consider recent cases of dishonesty in business when
formulating regulations.
Response. The comment is beyond the scope of this rulemaking.
Comment. One commenter expressed concern that inaccurate reporting,
inspection failures, and faulty equipment all occur in the nuclear
transport industry and may contribute to mishaps in transit.
Response. The NRC is aware of the potential for accidents in
transporting nuclear material and has considered the accident history
of nuclear transportation in estimating the risks of shipping. The NRC
believes that this rule provides adequate protection of the public and
workers in normal transport conditions and in accident conditions.
Comment. One commenter recommended that all radioactive shipments
be tracked, labeled, and publicly reported, including shipments being
made in secret without the consent of the American public.
Response. The NRC acknowledges the commenter's suggestion about
tracking, labeling, and reporting shipments. Current regulations
include requirements for labels and markings for packages that contain
radioactive materials. There are notification requirements for NRC
licensees applicable to shipments of spent nuclear fuel. Current NRC/
DOT requirements for tracking and labeling radioactive shipments
provide adequate protection of public health and safety.
Comment. Several commenters were concerned about the public
reporting requirements pertaining to the shipping of radioactive
materials. Two commenters believe that NRC should publicly report all
radioactive shipments.
Response. The NRC has regulations in 10 CFR part 73 (Physical
Protection of Plants and Materials) that deal with the reporting of
shipments of spent fuel nuclear fuel. This rule deals only with part
71; therefore, these comments are beyond the scope of this rulemaking.
Comment. Several commenters expressed concern with the tracking and
labeling aspects of the proposed rule. Two commenters urged the NRC to
track, label, and publicly report all radioactive shipments. One
commenter believes that the words ``radioactive materials'' should not
be removed from shipping placards because personnel and volunteers
understand the plain English warning better than technical language.
This commenter also suggested that the warnings be written in several
languages. In addition, one commenter stated that the standard symbol,
the black and yellow ``windmill'' for radiation, should adorn all
containers.
Response. Tracking and labeling shipments are part of the
responsibility of the shipper of the licensed material in accordance
with NRC and DOT regulations. Reporting all radioactive shipments would
be an administrative burden with minimal benefit. The NRC's regulations
do require a shipper to provide advance notification of a shipment of
spent nuclear fuel to both the NRC and to the Governor or designee of a
State through which the shipment would be passing. The information is
considered safeguards information and cannot be released to the public
until after a shipment has been completed.
Comment. One commenter expressed support for NRC's acknowledging
DOT's responsibility to ensure the safe shipment of spent nuclear fuel.
Response. The comment is acknowledged. No further response is
required.
[[Page 3709]]
Comment. One commenter requested a clarification of the current
status of DOT's regulations for international shipments regarding
exempt quantities and concentrations.
Response. This request has been forwarded to DOT for consideration.
The commenter should refer to DOT's proposed rule found at 67 FR 21328
dated April 30, 2002.
Comment. One commenter expressed concern with how the proposed
regulations fit into the hierarchy of Federal, State, and local
regulations. The commenter noted that DOT regulations expressly preempt
and supersede State and local regulations.
Response. The State regulations augment the overall national
program for the protection of public health and safety of citizens from
any hazards incident to the transportation of radioactive materials.
States usually adopt the Federal transportation regulations by
reference. The combined efforts of DOT, NRC, and the Agreement States
assure that the applicable Federal regulations are observed with
respect to packaging and transportation of radioactive materials on a
nationwide basis. This is accomplished through DOT, NRC, and State and
local government inspection and enforcement efforts.
Comment. One commenter expressed concern that the DOT definition of
``radioactive material'' is now defined as ``any material having a
specific activity greater than 70 Bq per gram (0.002 micro curie per
gram).'' According to the commenter, the effect of this new definition
would be to enable much more radioactivity to be exempt, thus allowing
more radioactive material to move unregulated in commerce.
Response. This referenced definition change also exists in the NRC
final rule. As described in the background section of this rule, NRC
has analyzed the impact on dose to the public from changing the
definition of ``radioactive material'' from the current definition 70
Bq/g (0.002 [mu]Ci/g) for all radionuclides to radionuclide-specific
exemption values. After considering transport scenarios, NRC concluded
that the new radionuclide-specific definition would result in an
overall reduction in dose to the public when compared to the current
definition.
Comment. One commenter noted that, in Table 1, the listings for Th
(nat) and U (nat) (68 FR 21482) do not refer to footnote b. Because
this is inconsistent with the text of the preamble, the commenter
concluded that it is a typographical error that should be corrected.
Response. The comment is acknowledged and was considered in
developing the final rule.
Comment. One commenter urged the NRC to consider ``the
relationships between and among the exposures associated with these
packaging, container, and transportation regulations and all other
sources of radiation exposures,'' to protect the public from ``adverse
impacts on their health and genetic integrity.''
Response. The comment is acknowledged and has been considered in
developing the final rule.
Comment. Three commenters expressed concern with the role of State
and local governments. One commenter believes that certain States are
already burdened with unusually high concentrations of hazardous and
radioactive materials transport. Another commenter asked about ``the
status of non-Agreement States with respect to compatibility'' and also
wanted further ``explanation of the extent to which a State or
Agreement State may deviate from NRC program elements, definitions, and
standards.'' One commenter stated that county sheriffs and the proper
State officials should be notified in advance of spent nuclear fuel
shipments scheduled to pass through their jurisdictions.
Response. It is NRC practice to seek input and comments from State
and local governments on any NRC proposed rules. For example, in
December 2000, the NRC staff forwarded the part 71 proposed rule to the
Agreement States for comment before sending the rule to the Commission.
Once the rule is published for public comments, NRC considers comments
from all State and local governments, and as such, they play an
important role in the NRC regulatory process. State officials
designated by the Governor are notified in advance of spent nuclear
fuel shipments made by NRC licensees, which pass through their
respective States.
Comment. Several commenters criticized the proposed rule for
acquiescing to the desires of the nuclear and radiopharmaceutical
industries to weaken transport regulations at the expense of increased
public risk.
Response. The proposed rule was developed to maintain compatibility
with the IAEA transportation standards as well as to issue other NRC-
initiated changes. Part 71 has been revised twice in the past 20 years
to stay compatible with IAEA regulations. The risk to the public from
transportation of radioactive materials were considered in the
development of the NRC regulations.
Comment. Two commenters expressed concern over implications for
worker safety. These commenters asked if workers would be protected
from and informed of leaks and whether there is sufficient money to pay
lawsuit damages. They stated that exposure to the transport vehicle
itself should not exceed 10 millirems/year, and all crew compartments
should be heavily shielded to reduce exposure. One commenter then
asserted that workers should be trained to handle radioactive materials
and informed of the risks involved.
Response. NRC radioactive material transportation regulations have
always been issued and enforced to protect the worker and the public
health and safety. When shippers of radioactive material follow these
regulations, they are taking the protective measures called for in NRC
(and DOT) regulations to protect the crew and public. The NRC and DOT
regulations require worker training.
Comment. Several commenters believe that the proposed regulations
increased public risk and weakened protection of public health. One
commenter stated that additional independent oversight of the transport
casks should be conducted regarding quality control to determine
whether they are adequate for cross-country transport. This commenter
also believes that the testing criteria for containers should be more
demanding and require real-world conditions. Another commenter stated
that nuclear shipments should be transported at off-peak hours and also
supported the creation of a ``vehicle-free'' buffer zone ahead and
behind the shipment.
Response. The commenters did not specify how the proposed
rulemaking would increase public risk and weaken protection of public
health. When NRC developed the proposed rule, potential impacts were
carefully considered. NRC does not believe that any part of the
proposal will result in a significant impact on public health and
safety. NRC's quality assurance programs and inspections determine when
additional oversight is warranted. The request for additional and more
demanding testing is not specific; it does not specify how and why
particular testing procedures are inadequate. These procedures have
been carefully verified by NRC to ensure adequate safety.
NRC does not support the commenter's suggestion to transport at
``off-peak'' hours and use a buffer zone as an NRC safety requirement.
There is no safety basis to justify restricting travel only to off-peak
hours, and creating (and enforcing) buffer zones could result in
greater traffic impacts
[[Page 3710]]
and safety issues. Moreover, using these restrictions is not warranted
based on the more than 1300 shipments without incident.
Comment. One commenter urged the NRC to prohibit transport of long-
lived spent nuclear fuel via air or via barge across large waterways.
The commenter also urged NRC to disallow the transport of such fuel in
combination with people, animals, or plants.
Response. Existing NRC and DOT regulations establish requirements
that must be met for safe shipment of spent nuclear fuel by
transportation modes (i.e., truck, barge, or air). The commenter's
second recommendation is noted, but it is beyond the scope of the
proposed rule.
Comment. One commenter stated that dumping radioactive material
into oceans or landfills and incineration of such materials should
never be allowed.
Response. The comment is acknowledged. However, it is beyond the
scope of this rulemaking, and therefore no further response is
required.
Comment. One commenter suggested that NRC, in concert with other
agencies, identify and recover formerly regulated nuclear materials
that have been deregulated or have escaped from control in the past.
Response. This comment is beyond the scope of this rule.
Comment. One commenter requested an explanation of how NRC's
official proposal on the changes in packaging and transporting of
radioactive materials would affect industrial radiology.
Response. Generally, industrial radiography cameras are designed to
meet NRC requirements for Type B transportation packages. Of the 11
IAEA adoption issues and the 8 NRC-initiated issues, none have a
significant impact upon the transport package design requirements for
radiography cameras.
Comment. One commenter expressed support for compatibility among
the Agreement States. This commenter indicated that it is appropriate
for States to have the ability to develop materials necessary for
intrastate shipments. However, for interstate shipments, the commenter
stated that it is necessary for one State to be compatible with the
rest of the country for the country to be compatible with the world.
Response. NRC notes that the commenter's views are consistent with
the Commission's Policy Statement on the Adequacy and Compatibility of
Agreement State Programs, which became effective on September 3, 1997
(62 FR 46517).
Comment. Several commenters urged NRC to improve its scientific
understanding and bases for the proposed rulemaking. Two commenters
suggested that NRC complete the comprehensive assessments of TS-R-1 and
future IAEA standards, the PPS, and real cask tests before proceeding
with this rulemaking.
Response. NRC believes it has an adequate technical basis to make
determinations on the adoption of regulatory changes to address the
issues that are the subject of this rulemaking. The ongoing PPS is
beyond the scope of this rulemaking.
III. Discussion
This section is structured to present and discuss each issue
separately (with cross references as appropriate). Each issue has four
parts: Summary of NRC Final Rule, Affected Sections, Background, and
Analysis of Public Comments on the Proposed Rule.
A. TS-R-1 Compatibility Issues
Issue 1. Changing Part 71 to the International System of Units (SI)
Only
Summary of NRC Final Rule. The NRC has decided to continue using
the dual-unit system (SI units and customary units) in part 71. This
will not conflict with TS-R-1, which uses SI units only, because TS-R-1
does not specifically prohibit the use of a dual-unit system.
We have decided not to change part 71 to use SI units only nor to
require NRC licensees and holders and applicants for a Certificate-of-
Compliance (CoC) to use SI units only because doing so will conflict
with NRC's Metrication Policy (61 FR 31169; June 19, 1996) which allows
a dual-use system. The NRC did not make metrication mandatory because
no corresponding improvement in public health and safety would result;
rather, costs would be incurred without benefit. Moreover, as noted in
the proposed rule (67 FR 21395-21396), the change to SI units only
could result in the potential for adverse impact on the health and
safety of workers and the general public as a result of unintended
exposure in the event of shipping accidents, or medical dose errors,
caused by confusion or erroneous conversion between the currently
prevailing customary units and the new SI units by emergency responders
or medical personnel.
Affected Sections. None (not adopted).
Background. TS-R-1 uses the SI units exclusively. This change is
stated in TS-R-1, Annex II, page 199: ``This edition of the Regulations
for the Safe Transport of Radioactive Material uses the International
System of Units (SI).'' The change to SI units exclusively is evident
throughout TS-R-1. TS-R-1 also requires that activity values entered on
shipping papers and displayed on package labels be expressed in SI
units (paragraphs 543 and 549). Safety Series No. 6 (TS-R-1's
predecessor) used SI units as the primary controlling units, with
subsidiary units in parentheses (Safety Series 6, Appendix II, page
97), and either unit was permissible on labels and shipping papers
(paragraphs 442 and 447).
The NRC Metrication Policy allows a dual-unit system to be used (SI
units with customary units in parentheses). The NRC Metrication Policy
was designed to allow market forces to determine the extent and timing
for the use of the metric system of measurements. The NRC is committed
to work with licensees and applicants and with national, international,
professional, and industry standards-setting bodies (e.g., American
National Standards Institute (ANSI), American Society for Testing and
Materials (ASTM), and American Society of Mechanical Engineers (ASME))
to ensure metric-compatible regulations and regulatory guidance. The
NRC encouraged its licensees and applicants, through its Metrication
Policy, to employ the metric system wherever and whenever its use is
not potentially detrimental to public health and safety, or its use is
economic. The NRC did not make metrication mandatory by rulemaking
because no corresponding improvement in public health and safety would
result, but rather, costs would be incurred without benefit. As a
result, licensees and applicants use both metric and customary units of
measurement.
According to the NRC's Metrication Policy, the following documents
should be published in dual units: new regulations, major amendments to
existing regulations, regulatory guides, NUREG-series documents, policy
statements, information notices, generic letters, bulletins, and all
written communications directed to the public. Documents specific to a
licensee, such as inspection reports and docketed material dealing with
a particular licensee, will be issued in the system of units employed
by the licensee.
Currently, part 71 uses the dual-unit system in accordance with the
NRC Metrication Policy.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
[[Page 3711]]
Comment. Eight commenters stated they appreciated the NRC's
decision to maintain both the international and the familiar system of
becquerels and curies and sieverts and rem.
Response. No response is necessary.
Issue 2. Radionuclide Exemption Values
Summary of NRC Final Rule. The final rule adopts, in Sec.Sec.
71.14, 71.88 and Appendix A, Table A-2, the radionuclide activity
concentration values and consignment activity limits in TS-R-1 for the
exemption from regulatory requirements for the shipment or carriage of
certain radioactive low-level materials. In addition, the final rule
provides an exemption from regulatory requirements for natural material
and ores containing naturally occurring radionuclides that are not
intended to be processed for use of these radionuclides, provided the
activity concentration of the material does not exceed 10 times the
applicable values. These amendments conform part 71 with TS-R-1 and
with DOT's parallel IAEA compatibility rulemaking for CFR 49.
During the development of TS-R-1, it was recognized that there was
no technical justification for the use of a single activity-based
exemption value for all radionuclides for defining a material as
radioactive for transportation purposes (a uniform activity
concentration basis) and that a more rigorous technical approach would
be to base radionuclide exemptions on a uniform dose basis. The values
and limits in TS-R-1, and adopted in Appendix A, Table A-2, establish a
consistent dose-based model for minimizing public exposure. Overall,
NRC's analysis shows that the new system would result in lower actual
doses to the public than the uniform activity concentration basis
system. NRC's regulatory analysis indicated that adopting the
radionuclide-specific exemption values contained in TS-R-1 is
appropriate from a safety, regulatory, and cost perspective. Moreover,
the final rule assures continued consistency between domestic and
international regulations for the basic definition of radioactive
material in transport.
Affected Sections. Sections 71.14, 71.88, and Appendix A.
Background. The DOT previously used an activity concentration
threshold of 70 Bq/g (0.002 [mgr]Ci/g) for defining a material as
radioactive for transportation purposes. DOT regulations applied to all
materials with activity concentrations that exceeded this value.
Materials were exempt from DOT's transportation regulations if the
activity concentration was equal to or below this value. The 70-Bq/g
(0.002-[mgr]Ci/g) activity concentration value was applied collectively
for all radionuclides present in a material.
In Sec. 71.10, the NRC used the same activity concentration
threshold as a means of determining if a radioactive material was
subject to the requirements of part 71. Materials were exempt from the
transportation requirements in part 71 if the activity concentration
was equal to or below this value. Although the materials may be exempt
from any additional transportation requirements under part 71, it is
important to note that the requirements for controlling the possession,
use, and transfer of materials under parts 30, 40, and 70 continue to
apply, as appropriate, to the type, form, and quantity of material.
Basically, the radionuclide exemption values mean that licensed low
radioactivity materials are not required to be handled as hazardous
materials while they are being transported. These exemption values do
not mean that these materials are released from other regulatory
controls, including the controls that apply to the disposal or release
of radioactive material.
During the development of TS-R-1, it was recognized that there was
no technical justification for the use of a single activity-based
exemption 70-Bq/g (0.002-[mgr]Ci/g) value for all radionuclides. It was
concluded that a more rigorous technical approach would be to base
radionuclide exemptions on a uniform dose basis, rather than a uniform
activity concentration basis.
By 1994, the IAEA had developed Safety Series No. 115 (also known
as Basic Safety Standard, or BSS) and a set of principles for
determining when exemption from regulation was appropriate. One
exemption criterion was the effective dose expected to be incurred by a
member of the public from a practice (e.g., medical use of
radiopharmaceuticals in nuclear medicine applications) or a source
within a practice should be unlikely to exceed a value of 10 [mgr]Sv (1
mrem) per year. IAEA researchers developed a set of exposure scenarios
and pathways which could result in exposure to workers and members of
the public. These scenarios and pathways were used to calculate
radionuclide exemption activity concentrations and exemption activities
which would not exceed the recommended dose.
To investigate the exemption issue from a transportation
perspective during the development of TS-R-1, IAEA Member State
researchers calculated the activity concentration and activity for each
radionuclide that would result in a dose of 10 [mgr]Sv (1 mrem) per
year to transport workers under various BSS and transportation-specific
scenarios. Due to differences in radionuclide radiation emissions,
exposure pathways, etc., the resulting radionuclide-specific activity
concentrations varied widely. The appropriate activity concentrations
for some radionuclides were determined to be less than 70 Bq/g (0.002
[mgr]Ci/g), while the activity concentrations for others were much
greater. However, the calculated dose to transport workers that would
result from repetitive transport of each radionuclide at its exempt
activity concentration was the same ((10 [mgr]Sv) (1 mrem)) per year.
For the single activity-based value, the opposite was true (i.e., the
exempt activity concentration was the same for all radionuclides (70
Bq/g) (0.002 [mgr]Ci/g)), but the resulting doses under the same
transportation scenarios varied widely, with annual doses ranging from
much less than 10 [mgr]Sv (1 mrem) per year for some radionuclides to
greater than 10 [mgr]Sv (1 mrem) per year for others. A comparison of
the transportation scenario doses resulting from the single (70 Bq/g
(0.002 [mgr]Ci/g)) activity concentration value and the radionuclide-
specific activity concentration values shows that the radionuclide
activity concentration values reduced the variability in doses that
were likely to result from exempt transport activities.
The basis for the exemption values indicates that materials with
very low hazards can be safely exempted from the transportation
regulations (see draft Advisory Material for the Regulations for the
Safe Transport of Radioactive Material, TS-G-1.1, paragraphs 107.5 and
401.3). If the exemptions did not exist, enormous amounts of material
with only slight radiological risks (materials which are not ordinarily
considered to be radioactive) would be unnecessarily regulated during
transport.
Some of the lower activity concentration values might include
naturally occurring radioactive material (NORM). As an example, ores
may contain NORM. Regarding the transport of NORM, one petroleum
industry representative stated that there are no findings that indicate
the current standard fails to protect the public, and that there is no
benefit in making the threshold more stringent. Further, it would have
a significant impact on their operations. Other similar comments were
received during the public meetings. The overall impact would be that
some material formerly not subject to the radioactive material
transport regulations may need to be transported as radioactive
material and therefore
[[Page 3712]]
meet the corresponding applicable DOT transport requirements.
IAEA recognized that application of the activity concentration
exemption values to natural materials and ores might result in
unnecessary regulation of these shipments and established a further
exemption for certain types of these materials. Paragraph 107(e) of TS-
R-1 further exempts: ``Natural material and ores containing naturally
occurring radionuclides which are not intended to be processed for use
of these radionuclides provided the activity concentration of the
material does not exceed 10 times the values specified in paragraphs
401-406.''
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. One commenter opposed the reuse of radioactive materials
in other products, arguing that this is not based on sound science, but
on commercial judgment. Several commenters expressed general objections
to the proposal to exempt certain amounts of radionuclides from
transportation regulatory control and urged NRC to help prevent more
radioactive waste from being deregulated. Seven commenters stated that
adopting these exemptions would remove a significant barrier to the
purposeful release of radioactive materials from nuclear power and
weapons production into raw materials that can be used to make daily
items (e.g., hip replacements, braces, and toothbrushes) that come into
contact with members of the public.
Another commenter stated that the exempted levels could potentially
provide a back door to recycle and release of radioactive material.
One commenter said that the NRC's stated objectives to facilitate
nuclear transportation and harmonize international standards should not
supersede the NRC's mandate to protect public health and safety. The
commenter also stated that the proposed regulations do not do enough to
protect public health. The commenter opposed the technically
significant motive for adopting exemption values, which is to
facilitate radioactive ``release'' and ``recycling'' or dispersal of
nuclear waste into daily commerce and household items.
One commenter stated that NRC regulations should not treat
radioactive materials like nonradioactive materials. Two other
commenters criticized the proposed regulations for treating radioactive
substances as if they were not radioactively contaminated.
Response. The transportation exemption values do not establish
thresholds for the release of radioactive material to unlicensed
parties or to the environment. They do not relieve the recipient from
regulations that apply to the use or release of that material. Also,
the transportation regulations do not authorize the possession of
licensed material (Sec. 71.0(c)). Thus, no unauthorized party may
receive or possess radioactive material just because the material is
exempted from transportation requirements. Radioactive material
transported under the rule remains subject to separate regulatory
safety requirements regarding possession, use, transfer, and disposal.
Comment. One commenter stated that the use of ``or'' in proposed
Sec. 71.14(a)(2) (67 FR 21448) suggests that there is no consignment
limit if the exempt activity concentration limits are not exceeded. NRC
was asked to replace ``or'' by ``and'' to prevent deliberate dilution
of radioactive material to obtain exemption from transport regulations.
Response. The comment is correct in that the consignment activity
limit does not apply to materials that do not exceed the exempt
activity concentration. Under the final rule, the transport regulations
apply only to radioactive material for which both the activity
concentration for an exempt material and the activity limit for an
exempt consignment are exceeded, so the use of ``or'' in the regulatory
text is correct. When describing materials that are subject to the
regulations, ``and'' is the correct term; when describing materials
that are not subject to the regulations, ``or'' is the correct term.
Because Sec. 71.14 defines materials that are not subject to the
regulations, ``or'' is the correct term.
Material consignments that exceed the exempt activity
concentration, but not the exempt consignment limit, are not regulated
in transport due to the small quantity of material being transported.
Material consignments that exceed the exempt consignment limit, but not
the exempt activity concentration, are not regulated in transport due
to the low radioactivity concentration of the material being
transported. The NRC has no information to support the notion that
radioactive material is diluted to obtain exemption from transport
regulations. The NRC does not propose any regulatory action in this
regard.
Comment. One commenter expressed concern both that the proposed
rule would exempt radionuclide values at various levels and that an
international body created these exemption levels.
Response. The activity concentration exemption values do vary by
radionuclide. However, the doses to the public estimated to occur from
using these values under the transport scenarios are low. The U.S.
participated in assessing the dose impacts from the use of the
exemption values in transport.
Comment. Another commenter asked if it is really necessary for NRC
to adopt the entire IAEA rule to accomplish its goals.
Response. There are a number of specific goals associated with this
rulemaking, one of which is harmonization of NRC regulations with
IAEA's TS-R-1 and DOT regulations. NRC is not adopting TS-R-1 in its
entirety in this rulemaking. However, with respect to revising
exemption values, the NRC staff believes adoption of the exemption
values from TS-R-1 is warranted to maintain consistency between
domestic and international regulations.
Comment. One commenter asked if the NRC told DOT that the American
public has rejected these proposed standards three times in the past
decade, and if DOT has advised IAEA of these objections. The commenter
said that if the IAEA has not been informed of the American public's
resistance to these regulations, NRC needs to inform the agency (DOT
and IAEA) immediately.
Response. The NRC acknowledges this comment, including both the
NRC's and DOT's earlier opposition to the IAEA proposed exemption
values. This rule is the first time that IAEA exemption values are
adopted and are being carried out for maintaining compatibility with
international transportation regulations.
Comment. One commenter asked about the amount of money being spent
regulating levels below the exemption values. The commenter asked if
more money would be spent attempting to verify the proposed exemption
values than would be saved by deregulating them. The commenter wanted
to know if there is any guarantee that money saved by deregulating
levels below the exemption values will be spent on improving public
safety in other areas.
Response. The NRC believes the benefits of the exemption values
will outweigh the costs. NRC analyses lead the NRC staff to believe
that the increase in regulatory efficiency between regulatory agencies
and the facilitation of international shipments make the exemption
values advantageous overall. Further, as part of this rulemaking, NRC
specifically requested information on the costs and benefits of the
proposed
[[Page 3713]]
changes. To the extent this information was received, it was considered
in the development of NRC's position. Lastly, it is beyond the scope of
this rulemaking to guarantee that any money saved will be spent on
improving public safety elsewhere.
Comment. One commenter suggested that the NRC could not determine
costs or savings from the proposed radionuclide exemption values, in
part because the NRC does not know what amounts will be exempted. The
commenter also explained that although NRC could attempt to do
projections based on the current industry, NRC could not know what
amounts would be exempted in the future.
Response. The NRC fully realizes the difficulties associated with
predicting the impacts of implementing the exemption values. The NRC
also agrees that it is difficult to predict what amounts would be
exempted under this final rule, just as it is difficult to assess the
amount of material exempted under the current regulations. However, a
large majority of commercial radioactive materials are shipped in
highly purified forms that far exceed the exemption levels. NRC expects
this would continue to be the case under the exemption values. For all
of these reasons, the NRC staff explicitly asked for data on the
anticipated impacts of the proposed rule. The NRC staff used these data
to aid decisionmaking. In general, the NRC expects that the increase in
regulatory efficiency among regulatory agencies and the facilitation of
international shipments will outweigh any increased costs of shipments
resulting from the changes in the exemption values.
Comment. One commenter requested that a cost-benefit analysis be
done to account for both the proposed rule's complexity and its
enforcement difficulties. The commenter notes that no cost-benefit
analysis had been done on this issue and that the NRC chose it
subjectively.
Response. The draft regulatory analysis considered the benefits and
costs associated with adoption of the radionuclide exemption values
from TS-R-1 using the best available information. In addition, the NRC
decided to adopt the dose-based exemption values because the NRC
believes these values would actually reduce exposure in transport by
establishing a consistent dose-based model for minimizing public
exposure. This benefit is in addition to the expected harmonization and
financial benefits. NRC disagrees with the commenter's assertion that
the exemption values were chosen subjectively. NRC used the best
available information and gathered as much information as possible from
the public, the regulated community, and outside experts. The purpose
of this rulemaking, with its public meetings and public comment period,
is to ensure that all affected parties have adequate opportunity to
register their comments and provide supporting materials to justify
their position (and thus better influence the development of NRC's
final position).
Comment. Another commenter stated that the technical benefits of
the proposed rule do not outweigh the associated costs and efforts.
Response. Because NRC staff are unclear what the commenter means by
``technical benefits,'' NRC cannot specifically respond to this
comment. Overall, NRC believes that the benefits that will accrue with
adoption of exemption values from TS-R-1 (e.g., harmonization with
other regulatory agencies and facilitation of international shipments)
will outweigh the costs (e.g., administrative changes, determining
whether packages are exempt, and regulating previously exempt
packages).
Comment. One commenter opposed the proposed exemption values
because they were not derived directly and did not directly involve
public input or a cost-benefit analysis.
Response. A preliminary RA that evaluated possible costs and
benefits was conducted as part of the development of this rule.
Additional information obtained during the rulemaking process was
considered in determining NRC's final position on adopting the TS-R-1
exemption values.
Comment. One commenter stated that, although the revised limits are
not expected to create any significant burden to the Naval Nuclear
Propulsion Program, use of the new limits could create a cumbersome
work practice for some shipments. All low-level shipments that are
currently exempt will require a detailed evaluation to ensure that
activity concentrations for each radionuclide are acceptable. For
example, thoriated tungsten weld rods and soil from site excavations
would require individual isotope analyses at an additional expense. The
commenter stated that the current 70-Bq/g activity concentration limit
for domestic shipments should be retained.
Response. The comment is consistent with others from the shipping
community (i.e., the radionuclide activity concentration and activity
exemption values are likely to be more cumbersome to work with but do
not pose an excessive burden). The NRC agrees that expenses may be
involved in achieving compliance with these values but notes that
expenses are also associated with determining compliance with the
current 70-Bq/g (0.002-[mu]Ci/g) value. Most shipments of radioactive
materials involve materials that have been processed to concentrate
radioactivity. These materials are known by shippers to greatly exceed
the exemption values, and are packaged and transported in accordance
with the radioactive material transporation safety regulations. Thus
the exemption values are irrelevant to the majority of radioactive
material shipments, such as most shipments in the Naval Nuclear
Propulsion Program and most shipments in industry as well. The
exemption values are relevant to shipments of low activity
concentration. For these shipments, shippers will need to establish
either by process knowledge or analysis whether a shipment exceeds the
exemption values and is regulated in transport as a radioactive
hazardous material, or does not exceed the exemption values and may be
shipped as non-hazardous material (regular freight). Most shipments
that minimally exceed the exemption values are likely to be transported
as limited quantities, which would impose a minimal regulatory burden
on shippers. Overall, NRC believes that the benefits that will accrue
with adoption of exemption values from TS-R-1 (e.g., harmonization with
other regulatory agencies and facilitation of international shipments)
will outweigh the costs (e.g., administrative changes, determining
whether packages are exempt, and regulating previously exempt
packages).
Comment. Two commenters stated that the proposed rule would
increase industry's regulatory burden. In particular, the NRC was told
that the proposed rule is too conservative and would unnecessarily
burden industry, particularly in the case of bulk shipments of
contaminated materials. The proposed exemption thresholds would
increase worker exposure to radioactive materials.
Response. NRC acknowledges that the exemption values impose some
new complexity and economic burden on industry. However, NRC believes
that the increase in costs will be minimal. The NRC believes that the
exemption values represent a good balance between economic and public
health interests. From an economic perspective, the increased costs of
the exemption values are outweighed by the benefits of conforming to
other regulatory agencies and facilitating international shipments. NRC
staff recognizes that preshipment requirements under the exemption
values may increase some low-level exposures, but the NRC still expects
that
[[Page 3714]]
the shift to a consistent set of dose-based exemption values will
minimize the potential dose to transport workers.
Comment. One commenter stated that, although cost reduction was one
incentive for the rule, the proposed rule as written was so complicated
that enforcement costs would rise.
Response. NRC acknowledges the comment and, as previously
discussed, NRC believes that any additional enforcement or other costs
will be minimal due to the anticipated benefits of having only one set
of shipping requirements, as well as the cost savings that would result
from moving some materials outside the scope of transport regulation.
Comment. Two commenters stated that the proposed regulations failed
to properly implement IAEA exemption values regarding naturally
occurring radioactive material, which would dramatically expand the
universe of regulated materials and increase the burden on the
regulated community. One commenter stated that other agencies, such as
the Occupational Safety and Health Administration (OSHA), afford
adequate protection from naturally occurring radioactive materials for
workers and the public, and therefore NRC should not enter this
regulatory arena. This commenter also stated that the proposed
exemption values would also lead to a conflict with the Resources
Conservation and Recovery Act (RCRA), which stipulates that waste
disposal sites may not accept radioactive materials of more than 70 Bq/
g.
Another commenter specifically noted that the NRC has not
implemented the exemption provisions for phosphate ore and fertilizer;
zirconium ores; titanium minerals; tungsten ores and concentrates;
vanadium ores; yttrium and rare earths; bauxite and alumina; coal and
coal fly ash. The commenter urged NRC to consider the activity
concentration of the parent nuclide in determining exemption values.
Response. Section 71.14(a)(1) provides the same exemption for low
level materials (e.g., natural materials and ores) that IAEA provides
in TS-R-1 paragraph 107(e). The exemption multiple for activity
concentration (10 times the values listed in 10 CFR part 71, Table A-2)
applies to natural material and ores containing naturally occurring
radionuclides which are not intended to be processed for use of these
radionuclides. If the materials identified in the comment meet the
definition and are not being processed to use radionuclides, the
exemption multiple would apply. Thus, the burden indicated by the
commenter would not occur.
The activity concentration for exempt material applies to each
radionuclide listed in Table A-2. For radionuclides in secular
equilibrium with progeny, the listed activity concentration applies to
the listed radionuclide (as parent), and was determined considering the
contribution from progeny. Table A-2, as published on April 30, 2002;
67 FR 21472, contains several typographical errors, including the
omission of the reference to footnote (b) for the U (nat) and Th (nat)
radionuclides. These errors have been corrected in this final rule.
Comment. One commenter was concerned that the exemption values in
TS-R-1 could result in the unnecessary regulation of certain materials
that are currently exempt from NRC regulation under Sec. 40.13. The
commenter urged NRC to allow unimportant quantities to remain exempt.
The commenter was concerned that the public and operators of RCRA
disposal facilities may question the safety of materials that were
previously exempt but are not exempt under the new regulations. The
commenter pointed out that the actual risk would not change because
RCRA will not change.
Response. Materials that are exempt (i.e., not licensed) under Sec.
40.13 are not subject to part 71 under the current or final
transportation regulations. Nothing in this final rule affects the
exemption status of materials subject to Part 40.
RCRA sites can continue to use the 70-Bq/g (0.002-[mu]Ci/g) value
as a material acceptance criterion at their option. The final rule
establishes new exemption values for radioactive materials in transport
that differ from 70 Bq/g (0.002 [mu]Ci/g) that might be used (for
nontransport purposes) at RCRA sites. However, the final rule does not
preclude the shipment of materials to RCRA sites in a manner that would
satisfy both transportation and site safety regulations.
Comment. Ten commenters expressed opposition to the exemption
values. One commenter argued that the proposed guidelines should allow
no exemptions. Two commenters stated that the proposed exemptions would
negatively impact public health. Two commenters argued that the
redefinition would pose a threat to public health. Two commenters
opposed weakening regulations that would reduce the public safety and
health through new definitions or accepted concentration values. Two
commenters emphasized that there is no justification for increasing
allowable concentrations because there are ramifications beyond
transportation, and that using a dose-based system is less measurable,
enforceable, and justifiable.
Some commenters added that if NRC needed to adopt risk-based
standards, NRC should adopt the standards that would reduce the
allowable exemptions. One commenter criticized the proposed rule for
increasing the allowable contamination in materials. One commenter
disagreed with the current 70 bequerels-per-gram exemption level and
urged NRC to change only the exemption levels to make them more
protective for isotopes whose exempt concentrations go down.
One commenter also stated that NRC had not actively participated in
determining the proposed exemption values.
Response. NRC disagrees with the comment that no exemptions should
be allowed. Because almost all materials contain at least trace
quantities of radioactivity, if there were no exemptions, essentially
all materials transported in commerce would be treated as radioactive
materials. This would entail considerable expense and impact on
commerce without commensurate benefit to public health and safety.
The NRC disagrees that the proposed exemptions would negatively
impact public health. The NRC's analysis of the radionuclide-specific
exemption values indicates the overall dose impact of their adoption
would be low (much less than background levels), and lower than that of
the single-value exemption currently in place. Please see the
Background section under this issue for further details.
The NRC acknowledges the comment that there is no justification for
increasing allowable concentrations. However, the NRC believes the
benefits of the exemption values will outweigh the costs. NRC analyses
lead the NRC staff to believe that the increase in regulatory
efficiency between regulatory agencies and the facilitation of
international shipments make the exemption values advantageous overall.
The NRC finds the low uniform-dose approach that was used in the
development of the exemption values to be acceptable.
Although additional measurements may be necessary under the new
requirements, the industry has not indicated that these requirements
pose an excessive burden. The NRC does not believe the radionuclide
exemption values would be less enforceable than the current single
exemption value.
Lastly, as a working participating member of the IAEA, both NRC and
DOT staff participated in the development of the exemption values.
[[Page 3715]]
Comment. One commenter requested information on calculations for
dose impacts to members of the public, particularly regarding recycling
and the possibility of exempting materials that pose a radiation hazard
to the public.
Response. An assessment of public dose that might result from
adopting the exempt activity concentrations and exempt activities per
consignment under transportation scenarios may be found at the
following reference: A. Carey et al. The Application of Exemption
Values to the Transport of Radioactive Materials. CEC Contract CT/PST6/
1540/1123 (September 1995). The NRC has performed no assessment
regarding recycling because that is beyond the scope of this
rulemaking.
Comment. A commenter requested the risk and biokinetic data
supporting the proposed exemption values. The commenter also wanted to
know more about who determines what data NRC uses, including the
physiological data used to justify the change in dose models.
Response. The basic radiological protection data used in the
development of the exempt activity concentrations and exempt activities
per consignment may be found at the following reference: International
Basic Safety Standards for Protection Against Ionizing Radiation and
for the Safety of Radiation Sources, Safety Series No. 115, IAEA 1996.
Comment. Two commenters stated that it is unclear how or why the
risk decreases for 222 of the 382 listed radioisotopes, when the
allowable concentrations for those radioisotopes increase to above 70
becquerels. The commenters asked how the ``risk or dose goes down''
while some exempt quantities could lead to more than the ``worker doses
to members of the public from unregulated amounts of exempt quantities
of radioisotopes.''
Response. Under the previous system, radioactive materials
exceeding the 70-Bq/g (0.002-[mu]Ci/g) activity concentration were
regulated in transport. Although the 70-Bq/g (0.002-[mu]Ci/g) value
applied to all radionuclides, different radionuclides resulted in
different doses to the public when transported at that activity
concentration (as calculated using the transport scenarios). The
transport scenario doses for many radionuclides when transported at 70
Bq/g (0.002 [mu]Ci/g) are less than the reference dose of 0.01 mSv/y (1
mrem/y). However, for other radionuclides, the transport scenario doses
at 70 Bq/g (0.002 [mu]Ci/g) are greater than the reference dose of 0.01
mSv/y (1 mrem/y). Under the radionuclide-specific approach, the
calculated doses are more representative, and the average dose
(considering all radionuclides) is lower than under the 70-Bq/g (0.002-
[mu]Ci/g) approach. Overall, the NRC's analysis shows that the new
system would result in lower actual doses to the public than the
current system.
Comment. Another commenter urged NRC to either make exemption
values more stringent or not adopt any new values at all.
Response. The comment provides no justification to make the
exemption values more stringent. The IAEA and other Member States have
adopted the new system. Failure to adopt the new system would put the
U.S. at a competitive disadvantage in international commerce without
commensurate benefit to public health and safety and would allow the
continued shipment of exempt materials that are calculated to produce
higher doses to workers and members of the public.
Comment. One commenter asked that NRC provide a separate activity
concentration threshold, and suggested 2,000 picocuries per gram, for
samples collected for laboratory analysis in situations where relevant
data is unavailable. The commenter believes that the current proposed
threshold of 2.7 picocuries per gram is too restrictive for samples
acquired for laboratory analysis.
Response. Although data is apparently unavailable for the samples
the commenter refers to, it appears the samples are minimally
radioactive and, therefore, could be shipped as a limited quantity, one
of the least burdensome shipments. As we received no other comment on
this issue, the commenter's concern does not appear to be widespread.
The NRC has concluded that the information and justification provided
do not warrant the introduction of a provision in part 71 that would
not be compatible with TS-R-1.
Comment. One commenter asked that NRC provide for expeditious
transportation of discrete solid sources encountered in public areas.
The commenter noted that part 71 currently permits a source of up to
2.7 millicuries to be transported as a limited quantity, even if no
relevant data about the source is available. The commenter then asked
NRC to retain this arrangement for sources encountered in public areas
because it has been a useful provision.
Response. The quantities involved (2.7 mCi) would not normally
require NRC-certified packaging, thus the current part 71 rulemaking
would have little bearing upon them. The NRC understands that DOT has a
system of exemptions in place, which has been coordinated with State
regulators, to facilitate the safe and timely transport of sources
discovered in the public domain.
Comment. One commenter asked about the proposed mechanism for
approving nondefault exemption values. Some commenters requested
further information on how default exemption values could be calculated
from the A1 and A2 values.
Response. The scenarios used to develop the exemption values were
selected to model exposures that could result from relatively close
distances and long duration exposure times to exempt materials. The
scenarios used in the Q-system were selected to model exposures that
could result from shorter-term exposure to the contents of a damaged
Type A package following an accident. Because of the differences in the
exposure scenarios and the resulting differences in the equations used
to calculate the values, the Q-system cannot be used to calculate
activity limits for exempt consignments or exempt activity
concentrations.
Comment. One commenter stated that the landfill disposal of NORM is
outside NRC jurisdiction when technologically advanced NORM is involved
with RCRA-regulated hazardous constituents. The commenter explained
that numerous RCRA landfills around the country have adopted the EPA-
and State-approved programs for the disposal of NORM. The commenter
wondered how the proposed changes in radionuclide exemption values
would affect the regulations governing these landfills.
Response. Part 71 has no direct effect on the regulations governing
the licensing or operation of landfills. The comment is beyond the
scope of this rulemaking.
Comment. Two commenters opposed the regulation of NORM ores and
natural materials, including materials derived from those substances,
because it does not include appropriate exemptions and will result in
unjustified increased costs and transportation burdens and liabilities.
Response. This rule does not extend NRC's scope of regulation of
radioactive material. If a material, such as NORM, was not previously
subject to NRC regulation, it would not be subject to regulation under
this final rule. For regulatory consistency, both DOT and NRC publish
the radionuclide exemption tables, including the 10 times exemptions
for natural materials and ores containing NORM. Also, part 71 only
applies to material licensed by
[[Page 3716]]
the NRC, and NRC does not regulate NORM.
Comment. One commenter suggested that NRC reevaluate the proposed
factor for the allowance of NORM. This commenter recommended that NRC
consider using a factor of 100 rather than 10, because many materials
are not hazardous and do not require more stringent shipping
regulations.
Response. The comment does not provide compelling data to support
the requested change. Furthermore, the requested change would result in
the U.S. being noncompatible with international transportation
regulations. Therefore, no change is made.
Comment. One commenter stated that this rule has taken the focus
off of more important issues in place of issues that are of less
concern, such as the regulation of NORM. The commenter stated that
lowering exemption values could distract attention from materials that
would otherwise be of concern to law enforcement, particularly
regarding transportation across U.S. borders.
Response. The exemption values are considered by shippers when
preparing radioactive materials for transport. The NRC staff does not
believe these rule changes will affect law enforcement activities.
Comment. One commenter was concerned that ``uranium and thorium
levels in phosphate, gypsum, and coal cannot be considered safe simply
because they are naturally occurring. The commenter added that from a
public health point of view, there is no need to determine whether
alpha emissions above the 70-Bq/g (0.002-[mu]Ci/g) threshold are
naturally occurring or man-made, their effect on somatic cells and germ
cells is the same.'' The commenter was concerned that NRC has not
proposed sufficient regulations regarding the ``shipment of ores and
fossil fuels with regard to radioactive levels of naturally occurring
radionuclides.'' The commenter requested that NRC provide an analysis
of the ``regulatory burden of radionuclide HMR on the fertilizer,
construction, and fossil-fuel energy industries.''
Response. NRC's transportation regulations apply to NRC licensees
that transport licensed material and require that licensees comply with
U.S. DOT Hazardous Materials Regulations. The DOT regulations
previously included the 70-Bq/g (0.002-[mu]Ci/g) value in the
definition of radioactive material, and materials determined to be less
than that activity concentration did not satisfy DOT's definition of a
radioactive material and were not regulated as hazardous material in
transport. The DOT definition applied regardless of whether the
material was naturally occurring or not.
With regard to burden, this rule adopts a change in the
transportation exemption for radioactive materials from a single value
to radionuclide-specific values. In its proposed rule, NRC requested
specific information on the impact of that change. The information
provided to NRC is presented in the regulatory analysis accompanying
this rule.
Comment. One commenter suggested that NRC not use the wording in
Sec. 71.14(a)(1), ``Natural materials * * * that are not intended to be
processed for the use of these radionuclides * * *,'' because it
unreasonably requires the shipper to know the intended use of the
material. The commenter emphasized that NRC should base transport
regulations solely on the radiological properties of the material
shipped.
Response. This provision applies to a subset of the industry that
processes an ore that contains radioactive material, not for the
radioactive material, but for some other element, mineral, or material.
For example, this provision would apply to the processing of an ore
during which thorium or uranium was produced incidentally in a waste
stream, but would not apply to the processing of an ore to extract
thorium or uranium for use or sale. NRC staff believes the industry can
reasonably be expected to determine the intent for processing the ore
when that ore is shipped to a consignee.
Comment. One commenter indicated that, should the exemption values
be adopted in a way that departs from IAEA, newly regulated entities
could face high monetary penalties for failure to comply with the
regulations due to DOT's enforcement penalty policies. The commenter
noted that DOT regulations preempt and supersede State and local
regulations, so these regulations make it more difficult for people to
protect themselves from the dangers of exposure to radiation.
Response. The NRC staff believes the rule adopts the exemption
values in a manner that is compatible with the IAEA regulations and
with a parallel DOT final rule.
Comment. One commenter asked the NRC if States whose regulations
are more protective than the proposed rule would have to abandon those
regulations if NRC adopted the proposed rule.
Response. States do not have regulations that are more protective
than those in this rulemaking for the transportation of radioactive
materials. State regulations in this area are essentially identical to
those of the Federal government to eliminate any conflicts,
duplications, gaps, or other conditions that would jeopardize an
orderly pattern in the regulation of radioactive materials on a
nationwide basis.
Comment. One commenter stated that there is no way to know how much
is being exempted in terms of curies or becquerels because there is no
limit on the number of negligible doses from exemptions.
Response. The dose criteria used in determining the activity
concentrations for exempt materials ensure that the doses (from either
single or multiple sources) do not reach unacceptable levels, and will
therefore be far below public dose limits. Quantifying exempted
materials (i.e., those materials that are not regulated as radioactive
material in transport) would impose a significant burden without
commensurate benefit to public health and safety.
Comment. One commenter expressed concern that, for some members of
the public, exposure could be over 100 millirem per year. The commenter
understood from the proposed rule that the dose-based exemption values
are designed to deal with transport worker exposures in the range of 25
to 50 millirem per year. The commenter requested information about how
the expected annual dose to transport workers changes under the
proposed rule, particularly if it increases or decreases.
Response. The NRC staff notes that exposures to members of the
public are more likely to be over 1 mSv (100 mrem) per year under the
current single exemption value than under the radionuclide-specific
system. However, these are dose estimates; the transport scenarios used
to estimate these doses overstate actual doses by overstating exposure
periods in a year (50-400 hrs/yr) and exposure distances [less than
1.52 m (5 ft)] to radioactive materials in transport.
For those radionuclides with a relatively low estimated dose for
transport at 70 Bq/g (0.002 [mu]Ci/g) under the transport scenarios,
the estimated dose will increase under the dose-based exemptions; for
those radionuclides with a relatively high estimated dose for transport
at 70 Bq/g (0.002 [mu]Ci/g) under the transport scenarios, the
estimated dose will decrease under the dose-based exemptions. Even in
those instances where the estimated dose increases under the final
rule, the dose remains low and the average dose (considering all
radionuclides) is lower under the radionuclide-specific system.
[[Page 3717]]
Comment. One commenter questioned the composition of a list of 20
representative nuclides used to estimate the average annual dose per
radionuclide. The commenter asserted that, among the 20 representative
nuclides, a minority of nuclides whose doses decrease in the proposed
regulations were overrepresented. The commenter stated that most of the
dose concentrations increase, some of them dramatically.
Response. The 20 radionuclides referred to were chosen to be
representative of the radiation types (alpha, betas of various
energies, and gamma) most commonly encountered in transport and were
used to provide a representative measure of the proposed rule's likely
impact.
Although the radionuclide activity concentration values more often
exceed 70 Bq/g (0.002 [mu]Ci/g) than fall below it, the distribution of
all the new exemption values centers just above 70 Bq/g (0.002 [mu]Ci/
g).
It is recognized that the exempt activity concentration for some
radionuclides (those radionuclides with very low doses under the
transport scenarios when transported at 70 Bq/g (0.002 [mu]Ci/g)) will
increase under a dose-based exemption system. However, the measure of
impact from the change in exemption values is the estimated dose, and
that remains low, even for radionuclides where the exempt activity
concentration increases above 70 Bq/g (0.002 [mu]Ci/g). The radiation
protection benefit from the radionuclide-specific approach is that the
highest potential doses are reduced as well as the average dose from
all radionuclides.
Comment. One commenter noted that there is no precedent for exempt
quantities in NRC regulations and that this will create a new category.
The commenter questioned the logic of creating such a category.
Response. The DOT transportation safety regulations for radioactive
materials have always had a de facto ``exemption value'' built into the
definition of ``radioactive material.'' NRC regulations either
replicate or include references to DOT regulations. Any material with
an activity below the 70-Bq/g (0.002-[mu]Ci/g) threshold was not
defined as radioactive for the purposes of the regulations and
therefore was not subject to the regulations (i.e., exempt). Without
the exempt activity for consignments value, any quantity of material
that exceeded the exempt activity concentration, no matter how small,
would be regulated in transport as radioactive material. The exempt
consignment value is included to prevent the regulation of trivial
quantities of material as hazardous material in transport.
Comment. One commenter stated that the threat of terrorism should
be taken into account when exempting radionuclides from transport
regulations and changing container regulations.
Response. The nature of exempt materials is that they are either of
very low activity concentration or very low total activity. In both
cases, these materials present little hazard and would not be
attractive as targets for terrorist activities.
Comment. One commenter expressed concern that the revised exempt
concentrations in Table A-2 are a significant change in the
requirements for the transportation of unimportant quantities of source
materials.
Response. Although the comment expresses concern that the exempt
activity concentration values represent a significant change in the
requirements for unimportant source material, it does not provide data
or justification for this statement. NRC acknowledges that the
internationally developed transportation exemption values do not align
precisely with preexisting, domestic requirements in NRC regulations in
10 CFR part 30 or part 40 that were developed for other licensing
purposes. However, the current 70-Bq/g (0.002-[mu]Ci/g) exemption value
does not align precisely with part 30 or part 40 requirements either.
In most cases, the differences in the regulatory requirements do not
appear to be that significant, and the industry has not provided data
that demonstrate that the impact from the change for actual shipments
would be significant. NRC has no basis to change its conclusion in the
final RA that the overall benefits of achieving compatibility by
adopting the exemption values outweigh the associated costs, or its
belief that permitting natural materials and ores to be shipped at 10
times the Table A-2 values minimizes the impacts.
Comment. Five commenters supported NRC's efforts in the proposed
rule. One of these commenters supported lower concentrations for the
radioactive isotopes because the proposed rulemaking increases public
risk. Another stated that it was important to ensure consistency
between international and domestic regulations and that while
individual radionuclide levels may be raised or lowered by the proposed
rule, overall the estimated dose would be significantly lower. Another
commenter agreed with NRC's proposal to adopt the radionuclide
exemption values in TS-R-1, particularly the inclusion of exempt
consignment quantities in the regulations. Another commenter expressed
general support for ensuring consistency between domestic and
international regulations.
Response. NRC acknowledges the comments on revising radionuclide
exemption values. NRC staff agrees with the commenters who stated that
consistency between international and domestic regulations is a high
priority, and that the exemption values overall will result in lower
public exposure. However, while promulgating lower exemption levels
could reduce the already low public health risks, NRC believes that the
exemption values offer the best balance between economic and public
health concerns.
Comment. One commenter stated that the proposed exemption values
were too complex because it is too complicated to maintain more than
half of all exemption values at 70 Bq/g (0.002 [mu]Ci/g) and to reduce
those that are more protective.
One commenter said that there are no comparable exemptions in
existing regulations.
Response. The NRC does not believe that the proposal to maintain
more than half of the activity concentration exemption values at 70 Bq/
g (0.002 [mu]Ci/g), while reducing the activity concentration exemption
values for the remaining radionuclides, is warranted because the
resulting exemption system would be inconsistent, have no defined dose
basis, and would be incompatible with that of the IAEA and other Member
States.
The final rule introduces exemptions from the application of the
hazardous materials transportation regulations for materials in
transit. However, the definition of ``radioactive materials'' in the
transportation regulations has, for decades, contained a minimum
activity concentration value (i.e., any material with an activity
concentration less than 70 Bq/g (0.002 [mu]Ci/g)); effectively, the
definition has contained an exemption value. The final rule changes the
structure of the exemption from a single activity concentration value
applicable to all radionuclides to individual activity concentration
and consignment activity values that are specified for each
radionuclide.
Comment. Several commenters expressed concern about the health
effects of these regulations. One commenter opposed reliance on the
ICRP arguing that ICRP does not take into consideration important
information on the health impacts of radiation such as synergism with
other contaminants in the environment and the bystander effect, in
which cells that are near cells that are hit, but are not
[[Page 3718]]
themselves hit by ionizing radiation, exhibit effects of the exposure.
One commenter stated that the NRC did not consider the new evidence
that low doses of radiation are more harmful per unit dose than was
previously known. This commenter further noted that there are
synergistic effects and other types of uncertainties in radiation
health effects. Three commenters opposed the radionuclide exemption
value tables citing the use of outdated data, lack of data, and/or the
lack of calculations for more than 350 radionuclides. One commenter
stated that NRC radiation standards are outdated and should be subject
to rigorous review, including independent outside experts. One
commenter stated that ICRP does not represent the full spectrum of
scientific opinion on radiation and health and does not take into
account certain health impacts of radiation. One commenter noted that
ICRP and IAEA risk models only look at fatal cancers and ignore
nonfatal cancers, years of lost life, and the bystander effect. The
commenter also asserted that these agencies' reports do not accurately
reflect risk and that low levels of radiation are more damaging than
the models are predicting.
Response. The Board of Governors of the International Atomic Energy
Agency stated in 1960, that ``The Agency's basic safety standards * * *
will be based, to the extent possible, on the recommendations of the
International Commission on Radiological Protection (ICRP).'' The ICRP
is a nongovernmental scientific organization founded in 1928 to
establish basic principles and recommendations for radiation
protection; the most recent recommendations of the ICRP were issued in
1991 (International Commission on Radiological Protection, 1990
Recommendations of the International Commission on Radiological
Protection, Publication No. 60, Pergamon Press, Oxford and New York
(1991)). The IAEA Basic Safety Standards (from which the exemption
values are taken) were developed with full IAEA Member State
participation (including the U.S.) and have taken the ICRP
recommendations into account. NRC rejects the comment that the data
used to develop the exemption values are outdated or inadequate. In
general, NRC believes ICRP reports provide a widely held consensus view
by international scientific authorities on radiation dose responses and
accepts their principal conclusions. Furthermore, the NRC notes that
fundamental research into radiation dose effects is beyond the scope of
this rulemaking. For that information, NRC relies on national and
international scientific authorities.
Comment. The NRC was criticized by commenters for not having
developed and pursued actual transport exposure scenarios for every
radionuclide to justify the exemptions. One commenter also noted that
although NRC has not carried out calculations for transportation
scenarios for over 350 of the listed radionuclides, individual exempt
concentration and quantity values have been assigned to each
radionuclide. The commenter further concluded that NRC has technical
data to support the conclusion that these exemption values will pose no
risk to the public. Another commenter stated that it was unclear why
NRC performed calculations for only 20 of the 350 isotopes. The
commenter noted that because NRC only modeled 20 of the radionuclides,
NRC has not collected complete data for the other radionuclides;
otherwise, they would have been also modeled. The commenter further
stated that NRC should either lower the exemption values or withdraw
the values and perform further studies.
Response. NRC selected a subset of 20 radionuclides believed to be
representative of the most commonly transported radionuclides. Exempt
activity concentration and consignment activity values were calculated
for all the radionuclides listed in Table A-2, not just the 20 selected
to be used in NRC's impact analysis. NRC used the 20 radionuclides to
illustrate that the impact from activity concentration exemption values
for materials commonly transported in significant quantities is less
than that from the current single exemption value.
Comment. One commenter expressed concern that NRC had arbitrarily
determined the radionuclide values.
Response. The A1 and A2 values in Table A-1
and the exempt activity concentration values and exempt activity values
in Table A-2 are not arbitrary values. The derivation of these values
is dose based and provided in the references in TS-R-1.
Comment. One commenter expressed opposition to the exemption values
because they raised the allowable exempt concentrations and allowed for
exempt quantities, which are currently not permitted.
Response. The current definition of radioactive material is
specified only in terms of a minimum activity concentration.
Conceivably, this leads to the regulation of any quantity of material
that exceeds that activity concentration, even minute quantities, as a
radioactive material in transport. To address this issue, an activity
limit for exempt consignments has been introduced that specifies a
minimum activity that must be exceeded for a material to be regulated
as a radioactive material in transport.
As with the exempt activity concentration values, the exempt
activity values in Table A-2 were taken from the BSS exemption values.
The doses associated with the use of these exempt activity values were
estimated using the same scenarios used for assessing the impact of the
exempt activity concentration values. The results are that doses are
low, and that for 19 of the 20 representative radionuclides examined,
the dose from the radionuclide exempt activity value is less than that
from the exempt activity concentration value.
Comment. One commenter asked if there is any possibility that NRC
could simply decline to adopt the sections of the proposed rules that
relate to radionuclide exemption values.
Response. NRC's and DOT's approach in this compatibility rulemaking
is to adopt the provisions of IAEA's TS-R-1 as proposed unless adoption
would pose a significant detriment to radioactive material transport
commerce, or is unjustified. The NRC has determined that the exemption
change is justified based on its regulatory analysis and public
comments.
Comment. One commenter stated that NRC should ensure that no member
of the public would receive a dose above 1mrem/year from any practice
or source, and should clarify what is meant by ``practice'' and
``source.'' One commenter stated that the current HMR standard of 70
Bq/g (0.002 Ci/g) should be maintained as the minimum standard for the
protection of public health and transport worker safety. The commenter
opposed the replacement of this standard with the radionuclide-specific
values per the IAEA's TS-R-1 for the following reasons:
(1) There is no radiation risk level which is sufficiently low as
to be of no regulatory concern;
(2) There are no collective radiological impacts which are
sufficiently low as to be of no regulatory concern; and
(3) No one will be able to determine if proposed exempt sources are
safe.
One commenter noted that the current and proposed regulations have
50 and 23 millirem being average doses, respectively. To adequately
protect public health, the average dose should be no more than one
millirem. One commenter stated the assumptions and
[[Page 3719]]
scenarios that NRC and DOT used to justify the adoption of these
exemption values fail to prove that these exemptions will have either
no or an insignificant effect.
One commenter stated that the proposed exemption values are based
on unrealistic models. The commenter said that the exempt levels do not
appear to reflect the material's longevity in the environment and
hazard to living creatures. One commenter stated that the standards
should be based on the most vulnerable members of the population, and
NRC should adopt stricter values. Two commenters argued that, using the
existing dose models, some of the exempt quantities could lead to high
public doses from unregulated amounts of exempt quantities of
radioisotopes. Another commenter opposed reliance on computer model
scenarios that may not be realistic to project doses, citing that this
lack of realism to justify certain exposure scenarios is inadequate.
One commenter stated that it is unclear in the proposed regulations
what the exact dose impact will be in converting from an empirical
exemption value to a dose-based exemption value. The commenter's
understanding is that while there is a reduction in dose for the
results that were calculated, the standard deviation and median dose
values both decrease. One commenter was concerned that the proposed
exemption values are not adequately protective for transportation
scenarios, because the IAEA transportation exemption values for some
radionuclides are too high to meet safety goals. The commenter added
that the average annual dose for a representative list of 20
radionuclides (see April 30, 2002; 67 FR 21396) is too high to be safe.
Some commenters stated that NRC should tighten controls on radioactive
materials instead of loosening them because NRC admitted that the
proposed increases in exempt concentrations of radioactive materials
would reduce public safety, One commenter stated that the public is
told not to worry about the proposed exemption values because it will
only be exposed to one millirem of radioactive material. However, the
commenter noted that the 20 most commonly shipped materials with the
new exemption values are at 23 millirem. Therefore, the commenter was
confused about what it meant to only be exposed to one millirem of
radioactive material. One commenter stated that the proposed exemption
values would not enforce the principle of limiting exposure to less
than 1 mrem/yr. Four other commenters opposed the proposed definition
of ``radioactive materials,'' one doing so in the name of national
security. This commenter argued that there are no low-level nuclear
wastes and that there is no safe threshold for exposure to radioactive
materials.
Response. The terms ``practice'' and ``source'' are used in the
context of the IAEA's BSS, and have the meanings provided in the
glossary of that document.
A criterion for the BSS exemption of practices ``without further
consideration'' (Schedule I, paragraph I-3) is that the effective dose
expected to be incurred by any member of the public due to the exempted
practice is of the order of 0.01 mSv (1 mrem) or less in a year.
Estimates of doses resulting from the use of the exemption values in
the transport scenarios have been specifically examined and may result
in doses that exceed 0.01 mSv/yr (1 mrem/yr) (an average of 0.23 mSv/yr
(23 mrem/yr) for 20 commonly transported radionuclides). However, the
dose estimates for the use of the exempt activity concentration values
are less than those resulting from the use of the current 70-Bq/g
(0.002-[mu]Ci/g) activity concentration (an average of 0.5 mSv/yr (50
millirem/yr) for the same 20 radionuclides). The NRC staff notes that
there have been no adverse public health impacts identified from the
use of the current exemption value. Because the annual doses estimated
to result from the use of the radionuclide-specific exemption values
are low, and on average are lower than the dose estimates for the
current 70-Bq/g (0.002-[mu]Ci/g) activity concentration, the NRC staff
believes that changing from the 70-Bq/g (0.002-[mu]Ci/g) value to the
radionuclide-specific exemption values will result in no adverse impact
on public health and safety.
In addition, the transport scenarios are based on exposure periods
(40-500 hours per year) and exposure distances (less than 1.52 m (5
ft)) that overstate actual exposures to workers and greatly overstate
actual exposures to the public. The models used to develop the
exemption values consider the exposure pathways that are significant
for assessment of impact on public health and safety, including
external exposure, inhalation and ingestion, and contamination of the
skin.
The length of the exposure periods and the close distance
assumptions make multiple exposures for the full duration at those
distances to multiple radionuclides very unlikely. The dose estimates
are sufficiently low that NRC believes any actual multiple exposures
would also be acceptably low (well below regulatory limits). Neither
NRC nor DOT has any information to suggest that multiple exposures to
materials regulated under the current 70-Bq/g (0.002-[mu]Ci/g) minimum
activity concentration is of concern.
The NRC believes that regulatory efficiency requires that exemption
values be established for determining when material in transport should
be subject to radioactive material transport safety regulations. The
NRC believes adoption of the radionuclide-specific exemption values is
warranted because it achieves international compatibility without
negative public health impact or undue burden.
Comment. One commenter stated that the proposed regulations were
unclear as to the exact definition of ``per radionuclide.''
Response. The term ``per radionuclide'' means that the doses
estimated to result from the use of the exemption values were
determined for each radionuclide.
Comment. One commenter expressed the lack of understanding of the
concept of the ``millirem.'' To this end, the commenter said that
``millirem'' is a fluid, unenforceable, and unverifiable term.
Response. The term ``millirem'' is a combination of the prefix
``milli,'' meaning one-thousandth, and ``rem,'' an acronym for Roentgen
Equivalent Man, a radiation dosimetry unit. Units of radiation doses,
including rem, are defined in Sec. 20.1004.
Comment. One commenter requested that NRC track, label, and
publicly report all radioactive shipments of any kind, and reject the
exemption tables. The commenter believed that ``harmonization'' was not
an adequate justification for increasing public risk.
Response. The NRC believes that the current regulations require
appropriate measures for hazard communication during transportation. As
noted previously, the public risk from the transportation of exempt
materials, as measured by the average dose, will actually decrease.
Comment. One commenter stated that the new exemption values will
result in bulk shipments of decommissioning soil and debris being
classed as LSA (Low Specific Activity) rather than being exempted from
regulation. The commenter quantified the percentage of his shipments
that would now be classed as LSA. The commenter stated that the
increase in LSA-classified shipments will result in minimal additional
costs.
Response. No response is required.
[[Page 3720]]
Comment. One commenter expressed opposition to the changes in
definitions that could include changing exemption values, particularly
because this is not subject to an EA.
Response. This rule adopts the TS-R-1 exempt material activity
concentrations and exempt consignment activity limits as found in Table
A-2 of the proposed rule. In essence, use of both of these values will
replace the current definition for ``radioactive material'' found in 49
CFR 173.403, and applied in current 10 CFR 71.10. Within the revision
to part 71, reference to the exemption values will be added to the new
Sec. 71.14, ``Exemption for low-level materials,'' to provide an
exemption from NRC requirements during the transportation of these
materials. Estimated impacts from this revision are included in the EA
prepared to support this rulemaking.
Comment. One commenter stated that the redefinition would pose a
threat to national security.
Response. NRC does not believe adoption of the exemption values for
radioactive materials in transport will have any bearing on national
security.
Comment. One commenter expressed concern that the NRC proposed
regulations could increase the variety of materials that are regulated
as ``radioactive'' for transportation purposes.
Response. It is possible that materials that were not regulated
under the previous DOT definition based on 70 Bq/g (0.002-[mu]Ci/g)
would be newly regulated under the exemption values. However, a
material consignment must exceed both the activity concentration for
exempt material and the activity limit for exempt consignment to be
regulated under the final DOT and NRC regulations. It is NRC's position
that regulation of such material consignments as radioactive material
in transport is appropriate.
Comment. One commenter asked the NRC to explain how NRC's official
proposal on the changes in packaging and transporting of radioactive
materials would affect industrial radiography.
Response. The final rule does not affect the transportation of
standard industrial radiography devices.
Comment. One commenter stated that in ``no case should NRC part 71
definitions be relaxed or downgraded merely to provide ``internal
consistency and compatibility with TS-R-1.''' The commenter stated that
those who ``wish to engage in trans-boundary trade in nuclear materials
can be required to meet stiffer U.S. import requirements'' than those
elsewhere in the world. The existing NRC staff justification is ``a
very lame dog that won't hunt,'' and regulatory relaxation is ``both
arbitrary and capricious and unacceptable.'' The commenter stated that
NRC should have definitions with full clarity, and no changes should be
allowed that reduce safety levels or relax requirements. The commenter
was especially troubled with the proposed change to ``radioactive
material'' because this change would ``allow shipments of radioactively
contaminated materials that are declared to be exempted according to
the concentrations and consignment limits shown in the Exemption
Tables.''
Response. NRC believes that the amended definitions and new
adoptions to support definitions for individual Issues are sufficiently
justified and not arbitrary and capricious.
Issue 3. Revision of A\1\ and A\2\
Summary of NRC Final Rule. The final rule adopts, in Appendix A,
Table A-1 of part 71, the new A1 and A2 values
from TS-R-1, except for molybdenum-99 and californium-252. The final
rule does not include A1 and A2 values for the 16
radionuclides that were previously listed in part 71 but which do not
appear in TS-R-1.
The A1 and A2 values were revised by IAEA
based on refined modeling of possible doses from radionuclides. The NRC
believes that these changes are based on sound science, incorporating
the latest in dosimetric modeling and that the changes improve the
transportation regulations. The regulatory analysis indicates that
adopting these values is appropriate from a safety, regulatory, and
cost perspective. Further, adoption of the new A1 and
A2 values will be an overall benefit to public and worker
health and international commerce by ensuring that the A1
and A2 values are consistent within and between
international and domestic transportation regulations. The NRC is not
adopting the A1 value for californium-252 because the IAEA
is considering changing the value that appears in TS-R-1 back to what
presently appears in part 71. The NRC is not adopting the A2
value for molybdenum-99 for domestic commerce because this would result
in a significant increase in the number of packages shipped, and
therefore in potential occupational doses, due to the lower
A2 value in TS-R-1.
Affected Sections. Appendix A.
Background. The international and domestic transportation
regulations use established activity values to specify the amount of
radioactive material that is permitted to be transported in a
particular packaging and for other purposes. These values, known as the
A1 and A2 values, indicate the maximum activity
that is permitted to be transported in a Type A package. The
A1 values apply to special form radioactive material, and
the A2 values apply to normal form radioactive material. See
Sec. 71.4 for definitions.
In the case of a Type A package, the A1 and
A2 values as stated in the regulations apply as package
content limits. Additionally, fractions of these values can be used
(e.g., 1x10-\3\ A2 for a limited quantity of
solid radioactive material in normal form), or multiples of these
values (e.g., 3,000 A2 to establish a highway route
controlled quantity threshold value).
Based on the results from an updated Q-system (see draft Advisory
Material for the Regulations for the Safe Transport of Radioactive
Material, TS-G-1.1, Appendix I), the IAEA adopted new A1 and
A2 values for radionuclides listed in TS-R-1 (see paragraph
201 and Table I). IAEA adopted these new values based on calculations
which were performed using the latest dosimetric models recommended by
the ICRP in Publication 60, ``1990 Recommendations of the ICRP.'' A
thorough review of the Q-system also included incorporation of data
from updated metabolic uptake studies. In addition, several refinements
were introduced in the calculation of contributions to the effective
dose from each of the pathways considered. The pathways themselves are
the same ones considered in the 1985 version of the Q-system: External
photon dose, external beta dose, inhalation dose, skin and ingestion
dose from contamination, and dose from submersion in gaseous
radionuclides. A thorough, up-to-date radiological assessment was
performed for each radionuclide of potential exposures to an individual
should a Type A package of radioactive material be involved in an
accident during transport. The new A1 and A2
values reflect that assessment.
While the dosimetric models and dose pathways within the Q-system
were thoroughly reviewed and updated, the reference doses were
unchanged. The reference doses are the dose values which are used to
define a ``not unacceptable'' dose in the event of an accident.
Consequently, while some revised A1 and A2 values
are higher and some are lower, the potential dose following an accident
is the same as with the previous A1 and A2
values. The general A value radiological criteria are: effective or
committed effective dose to a person should not exceed 50 mSv (5 rem);
the dose or committed dose received by individual organs should not
exceed 0.5 Sv (50 rem) (see IAEA
[[Page 3721]]
TS-G-1.1 for further details on Q-system dosimetric models and
assumptions). Changes in the A values do not change the reference dose
values. The revised dosimetric models are used internationally to
calculate doses from individual radionuclides, and these refinements in
the pathway calculations resulted in various changes to the
A1 and A2 values. In other words, where an
A1 or A2 value has increased, the potential dose
is still the same--the use of the revised dosimetric models just shows
that a higher activity of that radionuclide is actually required to
produce the same reference dose. Conversely, where an A1 or
A2 value has decreased, the revised models show that less
activity of that nuclide is needed to produce the reference dose.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. One commenter stated that the NRC should not reduce the
numbers and types of material subject to shipping regulations. The
commenter was concerned that the proposed rule would:
(1) Exempt numerous radionuclide shipments from any regulation;
(2) Increase worker exposure and the difficulty of enforcement;
(3) Create an inconsistency with other Federal radionuclide
standards; and
(4) Otherwise reduce the protections afforded the public during
radionuclide transportation.
Another commenter stated that the revisions' rationale does not
justify such weakening, that inconsistency with IAEA standards is an
inadequate justification for the proposed changes because there has
been no demonstration that inconsistencies have caused any difficulty.
Finally, one commenter stated that increasing the A1 and
A2 values should not be allowed and added that conforming
with IAEA regulations is an insufficient justification to increase
``levels of exposure to American citizens.'' Further, the commenter
stated that avoiding ``negative impacts on the nuclear industry are not
justifiable reasons for NRC to relax any standards for protection of
the public.''
Response. The NRC disagrees with the first commenter. The final
rule does not exempt numerous radionuclide shipments, nor increase
worker exposure, nor reduce protection to the public, nor create an
inconsistency with other Federal standards.
The NRC disagrees with the second commenter that the final rule
weakens the regulations. Conforming NRC regulations to the IAEA
regulations is not the sole justification; it is also adopting sound
science, incorporating the latest in dosimetric modeling and that the
changes improve the transportation regulations. The regulatory analysis
indicates that adopting these values is appropriate from a safety,
regulatory, and cost perspective.
Comment. One commenter suggested that the NRC organize the
A1 and A2 tables to be sorted alphabetically by
name rather than symbol, because the people who will use these tables
most frequently will be more familiar with the spelling of the name
rather than the chemical symbol. In addition, using the full name will
make the tables easier to use and will be more consistent with the June
1, 1998, Presidential memo, ``Plain Language in Government Writing.''
Response. The comment is acknowledged; however, the tables will
remain sorted as proposed to maintain consistency with the current DOT
and IAEA regulations.
Comment. One commenter stated that the dose to workers could
increase due to their need to handle more packages. The commenter also
stated that the demand for molybdenum-99, the principal isotope used in
medical imaging, would likely increase with the aging population.
Response. The proposed A1 and A2 values
should result in only a minimal change in occupational risk. The
proposed A1 and A2 values are based on the same
reference doses as the current values, and only the dosimetric models
were revised, leading to the updated values. In general, the proposed
A1 and A2 values are within a factor of about
three of the current values; very few radionuclides have proposed
A1 and A2 values that are outside this range.
Currently in part 71, the A2 value for Mo-99 is 0.5 TBq
(13.5 Ci) for international transport and 0.74 TBq (20 Ci) for domestic
transport. The NRC originally proposed an A2 value of 0.6
TBq (16.2 Ci) for Mo-99, but commenters suggested that adopting the
lower A2 value for domestic use would only result in an
increase in the number of packages shipped and, thus, in a potential
increase in occupational dose. Therefore, NRC will retain the current
Mo-99 A2 value of 0.74 TBq (20 Ci) for domestic shipments.
Comment. One commenter indicated that the proposed A1
and A2 values were ``far reaching.'' The commenter was
concerned by the lack of data supporting these significant changes but
generally supported the changes.
Response. NRC does not believe that the proposed changes to the
A1 and A2 values are ``far reaching.'' NRC does
not believe there is a lack of data on the proposed changes to the
A1 and A2 values. Instead, the information on the
Q-system, the details of the exposure pathways, and the actual IAEA
A1 and A2 values are contained in the guidance
document for TS-R-1, TS-G 1.1, and Safety Series 7.
The revisions of the A1 and A2 values are
based on a reexamination/new assessment of the dosimetric models used
in deriving the content limits for Type A packages. The overall impact
of the reexamination resulted in improved methods for the evaluation of
the content limits for special form (denoted by A1) and
nonspecial form (denoted by A2) radioactive material.
Internationally, as increased knowledge and scientific methods are
gained and applied in the areas of health physics, radioactive material
packaging, and radioactive material transportation, it is appropriate
to take advantage of that knowledge and information and apply it to the
IAEA regulations. This has occurred with the revision of the
A1 and A2 values. The IAEA applied the newly-
revised Q-system to the same uptake scenarios it used for the 1985
regulations. Thus, the same dose criteria, which were used in the
assessment of the 1985 A1 and A2 values, were
also used to determine the new A1 and A2 values
in TS-R-1.
While some of the A1 and A2 values have
increased, some values remain unchanged, and some values decreased, the
overall safety implications for TS-R-1 remain the same as those used in
the 1985 IAEA regulations.
Within the Q-system, a series of exposure routes are considered
which may result in radiation exposure to persons near a Type A package
of radioactive material that has been involved in an accident. The
exposure routes include external photon dose, external beta dose,
inhalation dose, skin and ingestion dose due to contamination transfer,
and submersion (exposure to vapor/gas) dose.
Comment. One commenter requested more explanation of the
implications of revision of the A1 and A2 values.
The commenter requested simple summaries for both special form and
normal materials.
Response. See response to the preceding comment. Special form
radioactive material and normal form radioactive material are defined
in Sec. 71.4. In general, special form radioactive material is
subjected to various tests found in Sec. 71.75, ``Qualification of
special form radioactive material.'' These materials
[[Page 3722]]
are known to be nondispersible (will not disperse contamination). Thus,
in a transportation scenario, special form radioactive material could
be considered relatively safer in transport by the fact that it poses
only a direct radiation hazard (and not a contamination hazard). On the
other hand, radioactive material that has not been tested to the
requirements of Sec. 71.75 or has not passed these tests has not
qualified to be considered special form radioactive material. Such
material is called nonspecial form (commonly known as normal form)
radioactive material. In general, these materials pose both a radiation
and contamination hazard in that they are considered to be dispersible.
As an example, consider the A1 and A2 values for
actinium-227 (A1 = 9E-1 TBq (2.4E1 Ci); A2 = 9E-5
TBq (2.4E-3 Ci)). Notice the tremendous difference between
A1 and A2. This example demonstrates that in
special form, a much larger amount of activity can be placed in a Type
A package because the special form material has been sealed or
encapsulated and has proven its robustness by passing the test
requirements of Sec. 71.75. The same encapsulation and testing is not
true for the nonspecial form (A2) value. This is where the
applicability of health physics and metabolic uptake come into
consideration for determining the A1 and A2
values for each individual radionuclide.
Comment. One commenter asked if the justification for the change is
the shift in accepted dose models from ICRP 26 and 30 to 60 and 66. The
commenter requested data supporting the shift in dose models.
Response. The most recent recommendations of the ICRP were issued
in 1991 (1990 Recommendation of the International Commission on
Radiological Protection, Publication No. 60, Pergamon Press, 1991).
Within TS-R-1, IAEA applied the values from ICRP 60 and 66, thus the
shift in dose models. This data can be found in the ICRP 60 and 66
documents.
Comment. One commenter noted that ICRP and IAEA risk models only
look at fatal cancers and ignore nonfatal cancers, years of lost life,
and the bystander effect. The commenter asserted that the ICRP and IAEA
reports do not accurately reflect risk and that low levels of radiation
are more damaging than the models are predicting.
Response. The NRC acknowledges this comment but notes that a
response to similar concerns expressed is provided in the first comment
of section II--Analysis of Public Comments, under the heading: Adequacy
of NRC Regulations and Rulemaking Process.
Comment. One commenter asked if these revisions would actually
expand the number of containers that have to meet test standards.
Response. Within part 71, NRC approves packages and shipping
procedures for fissile radioactive materials and for licensed materials
in quantities that exceed A1 or A2. NRC will
continue to apply the regulations in part 71 to Type B and fissile
radioactive material packages. NRC is not aware of an expansion of the
container inventory which will have to meet test standards due to an
increase in any individual A1 or A2 value.
Comment. One commenter said that the scientific basis for the
changes to the A1 and A2 values is understood and
justified. However, the commenter urged NRC to maintain the exception
(found in Table A-1 of Appendix A to part 71) to allow the domestic
A2 limit of 20 Ci for Mo-99, which, the commenter states, is
necessary to allow domestic manufacturers to continue to provide Mo-99
generators to the diagnostic nuclear medicine community. The commenter
said that changing the A2 limit to the TS-R-1 value would
result in an increase in the number of packages shipped and, thus, an
increase in the doses received by manufacturers, carriers, and end
users.
Response. NRC agrees with this commenter concerning the revision to
the A1 and A2 values and the scientific
background used to support the changes. Further, the commenter has
indicated that the TS-R-1 A2 value for molybdenum-99 would
increase the number of packages shipped and, thus, an increase the
radiation exposure to various workers. Accordingly, to reduce these
concerns NRC will retain the current A2 value for
molybdenum-99 (7.4E-1 TBq; 2.0E1 Ci) as stated in the proposed rule and
as found in Table A-1 for domestic transport. NRC is aware that by
adopting this value (as opposed to the current value for molybdenum-99
in TS-R-1), the number of shipments of molybdenum-99 and the associated
radiation exposure may be reduced.
Comment. One commenter indicated that revising the A1
and A2 values might have an adverse impact on currently
certified casks. The commenter stated that the proposed regulation does
not ensure that transport casks certified under previous revisions will
still be usable without modification or analysis in the future.
Response. Although NRC staff could revise cask certificates if
necessary, no changes are known to be needed to accommodate the revised
A1 and A2 values.
Comment. One commenter stated that because DOE is the principal
shipper of californium-252 under the current exemption value, the
potential impacts to industry could not be assessed.
Response. NRC is aware of the limited and safe transportation of
californium-252 by DOE.
Comment. One commenter stated that by omitting the A1
and A2 values for 16 radionuclides, the Commission would
have to set these values upon future request of a licensee. The
commenter recommended that the NRC not delete these values from part
71, Appendix A, to save NRC the cost and resources necessary to
establish these values in the future.
Response. NRC agrees that more time and effort may be needed to
reintroduce these 16 radionuclides into Appendix A at some time in the
future, as compared to retaining their names and symbols but not
publishing actual A1 and A2 values for them.
Instead, the reference to the general values for A1 and
A2 provided in Table A-3 would be used without NRC approval
for shipping these radionuclides. Further, to maintain consistency/
harmonization with future IAEA transport standards, NRC may adopt a
revised list of A1 and A2 values, should there be
revisions to Table 1 in future editions of the IAEA transport
standards.
Comment. Four commenters agreed with NRC's efforts to revise
A1 and A2 values.
Response. The NRC acknowledges these comments.
Comment. Several commenters disagreed with the NRC staff's
position. One commenter opposed weakening the present standard of
radiation protection during transportation, particularly because NRC is
proposing to ship radioactive wastes to a repository. Another commenter
expressed concern that many, if not most, of the A1 and
A2 values, both current and proposed in the NRC's part 71
regulations, appear to have been arbitrarily chosen and are unsafe.
Another commenter stated that any additional costs ``must be borne by
licensees and beneficiaries of use of materials.'' Another commenter
asked the NRC not to adopt the exemption values contained in Table 2 of
TS-R-1.
Response. NRC does not consider the adoption of the A1
and A2 values from TS-R-1 to be a weakening of the present
standards for packaging and transporting radioactive material. The NRC
believes the revision of the A1 and A2 values to
be based on sound science and that it provides adequate protection to
the public and workers. Furthermore,
[[Page 3723]]
there is not a direct connection between adopting the revised
A1 and A2 values into part 71 and the package
standards and safety requirements which will be imposed on the
transport packages for high-level waste en route to a geologic
repository.
The process used to determine the appropriate A1 and
A2 value assigned to each radionuclide is based on several
factors. These include the type of radiation emitted by the
radionuclide e.g., alpha, beta, or gamma), the energy of that radiation
i.e., strong alpha emitter, strong gamma emitter, weak beta emitter,
etc.), and the form of the material (nondispersible as applied to
special form radioactive material, or dispersible as applied to
nonspecial form radioactive material). All of these factors have been
modeled in the IAEA's Q-system to determine the appropriate value to be
assigned to each radionuclide. Thus, the values have not been
arbitrarily obtained, and they are safe. Further, the revision to the
A1 and A2 values in TS-R-1 has maintained the
same level of safety as was applied in determining the A1
and A2 values for the radionuclides in the 1985 IAEA
transportation standards. Thus, there is no weakening of the intended
safety aspects of the new A1 and A2 values.
Comment. Several commenters noted various typographical errors. The
first commenter noted that Footnote 2 to Table A-1 is incorrect and
should instead read, ``See Table A-4.'' The second commenter noted an
error in the proposed Table A-1 for the A2 (Ci) value for
Pu-239, suggesting that the correct value should be 2.7 x
10-2 Ci, as evidenced from the A2 (TBq) value for
Pu-239 and the similar Table 1 in the IAEA TS-R-1 regulations and Table
10A in the proposed DOT regulations.
Response. NRC acknowledges the comment, and corrections have been
made to the final rule.
Comment. One commenter addressed changing a number of the
radionuclide values. The commenter suggested that the radionuclide Al-
26 value for specific activity in 10 CFR part 71, Table A-1, should be
changed from 190 Ci/g to 0.019 Ci/g. The A1 and
A2 values in both 10 CFR part 71 Table A-1 and 49 CFR
173.435 for Ar-39 appear reversed from that listed in IAEA TS-R-1. The
radionuclide Be-10 value for specific activity in 10 CFR part 71 Table
A-1 should be changed from 220 Ci/g to 0.022 Ci/g. The radionuclide Cs-
136 value for specific activity in 49 CFR 173.435 should be changed
from 0.0027 TBq/g to 270 TBq/g. The radionuclide Dy-165 value for
A2 (Ci) in 10 CFR part 71 Table A-1 should be changed from
0.16 to 16 Ci. The radionuclide Eu-150 (long-lived) value for
A1 (TBq) in 10 CFR part 71 Table A-1 and 49 CFR 173.435 is
not consistent with the IAEA TS-R-1 value of 0.7. The radionuclide Fe-
59 value for A2 (TBq) in 10 CFR part 71 Table A-1 is in
error. The radionuclide Ho-166m value for A2 (TBq) in 10 CFR
part 71 Table A-1 should be 0.5. The radionuclide K-43 value for
A2 (TBq) in 10 CFR part 71 Table A-1 should be 0.6. The
radionuclide Kr-81 value for A1 (TBq) in 49 CFR 173.435
should be 40, A1 (Ci) in 49 CFR 173.435 should be 1100. The
radionuclide Kr-85 value for A2 (TBq) in 49 CFR 173.435
should be 10; A2 (Ci) in 49 CFR 173.435 should be 270. The
radionuclide La-140 value for A2 (Ci) in 49 CFR 173.435
should be 11. The radionuclide Lu-177 value for A2 (TBq) in
49 CFR 173.435 should be 0.7; A2 (Ci) in 49 CFR 173.435
should be 19. The radionuclide Mn-52 value for specific activity (Ci)
in 49 CFR 173.435 should be 4.4E+05. The radionuclide Np-236 (long-
lived) value for A1 (TBq) in IAEA TS-R-1 is 9; A2
(TBq) in IAEA TS-R-1 is 0.02, different from the values in both 49 CFR
173.435 and 10 CFR part 71, Table A-1. The radionuclide Pt-197m value
for A2 (TBq) in 49 CFR 173.435 should be 0.6; A2
(Ci) in 49 CFR 173.435 should be 16. The radionuclide Pu-239 value for
A2 (Ci) in 10 CFR part 71, Table A-1, should be 0.027. The
radionuclide Pu-240 value for specific activity (Ci) should be 0.23 Ci/
g. The radionuclide Ra-225 value for A2 (Ci) in 10 CFR part
71, Table A-1, should be 0.11. The radionuclide Ra-228 value for
A2 (TBq) in 10 CFR part 71, Table A-1, should be 0.02. The
radionuclide Rh-105 value for A2 (Ci) in 10 CFR part 71,
Table A-1, is in error. The radionuclide Sc-46 value for A1
(TBq) in 10 CFR part 71, Table A-1, should be 0.5. The radionuclide Sn-
119m value for A2 (TBq) in 10 CFR part 71, Table A-1, should
be 30. The radionuclide Sn-126 value for specific activity (TBq) in 10
CFR part 71, Table A-1, should be 0.001. The radionuclide H-3 value for
A2 (TBq) in 10 CFR part 71, Table A-1, should be 40. The
radionuclide Ta-179 value for A1 (TBq) in 10 CFR part 71,
Table A-1, should be 30. The radionuclide Tb-157 value for
A1 (TBq) in 10 CFR part 71, Table A-1, should be 40; value
for specific activity (TBq) in 10 CFR part 71, Table A-1, should be
0.56 TBq/g. The radionuclide Tb-158 value for A2 (Ci) in 10
CFR part 71, Table A-1, should be 27; value for specific activity (TBq)
in 10 CFR part 71, Table A-1, should be 0.56 TBq/g.
The radionuclide Tb-160 value for A1 (Ci) in 10 CFR part
71, Table A-1, should be 27. The radionuclide Tc-96 value for
A1 (TBq) in 10 CFR part 71, Table A-1, should be 0.4. The
radionuclide Tb-96m value for A1 (TBq) in 10 CFR part 71,
Table A-1, should be 0.4; value for A2 (TBq) in 10 CFR part
71, Table A-1, should be 0.4. The radionuclide Tc-97 value for specific
activity (TBq) in 10 CFR part 71, Table A-1, should be 5.2E-05; value
for specific activity in 10 CFR part 71, Table A-1, should be 0.0014.
The radionuclide Te-125m value for A2 (Ci) in 10 CFR part
71, Table A-1, should be 24. The radionuclide Te-129 value for
A1 (TBq) in 10 CFR part 71, Table A-1, should be 0.7; value
for A2 (TBq) in 10 CFR part 71, Table A-1, should be 0.6.
The radionuclide Te-132 value for A1 (TBq) in 10 CFR part
71, Table A-1, should be 0.5. The radionuclide Th-227 value for
A2 (Ci) in 10 CFR part 71, Table A-1, should be 0.14. The
radionuclide Th-231 value for A2 (TBq) in 10 CFR part 71,
Table A-1, should be 0.02. The radionuclide Th-234 value for
A1 (TBq) in 10 CFR part 71, Table A-1, should be 0.3. The
radionuclide Ti-44 value for A1 (TBq) in 10 CFR part 71,
Table A-1, should be 0.5; value for A2 (TBq) in 10 CFR part
71, Table A-1, should be 0.4, value for A2 (Ci) in 10 CFR
part 71, Table A-1, should be 10. The radionuclide Tl-200 value for
A1 (TBq) in 10 CFR part 71, Table A-1, should be 0.9. The
radionuclide Tl-204 value for A2 (TBq) in 10 CFR part 71,
Table A-1, should be 0.7. The radionuclide U-230, U-232, U-233, and U-
234 values for medium and slow lung absorption, and U-236 values for
slow lung absorption are not consistent with IAEA TS-R-1. The comment
points out that the Table values published in the Federal Register for
the proposed rule did not match TS-R-1.
Response. NRC accepts the comment and has updated the values in the
final rule, Table A-1, to be consistent with TS-R-1. Appropriate
changes have been made in the final rule.
Comment. Three commenters stated that the A2 value for
molybdenum-99 and the A1 and A2 values for
californium-252 should be retained for domestic use only packages.
Response. NRC agrees with the comment. (See 67 FR 21399; April 30,
2002, for more details.)
Issue 4. Uranium Hexafluoride (UF6) Package Requirements
Summary of NRC Final Rule. The final rule provides, in new Sec.
71.55(g), a specific exception for certain uranium hexafluoride
(UF6) packages from the requirements of Sec. 71.55(b). The
exception allows UF6 packages to be evaluated for
criticality safety without considering the in leakage of water into
[[Page 3724]]
the containment system provided certain conditions are met, including
that the uranium is enriched to not more than 5 weight percent uranium-
235. The rule makes part 71 compatible with TS-R-1, paragraph 677(b).
Other uranium hexafluoride package requirements in TS-R-1 (paragraphs
629, 630 and 631) do not necessitate changes for compatibility because
NRC uses analogous national standards and addresses package design
requirements in its design review process.
The specific exception being placed into the regulations for the
criticality safety evaluation of certain uranium hexaflouride packages
does not alter present practice which has allowed the same type of
evaluation under other more general regulatory provisions. NRC has
decided to provide this specific exception: (1) To be consistent with
the worldwide practice and limits established in national and
international standards (ANSI N14.1 and IS 7195) and current U.S.
regulations (49 CFR 173.417(b)(5)); (2) because of the history of safe
shipment; and (3) because of the essential need to transport the
commodity.
Affected Sections. Section 71.55.
Background. Requirements for UF6 packaging and
transportation are found in both NRC and DOT regulations. The DOT
regulations contain requirements that govern many aspects of
UF6 packaging and shipment preparation, including a
requirement that the UF6 material be packaged in cylinders
that meet the ANSI N14.1 standard. NRC regulations address fissile
materials and Type B packaging designs for all materials.
TS-R-1 contains detailed requirements for UF6 packages
designed for transport of more than 0.1 kilogram (kg) UF6.
First, TS-R-1 requires the use of the International Organization for
Standardization (ISO) 7195, ``Packaging of Uranium Hexafluoride for
Transport.'' Second, TS-R-1 requires that all packages containing more
than 0.1 kg UF6 must meet the ``normal conditions of
transport'' drop test, a minimum internal pressure test, and the
hypothetical accident condition thermal test (para 630). However, TS-R-
1 does allow a competent national authority to waive certain design
requirements, including the thermal test for packages designed to
contain greater than 9,000 kg UF6, provided that
multilateral approval is obtained. Third, TS-R-1 prohibits
UF6 packages from using pressure relief devices (para 631).
Fourth, TS-R-1 includes a new exception for UF6 packages
regarding the evaluation of criticality safety of a single package.
This new exception (para 677(b)) allows UF6 packages to be
evaluated for criticality safety without considering the in leakage of
water into the containment system. Consequently, a single fissile
UF6 package does not have to be subcritical assuming that
water leaks into the containment system. This provision only applies
when there is no contact between the valve body and the cylinder body
under accident tests, and the valve remains leak-tight, and when there
are quality controls in the manufacture, maintenance, and repair of
packages coupled with tests to demonstrate closure of each package
before each shipment.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC responses for this issue
follows:
Comment. Five commenters expressed support for the proposed changes
to UF6 package rules that continue the current practice of
moderator exclusion for UF6. One commenter cited the strong
safety record applying these rules as evidence that the practice is
adequate. Two commenters objected to the 5 percent enrichment limit
provision in proposed Sec. 71.55(g), and a third commenter expressed
concern with the enrichment limit. One commenter noted that the safety
case for the specific enrichment to use can be a part of the package
certification application and, therefore, does not need to be specified
by rule. The same commenter further noted that arguments that water in
leakage is not a realistic scenario for a UF6 cylinder
regardless of enrichment and that the 5 percent limit, if imposed for
transportation, could have very high cost implications in light of
pending decisions to use higher enrichments in the fuel cycle. One
commenter suggested that the rule retain the limit of 5 percent for the
existing ANSI N14.1 Model 30B cylinder, but that the rule also contain
provisions that permit greater than 5 percent enrichments in an
``improved UF6 package with special design features'' to
accommodate future industry plans.
Response. The NRC's decision to exempt uranium hexafluoride
cylinders from Sec. 71.55(b) with a limiting condition of 5 weight
percent enriched uranium was made based on:
(1) Consistency with the worldwide practice and limits established
in national and international standards (ANSI N14.1 and IS 7195) and
current U.S. regulations (49 CFR 173.417(b)(5));
(2) The history of safe shipment; and
(3) The essential need to transport the commodity.
The NRC staff believes that further expansion of the practice of
authorizing shipment of materials in packages that do not meet Sec.
71.55(b), without a strong technical safety basis and without full
understanding of the potential reduction in safety margins, is not
prudent or necessary at this time. In addition, provisions are
available to request approval of alternative package designs that could
be used for the shipment of uranium hexafluoride with uranium
enrichments greater than 5 weight percent under the provisions of Sec.
71.55(b) or Sec. 71.55(c). Merits of a new or modified design that
included special design features could be reviewed and approved under
the provisions of Sec. 71.55, including Sec. 71.55(c).
Because package certification is directly tied to the regulations,
any assessment of the safety of enrichments greater than 5 weight
percent uranium-235, considering the potential or probability of water
in leakage, would not be part of the safety case of an application if
the enrichment limit is not included as part of the regulation.
Although it is correct that the water in leakage scenario is not
changed for enrichments less than or greater than 5 weight percent, it
is not clear that the safety margins against accidental nuclear
criticality for all enrichments would be the same if water were
introduced into the containment vessel accidentally. Because these
margins are undefined at this time, it does not seem prudent or
necessary to modify the regulatory standard that was based on worldwide
practice in existence today. Future changes in the fuel cycle that
could necessitate transport of enrichments greater than 5 weight
percent uranium-235 could result in new packages designed to meet the
normal fissile material package standards in Sec. 71.55(b), as are
required for other commodities, or could include special design
features that would enhance nuclear criticality safety for transport
for approval under the provisions of Sec. 71.55(c). Alternatively, a
safety assessment could be developed for possible transport of
enrichments greater than 5 weight percent to support some future
rulemaking to modify Sec. 71.55(g) to increase the enrichment
limitation.
For the previously mentioned reasons, the NRC staff has retained
the 5 percent enrichment limit in the final rule.
Comment. One commenter stated an opinion that all UF6
packages should have overpacks and noted that the proposed rule should
resolve this issue.
Response. The NRC staff does not agree with the position that all
UF6 packages be required by rule to
[[Page 3725]]
incorporate an overpack. Design and performance standards for fissile
UF6 packages are stated in part 71, and design and
performance standards for nonfissile UF6 packages appear in
DOT regulations. Use of specific design features (e.g., overpacks) to
meet regulatory standards is left to designers.
Comment. One commenter expressed concern that NRC had not provided
data to back up its proposal to ``relax the current packaging
requirements'' in Sec. 71.55(b) for UF6. The commenter
stated that NRC should not adopt this proposal unless it can provide
justification for doing so. The commenter was also concerned that NRC's
EA does not address any impacts associated with this proposal.
Response. The NRC staff disagrees with the commenter's assertion
that adoption of Sec. 71.55(g) is a relaxation of current packaging
requirements in Sec. 71.55(b). As noted by the commenter, NRC's
proposed rule (67 FR 21400) explains that the new Sec. 71.55(g)
provisions are consistent with existing worldwide practice for UF6
packages. This worldwide practice has been in use since its development
in the 1950s, and the functioning of the nuclear fuel cycle in the U.S.
relies upon transport of this commodity. The exception was limited to 5
weight percent enriched uranium consistent with the worldwide practice
and limits established in national and international standards (ANSI
N14.1 and IS 7195) and current U.S. regulations (49 CFR 173.417(b)(5)).
The new regulatory text replaces the more general ``special features''
allowances with a more explicit provision pertaining to certain
UF6 packages.
Comment. Two commenters expressed opposition for the relaxation of
testing for radioactive transport containers. One commenter stated that
the drop test, minimum internal pressure test, and the hypothetical
accident condition test must be accompanied by the thermal test to
assure public protection in the event of an accident. One commenter
cited both the Baltimore tunnel fire and the Arkansas bridge incident
as justifications for not allowing any exemptions.
Response. The NRC staff reviewed these comments and determined that
they concern the nonfissile UF6 packaging issues discussed
in Issue 6 in the DOT's proposed rulemaking (April 30, 2002; 67 FR
21337), not the fissile UF6 package matters in Issue 4 in
the related NRC proposed rulemaking. The NRC staff noted that the
commenter's letter was jointly addressed to NRC and DOT for resolution
in their final rule.
Issue 5. Introduction of the Criticality Safety Index Requirements
Summary of NRC Final Rule. The final rule adopts the TS-R-1
(paragraphs 218 and 530). Paragraph 218 results in NRC incorporating a
Criticality Safety Index (CSI) in part 71 that is determined in the
same manner as current part 71 ``Transport Index for criticality
control purposes,'' but now it must be displayed on shipments of
fissile material (paragraphs 544-545) using a new ``fissile material''
label. NRC's adoption of TS-R-1 (paragraph 530) increases the CSI-per
package limit from 10 to 50 for fissile material packages in
nonexclusive use shipments. (The previous Transport Index criticality
limit was 10.) The TI is determined in the same way as the ``TI for
radiation control purposes'' and continues to be displayed on the
traditional ``radioactive material'' label. The basis for these changes
that makes part 71 compatible with TS-R-1 is that NRC believes the
differentiation between criticality control and radiation protection
would better define the hazards associated with a given package and,
therefore, provide better package hazard information to emergency
responders. The increase in the per package CSI limit may provide
additional flexibility to licensees by permitting the increased use of
less expensive, nonexclusive use shipments. However, licensees will
still retain the flexibility to ship a larger number of packages of
fissile material on an exclusive use conveyance. The adoption of the
CSI values would make part 71 consistent with TS-R-1 and, therefore,
would enhance regulatory efficiency.
Affected Sections. Sections 71.4, 71.18, 71.20, 71.59.
Background. Historically, the IAEA and U.S. regulations (both NRC
and DOT) have used a term known as the Transport Index (TI) to
determine appropriate safety requirements during transport. The TI has
been used to control the accumulation of packages for both radiological
safety and criticality safety purposes and to specify minimum
separation distances from persons (radiological safety). The TI has
been a single number which is the larger of two values: the ``TI for
criticality control purposes''; and the ``TI for radiation control
purposes.'' Taking the larger of the two values has ensured
conservatism in limiting the accumulation of packages in conveyances
and in-transit storage areas.
TS-R-1 (paragraph 218) has introduced the concept of a CSI separate
from the old TI. As a result, the TI was redefined in TS-R-1. The CSI
is determined in the same way as the ``TI for criticality control
purposes,'' but now it must be displayed on shipments of fissile
material (paragraphs 544 and 545) using a new ``fissile material''
label. The redefined TI is determined in the same way as the ``TI for
radiation control purposes'' and continues to be displayed on the
traditional ``radioactive material'' label.
TS-R-1 (paragraph 530) also increased the allowable per package TI
limit (for criticality control purposes (new CSI)) from 10 to 50 for
nonexclusive use shipments. No change was made to the per package
radiation TI limit of 10 for nonexclusive use shipments. As noted
above, a consolidated radiation safety and CSI existed in the past. In
this consolidated index, the per package TI limit of 10 was
historically based on concerns regarding the fogging of photographic
film in transit, because film might also be present on a nonexclusive
use conveyance. Consequently, when the single radiation and criticality
safety indexes were split into the TI and CSI indexes, the IAEA
determined that the CSI per package limit, for fissile material
packages that are shipped on a nonexclusive use conveyance, could be
raised from 10 to 50. The IAEA believed that limiting the total CSI to
less than or equal to 50 in a nonexclusive use shipment provided
sufficient safety margin, whether the shipment contains a single
package or multiple packages. Therefore, the per package CSI limit, for
nonexclusive use shipments, can be safely raised from 10 to 50, thereby
providing additional flexibility to shippers. Additionally, no change
was made to the per package CSI limit of 100 for exclusive use
shipments.
The NRC believes the differentiation between criticality control
and radiation protection would better define the hazards associated
with a given package and, therefore, provide better package hazard
information to emergency responders. The increase in the per package
CSI limit may provide additional flexibility to licensees by permitting
the increased use of less expensive, nonexclusive use shipments.
However, licensees will still retain the flexibility to ship a larger
number of packages of fissile material on an exclusive use conveyance.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment 1. One commenter requested a basic explanation of the CSI
[[Page 3726]]
and TI. The commenter questioned if the proposed changes would increase
public risk. Another commenter asked for clarification on how NRC would
calculate CSI for radiological shipments to ensure that a shipment is
under limits.
Response. The requested explanation was provided during the June 4,
2001, public meeting at which the first comment was made (see NRC
rulemaking interactive Web site at http://ruleforum.llnl.gov. In
addition, the proposed rule contains background on the CSI; regarding
increased public risk. The draft RA concluded the change is appropriate
from a safety perspective. Also, see Background discussion for this
issue.
Comment. One commenter expressed opposition to the text that would
restrict accumulations of fissile material to a total CSI of 50 in
situations where radioactive materials are stored incident to
transport. The commenter added that this would effectively remove the
ability to transport internationally and/or by multiple modes under
exclusive use conditions and would negatively impact the international
movement of fissile materials under nonproliferation programs. The
commenter further noted that this provision would apply only to
shipments to or from the U.S., thus creating a disadvantage for
American businesses in the international market.
Response. The NRC agrees with these comments. The intent of the
storage phrase was to permit segregation of groups of stored packages,
consistent with IAEA and DOT requirements, but the NRC staff believes
that the proposed text did not accommodate that practice. DOT
requirements restrict accumulation of packages during transport, based
on summing the packages' CSI or TI, including during storage incident
to transport. In light of the division of regulatory responsibilities
explained in the NRC-DOT Memorandum of Understanding (44 FR 38690; July
2, 1979), the NRC exemptions for carriers-in-transit in 10 CFR 70.12,
and DOT's proposed 49 CFR 173.457 (67 FR 21384; April 30, 2002), the
NRC staff believes that storage in transit provisions proposed in
Sec.Sec. 71.59(c)(1), 71.22(d)(3), and 71.23(d)(3) are unwarranted. The
NRC has deleted the phrase ``or stored incident to transport'' from
these sections.
Comment. One commenter stated that in proposed Sec.Sec. 71.59(
c)(1), (2) and (3), and 71.55(f)(3), the values of 50.0 and 100.0
should be changed to 50 and 100 to be consistent with the application
of the CSI.
Response. The NRC staff did not intend nor does it believe that
there is a substantive difference between ``50'' and ``50.0'' as used
in part 71. In proposing to use the decimal place, the NRC staff was
attempting to increase precision when the CSI is exactly 50.0 and
promote consistency as the CSI is by definition rounded to the nearest
tenth. However, the NRC staff noted that both DOT's proposed rule and
IAEA TS-R-1 use ``50'' without a decimal place. The NRC staff agrees
that consistency amongst the three rules is desirable unless a reason
exists for differentiating. Accordingly, conforming changes have been
made to the part 71 final rule.
Comment. One commenter expressed opposition to the rounding of the
CSI provision in the proposed rule, because it is inconsistent with TS-
R-1 and places additional limits on the array size of shipments.
Response. The commenter correctly observes that Sec. 71.59(b)
requires all nonzero CSIs to be rounded up to the first decimal place
and that the corresponding TS-R-1 requirement (paragraph 528) does not
require such rounding. Rounding up the CSI is necessary to ensure that
an unanalyzed number of packages are not transported together; rounding
a CSI down would permit such situations. The NRC staff notes that this
U.S. provision predates the currently contemplated changes for
compatibility with TS-R-1 (viz., the existing U.S. domestic regulations
are also different than the 1985 IAEA transport regulations in this
respect).
Consistent with the NRC proposal, the IAEA's implementing guidance
for TS-R-1 (i.e., TS-G-1.1 at para. 528.3) states, ``The CSI for a
package * * * should be rounded up to the first decimal place'' and
``the CSI should not be rounded down.'' The NRC staff noted that the
IAEA's guidance, however, does observe that use of the exact CSI value
may be appropriate in cases when rounding results in less than the
analyzed number of packages to be shipped.
The NRC staff believes that the rule is compatible with IAEA TS-R-
1. Furthermore, because the domestic convention on rounding predates
this rulemaking for compatibility with 1996 TS-R-1, and because the
statements of consideration did not explicitly discuss the rounding
practice, the potential elimination of the rounding practice is beyond
the scope of the current rulemaking action.
Comment. Three commenters expressed agreement with NRC's proposed
position. One of the three commenters expressed support for the NRC's
CSI proposal, reasoning that it provides more accurate communication
regarding radioactive material in transport, especially in conjunction
with the TI for radiation exposure. The commenter noted that the CSI is
important to ensure consistency between domestic and international
movements of fissile material. Another commenter stated that use of the
CSI would ``remove a source of confusion with the old TI values. The
resulting enhancement of the safety of shipments makes the extra
efforts necessary to implement these proposals worthwhile.''
Response. No response is necessary.
Comment. One commenter stated that the CSI ``should be set so as to
maximize protective benefit for workers and the public without regard
for added costs to licensees and users.'' The commenter added that
there doesn't seem to be a ``strong argument against adoption'' of the
IAEA CSI but then stated that the increase from 10 to 50 per package
does not have adequate justification. Further, the commenter stated
that if cost reduction for licensees is the only reason for this
change, then the proposal is unacceptable.
Response. The CSI is derived to prevent nuclear criticality for
single packages and arrays of packages, both in incident-free and
accident conditions of transport. Therefore, the NRC staff has
determined that the application of the CSI does support protection of
workers and the public. The basis for increasing the accumulation of
packages from 10 TI under the old system to 50 CSI in the new system is
given in the proposed rule (at 67 FR 21401), and it is not a solely
economic basis. Specifically, the limit of 10 TI was based on radiation
damage to film, so when the TI and CSI were split in 1996, a separate
limit on package accumulation based on criticality prevention, of 50
CSI, became warranted.
Issue 6. Type C Packages and Low Dispersible Material
Summary of NRC Final Rule. The final rule does not adopt the Type C
or Low dispersible material (LDM) requirements for plutonium air
transport as introduced in the IAEA TS-R-1. NRC decided not to adopt
Type C or LDM requirements because the U.S. regulations in Sec.Sec.
71.64 and 71.71 governing plutonium air transportation to, within, or
over the United States contains more rigorous packaging standards than
those in the IAEA TS-R-1. Furthermore, the NRC's perception is that
there is a lack of current or anticipated need for such packages, and
NRC acknowledges that the DOT import/export provisions permit use of
IAEA regulations.
[[Page 3727]]
Affected Sections. None (not adopted).
Background. TS-R-1 introduced two new concepts: the Type C package
(paragraphs 230, 667-670, 730, 734-737) and the LDM. The Type C
packages are designed to withstand severe accident conditions in air
transport without loss of containment or significant increase in
external radiation levels. The LDM has limited radiation hazard and low
dispersibility; as such, it could continue to be transported by
aircraft in Type B packages (i.e., LDM is excepted from the TS-R-1 Type
C package requirements). United States regulations do not contain a
Type C package or LDM category but do have specific requirements for
the air transport of plutonium (Sec.Sec. 71.64 and 71.74). These
specific NRC requirements for air transport of plutonium would continue
to apply.
The Type C requirements apply to all radionuclides packaged for air
transport that contain a total activity value above 3,000 A1
or 100,000 A2, whichever is less, for special form material,
or above 3,000 A2 for all other radioactive material. Below
these thresholds, Type B packages would be permitted to be used in air
transport. The Type C package performance requirements are
significantly more stringent than those for Type B packages. For
example, a 90-meter per second (m/s) impact test is required instead of
the 9-meter drop test. A 60-minute fire test is required instead of the
30-minute requirement for Type B packages. There are other additional
tests, such as a puncture/tearing test, imposed for Type C packages.
These stringent tests are expected to result in package designs that
would survive more severe aircraft accidents than Type B package
designs.
The LDM specification was added in TS-R-1 to account for
radioactive materials (package contents) that have inherently limited
dispersibility, solubility, and external radiation levels. The test
requirements for LDM to demonstrate limited dispersibility and
leachability are a subset of the Type C package requirements (90-m/s
impact and 60-minute thermal test) with an added solubility test, and
must be performed on the material without packaging for nonplutonium
materials. The LDM must also have an external radiation level below 10
mSv/hr (1 rem/hr) at 3 meters. Specific acceptance criteria are
established for evaluating the performance of the material during and
after the tests (less than 100 A2 in gaseous or particulate
form of less than 100-micrometer aerodynamic equivalent diameter and
less than 100 A2 in solution). These stringent performance
and acceptance requirements are intended to ensure that these materials
can continue to be transported safely in Type B packages aboard
aircraft.
In 1996, the NRC communicated to the IAEA that the NRC did not
oppose the IAEA adoption of the newly created Type C packaging
standards (letter dated May 31, 1996, from James M. Taylor, EDO, NRC,
to A. Bishop, President, Atomic Energy Control Board, Ottawa, Canada).
However, Mr. Taylor stated in the letter that to be consistent with
U.S. law, any plutonium air transport to, within, or over the U.S. will
be subject to the more rigorous U.S. packaging standards. Industry
needs to be aware of changes or potential changes based on new IAEA
standards.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. Four commenters expressed support for NRC's proposal to
not adopt the requirements for Type C packages and LDM. One commenter
also expressed support for the NRC's decision to ensure that there is a
mechanism for reviewing validations of foreign approvals. One commenter
stated that the IAEA specification is too broad and that NRC and DOT
should work with IAEA to reduce the scope to a few packages containing
fissile oxides of plutonium, but there is no need for this package to
transport Class 7 materials.
Two commenters stated that the benefits did not justify the costs
of the proposed changes and strongly supported the NRC position not to
adopt the Type C requirements. One commenter stated that many parties
are asking IAEA to modify the Type C requirements. The commenter urged
NRC to see how these change proposals will affect the Type C
requirements before adopting them into the U.S. regulations.
Additionally, the commenter stated that the need for Type C packages
for all radioactive material has not been demonstrated.
Response. The NRC staff acknowledges these comments that endorse
the position to not adopt Type C package requirements at this time, for
the reasons specified in the proposed rule (67 FR 21402). The NRC staff
agrees that Type C issues will likely receive further consideration in
future IAEA rule cycles. No further response is necessary.
Comment. Two commenters stated that the threat of terrorism should
be taken into account when exempting radionuclides from transport
regulations and changing container regulations. One commenter stated
that the fact of the September 11, 2001, attacks needs to be accounted
for with upgraded Types B and C testing, which are currently believed
to be insufficient. The commenter added that these tests should
``assure the highest probability that packages will survive
unbreached.''
Response. The NRC acknowledges the concern expressed regarding the
threat of terrorism. However, the NRC does not propose adopting Type C
and LDM requirements at this time. The NRC staff notes that the IAEA is
conducting further evaluations on Type C package requirements, which
may result in other changes for safety and security purposes. Also, see
Section II, above, for general comments on terrorism.
Comment. One commenter asked if workers will be protected and
notified when handling Type C packages and plutonium, and whether they
will be notified that there will be increased hazards once the proposed
rule is effective.
Response. The requested information on worker protection was
provided at the public meeting at which the comment was made.
Application of DOT's regulations, including hazardous materials
training requirements, package radiation limits, and contamination
limits, will protect workers for Type C packages just as for other
shipments. In addition, the robustness of the packaging would provide
protection in accidents. Thus, changes to the probability or
consequences of releases in accidents do not result from proposed
changes to Type C packages. The NRC does not propose adopting IAEA Type
C or LDM standards at this time, and domestic regulations were not
revised.
Comment. One commenter recommended that the NRC ``adopt these
provisions in order to better the goal of compatibility with IAEA
regulations.'' This commenter continued by stating that ``industry
would then have a basis for developing such a package if desirable.''
Response. These comments recommend adoption of Type C standards in
the interest of the goal of IAEA compatibility and speculate that a
domestic Type C package regulation and certification might be desirable
in the future. The NRC staff does not believe that deferring domestic
rules on Type C packages makes U.S. regulations incompatible with IAEA
regulations (viz., the U.S. and IAEA rules are not identical but they
are compatible). The NRC staff believes there is not a need to adopt
Type C standards at this time because of the reasons specified in the
proposed rule (67 FR 21402) and
[[Page 3728]]
(a) The perception of a lack of a current or anticipated need,
(b) The DOT import/export provisions that permit use of IAEA
regulations, and
(c) The existing U.S. regulations and laws covering plutonium air
transport.
This can be reevaluated during future periodic rulemakings for IAEA
compatibility, as necessary. In addition, the proposed rule stated that
upon request from DOT, NRC would perform a technical review of Type C
packages against IAEA TS-R-1 standards. The comments do not indicate a
current need; therefore, the NRC staff has decided to retain the
position explained in its proposed rule to not adopt Type C or LDM
requirements.
Comment. One commenter said that air transport of plutonium and
other radionuclides should be prohibited under all circumstances. The
commenter stated that ``low dispersible materials'' is a faulty concept
regarding air transport and urged NRC to abandon this concept.
Response. The NRC staff disagrees with the comments that air
transport of plutonium and other radionuclides should be prohibited
under all circumstances. These practices are recognized in multiple
U.S. laws and regulations, and have been carried out with an excellent
safety record. Consistent with the position expressed in the proposed
rule, the NRC decided not to adopt the low dispersible material
provisions at this time.
Issue 7. Deep Immersion Test
Summary of NRC Final Rule. The final rule adopts the requirement
for an enhanced water immersion test (deep immersion test) which is
applicable to any Type B or C packages containing activity greater than
105A2. The purpose of the deep immersion test is
to ensure package recoverability. The basis for expanding the scope of
the deep immersion test to include additional Type B or C packages
containing activity greater that 105A2 was due to
the fact that radioactive materials, such as plutonium and high-level
radioactive waste, are increasingly being transported by sea in large
quantities. The threshold defining a large quantity as a multiple of
A2 is considered to be a more appropriate criterion to cover
all radioactive materials and is based on a consideration of potential
radioactive exposure resulting from an accident. Also, the NRC is
retaining the current test requirements in Sec. 71.61 of ``one hour w/o
collapse, buckling or leakage of water.'' The NRC is retaining this
acceptance criterion of ``w/o collapse, buckling, or leakage'' as
opposed to the acceptance criterion specified in TS-R-1 of only ``no
rupture'' of the containment. NRC has determined that the term
``rupture'' cannot be determined by engineering analysis and the term
``w/o collapse, buckling or leakage of water'' is a more precise
definition for acceptance criterion.
Affected Sections. Sections 71.41, 71.51, 71.61.
Background. TS-R-1 expanded the performance requirement for the
deep water immersion test (paragraphs 657 and 730) from the
requirements in the IAEA Safety Series No. 6, 1985 edition. Previously,
the deep immersion test was only required for packages of irradiated
fuel exceeding 37 PBq (1,000,000 Ci). The deep immersion test
requirement is found in Safety Series No. 6, paragraphs 550 and 630,
and basically stated that the test specimen be immersed under a head of
water of at least 200 meters (660 ft) for a period of not less than 1
hour, and that an external gauge pressure of at least 2 MPa (290 psi)
shall be considered to meet these conditions. The TS-R-1 expanded
immersion test requirement (now called enhanced immersion test) now
applies to all Type B(U) (unilateral) and B(M) (multilateral) packages
containing more than 10 \5\ A2, as well as Type C packages.
In its September 28, 1995 (60 FR 50248), rulemaking for part 71
compatibility with the 1985 edition of Safety Series No. 6, the NRC
addressed the new Safety Series No. 6 requirement for spent fuel
packages by adding Sec. 71.61, ``Special requirements for irradiated
nuclear fuel shipments.'' Currently, Sec. 71.61 is more conservative
than Safety Series No. 6 with respect to irradiated fuel package design
requirements. It requires that a package for irradiated nuclear fuel
with activity greater than 37 PBq (10 \6\ Ci) must be designed so that
its undamaged containment system can withstand an external water
pressure of 2 MPa (290 psi) for a period of not less than 1 hour
without collapse, buckling, or inleakage of water. The conservatism
lies in the test criteria of no collapse, buckling, or inleakage as
compared to the ``no rupture'' criteria found in Safety Series No. 6
and TS-R-1. The draft advisory document for TS-R-1 (TS-G-1.1,
paragraphs 657.1 to 657.7) recognizes that leakage into the package and
subsequent leakage from the package are possible while still meeting
the IAEA requirement.
The Safety Series No. 6 test requirements were based on risk
assessment studies that considered the possibility of a ship carrying
packages of radioactive material sinking at various locations. The
studies found that, in most cases, there would be negligible harm to
the environment if a package were not recovered. However, should a
large irradiated fuel package (or packages) be lost on the continental
shelf, the studies indicated there could be some long-term exposure to
man through the food chain. The 200-meter (660-ft) depth specified in
Safety Series No. 6 is equivalent to a pressure of 2 MPa (290 psi), and
roughly corresponds to the continental shelf and to depths that the
studies indicated radiological impacts could be important. Also, 200
meters (660 ft) was a depth at which recovery of a package would be
possible, and salvage would be facilitated if the containment system
did not rupture. (Reference Safety Series No. 7, paragraphs E-550.1
through E-550.3.)
The expansion in scope of the deep immersion test was due to the
fact that radioactive materials, such as plutonium and high-level
radioactive wastes, are increasingly being transported by sea in large
quantities. The threshold defining a large quantity as a multiple of
A2 is considered to be a more appropriate criterion to cover
all radioactive materials and is based on a consideration of potential
radiation exposure resulting from an accident.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. One commenter stated that a 1-hour test is ``wholly
inadequate as a risk basis, given that as many as 100,000 shipments of
highly irradiated `spent' fuel are anticipated to being moved
transcontinentally on highways and railroads.'' The commenter added
that ``barge shipments should be prohibited outright.'' Finally, the
commenter recommended more stringent immersion testing for shipping
canisters.
Response. The NRC acknowledges the comment. However, the NRC
believes it is already moving towards more stringent standards with
this rule. The 1-hour test is sufficient to demonstrate structural
integrity and prevent inleakage. Most hydrostatic testing of components
are for durations much less than 1 hour. A test duration of 1 hour is
reflective of a practical requirement that will ensure the desired
package performance. While a longer duration test may appear to be more
reflective of the actual immersion times that might exist following an
accident, the duration of the test must be considered in conjunction
with the purpose of the test and the acceptance criteria specified for
successfully passing the test.
[[Page 3729]]
The purpose of the deep immersion test, as described in IAEA TS-G-
1.1, paragraphs 657.1 to 657.7, is to ensure package recoverability.
The acceptance criterion specified in TS-R-1 is that there be no
``rupture'' of the containment system. As described in the rule, NRC
believes that a more precisely defined acceptance criterion of no
``collapse, buckling, or inleakage of water'' is preferable. Type B
package designs that are capable of withstanding a 1-hour test without
``collapse, buckling, or inleakage of water'' are likely to be
sufficiently robust that a longer duration test would not produce
significantly greater structural damage.
Comment. One commenter suggested that the deep immersion test
should consider the possibility that the cask could already be damaged
or ruptured at the time of immersion. The commenter asked if there has
been an analysis of the dissemination of radionuclides at high
pressures for partially or completely ruptured casks. The commenter
stated that this issue is relevant due to the frequent transportation
of radioactive waste across the Great Lakes and between the U.S. and
other nations, such as Russia.
Response. The acceptance criterion for the deep immersion test is
no ``collapse, buckling, or inleakage of water.'' If a cask is already
damaged or ruptured at the time of immersion, then the immersion test
becomes a moot point because the acceptance criterion cannot be met.
Studies have been performed, including the IAEA-sponsored Coordinated
Research Project on ``Severity, probability and risk of accidents
during the maritime transport of radioactive material,'' that examined
the potential radiological consequences of such accidents. The report
of the Coordinated Research Project, IAEA-TECDOC-1231, is available
online at: http://www.iaea.org/ns/rasanet/programme/radiationsafety/transportsafety/Downloads/Files2001/t1231.pdf.
Comment. One commenter stated that if older, previously certified
packages can no longer be ``grandfathered,'' it will take significant
effort to show that these packages meet the deep immersion test and
will result in little safety benefit for the shipments.
Response. The commenter's connection between immersion testing and
grandfathering (see Issue 8) of existing certified packages is not
obvious. Under current NRC regulations (Sec. 71.61), a package for
irradiated nuclear fuel with activity greater than 37 PBq
(106 Ci) must meet the immersion test requirement. Under the
revised requirement, these same packages could be used for shipment of
irradiated nuclear fuel containing activity greater than 105
A2 and would not require additional immersion testing
(because the packages must already comply with the test requirement).
Comment. Three commenters expressed support for NRC's position on
this issue. One commenter stated that the proposed rule's deep
immersion test provisions would increase cask safety.
Response. No response is required.
Comment. One commenter urged the NRC to require more stringent
testing procedures for both old and new shipping containers (including
longer drops; greater crash impacts; longer and higher pressure water
submersion; leakage resistance; higher, longer, more intense fire
temperatures; and much greater explosive forces). Another commenter
requested that NRC change its standards so that casks damaged in
sequential tests would be required to survive immersion at depths
greater than those in the proposed rule.
Response. The NRC acknowledges this comment but believes that it
has adequate package testing requirements in the rule.
Comment. One commenter asked if containers that were not currently
certified to carry over one million curies would become authorized to
carry over one million curies under the proposed rule.
Response. If a package design is not currently certified to carry
over one million curies, its status will not be changed by this
rulemaking. Any restrictions on a package design imposed through the
NRC-issued CoC remain unaffected.
Comment. One commenter stated that the cost of compliance was
grossly underestimated, particularly for demonstrating cask integrity
at 200 meters.
Response. NRC staff appreciates the comment and fully understands
the importance of accurate cost data. As part of the proposed
rulemaking, the NRC specifically requested cost-benefit information on
this issue as well as a number of other issues. To the extent NRC
received data from public comments, these data were considered in
developing its final decision.
Comment. One commenter asked if the deep immersion test would apply
to all packages shipped across Lake Michigan.
Response. Under the proposed rule, the deep immersion test would be
applied to any Type B or C package that contains greater than
105 A2, regardless of the transport mode.
Therefore, the immersion test requirement would be applicable to all
shipments involving a package with an activity exceeding 105
A2, including any across Lake Michigan.
Comment. One commenter asked if the deep immersion test actually
requires a physical test. If the deep immersion test did not actually
require a physical test, the commenter asked NRC to clarify what it
means by ``test.'' The commenter also wanted NRC to clarify to what the
test specifically applies.
Response. As cited in the IAEA advisory document TS-G-1.1,
paragraph 730.2: ``The water immersion test may be satisfied by
immersion of the package, a pressure test of at least 2 MPa, a pressure
test on critical components combined with calculations, or by
calculations for the whole package.'' In answer to the commenter's
specific question, a physical test is not required, and calculational
techniques may be used. Regarding what the test specifically applies
to, ST-2, Section 730.3, states that: ``The entire package does not
have to be subjected to a pressure test. Critical components such as
the lid area may be subjected to an external gauge pressure of at least
2 MPa and the balance of the structure may be evaluated by
calculation.'' Thus, testing may be performed physically, by analysis,
or by a combination of the two.
Comment. One commenter stated that industry supports the NRC
position on deep immersion testing.
Response. The comment is acknowledged.
Comment. One commenter expressed concern that the deep immersion
test only requires that packages be submerged for 1 hour. The concern
is based on the belief that it is unlikely a package could be recovered
within an hour following a real accident.
Response. The 1-hour time limit only applies to the immersion test
and is the minimum time that the package shall be subjected to the test
conditions. It is not expected that a package could be recovered within
1 hour of an accident involving submergence of the package. In fact, in
the IAEA advisory document TS-G-1.1, paragraph 657.7 states:
``Degradation of the total containment system could occur with
prolonged immersion and the recommendations made in the above
paragraphs (657.1 through 657.6) should be considered as being
applicable, conservatively, for immersion periods of about 1 year,
during which recovery should readily be completed.''
Comment. One commenter asked NRC to clarify its assertion that the
immersion test is stricter than the IAEA's test because the NRC's
language
[[Page 3730]]
does not allow collapse, buckling, or any leakage of water.
Response. TS-R-1, paragraph 657, states, in part, that for a
package subjected to the enhanced water immersion test (NRC uses the
term deep immersion test), there would be no ``rupture of the
containment system.'' The term rupture is not a defined engineering
term in the IAEA literature related to TS-R-1. Further, the IAEA
advisory document TS-G-1.1, paragraph 730.3, states, in part, that some
degree of buckling or deformation is acceptable during the enhanced
water immersion test. Lacking specificity to the term rupture, the NRC
imposed specific, and it believes conservative, requirements that do
not allow collapse, buckling, or inleakage of water for a package
undergoing the deep immersion test.
Issue 8. Grandfathering Previously Approved Packages
Summary of NRC Final Rule. The final rule adopts the following
grandfathering provisions for previously approved packages in section
71.13:
(1) Packages approved under NRC standards that are compatible with
the provisions of the 1967 edition of Safety Series No. 6 may no longer
be fabricated, but may be used for a 4-year-period after adoption of a
final rule;
(2) Packages approved under NRC standards that are compatible with
the provisions of the 1973 or 1973 (as amended) editions of Safety
Series No. 6 may no longer be fabricated; however, may still be used;
(3) Packages approved under NRC standards that are compatible with
the provisions of the 1985 or 1985 (as amended 1990) editions of Safety
Series No. 6, and designated as ``-85'' in the identification number,
may not be fabricated after December 31, 2006, but may be continued to
be used; and
(4) Package designs approved under any pre-1996 IAEA standards
(i.e., packages with an ``-85'' or earlier identification number) may
be resubmitted to the NRC for review against the current standards. If
the package design described in the resubmitted application meets the
current standards, the NRC may issue a new CoC for that package design
with a ``-96'' designation.
Thus, the final rule adopts, in part, the provisions for
grandfathering contained in TS-R-1. The NRC believes that packages
previously approved under the 1967 edition of Safety Series No. 6 lack
the enhanced safety enrichments which have been incorporated in the
packages approved under the provisions of the 1973, 1973 (as amended),
1985 and 1985 (as amended) editions of Safety Series No. 6. For
example, later designs demonstrate a greater degree of leakage
resistance and are subject to quality assurance requirements in subpart
H of part 71. Furthermore, NRC believes that by discontinuing the use
of package designs that have been approved to Safety Series No. 6,
1967, for both domestic and international transport of radioactive
material, it will ensure safety during transportation and thus will
increase public confidence. However, NRC has not adopted the immediate
phase out of 1967-approved packages as the IAEA has, Instead, NRC
implemented a 4-year transition period for the grandfathering provision
on packages approved under the provisions of the 1967 edition of Safety
Series No. 6. This period provides industry the opportunity to phase
out old packages and phase in new ones, or demonstrate that current
requirements are met. NRC recognizes that when the regulations change
there is not necessarily an immediate need to discontinue use of
packages that were approved under previous revisions of the
regulations. The final rule includes provisions that would allow
previously-approved designs to be upgraded and to be evaluated to the
newer regulatory standards. Note that in 1996, IAEA first published
that the 1967-approved packages would be eliminated from use. Thus,
with the final rule 4-year phase out of these older packages, industry
will have had 12 years (i.e., until 2008) to evaluate its package
designs and prepare for the eventual phase out.
Affected Sections. Section 71.13.
Background. Historically, the IAEA, DOT, and NRC regulations have
included transitional arrangements or ``grandfathering'' provisions
whenever the regulations have undergone major revision. The purpose of
grandfathering is to minimize the costs and impacts of implementing
changes in the regulations on existing package designs and packagings.
Grandfathering typically includes provisions that allow: (1) Continued
use of existing package designs and packagings already fabricated,
although some additional requirements may be imposed; (2) completion of
packagings that are in the process of being fabricated or that may be
fabricated within a given time period after the regulatory change; and
(3) limited modifications to package designs and packagings without the
need to demonstrate full compliance with the revised regulations,
provided that the modifications do not significantly affect the safety
of the package.
Each transition from one edition of the IAEA regulations to another
(and the corresponding revisions of the NRC and DOT regulations) has
included grandfathering provisions. The 1985 and 1985 (as amended 1990)
editions of Safety Series No. 6 contained provisions applicable to
packages approved under the provisions of the 1967, 1973, and 1973 (as
amended) editions of Safety Series No. 6. TS-R-1 includes provisions
which apply to packages and special form radioactive material approved
under the provisions of the 1973, 1973 (as amended), 1985, and 1985 (as
amended 1990) editions of Safety Series No. 6.
TS-R-1 grandfathering provisions (see TS-R-1, paragraphs 816 and
817) are more restrictive than those previously in place in the 1985
and 1985 (as amended 1990) editions of Safety Series No. 6. The primary
impact of these two paragraphs is that packagings approved under the
1967 edition of Safety Series No. 6 are no longer grandfathered; i.e.,
cannot be used. The second impact is that fabrication of packagings
designed and approved under Safety Series No. 6 1985 (as amended 1990)
must be completed by a specified date. Regarding special form
radioactive material, TS-R-1 paragraph 818 does not include provisions
for special form radioactive material that was approved under the 1967
edition of Safety Series No. 6. Special form radioactive material that
was shown to meet the provisions of the 1973, 1973 (as amended), 1985,
and 1985 (as amended 1990) editions of Safety Series No. 6 may continue
to be used. However, special form radioactive material manufactured
after December 31, 2003, must meet the requirements of TS-R-1. Within
current NRC regulations, the provisions for approval of special form
radioactive material are already consistent with TS-R-1.
In TS-R-1, packages approved under Safety Series No. 6, 1973 and
1973 (as amended) can continue to be used through their design life,
provided the following conditions are satisfied: (1) Multilateral
approval is obtained for international shipment; (2) applicable TS-R-1
quality assurance (QA) requirements and A1 and A2 activity limits are
met; and (3) if applicable, the additional requirements for air
transport of fissile material are met. While existing packagings are
still authorized for use, no new packagings may be fabricated to this
design standard. Changes in the packaging design or content that
significantly affect safety require that the package meet current
requirements of TS-R-1.
[[Page 3731]]
TS-R-1 further states that those packages approved for use based on
the 1985 or 1985 (as amended 1990) editions of Safety Series No. 6 may
continue to be used with unilateral approval until December 31, 2003,
provided the following conditions are satisfied: (1) TS-R-1 QA
requirements and A1 and A2 activity limits are
met; and (2) if applicable, the additional requirements for air
transport of fissile material are met. After December 31, 2003, use of
these packages for foreign shipments may continue under the additional
requirement of multilateral approval. Changes in the packaging design
or content that significantly affect safety require that the package
meet current requirements of TS-R-1. Additionally, new fabrication of
this type of packaging must not be started after December 31, 2006.
After this date, subsequent package designs must meet TS-R-1 package
approval requirements.
Analysis of Public Comments on the Proposed Rule
The NRC notes that although there were a significant number of
comments reflecting opposition to the proposed grandfathering change to
the regulation, the majority of these comments were received from two
commenters representing the same company. The remaining comments
reflected opinions ranging from strong opposition to any grandfathering
of designs to full support for the proposed rule change. Accordingly,
following discussions with the DOT, NRC changed the transition period
from 3 years in the proposed rule to 4 years in the final rule. With
the effective date of this final rule being October 1, 2004, the
transition period is almost 5 years. A review of the specific comments
and the NRC staff's responses for this issue follows.
Comment. One commenter stated that the IAEA standards are consensus
based and that NRC must recognize they do not necessarily consider the
risk-informed, performance-based aspects of regulations that are
developed in the United States. The commenter added that NRC
regulations should also provide allowance for domestic-only
applications, which would include, for example, the grandfathering
provision. While the IAEA provisions must apply to international
shipments, for domestic-only shipments the grandfathering provision
would allow the continued use of existing packages manufactured to the
1967 standard, but prohibit the manufacture of any new packages.
Response. The NRC staff finding is to phase out those packages
approved to Safety Series No. 6, 1967 Edition, over a 4-year period
after October 1, 2004. The NRC believes this time period allows
industry adequate time to phase out old packages, phase in new ones, or
resubmit a package design for review against the current standards. NRC
considers it undesirable to be incompatible with IAEA with respect to
this provision. In eliminating the grandfathering of these older
designs, the IAEA concluded and NRC agrees that the continuance of
packages that could not be shown to meet updated standards was no
longer justified. As described, certain packages approved under the
1967 edition of the regulations may lack safety enhancements that later
designs have incorporated. The NRC acknowledges the comment about risk-
informed, performance-based regulations but notes that the
applicability of this change was not justified.
Comment. One commenter suggested that NRC require far more
stringent testing procedures for both old and new shipping containers
(longer drops; greater crash impacts; longer and higher pressure water
submersion; leakage resistance; higher, longer, more intense fire
temperatures; and much greater explosive forces). Another commenter
stated that ``packages and containers should be subject to upgraded
safety testing and more rigorous standards than have been required in
the past,'' especially after the events of September 11, 2001.
Response. The NRC acknowledges these comments and notes that the
commenters did not provide justification for the proposed changes.
Packages designed to regulations that are based on the 1973 and later
editions of Safety Series 6, in general, may include safety
enhancements, including designs, that demonstrate a greater degree of
leakage resistance. Major changes in the physical test parameters for
Type B packages are not being considered at this time, either by NRC or
the IAEA. NRC is confident that packages designed to meet the current
Type B standards provide a high degree of safety in transport, even
under severe transportation accidents.
Comment. One commenter objected to any grandfathering of casks. The
commenter stated that ``it will be a number of years before appreciable
amounts of `spent' fuel can be transported for more permanent
disposition'' and that this ``gives a substantial window of time for
design, development, and proof testing of new, better shipping casks.''
Response. The NRC and DOT have in place comprehensive regulations
that will support the safety of a large scale shipping campaign to a
central geologic repository should one ever be built. Such safety is
reliant upon the use of certified casks with robust design and
regulations that address training of staff dealing with shipments and
use of routes that minimize potential dose to the public. The safety
record of shipments of spent fuel both here and overseas has been
excellent. NRC regulations are compatible with IAEA regulations with
respect to grandfathering previously approved designs. These provisions
allow continued use of designs approved to earlier regulatory
standards; however, the provisions include certain restrictions with
respect to package modifications and fabrication. These provisions have
been adopted to allow a transition to newer regulations while
maintaining a high level of safety in transport. Packages that were
approved to the 1967 IAEA standards are being phased out because they
may not include safety enhancements of later designs.
Comment. One commenter stated that accurate data are not currently
available to forecast cost-benefit impacts. The commenter urged NRC to
work with those who hold Type B packages to determine whether they want
to maintain these packages. A second commenter stated that the costs of
requiring the replacement of 1967-specification packages are
substantial and that the benefits of requiring the replacements for
domestic use are zero. The commenter also stated that the NRC should
allow usage periods to be extended long enough to ensure that the
``money's worth'' has been obtained. The commenters added that NRC
should not propose changes when no harm or hazard has been
demonstrated.
Response. The NRC has made the decision to begin a 4-year phase out
of packages that have been approved to Safety Series No. 6, 1967.
However, NRC will allow package designs to be submitted for review
against the current requirements (TS-R-1). Based on this pathway, over
the 4-year period (after effective date of the final rule), industry
can determine which Type B packages they choose to submit for review to
the current requirements or have them phased out of use for shipping.
NRC has no current plans to contact individual design holders of
affected package designs to suggest an action on their part.
In evaluating the cost and benefits associated with the proposed
phasing out of the 1967-based packages, the NRC staff considered that
these designs may fall into one of the following five categories:
[[Page 3732]]
(1) Package designs that may meet current safety standards with no
modifications but have not been submitted for recertification. This
category includes package designs for which there is probably
sufficient supporting technical safety basis to support certification
under current requirements. For example, test data and engineering
analyses probably exist and are still relevant to the current safety
standards.
Costs associated with these package designs include the following:
(a) Development of an application ($10-$50K); and
(b) Review costs for NRC certification ($20K for 135 hours--
nonspent fuel amendment).
The total costs might be expected to be in the range of $30-$70K
per package design.
(2) Package designs that can be shown to meet current safety
standards with probably relatively minor design changes.
Costs associated with these package designs include the following:
(a) Design analysis and physical testing for modifications ($10K-
$100K);
(b) Development of revised package application ($10K-$50K--based on
approximately 200 staff hours of work);
(c) Review costs for NRC certification ($20K--based on 135 staff
hours for review of nonspent fuel amendment requests); and
(d) Packaging modifications to fleet of packagings (minor--$200 per
packaging, major--$5K per packaging).
The total cost would be expected to be in the range of $40K to
$170K depending on the modifications in the design or testing
information. This does not include the costs for making the physical
changes in the packagings, which could vary significantly for different
package types and different design modifications, in addition to the
number of packagings that needed to be modified.
For packages in Categories 1 and 2, NRC staff believe that the
expense of recertifying the design should be reasonable and is small
when considering the length of time these package designs have already
been in service (longer than 20 years). There is additional financial
incentive for upgrading these designs, because upgrading would allow
additional packagings to be fabricated and allow certificate holders to
request a wide range of modifications, both to the package design and
the authorized contents.
(3) Package designs that may meet current safety standards but are
impractical to recertify.
This category is intended to capture the special nature of spent
fuel casks that were certified to the 1967 IAEA standards. These
package designs may be considered separately for several reasons,
including:
(a) Domestic regulatory design standards for spent fuel casks
existed before standards for other package types;
(b) QA requirements were applied to this type of package, whereas
other package types were not subjected to the same level of QA either
for design or fabrication; and
(c) These packages normally have a limited specific use and are,
therefore, not present in large numbers in general commerce.
For packages in this category, NRC staff will be willing to review
an application under the exemption provisions of Sec. 71.8 that
requests an exemption to specific performance requirements for which
demonstration is not practical. The applicant would be free to propose,
for example, additional operational controls that would provide
equivalent safety. The exemption request could use risk information in
justifying the continued use of these existing packagings.
Costs associated with these package designs include the following:
(a) Development of application, including risk information ($150K);
and
(b) NRC review costs ($40,000--based on 270 staff hours for a
``non-standard'' spent fuel package amendment request).
(4) Package designs that cannot be shown to meet current safety
standards.
Costs associated with these package designs include the following:
(a) Development of new designs ($100-150K);
(b) Analysis and physical tests ($50K for prototype + 100K);
(c) Development of package application;
(e) NRC review costs ($40,000--based on 270 staff hours for review
of new designs for nonspent fuel); and
(f) Fabrication costs ($50K per package).
The cost information for development of new designs and the
analysis and testing of these newly designed packages (Category 4) were
provided to NRC by industry commenters during the public comment
period.
(5) Packages for which the safety performance of the package design
under the current safety standards is not known. This is due primarily
to a lack of documentation available regarding the package design and
performance.
NRC staff believes it is appropriate to phase out the use of
designs that fall into Categories 4 and 5. NRC staff believes that
there are package designers that may be willing and able to develop new
designs provided there is a financial incentive. With the continued use
of packages that cannot be shown to meet current standards, there will
be no financial incentive to upgrade designs. In addition, most
packagings certified to the 1967 design standards are more than 20
years old. Although proper maintenance of transportation packagings is
required, it is not clear that the service life of many types of
packagings would justify continued use.
The cost estimates associated with NRC review are based on
historical information gathered over years of performing technical
reviews of transportation package designs. There are many factors that
significantly influence the review time associated with performing
staff technical reviews for new package designs and amendments. Some of
the most important factors are: quality of the application, design
margins in the package, and a clear and unambiguous demonstration that
the regulatory acceptance criteria have been met. The costs previously
cited are not considered maximum or minimum but are representative and
conservative averages based on receipt of a complete and high-quality
package application.
The estimates of costs associated with development of designs,
testing, and preparation of application are extrapolated from
information provided by commenters to the proposed rule.
Comment. One commenter stated that packages that were manufactured
to the 1967 safety standard should be allowed to continue in domestic
service, unless a safety problem is identified. This commenter provided
monetized data to show how expensive our proposed position could be.
Response. In the final rule published September 28, 1995 (60 FR
50254), NRC wrote: ``NRC believes that the international package
standards should be used by the United States for both domestic and
international shipments, to the extent practicable. However, based on a
history of safe use under earlier safety standards, and the absence of
unfavorable operational data, NRC will allow the continued use of
existing packages in domestic transport until the end of their useful
lives. NRC will not allow, however, the continued fabrication of
packages to the old designs. This action permits use of existing
packages. It does not perpetuate package designs that can be discarded
or upgraded to satisfy the new standards.''
Further, in the April 30, 2002 (67 FR 21405), proposed rule, NRC
wrote ``The NRC recognizes that when the
[[Page 3733]]
regulations change there is not an immediate need to discontinue use of
packages that were approved under previous revisions of the
regulations. Part 71 has included provisions that would allow
previously-approved designs to be upgraded and to be evaluated to the
newer regulatory standards. NRC believes that packages approved under
the provisions of the 1967 edition of Safety Series No. 6, and which
have not been updated to later editions, may lack safety enhancements
which have been included in the packages approved under the provision
of the 1973, 1973 (as amended), 1985 and 1985 (as amended 1990)
editions of Safety Series No. 6. Therefore, the NRC believes that it is
appropriate to begin a phased discontinuance of these earlier packages
(1967-approved) to further improve transport safety.''
NRC adopted the 1985 IAEA standards on April 1, 1996 (60 FR 50248),
which allowed continued use of 1967 packages. In 1996, however, IAEA
published new regulations in TS-R-1 which discontinued grandfathering
these older designs. NRC agrees with IAEA's position that continuance
of these older designs is no longer justified. Therefore, to be
compatible with IAEA, NRC will begin a phased discontinuance of the
packages approved to Safety Series No. 6, 1967 after adoption of a
final rule.
The NRC has justified phasing out these designs based on the
following:
Safety standards have been upgraded three times since these designs
were initially evaluated and approved. In some cases, the documented
safety basis for these designs is substantially incomplete. Although
NRC knows of no imminent safety hazards posed by use of these packages,
it is judged to be prudent to be consistent with IAEA in phasing out
these designs. In addition, the performance of the package in a
transportation accident may not be known until a challenging accident
occurs.
Opportunity was provided to upgrade these designs to later
regulatory standards; however, applicants chose not to provide an
application to show that the designs met later safety standards. That
opportunity still exists and should be used by package owners that rely
on these packages for transporting their products.
Although there is a financial impact for phasing out these designs,
it is judged that there will also be a financial benefit to package
designers that choose to develop replacement packages that meet current
domestic and international safety standards.
Comment. One commenter stated that the proposed rule has no
discernible safety benefit to adopting TS-R-1 on this issue, there is
no direct economic information on the effect of implementing this
proposal, and NRC has requested cost-benefit information from the
regulated community.
Response. The NRC does not agree that there is no safety benefit in
adopting TS-R-1 provisions on grandfathering. The NRC believes that
packages approved to later safety standards (after 1967) may include
important safety enhancements. The grandfathering provision allows a 4-
year phase out period. Based on this pathway, over the impending 4-year
period (after effective date of the final rule), certificate holders
can determine which Type B packages they choose to have phased out or
reviewed to the current requirements. The commenter accurately notes
that NRC has solicited cost information regarding this proposal.
Comment. Three commenters stated that the proposed rule's effort to
phase out 1967-specification packages would negatively impact their own
business. One commenter argued that phasing out these packages would
have such a high cost that it would drive many small nuclear-shipping
businesses out of business with no ready successors. Another commenter
stated that phasing out these packages would cost about $20-$25 million
and could force some entities out of business, which could create an
unintended side-effect of orphaning over 1,000 radioactive sources of
considerable size. Another commenter discussed his business of
designing, manufacturing, servicing, shipping and disposing of devices
(principally calibrators and irradiators) that use Type B quantities of
Cobalt-60 or Cesium-137 sources, and the process of shipping
radioactive sources and how it relates to his business. The commenter
discussed the impact of phasing out 1967-specification packages. The
commenter argued that phasing out these packages for domestic shipments
would impose substantial economic, safety, and environmental costs
without any benefits.
Response. The NRC believes that packages approved under the
provisions of the 1967 edition of Safety Series No. 6, and which have
not been upgraded to later editions, may lack safety enhancements which
have been included in packages developed to later standards. NRC is
seeking to be compatible with the IAEA on the issue of grandfathering
and is not seeking to put shipping companies out of business.
Therefore, this final rule will phase out, 4 years after the rule
effective date, those packages that have been approved to Safety Series
No. 6, 1967. The NRC believes that many of the suggested orphaned
sources would qualify as Type A quantities and would not be negatively
impacted by the phase out of the 1967-approved packages.
Comment. One commenter opposed NRC's proposal on this issue because
it will have detrimental effects on his business. The commenter
explained that his company has 1,200 new packages built to the 1967
Safety Series No. 6 specifications that will be used in a contract that
runs through 2006. The company estimates that replacing these packages
would cost $5,000-$10,000 per package, which overall would devastate
the contract and be ruinous to the business. The commenter believes
that packages should be removed from service when they no longer meet
the safety requirements they were designed to meet or if a new safety
issue with the package is identified which would prevent the package
from meeting its intended safety function; neither of these conditions
have been identified for the package.
Response. With the adoption of the final rule, the opportunity
exists to have packages that were built to the 1967 Safety Series No. 6
specifications reevaluated to the current standards. Since August 1986,
fabrication of new packages to the old (1967) specifications has not
been authorized by NRC. The comment supports NRC's pre-1995 position
that, based on satisfactory performance, the 1967-type packages could
continue to be used. The new packages suggested in the comment are
assumed to have been fabricated in accordance with DOT regulations.
However, NRC's and DOT's current position, which is consistent with the
IAEA's on grandfathering, is to phase out the packages with these old
designs over a 4-year period. This time period will allow certificate
holders to determine which packages they will phase out or resubmit to
NRC for evaluation to the current standards. Industry needs to be aware
of changes or potential changes based on IAEA rules. Note in 1996, IAEA
first published that the 1967-approved packages would be eliminated,
and 5 years later (i.e., 2001) the international regulations were
implemented. Thus, with the 4-year phase out of the 1967-approved
packages, industry will have had 12 years (i.e., until 2008) to
evaluate their package designs, evaluate those designs that will not
meet the new standards, and prepare for the eventual phase out.
[[Page 3734]]
Comment. One commenter stated that eliminating 1967-specification
packages would cause severe harm. The commenter argued that many
businesses would have to requalify, relicense, and rebuild virtually
all of their current shipping containers at a very high cost. The
commenter noted that the RA did not take these costs into account. The
commenter argued that prohibiting the use of 1967-specification
packages would create thousands of orphan sources, creating a public
health risk, and that these sources could only be moved at very high
costs.
Response. The NRC notes that businesses may choose to requalify,
relicense, or rebuild their packages. Based on the long history
associated with grandfathering various packages, NRC believes that a 4-
year time period will allow certificate holders adequate opportunity to
make a responsible business decision as to which pathway to proceed--
phasing a package design out or resubmitting it for evaluation to the
current standards.
Comment. One commenter stated that certain containers excluded by
the proposed legislation couldn't be easily replaced because no
alternative packaging currently exists at comparable prices. The
commenter explained that designing, testing, and licensing a new
package is expensive (approximately $500,000) and usually takes over a
year to accomplish.
Response. The NRC acknowledges the comment about the cost and time
to design a new package. The staff notes that from the time TS-R-1
became effective to the date when NRC's grandfathering phase out
becomes effective will have been a significant and sufficient amount of
time for designers to learn about the new requirements, and to adopt
design and fabrication effort accordingly. As such new and conforming
packages would be available for use when needed by shippers.
Comment. One commenter stated that the RA lacks consideration of
costs to industry and health and safety benefits of the proposed
changes. The commenter believes that there were no arguments to be made
and that the only rationale would be harmonization with the IAEA, which
is not binding under U.S. law.
Response. The NRC disagrees that the only rationale for this
rulemaking is harmonization with the IAEA. NRC continues to believe
that harmonizing NRC's and DOT's regulations, when appropriate, will
prove beneficial to NRC, industry, and the general public. NRC believes
that packages approved to the 1967 standards lack safety enhancements
that were included in packages approved to later editions of Safety
Series No. 6 (i.e., 1973 and 1985).
Comment. One commenter stated that numerous participants in this
market sector are small entities within the meaning of the Regulatory
Flexibility Act and would be adversely affected by the proposed rule,
and neither agency's draft RA accounts for this fact.
Response. The NRC disagrees with this comment. The Commission
certified in Section XI of this notice that this rule will not have a
significant economic impact on a substantial number of small entities.
This rule affects NRC licensees, including operators of nuclear power
plants, who transport or deliver to a carrier for transport, relatively
large quantities of radioactive material in a single package. These
companies do not generally fall within the scope of the definition of
``small entities'' set forth in the Regulatory Flexibility Act or the
size standards adopted by the NRC (10 CFR 2.810).
Only one small entity commented on the proposed changes suggesting
that small entities would be negatively affected by the rule. Reviewing
records of licensed QA programs, NRC found that only 15 of the 127 NRC
licensed QA progams were small entities. Furthermore, of these 15
companies, NRC staff expects that only 2 or 3 would be negatively
affected by the final rule, given these companies' lines of business
and day-to-day operations. Based on this data, it is believed there
will not be significant economic impacts for a substantial number of
small entities.
Comment. One commenter asked how important this issue is to the
future success of small businesses that routinely transport Type B
quantities of radioactive materials domestically. The commenter found
it difficult to understand why some packages with proven safety records
would ``unjustly'' be phased out for domestic shipments in as little as
2 years after the proposed rule is issued.
Response. To be compatible with the IAEA on grandfathering, NRC has
made a decision to phase out those packages that may lack safety
enhancements found in other packages. This phase out will impact
packages approved to Safety Series No. 6, 1967, and will be completed 4
years after adoption of a final rule. This phase out is consistent with
NRC's belief that packages approved to the 1967 edition of Safety
Series No. 6 may lack safety enhancements that are included in packages
approved to later editions.
Comment. One commenter supported grandfathering casks made for the
1967 standards for domestic shipping and urged NRC to retain the
A2 value for molybdenum-99 and the A1 and
A2 values for californium-252, also for domestic shipping.
Response. NRC will retain the current A2 value for
molybdenum-99 (7.4E-1 TBq; 2.0E1 Ci) and the A2 value for
californium-252 (0.1 TBq; 2.7 Ci) (see Table A-1). The NRC is not
adopting the A1 value for californium-252 because the IAEA
is considering changing the value that appears in TS-R-1 back to what
presently appears in part 71. For reasons stated in the previous
response to comments, NRC will not allow grandfathering of packages
certified to the 1967 standard.
Comment. Because IAEA does not necessarily consider the risk-
informed, performance-based aspects of regulations that the NRC has
developed in the United States, a commenter suggested that the NRC
should consider the unique aspects of U.S.-only applications. The
commenter also suggested that the package identification number should
be revised to the appropriate identification number prefix together
with a suffix of ``-96'' provided that such packages shall be for
domestic use only and no additional packages be fabricated.
Response. The NRC does not agree with this suggestion because it
would allow continued use of B( ) packages for domestic use. NRC has
determined that only those packages that have enhanced safety features
(i.e., post-1967 package designs) will be allowed to be used and
manufactured beyond the 4-year phase-out period for all use (domestic
and international). When a package design designated as B( ) (i.e.,
approved to Safety Series No. 6, 1967) is submitted to NRC for review
to the current standards, the NRC may revise the package identification
number to designate the package design as a B, BF, B(U), B(M), etc, and
may assign the ``-96'' suffix to indicate that the design has met the
requirements of part 71. Those submitted package designs that do not
meet the current standard will not be assigned the ``-96'' suffix.
Comment. One commenter stated that adopting the revised
``grandfathering'' provision rule would have a significant impact on
the commenter's operations. The commenter highlighted how their
operational need to store fuel would cause unnecessary handling of
fuel, especially in light of design parameters to which their existing
containers must adhere. Replacement of certified containers with
satisfactory safety records is believed unnecessary by the commenter.
[[Page 3735]]
Furthermore, the commenter added that, if adopted, this proposal
would eliminate the flexibility to use M-130 containers on an ``as
needed'' basis. The commenter stated that these containers are safe and
asked that NRC consider allowing certified containers with satisfactory
safety records to continue to be ``grandfathered.''
Response. The NRC acknowledges the comment but notes that the
certificate holder could choose to request a recertification before use
beyond the 4-year phase-out period.
Comment. One commenter was concerned that, in departing from IAEA
grandfathering standards, NRC is placing the burden entirely on the
regulated industry to develop the justification for such a departure.
The commenter asserted that this is a problem because there was no
basis for having adopted the IAEA grandfathering standards in the first
place.
Response. In the interest of maintaining compatibility with the
IAEA regarding approved package designs to support the NRC's decision
to be consistent with IAEA on the grandfathering issue (i.e., phasing
out the Safety Series No. 6, 1967 package designs), and to allow only
those package designs with enhanced safety features to continue to be
used as viable packages, NRC will phase out the 1967-approved B( )
packages over a 4-year period after adoption of the final rule. Thus,
NRC does not agree with the comment ``departing from IAEA
grandfathering standards'' because NRC is making an effort to adopt the
IAEA grandfathering standards. The primary difference between the IAEA
and the NRC on this issue, however, is that IAEA has made an immediate
phase out of the 1967-approved packages, while NRC will phase out the
same packages over a 4-year period.
Comment. One commenter requested specific information on the types
and numbers of packages that would be affected and the timetable under
which packages would be excluded.
Response. The response to this comment is found at 67 FR 21406;
April 30, 2002. NRC does not require certificate holders or licensees
to submit information concerning the number of packages made to a
particular CoC.
Comment. One commenter stated that a regular 2-year reconsideration
of package design regulations will lead to a situation where package
designers and users will constantly be trying to keep up with ever-
changing regulations.
Response. NRC is aware of this concern and does not anticipate
major changes to the IAEA packaging standards every 2 years.
Additionally, NRC participates in the 2-year IAEA revision process and
will work with the IAEA and other member nations to assure that
proposed changes include appropriate justification with respect to cost
and safety.
Comment. One commenter disagreed with the proposed grandfathering
rule, stating that 1967-specification packages have operated
successfully for years and that there is no health or safety reason for
phasing them out. The commenter stated that extending the transition
period beyond 3 years would delay the negative economic impacts of
excluding these packages. The commenter did agree with the stricter
standards for new packages in the proposed legislation. The commenter
also agreed with the phase out of 1967-specification packages from
international sources.
Response. NRC agrees that the 1967-approved packages have appeared
to provide adequate performance in the past. However, these packages
lack the safety enhancements that other similar packages currently have
in place (i.e., post-1967 approved packages). Therefore, NRC believes
the time has come to phase out those package designs before a safety
issue occurs and to capitalize on those packages that have incorporated
the safety enhancements described in the proposed rule (67 FR 21406;
April 30, 2002). This phase out of the 1967 approved package designs is
consistent with the NRC's decision to be compatible with the IAEA on
the grandfathering issue.
Comment. One commenter expressed concern about the backfitting
issue and indicated that NRC should demonstrate that the basis for
IAEA's position is tenable in the U.S., or develop an independent
satisfactory basis for their position. The commenter stated that this
is particularly important with regard to grandfathering packages when
there may be different environments for international and domestic
shipments.
Response. The NRC does not support allowing the continued use of
the 1967-approved packages for domestic-use only. The NRC will continue
to phase out those package designs that currently meet Safety Series
No. 6, 1967, over a 4-year period after adoption of a final rule. This
approach is consistent with the NRC's desire to be compatible with the
IAEA on the grandfathering issue.
Comment. One commenter said that the proposed 3-year transition
period is too long.
Response. NRC has used the 3-year time line in previous rulemakings
and believes that this time period adequately supports those steps that
could be taken regarding grandfathering. However, NRC has worked with
the DOT and determined that a 4-year transition period would allow
certificate holders an additional year to determine the most effective
pathway for a particular design; namely, phase out old package designs,
phase in new package designs, or submit an existing package design for
review against the current standard.
Comment. One commenter was concerned that the proposed rule would
essentially remove from service any and all containers that could be
used to transport isotopes from DOE's Advanced Test Reactor for medical
or industrial use.
Response. As with other package designs approved to the 1967
standards, it is expected that certificate holders may request review
of these designs to the current regulatory standards.
Comment. Two commenters asserted that there is no safety benefit to
phasing out the 1967-specification packages. One of these commenters
noted that packages built to the 1967-specifications have an excellent
safety record and that NRC and DOT agree that the level of safety of
the 1967-specification is satisfactory. The commenter stated that the
phase out may be required for international shipping but not for
domestic shipping. The other commenter provided information on the high
cost of recertification and stated that these costs would likely drive
companies out of business.
Response. NRC is aware of the safety record of those packages
approved to Safety Series No. 6, 1967. However, NRC has made a decision
based on safety to be compatible with the IAEA on the issue of
grandfathering previously approved packages. Therefore, NRC will impose
a 4-year phase out of those package designs approved to the 1967
standards. While the IAEA has immediately terminated the use of 1967-
approved packages, the NRC has elected to terminate their use over a 4-
year period after adoption of a final rule. Any package design impacted
by the phase out may be submitted to NRC for review against the current
standards. While this review may be costly, it ensures package safety
during transport and is compatible with the IAEA.
Comment. One commenter asserted that the 1967-specification
packages may be impossible to replace at any cost because these devices
lack the ``QA Paper'' required under the NRC's regulations at 10 CFR
part 71. The commenter stated that these packages serve unique
functions and that phasing them out would leave thousands of Type B
sources stranded, and the cost of moving them would be prohibitive. The
commenter raised concerns about
[[Page 3736]]
exposure to these immovable packages and terrorism threats.
Response. NRC is aware that packages built to the 1967 standards
were not subject to QA requirements and that fabrication documents may
not be available. This is one reason why the NRC decided to incorporate
new standards in NRC regulations and discontinue use of the packages
certified to the 1967 standards.
Comment. One commenter said that currently approved DOT
specification packages should continue to be approved for domestic
shipments. The commenter based this suggestion on the fact that
packages that are currently accepted for use and proven to be safe
should continue to be used until they reach the end of their useful
life. The commenter did not believe that the costs that would be
associated with phasing out safely used transportation packages could
be justified on the basis of harmonization of regulations with TS-R-1.
Response. NRC has made a decision based on safety to phase out the
package designs that do not include the safety enhancements that other
packages currently maintain. Thus, the package designs that were
approved to Safety Series No. 6, 1967, will be phased out over a 4-year
period after adoption of the final rule. This approach is consistent
with the NRC decision to eliminate these types of packages for
transportation of radioactive materials. The safety enhancements for
post-1967 package designs can be found in the proposed rule (67 FR
21406; April 30, 2002).
Comment. One commenter urged the NRC to accept Competent Authority
Certificates for foreign-made Type B packages without requiring
revalidation by a U.S. Competent Authority. The commenter stated that
revalidation of foreign-made packages for which a country has issued a
Competent Authority Certificate other than the United States in
accordance with TS-R-1 is a redundancy that provides no additional
benefit.
Response. General license provisions in part 71 authorized use of
foreign-approved designs for import or export shipments provided that
DOT has revalidated the certificate. DOT may choose to request NRC
technical review of those designs. NRC experience has been that review
of those designs has been useful in identifying possible safety issues.
Comment. One commenter stated that there needs to be an effective
date applied to some or all of the proposed rule changes to grandfather
existing approved transport cask designs. Without that, all part 71 CoC
holders will be subject to backfit for compliance with no commensurate
safety benefit. The commenter urged NRC to perform a comprehensive
evaluation of what impact the proposed changes will have on existing
dual-purpose certificate holders if a grandfather clause is not
included in the rule.
Response. NRC is committed to working with DOT and the IAEA to
assure that future changes in package performance standards are limited
to those that are justified and are shown to be significant with
respect to safety.
Comment. One commenter urged NRC to provide a flexible CoC design
concept, which would permit internal packages whose dimensions and
weight fell within defined ranges (rather than being unique), to be
linked with one outerpack design of specific dimensions for shipment,
thus minimizing the number of separate CoCs to be obtained.
Response. Grandfathering provisions in Sec. 71.13 include certain
restrictions with respect to changes to previously approved designs.
However, for designs approved under the current regulations, a CoC can
be issued to show ranges for dimensions and weights at the request of a
certificate holder. The application for such a provision should include
an evaluation that shows that the ranges of weights and dimensions
would not negatively affect the performance of the package and its
ability to meet the requirements of part 71.
Comment. One commenter requested specification of the means by
which existing packages that were built before required compliance with
NRC QA standards can be qualified under the new regulations, without
requiring full, unobtainable ``QA Paper'' compliance.
Response. Packagings constructed to designs approved under the 1967
regulations were, in general, not subject to QA requirements in part
71. This was a consideration in NRC's decision to discontinue the use
of packages certified to the 1967 standards and to remain compatible
with IAEA on the grandfathering provisions. QA requirements in subpart
H of part 71 include provisions for existing packagings with respect to
QA.
Comment. One commenter suggested that NRC change the ``timely
renewal'' principle so as to enable holders of 1967-specification
packages that submit substantially complete applications for new or
requalified packages at least 1 year ahead of the ultimate phase-out
date to continue shipments past the phase-out deadline, pending NRC's
action on their request for certification or recertification.
Response. NRC does not agree with this comment or the suggested
approach. In 1996, IAEA rules indicated that package designs approved
to Safety Series No. 6, 1967, would be eliminated. The NRC is revising
its rules to maintain compatibility with these IAEA rules. Therefore,
the idea of phasing out these packages has been public knowledge for 7
years. IAEA rules regarding the elimination of the 1967-approved
packages were implemented in 2001 (5 years after being published). NRC
has posed a phase out of these package designs 4 years after adoption
of a final rule (i.e., in 2008). Thus, the overall timeframe already
encompasses 12 years, which is more than ample time to submit design
upgrades and have them approved by the NRC.
Comment. Two commenters expressed support for the proposed rule on
this issue. One commenter encouraged NRC to accept the IAEA
transitional requirements including the phase out of Type B
specification packages and the termination of authorization of Safety
Series 6 (1967) packages. The commenter said that these packages were
not designed and constructed according to standards where their
continued use would be consistent with the intent of the regulations.
Response. NRC acknowledges these comments. NRC will phase out the
packages designed to Safety Series No. 6, 1967, 4 years after adoption
of the final rule.
Comment. One commenter expressed support for NRC's proposal to
allow continued safe use of existing packaging through incorporation of
the TS-R-1 transitional arrangement provisions.
Response. NRC acknowledges this comment.
Comment. One commenter suggested that changes to A1 and
A2 exemption values were relevant to grandfathering
transport casks. The commenter believed that the NRC grandfathering
proposal could adversely impact currently certified casks by not
guaranteeing that casks certified under previous revisions ``will still
be usable without modification or analysis in the future.''
Response. The A1 and A2 values were last
changed in part 71 in 1995 (see 60 FR 50248; September 28, 1995) to
make the NRC regulations compatible with Safety Series No. 6, 1985.
With those changes and the adoption of new LSA definitions came the
awareness that a licensee, when using a CoC-controlled transport
container, had to apply the new A1 or A2 value
for a given radionuclide, determine the appropriate LSA limit, yet not
exceed the activity
[[Page 3737]]
limit for which the transport package was tested, and which was based
on the old (pre-September 28, 1995) A values. A very similar scenario
also exists regarding the new A1 and A2 values
and the existing transport containers. In other words, the new
A1 and A2 values would be used as the limits for
a shipment by a licensee, but the transport container's activity limit
would still be based on the pre-September 28, 1995, A values. Should a
package design be submitted for review to the current part 71, that
design would be subject to the current (i.e., TS-R-1) A1 and
A2 values that are part of this final rule. Thus, while NRC
is aware of the commenter's concern, industry has already had to
respond to a similar situation after April 1, 1996, when the September
28, 1995, final rule became effective.
Comment. One commenter expressed support for the phase out of the
1967-specification containers for international shipping to comply with
IAEA regulations. However, the commenter opposed the phase out for
domestic shipping, arguing that as long as these packages are
performing their function safely, then there is no benefit to the phase
out and extremely high economic costs. The commenter stated that there
would be huge environmental costs to the creation of hundreds or
thousands of new orphan sources. The commenter stated that there would
be large economic costs of these orphan sources because they will have
to be kept secure. The commenter noted that no facility in possession
of one of these devices will ever be able to terminate its license or
perform a close-out radiation survey, and sale or shutdown will be
impossible.
Response. The NRC has made a decision to phase out those package
designs that have been approved to Safety Series No. 6, 1967, for both
domestic and international transport of radioactive material. NRC
believes that package designs that include the safety enhancements (see
67 FR 21406; April 30, 2002) better suit the goals of the NRC and its
desire to ensure safe transport of all radioactive materials. NRC will
work closely with those licensees who may have sources that cannot be
easily transported as a direct result of this rule to provide a
suitable resolution. This could result in economic incentives for
package designers to develop new packages to retrieve orphan sources.
This could also result in the development and certification of a new
generation of Type B packages that could meet current safety standards
and fulfill that need for transport of certain radiation sources.
Comment. One commenter discussed the economic impacts of phasing
out 1967-specification packages on the entire nuclear waste-shipping
industry, estimating the total costs to the sector at over $1 billion.
The commenter argued that these estimates refuted the projection in
both NRC's and DOT's rulemaking notices, and the NRC's draft RA that
did not expect any significant costs to be associated with the
implementation of the rule. To arrive at this estimate, the commenter
predicted three possible outcomes and discussed these scenarios in the
comment letter. In two scenarios, the customers would have to design
and construct new containers and ship them at high costs. The commenter
discussed these costs in detail. In the third scenario, large amounts
of radioactive sources would be orphaned and would remain immovable
indefinitely.
Response. Based on the information provided by this commenter and
others regarding the costs of replacement packages, the NRC developed
an estimated cost of impacts, as previously described. The estimate is
based on either showing that the old designs meet current standards or
replacing older designs. The NRC does not have sufficient information
to substantiate the large costs estimated in this comment, partly
because NRC does not collect information regarding the number of
individual packagings fabricated to each design. However, based on
staff's knowledge, the following financial impacts specified in the
comment may not be reasonable:
1. The commenter claims that the cost of design, testing, and
licensing of new designs is estimated as $12 to $98 million. Based on
the assessment provided, even assuming that about half of the current
1967-based designs do not meet current safety standards and would need
to be phased out, the total costs to industry would not approach these
values. The derivation of these values cannot be substantiated by
information available to the NRC.
2. Cost of construction of new overpacks is stated as $7 to $13
million. These costs do not seem consistent with NRC knowledge of the
number of overpack designs currently in use.
3. Loss of existing overpacks and the loss of value of existing
devices are estimated from $500 to over $1,000 million. The derivation
of this value cannot be substantiated by information available to the
NRC.
Comment. One commenter stated that phasing out 1967-specification
containers would cause many nuclear-shipping firms to go out of
business, which would create thousands of orphan sources that are
unshippable and unmovable. The commenter stated that NRC would be
responsible for storing and securing these sources indefinitely and
protecting worker and public safety. The commenter noted that this
could create national security concerns with the potential for theft by
terrorists. The commenter stated that as long as these sources are
immovable, an entity could not conduct a final radiation survey and
terminate its license, forcing the entity to remain indefinitely on NRC
or Agreement State rolls.
Response. The commenter provided no justification for the opinion
that shipping firms would be forced to go out of business. The NRC
believes that if this situation occurs, package designers would be
motivated to develop new packages to retrieve orphan sources. This
could result in the development and certification of a new generation
of Type B packages (that would incorporate the current package
standards) that could fulfill that need.
Comment. One commenter stated that new containers would be
adequate, if they could be feasibly built. The commenter also stated
that the existing containers are adequate. The commenter stated that
orphan sources created by ``sunset'' on use of existing 1967-
specification containers decrease protection of public health and
safety protection.
Response. Regarding transport of radioactive material, NRC believes
that phasing out those package designs approved to Safety Series No. 6,
1967, will assure transport safety due to the fact that the package
designs will have enhanced safety features that the 1967-approved
packages lack. Furthermore, NRC is aware that packagings built to the
1967 standards were not subject to QA requirements, and that
fabrication documents may not be available. NRC does not agree that
this fact (lack of QA paperwork) enhances public confidence. Public
confidence may be increased by removal of such packages from use in
shipping. NRC will work closely with licensees who may have a source
that has been impacted by the elimination of its package to ensure
that, on a case-by-case basis, a suitable resolution is determined.
Comment. One commenter stated that orphan sources should be
considered in risk assessments and in assessing the costs and benefits
of the proposed ban on 1967-specification containers. The commenter
believes that when these factors are taken into consideration, they
argue overwhelmingly against the proposed change.
[[Page 3738]]
Response. The comment is acknowledged. The phase out of the Safety
Series No. 6, 1967, packages will occur 4 years after adoption of the
final rule. Thus, should orphan sources result as consequence of this
rule, industry will have a minimum of 4 years to establish a program
and a means to eliminate them from its inventory.
Comment. One commenter stated that any modification of current
requirements must not operate to prevent a device built to be
transported in DOT Specification 20WC containers, and which has
integral shielding and housing that is part of its ``packaging'' for
regulatory purposes, from being shippable merely because it was not
constructed fully under the part 71 QA rubric. The commenter warns that
the device would become, overnight, an ``orphan source.''
Response. Applicability of NRC QA requirements is specified in
subpart H of part 71, including provisions for fabrication of
packagings approved for use before January 1, 1979. Substantive
technical changes to the QA provisions in part 71 are not being made as
part of this rulemaking. Transport of packages that were built for the
DOT Specification 20WC overpacks would require that the package, which
includes the device within the overpack, be evaluated and certified to
the new regulations after the 4-year phase-out period.
Comment. One commenter stated that the U.S. is not bound to IAEA
requirements for domestic shipping. The commenter notes that NRC and
DOT have already deviated from the IAEA standards on other domestic-
only issues.
Response. NRC acknowledges these comments and adds that the NRC has
made a decision based on safety considerations not to deviate from the
IAEA on the grandfathering issue for packages. Thus, the NRC will move
forward to phase out those packages approved to Safety Series No. 6,
1967.
Comment. One commenter stated that both NRC and DOT have
misassessed the impact of their proposals on small entities protected
by the Regulatory Flexibility Act, 5 U.S.C. 601 et seq. The commenter
stated that NRC fails to consider the many small entities that would be
adversely impacted by phasing out the 1967-specification packages. The
commenter also disagreed with DOT's argument that international
uniformity will help small entities by the discarding of dual systems
of regulation. The commenter noted that in the U.S., unlike in Europe,
many firms do not have to deal with international shipping at all. The
commenter disagreed with DOT's argument that the proposed phase-in
period of 2 years would provide a smooth transition to the NRC approval
process. The commenter believes that the 2-year window was not
adequate.
Response. The NRC acknowledges these comments. This commenter was
the only small entity that made comments on this issue. Therefore, it
is not clear to the NRC that many small entities would be adversely
affected by this phase out. Further, NRC has made a decision based on
safety considerations not to deviate from the IAEA on the
grandfathering issue for packages. The NRC will move forward to phase
out those packages over a 4-year period after adoption of the final
rule. This time period should allow all businesses to assess their
particular packages and either have them phased out or resubmit them to
the NRC for review to the current standards. (The NRC staff notes that
DOT has also decided to adopt a 4-year transition period for DOT
specification packages.)
Comment. One commenter stated that there is no reason to compel
removal of properly inspected, properly maintained 1967-specification
packages from service for U.S. domestic shipments of special form Type
B quantities of radioactive material. The commenter argued that
requiring owners and users to inspect and maintain older packages, or
to convert to newer packages, would ensure safety. The commenter
concurred that it is reasonable to ban further construction of 1967-
specification packages.
Response. The packages approved to Safety Series No. 6, 1967, may
lack the safety enhancements possessed by post-1967 approved packages.
Thus, NRC will phase out these packages over a 4-year period including
production of new packages to these old standards. Alternatively,
owners and users of older packages have the opportunity to submit an
application showing that the design, or a modified design, meets the
current regulations. Recertification of these designs then would allow
continued fabrication of additional packagings.
Comment. One commenter stated that NRC and DOT should not subscribe
to the useful lifetime limitations for shipping packages implicit in
the IAEA's intended biennial review of its regulations. The commenter
stated that the cost of such forced obsolescence on an ongoing basis
would raise the cost of transportation unwarrantedly.
Response. NRC believes that those packages approved to Safety
Series No. 6, 1967, do not reflect the current safety standards. Thus,
these packages will be eliminated over a 4-year period after adoption
of a final rule. NRC does not anticipate that the future biennial
changes within IAEA standards will be as significant as the changes
found in the 1996 TS-R-1 standards. Therefore, based on the summary of
the impact that will occur on various packages (see 67 FR 21406; April
30, 2002), NRC will move forward with the elimination of certain
packages for radioactive material transport.
Comment. One commenter noted that there is a potential for
substantial delay in approving new designs or recertifying existing
designs. The commenter stated that any ``sunset'' deadline on the use
of any package design being phased out under this proposal should
permit its continued use pending an ultimate decision by the NRC on
either recertification of the existing design or approval of a new
design, as long as (1) a good-faith, substantially complete application
for approval or recertification, as the case may be, has been filed
with the NRC at least 12 months before the nominal ``sunset date'' on
use of the existing design; and (2) the application for approval or
certification is clearly related in the application to a design which
is subject to the ``sunset'' provision.
Response. The NRC has published guidance for applicants to use
regarding package approval. The purpose of the guidance is to document
practices used by NRC staff to review applications for package
approval. This guidance is available in NUREG-1609, ``Standard Review
Plan for Transportation Packages for Radioactive Material,'' and NUREG-
1617, ``Standard Review Plan for Transportation Packages for Spent
Nuclear Fuel.'' Using this guidance will assist applicants to prepare a
suitable application which will facilitate NRC review and ensure that
such a review is concluded in a timely fashion. Note that these NUREG
documents are available full-text on the NRC Web site (www.nrc.gov/NRC/NUREGS/indexnum.html). Regarding the ``sunset'' issue, note that
eliminating the 1967 packages was first published by IAEA in 1996
(i.e., 7 years ago) and that the international regulations were
implemented 5 years later in 2001. Industry should be aware of pending
changes or possible changes based on IAEA rules. Therefore, including
an additional 4-year implementation period (i.e., to 2008 (at least))
makes at least 12 years that industry has had the opportunity to
evaluate its package designs, identify designs that may not meet the
new standards, and prepare for the eventual phase out. The commenter is
essentially requesting another year of
[[Page 3739]]
use while the paperwork is in review. NRC does not agree with this
approach.
Comment. One commenter asserted that if a specific ``sunset'' date
is chosen, it should be significantly longer than the ones proposed by
either NRC or DOT to date. The commenter also requested that NRC and
DOT should agree on a common ``sunset'' date.
Response. The NRC and DOT have adopted a suitable transition date
for eliminating packages approved to Safety Series No. 6, 1967. Both
agencies believe that a 4-year phase-out period is adequate.
Comment. One commenter urged that the NRC allow for a substantially
longer transitional time than now proposed. The commenter argued that
the time necessary to design, fabricate, test, and complete NRC's
review of a new CoC design would be much greater than the 2-year
transition period proposed by DOT. The commenter stated that this would
cause a shipping hiatus.
Response. The NRC published the issues paper at 65 FR 44360; July
17, 2000, which indicated the position on the issues associated with
compatibility with the IAEA on many different issues, including
grandfathering of those packages approved to Safety Series No. 6, 1967
(see Issue 8). Thus, as a minimum, industry has been aware of the
overall proposed impact of phasing out the 1967-approved packages for
quite some time. Both NRC and DOT believe that a 4-year phase out
period provides adequate time for industry to phase out old packages,
phase in new packages, or demonstrate that current requirements are
met. The 4-year phase out will commence with the adoption of the final
rule.
Comment. One commenter supported grandfathering casks made for the
1967 standards for domestic shipping and urged NRC to retain the
A2 value for molybdenum-99 and the A1 and
A2 values for californium-252. The commenter also stated
that the package identification number should be revised to the
appropriate identification number prefix together with a suffix of ``-
96'' provided that such packages shall be for domestic use only and no
additional packages shall be fabricated.
Response. The NRC acknowledges the comments about grandfathering
and A1 and A2 values for domestic shipping. For
the comment about the package identification number, the NRC does not
agree with this comment (see earlier response and response below).
Comment. One commenter stated that the unique 1967-packages that
cannot be easily replaced should not be replaced. The commenter
supported the general concept of phasing out older packages and agreed
that use of most 1967-certified packages should be discontinued. The
commenter discussed the high costs of requalifying packages as ruinous
for some businesses. The commenter argued that this would result in
many orphan sources.
Response. The NRC will move forward to phase out the Safety Series
No. 6, 1967, packages that may not have the built-in safety
enhancements that other (post-1967) packages maintain. The NRC will
work in the future on a case-by-case basis with licensees who may have
orphaned sources in their inventory as a result of this final rule.
Comment. One commenter stated that if packages can be shown to meet
the proposed regulations, the package identification number should be
revised to the appropriate identification number prefix together with a
suffix of ``-96'' provided that such packages shall be for domestic use
only and no additional packages be fabricated.
Response. The NRC staff disagrees with this comment. Inasmuch as
this would allow continued use of B( ) packages for domestic use, NRC
has determined that only those packages that have enhanced safety
features (i.e., post-1967 package designs) will be allowed to be used
and manufactured beyond the 4-year phase-out period for all use
(domestic and international). When a package design is designated as B(
) (i.e., approved to Safety Series No. 6, 1967) and is submitted to NRC
for review to the current standards, the NRC may revise the package
identification number to designate the package design as B, B(U), B(M),
etc, and may assign the ``-96'' suffix.
Issue 9. Changes to Various Definitions
Summary of NRC Final Rule. The final rule adopts the TS-R-1
definition of Criticality Safety Index (CSI). NRC believes this
provides internal consistency and compatibility with TS-R-1.
Additionally, the following definitions have been revised to improve
their clarity and maintain consistency with DOT: A1,
A2, Consignment, LSA-I, LSA-II, LSA-III, and Unirradiated
uranium. NRC believes that terms must be clearly defined so that they
can be used to accurately communicate requirements to licensees. By
modifying existing definitions and adding new definitions, the licensee
would benefit through more effective understanding of the requirements
of part 71.
Affected Sections. Section 71.4.
Background. The changes implemented by NRC in this rulemaking
require changes to various definitions in Sec. 71.4 to provide internal
consistency and compatibility with TS-R-1. These terms must be clearly
defined so that they can be used to accurately communicate requirements
to licensees. By modifying existing definitions and adding new
definitions, the licensee benefits from a more effective understanding
of the requirements of part 71.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. Four commenters generally supported the proposal. One
commenter specifically asked that NRC and DOT agree on the definition
of ``common terms'' before issuance of the final rules.
Response. The DOT and the NRC continue to coordinate rulemaking
efforts to ensure regulatory consistency.
Comment. One commenter stated that `` `Radioactive materials' and
`contamination' should not be redefined as presented in the draft rule;
the new definitions would expand exemptions and the deregulation and
recycling of more nuclear materials and wastes.'' Another commenter
expressed concern over the omission of a definition for
``contamination.'' See response to comment on non-fixed contamination
below.
Response. The comments appear to be addressing a DOT concern, as
NRC has not proposed to adopt a definition for ``contamination'' in
this rulemaking. Currently, NRC regulations in Sec. 71.87(i) refer to
the contamination levels found in DOT regulations. The NRC notes that
contamination levels/concerns are not criteria for packaging approval
within part 71. Rather, they are a factor in safe transport of an
actual package of radioactive material.
Comment. One commenter stated that the definition of ``person'' as
stated in Sec. 70.4 should be included under Sec. 71.4 so it is clear
that entities such as DOE are not a person under proposed Sec. 71.0(e).
Response. The NRC does not agree with this comment. ``Person'' is
defined within each part of Title 10. It is only these entities who
would make shipments of radioactive material under part 71. Therefore,
the NRC will rely on the existing definitions to support the
transportation activities found in part 71.
Comment. Three commenters stated that the definition of LSA-I and
LSA-II should agree with the proposed DOT definition. One commenter
provided specific information in objection to the proposed definitions
of LSA-I and LSA-II.
[[Page 3740]]
Response. NRC agrees that the definitions for LSA-I and LSA-II
should be consistent between the NRC and DOT regulations. Therefore,
NRC modified its regulations appropriately in Sec. 71.4 and changed the
definitions for LSA-I and LSA-II to agree with the definitions found in
DOT's final rule. Additionally, NRC noted that DOT adopted the TS-R-1
definition for LSA-III material. To maintain consistency between these
regulations, NRC also adopted DOT's definition for LSA-III.
Comment. One commenter stated that defining only the containment
system is broad enough to include the confinement system, because
defining them differently will be confusing.
Response. NRC acknowledges the comment.
Comment. Three commenters were concerned about the omission of a
definition for ``consignment.'' One commenter suggested that NRC use
the definition provided in the DOT proposed rule.
Response. NRC is adding a definition for ``consignment'' in Sec.
71.4 that is consistent with DOT.
Comment. Two commenters were concerned about the omission of a
definition for ``unirradiated uranium.''
Response. NRC is adding a definition for ``unirradiated uranium''
to Sec. 71.4 that is consistent with DOT.
Comment. Two commenters stressed the importance of including the
definition of ``non-fixed contamination.''
Response. NRC disagrees. Section 71.87(i) refers to the nonfixed
(removable) contamination regarding the contamination levels found in
DOT regulations in 49 CFR 173.443, Table 11. NRC notes that the
definition of ``nonfixed contamination'' has been removed from Sec.
173.403 in DOT's rule. Furthermore, the definition of contamination
from TS-R-1, including the definitions for fixed and nonfixed
contamination, have also been added to Sec. 173.403 in DOT's proposed
rule. Contamination controls are not a function of NRC package approval
as much as they are a factor in safe transport of a package. Thus, it
is appropriate to define contamination in DOT's regulations, but not in
the NRC's.
Comment. One commenter supported the proposed adoption of the
specified definitions, and also urged NRC to adopt the TS-R-1
definitions for confinement system, consignment, contamination, fixed
contamination, nonfixed contamination, shipment, and transport index.
The commenter also stated that NRC defined LSA-I differently from DOT,
and that NRC and DOT should ensure compatibility between the rules.
Response. See response to the previous comments in this issue. NRC
agrees that the definition of ``transport index (TI)'' should be
consistent between NRC and DOT regulations. Therefore, NRC modified
Sec. 71.4 to include a definition for TI that is consistent with DOT.
NRC does not agree, however, with the comment to adopt the TS-R-1
definition of TI, as the definition adopted provides more clarity and
explanation for the applicability of the TI.
Issue 10. Crush Test for Fissile Material Package Design
Summary of NRC Final Rule. The final rule adopts, in Sec. 71.73,
the TS-R-1 requirement for a crush test for fissile material package
designs and eliminated the 1000 A2 criterion, but maintained
the current part 71 testing sequence and drop and crush test
requirements.
By adopting TS-R-1, the weight and density criteria will apply to
fissile uranium material packages, and packages that were previously
exempted because of the 1000 A2 criterion will now require
crush testing. Adopting crush test requirements and eliminating the
1000 A2 criterion is appropriate because not adopting the
TS-R-1 requirements would result in an inconsistency between part 71
requirements and TS-R-1, which could affect international shipments,
and fissile material package designs would continue to not be evaluated
for criticality safety against a potential crush test accident
condition.
The NRC did not adopt the TS-R-1 test sequence requirements because
no new information existed to address concerns from a previous
rulemaking regarding the difference in test requirements between
essentially the same IAEA requirements contained in Safety Series No. 6
and part 71. The NRC chose to remain more conservative than the IAEA by
requiring both a drop and crush test, rather than one or the other as
TS-R-1 would permit.
Affected Sections. Section 71.73.
Background. The crush test requirements in TS-R-1 were broadened to
apply to fissile material package designs (regardless of package
activity). Previously, IAEA Safety Series No. 6 and part 71 required
the crush test for certain Type B packages. This broadened application
was created in recognition that the crush environment was a potential
accident force that should be protected against for both radiological
safety purposes (packages containing more than 1000 A2 in
normal form) and criticality safety purposes (fissile material package
design).
Under requirements for packages containing fissile material, TS-R-
1, paragraph 682(b), requires tests specified in paragraphs 719-724
followed by whichever of the following is the more limiting:
(1) The drop test onto a bar as specified in paragraph 727(b) and
either the crush test as indicated in paragraph 727(c) for packages
having a mass not greater than 500 kg (1100 lbs) and an overall density
not greater than 1000 kg/m3 (62.4 lbs/ft3) based
on external dimensions, or the 9-meter (30-ft) drop test as defined in
paragraph 727(a) for all other packages; or
(2) The water immersion test as specified in paragraph 729.
Both Safety Series No. 6, paragraph 548, and current Sec. 71.73
require the crush test for packages having a mass not greater than 500
kg (1100 lbs), an overall density not greater than 1000 kg/
m3 (62.4 lbs/ft3) based on external dimensions,
and radioactive contents greater than 1000 A2 not as special
form radioactive material. Under TS-R-1, the criterion for radioactive
contents greater than 1000 A2 was eliminated for packages
containing fissile material. The 1000 A2 criterion still
applies to Type B packages and is also applied to the IAEA newly
created Type C package category.
Full compliance with TS-R-1 requirements for fissile material would
require changes to the hypothetical accident conditions test sequencing
of Sec. 71.73 and would require performance of the 9-meter (30-ft) free
drop test or the crush test, but not both, as presently required by
Sec. 71.73. The TS-R-1 test requirements are essentially the same as
those contained in Safety Series No. 6 (1985 edition). NRC addressed
the difference between Safety Series No. 6 and Sec. 71.73 in a previous
rulemaking and concluded that the two tests evaluate different features
of a package, and both tests are necessary to determine whether a
package response is within applicable limits (final rule, 60 FR 50248;
Sept. 28, 1995).
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. One commenter stated that the additional cost of the crush
test for fissile material is estimated at about $5,000,000. This cost
is to design, certify, and manufacture replacement packages currently
in use for the shipment of uranium oxide. The commenter thought that
currently three
[[Page 3741]]
to five packages are in use that will need to be modified and
recertified.
Response. The information provided by the commenter was considered
in the development of NRC's rule.
Comment. One commenter recounted how they were almost crushed under
``a boulder the width of the highway in the Wyoming Wind River Range
some years ago'' and stated that ``No vehicle or container could have
withstood the impact of that boulder's fall from several hundred feet
above.'' The commenter also stated that based on such probable events,
crush tests must be mandatory, with the cost borne by licensee or user.
The commenter added that the NRC needs to implement more rigorous crush
and drop tests than its current standard so that it can ensure
container survival in the event of severe accidents. The commenter also
recommended that because the TS-R-1 document was not readily available,
it was ``ingenuous, at best, for the NRC to give the references to the
actual testing requirements in terms of TS-R-1 paragraph citations.''
Response. The recommendation to implement more rigorous crush and
drop tests than the current regulatory standards to ensure container
survival for severe accidents is noted, but was not justified, and is
outside the scope of the current rulemaking. Further, it should be
noted that TS-R-1 is readily available online at: http://www.pub.iaea.org/MTCD/publications/pdf/Pub1098_scr.pdf.
Comment: Three commenters advocated more stringent testing
procedures. Specifically, one commenter stated support for NRC's effort
to adopt crush tests for all fissile material packages regardless of
size or activity (while rejecting the IAEA's option of choosing to
perform either a drop or a crush test on a container). The commenter
also urged the NRC to use a physical (as opposed to a simulating test
using computer modeling) crush test with a full-size package to provide
a realistic testing environment. The commenter suggested that the NRC's
proposal should include all containers, including the DT-22 (which
failed the dynamic crush test) and the 9975 container (which failed the
30-foot drop test). Further, it was noted that the redesigned 9975
container has not yet been ``crush tested to show the results of high-
speed impact against an unyielding surface.'' For this unit, the
commenter urged NRC to require a physical, as opposed to a simulated,
crush test with a full-size package to provide a realistic testing
environment. The commenter also stated that the NRC needs to require
other testing and noted that ``neither the DT-22 nor the 9975 have been
sufficiently tested against fire.'' Also, the commenter contended that
the current test (i.e., burn at 1475 degrees Fahrenheit for 30 minutes)
ignores the fact of ``more than 20 materials routinely transported on
highways that burn at more than twice this temperature.'' Two
commenters suggested that this heat test be made more stringent and
realistic. NRC also needs to test these two containers for ``durability
to terrorist attack with a variety of weapons, such as mortars or anti-
tank missiles, under a variety of conditions.'' Furthermore, ``all Type
B containers should be subject to rigorous testing for terrorist
resistance.''
Another commenter expressed concern that the proposed rule would
allow the DP-22 package to be licensed and approved, despite the fact
that it does not meet either the drop or crush test requirements.
Another commenter expressed concern that crush testing is not
required for packages having a mass greater than 500kg, which includes
rail SNF waste packages. The commenter suggested that the NRC ``require
rail transportation casks be subject to crush testing (scaled up to
produce impact energies of the magnitude expected in a railway
accident).'' The commenter cited a 1995 report entitled ``Rail
Transportation of Spent Nuclear Fuel--A Risk Review'' that argued small
packages are shipped in large numbers and ``as a result demonstrate a
higher possibility of experiencing crush loads than large packages
would.'' In addition, the commenter cited how packages transported by
North American rail would have a high probability of experiencing
dynamic crushing in an accident.
Response. The comment regarding more rigorous testing for all Type
B packages for terrorist resistance is noted. Please refer to the
second comment in Section II, under the heading: Terrorism Concerns.
The comment regarding stringency of heat tests is noted but is outside
the scope of the current rulemaking. With respect to comments regarding
the DT-22 and 9975 container, NRC staff is not familiar with these
designs as they are used within the DOE program and are authorized
under DOE's package approval authority. These containers do not
currently have an NRC CoC. The NRC staff also is not familiar with the
DP-22 design that the commenter alludes to as it does not currently
have an NRC CoC. To receive an NRC CoC, it would have to meet the NRC's
testing requirements, including drop and crush test if required.
The comment regarding crush testing for packages greater than 500
kg (1100 lb) is acknowledged. The NRC has already gone beyond the IAEA
testing requirements in requiring that all Type B packages subject to
the crush test must also be subjected to the free drop test. Extending
the crush test to other Type B packages (i.e., those exceeding 500 kg
(1100 lbs)) is beyond the scope of the current rulemaking.
Regarding the comment on requiring physical crush testing, rather
than simulated tests, and the use of full scale packages for physical
testing, the NRC staff believes that the use of computer code analysis
of finite element models and the use of scale models for physical
testing are valid methods for demonstrating compliance with the NRC's
package testing requirements. It should be noted that these methods
should be NRC approved.
Comment. Three commenters questioned the requirements for both a
drop test and a crush test. One commenter requested that if both a
crush test and a drop test are required on packages that meet the
requirements for the crush test, the rules should specify that this
could be carried out on two different packages. The commenter explained
that it does not make sense to require both tests for the same package,
because in an accident scenario, a single package would not experience
both conditions.
Two commenters stated that packages should either pass a drop test
or the crush test, but not both. The first commenter said that the rule
should state that separate packages should be used for each test, and
that the same package should not be used to pass both tests in
sequence. The second commenter said that, ``A line for deciding which
test a package should undergo could be based on the gross weight of the
package.''
Response. The current requirements under Sec. 71.73(a) state that:
``Evaluation for hypothetical accident conditions is to be based on
sequential application of the tests specified in this section, in the
order indicated, to determine their cumulative effect on a package or
array of packages.'' However, Sec. 71.73(a) does specifically allow for
an undamaged specimen to be used for the immersion test of Sec.
71.73(c)(6). NRC staff is aware that IAEA regulations do not require
both the free drop and crush test on a single specimen, but has chosen
to remain more conservative in this regard. In the NRC rulemaking for
compatibility with IAEA Safety Series No. 6 (September 28, 1995; 60 FR
50248), NRC staff stated the position that: ``NRC is requiring both the
crush test and drop
[[Page 3742]]
test for lightweight packages to ensure that the package response to
both crush test and drop forces is within applicable limits.'' NRC
staff is not aware of any new information that would cause NRC to
deviate from that position.
NRC staff does not agree with the commenter's assertion that
performing a drop and crush test is a double drop test. In the drop
test from 9 meters (30 feet), the specimen itself is dropped onto an
unyielding surface; in the crush test (if required by both the package
weight and density criteria), a 500-kg (1100-lb) weight is dropped from
9 meters (30 feet) onto the specimen. These are two independent tests
that may have different outcomes depending on the package and the
location where maximum damage is expected to occur for each test.
Comment. Two commenters supported NRC's proposal regarding crush
test requirements. One commenter expressed support for the NRC's
proposal to accept the part of IAEA's rule change under TS-R-1 which
requires a crush test for fissile material packages regardless of size
or activity while rejecting the IAEA's option of performing either
crush or drop tests of containers.
Response. No response is necessary.
Issue 11. Fissile Material Package Design for Transport by Aircraft
Summary of NRC Final Rule. The final rule adopts TS-R-1, paragraph
680, Criticality evaluation, in a new Sec. 71.55(f) that only applies
to fissile material package designs that are intended to be transported
aboard aircraft. Section 71.55 specifies the general package
requirements for fissile materials, and the existing paragraphs of Sec.
71.55 are unchanged. Among other requirements, TS-R-1, paragraph 680,
requires that packages must remain subcritical when subjected to the
tests for Type C packages, because:
(1) The NRC has deferred adoption of the Type C packaging tests
(see Issue 6);
(2) TS-R-1, paragraph 680 requires Type C tests; and
(3) Paragraph 680 applies to more than Type C packages; only the
salient text of paragraph 680 was inserted into Sec. 71.55(f) and
applies to domestic shipments.
Adopting this change will provide regulatory consistency. Shippers
would have been required to meet the TS-R-1 air transport requirements
even if the NRC did not adopt them, because the International Civil
Aviation Organization had adopted regulations consistent with TS-R-1 on
July 1, 2001. U.S. domestic air carriers require compliance with the
ICAO regulations even for domestic shipments. Therefore, these changes
are expected to benefit industry by eliminating the need for two
different package designs.
Affected Sections. Section 71.55.
Background. TS-R-1 introduced new requirements for fissile material
package designs that are intended to be transported aboard aircraft.
TS-R-1 requires that shipped-by-air fissile material packages with
quantities greater than excepted amounts (which would include all NRC-
certified fissile packages) be subjected to an additional criticality
evaluation.
In TS-R-1, paragraph 680, requirements for packages to be
transported by air are in addition to the normal condition and accident
tests that the package must already meet. Thus:
Type A fissile package by air must:
(1) Withstand normal conditions of transport with respect to
release, shielding, and maintaining subcriticality (single package and
5xN array; \1\
---------------------------------------------------------------------------
\1\ N represents the maximum number of fissile material packages
that can be shipped on a single conveyance.
---------------------------------------------------------------------------
(2) Withstand accident condition tests with respect to maintaining
subcriticality single package and 2xN array); and
(3) Comply with TS-R-1, paragraph 680, with respect to maintaining
subcriticality (single package);
Type B fissile package by air must:
(1) Withstand normal conditions of transport and Type B tests with
respect to release, shielding, and maintaining subcriticality (single
package and 5xN array/normal and 2xN array/accident); and
(2) Comply with TS-R-1, paragraph 680, with respect to maintaining
subcriticality.
TS-R-1, paragraphs 816 and 817, state that fissile package designs
intended to be transported by aircraft are not allowed to be
grandfathered. Consequently, all of these fissile package designs will
be evaluated before their use.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. Four commenters supported the NRC's position on this
issue. One commenter supported NRC's proposal to ensure consistent
review of package designs affected by the requirements of the
International Civil Aviation Organization. Another commenter said
adoption of Type C packages should be scheduled for future
harmonization with IAEA regulations.
Response. The NRC believes the changes create a uniform regulatory
framework for the review of package designs for both national and
international air shipments.
B. NRC-Initiated Issues
Issue 12. Special Package Authorizations
Summary of NRC Final Rule. The final rule adopts, in Sec. 71.41,
special package authorizations that will apply only in limited
circumstances and only to one-time shipments of large components.
Special package authorization regulations are necessary because there
are no regulatory provisions in part 71 for dealing with nonstandard
packages, other than the exemption provisions and Sec. 71.41(c). The
NRC processing of one-time exemptions for nonstandard packages, such as
the Trojan reactor vessel, has required the expenditure of considerable
NRC resources. Further, the NRC's policy is to avoid the use of
exemptions for recurring licensing actions. Special package
authorization requirements will result in enhanced regulatory
efficiency by standardizing the requirements to provide greater
regulatory certainty and clarity, and will ensure consistent treatment
among licensees requesting authorization for shipment of special
packages.
Any special package authorization will be issued on a case-by-case
basis, and requires the applicant to demonstrate that the proposed
shipment would not endanger life or property nor the common defense and
security, following the basic process used by applicants to obtain a
CoC for nonspecial packages from NRC.
The applicant will be required to provide reasonable assurance that
the special package, considering operational procedures and
administrative controls employed during the shipment, would not
encounter conditions beyond those for which it had been analyzed and
demonstrated to provide protection. The NRC will review applications
for special package authorizations. Approval will be based on NRC staff
determination that the applicant will meet the requirements of subpart
D of 10 CFR part 71. If approved, the NRC will issue a CoC or other
approval (i.e., special package authorization letter).
NRC will consult with DOT on making the determinations required to
issue an NRC special package authorization.
Affected Sections. Section 71.41.
Background. The basic concept for radioactive material
transportation is that radioactive contents are placed in
[[Page 3743]]
an authorized container, or packaging, and then shipped. The packaging,
together with its contents, is called the package. In general, the
transportation regulations in TS-R-1, 10 CFR part 71, and 49 CFR are
based on the shipment of radioactive contents in a separate, authorized
packaging. There are a few exceptions. In cases involving larger
quantities of radioactive material, the content to be shipped may
itself be a container. A storage tank containing a radioactive residue
is an example. It is not necessary for the shipper to place the tank
within an authorized packaging if the shipper demonstrates that the
tank satisfies the requirements for the packaging. DOT and NRC have
jointly provided guidance on such shipments (see ``Categorizing and
Transporting Low Specific Activity Materials and Surface Contaminated
Objects,'' NUREG-1608, RAMREG-003, July 1998).
As older nuclear facilities are decommissioned, DOT and NRC are
being asked to approve the shipment of large components, including
reactor vessels and steam generators. These components may contain
significant quantities of radioactive material, but they are so large
that it may not be practical to fabricate authorized packagings for
them. Because the potential shipment of these components was not
contemplated when the NRC transportation regulations were developed,
the regulations do not specifically address them.
Large components can be shipped under DOT regulations if the
components meet the definition of Surface Contaminated Object (SCO) or
Low Specific Activity (LSA) material (see 49 CFR 173.403 for SCO and
LSA definitions). For example, steam generators that meet the DOT SCO
definition are exempt from part 71 and are shipped under 49 CFR,
following guidance provided in NRC Generic Letter 96-07 dated December
5, 1996. This method has been applied to several shipments of steam
generators and small reactor vessels to the low level waste disposal
facility at Barnwell, SC. NRC and DOT intend to continue employing this
approach and method for steam generators and similar components that
can be shipped under DOT regulations.
Large components that exceed the SCO and LSA definitions are
subject to part 71. An example is the Trojan reactor vessel which was
transported to the disposal facility on the Hanford Nuclear Reservation
near Richland, Washington. The Trojan Reactor Pressure Vessel (TRPV)
contained approximately 74 PBq (2 million Ci) in the form of activated
metal and 5.7 TBq (155 Ci) in the form of internal surface
contamination, and was filled with low-density concrete, and weighed
approximately 900 metric tons (1,000 tons). Normally, large curie
contents are required to be shipped in a Type B packaging, but the TRPV
was too large and massive to be shipped within another packaging.
Section 71.8 provides that NRC may grant any exemption from the
requirements of the regulations in part 71 that it determines is
authorized by law and will not endanger life or property nor the common
defense and security.
Currently, no regulatory provisions exist in part 71 for dealing
with nonstandard packages, other than the exemption provisions and Sec.
71.41(c). The NRC's practice is to avoid the use of exemptions for
recurring licensing actions. The new rule language will support this
practice.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. One commenter stated that relaxation of requirements
applicable to large packages could potentially reduce the cost of these
shipments for parties who must routinely demonstrate that all
shipments, including reactor vessels and larger reactor compartments,
are made in compliance with part 71. However, the commenter asked that
the NRC relax the restriction that a special package authorization may
be approved only for ``one-time shipments'' and allow a limited number
of shipments to be approved if they are of the same design to avoid
repetitious certification requests.
Response. The NRC believes that standardizing the special package
authorization process will increase efficiency during the review of
large shipment components. These special packages were not provided for
specifically in earlier regulations. Establishing a standard process
for authorization also will reduce the regulatory burden associated
with shipping these packages. The NRC envisions the process for special
package authorization to be similar to authorization for Type B
packages, with specific criteria for approval judged on a case-by-case
basis. The special package authorization is not intended for repeat or
routine shipments of components. It is reserved for those unique
instances where traditional packaging and approval methods are
impractical. Therefore, NRC is not extending special package
authorizations to multiple shipments of the same component.
Comment. One commenter opposed NRC's proposal to allow special
package exemptions stating that it would not be a responsible action by
NRC and could lead to further requests to loosen regulatory
restrictions in the future. The commenter cited the precedent of
Shippingport, Trojan, and Yankee Rowe as reason for the concern. The
commenter further stated that post-September 11, 2001, NRC ``should not
assume the legality or safety of any exemptions from full packaging
container requirements.'' The commenter added that the TS-R-1,
paragraph 312, ``is not in the public interest and should be changed''
and NRC should not allow this decision to remain with DOT. The
commenter stated that NRC itself admits that DOT uses altered
definitions to justify transporting special (large) components without
the amount of protection demanded of lesser components; this is
unacceptable and a failure by NRC to exercise its mandated
responsibility. The commenter also requested the NRC to provide a
definition of ``reasonable assurance.''
This commenter further stated that the ``shortcoming of dual
regulation is evident in the handoff of regulatory control from one
agency to another'' and added that it is unacceptable ``for NRC to wash
its hands of its responsibility for packaging and containers by handing
over authority to another agency.'' The commenter then asked if NRC
planned this as ``merely a cost reduction for licensees,'' and stated
that NRC needed to provide a justification for this proposal. The
commenter also questioned the safety of these shipments.
The commenter also stated that the NRC's focus on high-level waste
transport would result in the NRC ignoring allowances for exemptions
for lower activity materials and wastes. This would result in these
materials and wastes passing from a ``regulated status to exemption and
release into commerce or unregulated `disposal' and would `increase
risks to the public that NRC ignores.' The commenter ended by stating
that this ``is not an acceptable deregulation, is a capricious failure
to protect the general welfare, and is therefore contrary to law'' and
reiterated the ``objection to NRC's reliance on `performance-based risk
informed' regulation that permits less stringent requirements for
containment and for transportation.''
Response. The special package authorization does not reduce the
protection of public health and safety;
[[Page 3744]]
rather, it affects the process used to approve nonstandard packages.
The special package authorization requirement clearly states that the
overall safety in transport for shipments approved under special
package authorization will be at least (emphasis added) equivalent to
that which would be provided if all applicable requirements had been
met. The NRC is not adding a definition for the term ``reasonable
assurance'' because it is not used in a regulatory requirement.
It is important to repeat that NRC approval will be required for
special package authorizations. In addition, DOT regulations will be
modified to recognize NRC's special package authorizations. The process
efficiencies offered by special package authorizations result in more
effective and efficient regulation.
The special package authorization will reduce the need for
exemptions in the package approval process and will not result in the
disposal of radioactive material.
Comment. One commenter stated that the Trojan reactor shipment
should not be used as a precedent for special package approval. The
commenter reasoned that the Trojan reactor shipment was an easy
shipment due to its origin and destination.
Response. The NRC believes the Trojan reactor vessel shipment
indicates there is a need for special package approvals because it
represents a class of contents that, due to their size, mass, or other
unique factors, are impractical to transport within standard
radioactive material packaging. The origin and destination of the
Trojan shipment has no bearing on this rule.
Comment. One commenter requested more information about how the NRC
is going to approve special packages. The commenter stated that a
better explanation of this process would aid regulated bodies in
acquiring special package authorization.
Another commenter indicated that with the current proposal, ``the
special package authorization is not bounded and applicants do not have
a common basis for preparation of an application'' and requested that
the NRC staff establish general criteria against which special packages
can be evaluated.
One commenter suggested that NRC establish general criteria for the
special package authorization process.
One commenter stated that the ``special package'' designator should
be clearly defined in terms of package size or other appropriate
feature to ensure that the rule is applied correctly.
Response. The purpose of this change is to establish general
criteria for the authorization of special package designs without the
need for the licensee to request an exemption from the current
regulations. The NRC agrees that additional information on special
package approvals is needed. NRC intends to develop regulatory guidance
in this area before this rule is implemented. In the interim, any
applications for special package approvals will be considered on a
case-by-case basis.
Comment. One commenter requested the NRC to view every shipment of
a reactor vessel as a significant process requiring National
Environmental Policy Act (NEPA) review. The commenter argued that a
NEPA process would allow for public input in the process of
decommissioning a reactor vessel.
Response. A NEPA review will not be required for the new special
package authorizations. Package approvals authorized by our regulations
are specifically excluded from the requirement to prepare an EA
pursuant to NEPA (Sec. 51.22(c)(13)). In contrast, an EA for the Trojan
reactor vessel was thought to be necessary because the NRC did not rely
on specific package approval regulations, but rather relied on an
exemption from those requirements.
Comment. One commenter suggested that shipping retired reactor
vessels should be a separate issue from the exception process.
Response. The NRC disagrees that reactor vessels should be excluded
from special package authorization. The NRC believes reactor vessels
are an example of the type of shipment that would benefit from special
package authorization, because the authorization would follow a more
standardized and efficient design review process. NRC's package design
review process has been shown to provide adequate protection of public
health and safety.
Comment. One commenter stated that no additional limitations should
be applied to the conditions under which one could apply for a package
authorization. The commenter noted that the few packages that have been
authorized have moved without incident and without undue risk to the
public, workers, or the environment.
Response. Comment noted. No response necessary.
Comment. Five commenters supported the proposed provisions in Sec.
71.41(d) for special package authorizations. Two of these commenters
stated that this revision provides a consistent approach to dealing
with the transport of large pieces of equipment and nonstandard items,
and that the revision would improve the safety and cost effectiveness
of onsite and offsite transfers of large equipment items. Two other
commenters supported corresponding with DOT to eliminate duplicitous
exemptions, but urged the NRC to work closely to ensure the clear
implementation of this proposal.
Response. No response necessary.
Issue 13. Expansion of Part 71 Quality Assurance (QA) Requirements to
Certificate of Compliance (CoC) Holders
Summary of NRC Final Rule. The final rule adds the terms
``certificate holder'' and ``applicant for a CoC'' to subpart H, part
71 and adds a new section, Sec. 71.9, on employee protection. Adopting
these requirements will ensure that the regulatory scheme of part 71
will remain more consistent with other NRC regulations in that
certificate holders and applicants for a CoC will be responsible for
the behavior of their contractors and subcontractors.
This expansion is necessary to enhance NRC's ability to enforce
nonconformance by the certificate holders and applicants for a CoC.
Although CoC's are legally binding documents, certificate holders and/
or applicants and their contractors and subcontractors have not clearly
been brought into the scope of part 71 requirements. This is because
the terms ``certificate holder'' and ``applicant for a certificate of
compliance'' do not appear in part 71, subpart H; rather, subpart H
only mentions ``licensee'' in these regulations. Consequently, the NRC
has not had a clear basis to cite applicants for, and holders of CoC's
for violations of part 71 requirements in the same way it has
licensees.
The NRC also added a new section (Sec. 71.9) on employee protection
to part 71. The NRC believes that employee protection regulations
should be added to cover the employees of certificate holders and
applicants for a CoC to provide greater regulatory equivalency between
part 71 licensees and certificate holders.
Affected Sections. Sections 71.0, 71.1, 71.6, 71.7, 71.8 , 71.9,
71.91, 71.93, 71.100, and 71.101 through 71.137.
Background. On October 15, 1999 (64 FR 56114), the Commission
issued a final rule to expand the QA provisions of part 72, subpart G,
to specifically include certificate holders and applicants for a CoC.
In a Staff Requirements Memorandum (SRM) to SECY-97-214, the Commission
directed the staff to consider whether conforming changes to the QA
regulations in part 71 would be necessary because of the existence of
dual-purpose cask designs.
[[Page 3745]]
The 1999 rule requires that Part 72 licensees, certificate holders,
and applicants for a CoC are responsible for assuring that their
contractors and subcontractors (e.g., fabricators) are implementing
adequate QA programs. Similarly, by this final rule, part 71 licensees,
certificate holders, and applicants for a CoC are responsible under
Sec. 71.115 for assuring that their contractors and subcontractors
(e.g., fabricators) are implementing adequate QA programs.
Under part 71, the NRC reviews and approves applications for Type B
and fissile material packages for the transport of radioactive
material. The NRC's approval of a package is documented in a CoC.
Applicants for a CoC are currently required by Sec. 71.37 to describe
their QA program for the design, fabrication, assembly, testing,
maintenance, repair, modification, and use of the proposed package.
Further, existing Sec. 71.101(a) describes QA requirements that apply
to design, purchase, fabrication, handling, shipping, storing,
cleaning, assembly, inspection, testing, operation, maintenance,
repair, and modification of components of packagings that are important
to safety. Type B packages are intended to transport radioactive
material that contains quantities of radionuclides greater than the
A1 or A2 limits for each radionuclide (see
Appendix A to part 71 for examples of A1 or A2
limits). Fissile material packages are intended to transport fissile
material in quantities greater than the part 71, subpart C, general
license limits for fissile material (e.g., existing Sec.Sec. 71.18,
71.20, 71.22, and 71.24).
Although CoCs are legally binding documents, certificate holders or
applicants for a CoC and their contractors and subcontractors have not
clearly been brought into the scope of part 71 requirements. This is
because the terms ``certificate holder'' and ``applicant for a
certificate of compliance'' do not appear in part 71, subpart H;
rather, subpart H only mentions ``licensee'' in these regulations.
Consequently, the NRC has not had a clear basis to cite certificate
holders and applicants for a CoC for violations of part 71 requirements
in the same way it has licensees.
When the NRC has identified a failure to comply with part 71 QA
requirements by certificate holders or applicants for a CoC, it has
issued a Notice of Nonconformance (NON) rather than a Notice of
Violation (NOV). Although an NON and an NOV appear to be similar, the
Commission prefers the issuance of an NOV because:
(1) The issuance of an NOV effectively conveys to both the person
violating the requirement and the public that a violation of a legally
binding requirement has occurred;
(2) The use of graduated severity levels associated with an NOV
allows the NRC to effectively convey to both the person violating the
requirement and the public a clearer perspective on the safety and
regulatory significance of the violation; and
(3) Violation of a regulation reflects the NRC's conclusion that
potential risk to public health and safety could exist. Therefore, the
NRC believes that limiting the available enforcement sanctions to
administrative actions is insufficient to address the performance
problems observed in industry.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. Five commenters supported the NRC's proposed position on
this issue. One commenter recommended that NRC establish and apply a
uniform set of QA requirements. Another commenter added that it would
like to see the consistent application of QA requirements throughout
the regulations.
Response. Expansion of the QA provisions enhances NRC's ability to
enforce noncompliance and will ensure broader, uniform application of
QA requirements. However, extension of the requirement beyond part 71
is outside the bounds of this rulemaking.
Issue 14. Adoption of the American Society of Mechanical Engineers
(ASME) Code
Summary of NRC Final Rule. The NRC has decided not to incorporate
the ASME Code, section III, division 3 requirements into part 71.
Public Law 104-113 requires that Federal agencies use consensus
standards in lieu of government-unique standards, if this use is
practical or inconsistent with other existing laws. Because a major
revision to the ASME Code is forthcoming and because the changes in
that revision are not yet available for staff and stakeholder review,
the NRC staff considered it an imprudent use of NRC and stakeholder
resources to initiate rulemaking on the current ASME Code revision only
to have the ASME Code requirements change during the part 71
rulemaking.
Affected Sections. None (not adopted).
Background. Currently, no ASME Code requirements exist in part 71
for fabrication/construction of spent fuel transportation packages. The
NRC considered the adoption of the ASME Boiler and Pressure Vessel
(B&PV) Code, section III, division 3, for two reasons. First, previous
NRC inspections at vendor and fabricator shops (for fabrication of
spent fuel storage canisters and transportation casks) identified
quality control (QC) and QA problems. Some of these problems would have
been prevented with improved QA programs, and may have been prevented
had fabrication occurred under more prescriptive requirements such as
the ASME Code requirements. Second, Public Law 104-113, ``National
Technology Transfer and Advancement Act,'' enacted in 1996, requires
that Federal agencies use, as appropriate, consensus standards (e.g.,
the ASME B&PV Code), except when there are justified reasons for not
doing so.
With respect to conformance to Public Law 104-113, the ASME issued
a consensus standard in May 1997, entitled: ``Containment Systems and
Transport Packages for Spent Fuel and High Level Radioactive Waste,''
ASME B&PV Code, section III, division 3. The ASME Code requires the
presence of an Authorized Nuclear Inspector during construction to
ensure that the ASME Code requirements are met and the stamping of
components (i.e., the transportation cask's containment) constructed to
the ASME Code. NRC staff participated, and continues to participate, in
the ASME subcommittee that developed the ASME Code requirements. It is
the NRC staff's understanding, through participation in the
subcommittee, that the ASME Code document is undergoing extensive
review and modification and that a major revision will be issued.
Therefore, NRC staff believes that inclusion of the ASME Code in part
71 is not appropriate at this time.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. Four commenters expressed support for the decision not to
adopt the ASME code. One commenter said that these are voluntary
standards and should not be made into requirements.
Response. No response is required.
Issue 15. Change Authority for Dual-Purpose Package Certificate Holders
Summary of NRC Final Rule. The Commission does not reach a final
decision on the issue of change authority for dual-purpose package
[[Page 3746]]
certificate holders in this final rule. The NRC has determined that
implementation of this change would result in new regulatory burdens
and costs which could be significant. The Commission believes it needs
further input from stakeholders on the values and impacts of this
change before deciding whether to adopt a final rule providing change
authority for dual-purpose package certificate holders. The NRC staff
plans to conduct public meetings with appropriate stakeholders to
develop a final regulatory solution which it will propose to the
Commission. At that time, the Commission will either issue a final rule
resolving this issue, taking into account the comments received on the
proposed rule and in any future public meetings, or will withdraw 10
CFR part 71 subpart I of the proposed rule.
Affected Sections. None.
Background. The Commission approved a final rule to expand the
provisions of Sec. 72.48, ``Changes, Tests, and Experiments,'' to
include part 72 certificate holders and licensees (64 FR 53582; October
4, 1999). Part 72 certificate holders and licensees are allowed, under
Sec. 72.48, to make certain changes to a spent fuel storage cask's
design or procedures used with the storage cask and to conduct tests
and experiments without prior NRC review and approval. Part 71 does not
contain any similar provisions to permit a CoC holder to change the
design of a part 71 transportation package, without prior NRC review
and approval. The NRC has issued separate CoC's under parts 71 and 72
for dual-purpose spent fuel storage casks and transportation packages.
This has created a situation where an entity holding both a part 71 and
a part 72 CoC would be allowed under part 72 to make certain changes to
the design of a dual-purpose cask (i.e., changes that affected a
component or design feature that has a storage function) without
obtaining prior NRC approval. However, the entity would not be allowed
under part 71 to make changes to the design of this same dual-purpose
cask (package) if that component or feature also has a transportation
function without obtaining prior NRC approval, even when the same
physical component and change are involved (i.e., the change involves a
component that has both storage and transportation functions).
NRC staff recognized a need to consider making both part 72 and
part 71 more consistent in dealing with design changes of a minor
nature. Thus, in SECY-99-054,\2\ NRC staff recommended that an
authority similar to Sec. 72.48 be created for dual-purpose spent fuel
storage casks and transportation packages intended for domestic use
only. NRC staff also recommended that this authority be limited to the
part 71 CoC holder.
---------------------------------------------------------------------------
\2\ SECY-99-054; February 22, 1999, ``Plans for Final Rule-
Revisions to Requirements of 10 CFR parts 50, 52, and 72 Concerning
Changes, Tests, and Experiments.''
---------------------------------------------------------------------------
Since the proposed rule was published, the NRC has evaluated
comments received from the public and has conducted a detailed analysis
of the implementation of the change authority, as proposed. Based on
this analysis, the NRC has determined not to finalize subpart I, Type
B(DP) Package Approval, as proposed. Instead, the NRC will seek further
input on the values and impacts of this change and then decide whether
to proceed with a final rule.
Proposed Sec. 71.153 stated that the application for a Type B(DP)
package shall include an analysis of potential accidents, package
response to these potential accidents, and any consequences to the
public. Currently, under part 71, an applicant has to demonstrate,
either by test or analysis, that a package design can withstand the
cumulative effects of the Hypothetical Accident Conditions of a 30-foot
drop test, a 40-inch puncture test, a thermal test, and immersion tests
as described in Sec. 71.73 and Sec. 71.61, and meet Subpart E--Package
Approval Standards. Applicants are not required to perform an
independent analysis of potential transportation accidents specific to
that design and plans for use, project package responses to ``real
world'' transportation accidents, or determine the consequences to the
public from such accidents.
The NRC reviewed and considered the comments that were received
about this proposed change. The new process included the need to
establish a design specific accident assessment for the cask design
response to potential ``real world'' transportation accidents. Such an
accident analysis has not been required for a transportation cask
application before. Which accidents would be appropriate, for which
routes, under what conditions, for what duration, and with what
combinations of forces and assumptions, all would be questions that
would need to be answered by CoC applicants who have not been required
to perform such analysis for cask designs applications.
To provide new guidance for the development of an acceptable
accident analysis for a transportation cask, the NRC staff would need
to perform significant research on what types of accidents would be
required to be included. The NRC believes that such an analysis can be
performed; however, the NRC does not believe that it had fully
considered in the proposed rule the rigor, resources, and time that
such a requirement would require. The detailed associated cost
estimates had not been included in the RA for this part of the rule
change. The RA has been revised, and the costs of implementation for
CoC holders could be significantly higher than that reflected in the
proposed rulemaking. This additional regulatory burden had not been
accurately reflected in the draft RA. The Safety Analysis Report (SAR)
for part 71 applications is based, in part, on demonstrating compliance
with the Hypothetical Accident Conditions of part 71. Thus, there is
not a clear linkage between the SAR and regulatory conditions for
making changes to a design without NRC approval, such as a minimal
increase in the probability of an accident sequence or the creation of
accidents of a different type. Given these revised cost estimates, the
NRC is uncertain whether the benefits to be gained from this change
outweigh the costs. The NRC intends to explore this issue further
before deciding whether to proceed to a final rule.
The proposed Sec. 71.175, ``Changes,'' establishes methods to
determine if a proposed change to a Type B(DP) package can be made
without prior NRC approval. As stated in a public comment, the language
in this section mirrors that in Sec. 72.48. It should be noted that the
design and application process under part 72 does require that an
applicant perform an accident analysis as part of its application for
approval, but such a requirement has never been incorporated into part
71 as noted above.
The intent of subpart I was to allow a certificate holder
flexibility to make minor changes to the design of the package to be
consistent with the change authority provided under Sec. 72.48 for
spent fuel storage casks in a cost and time effective manner. The NRC
notes that transportation CoCs issued under part 71 do allow for many
changes to be made to package designs without NRC approval, provided
the changes do not impact upon compliance with part 71 standards. For
example, changes in the SAR for a transportation package, in general,
do not require NRC approval provided the changes do not affect the
conditions listed in the CoC or the ability of the package to meet the
requirements of part 71. Additionally,
[[Page 3747]]
packaging design drawings that are included as conditions in the CoC do
not need to specify fabrication details that are not important to
safety. In this way, changes may be made to nonsafety features without
modifying the drawings and without NRC review and approval. This is in
contrast to the approaches for part 72 CoCs. It is therefore important
that applications for package approval, including packaging design
drawings, are developed to focus on the safety features of the design.
The NRC notes that the current regulatory process for evaluating and
approving CoC amendments for transportation packaging may be more
efficient than developing a new regulatory infrastructure. To aid in
receiving high quality transportation applications, the NRC staff is
preparing an amended standard format and content regulatory guide.
The NRC has determined that implementation of the proposed change
process would result in new regulatory burdens and costs which could be
significant. The NRC also recognizes the concerns of public commenters
related to the potential benefits of allowing changes to the design of
a Type B(DP) package without prior NRC approval. The NRC staff will
work with appropriate stakeholders to determine whether a final rule is
the preferred method for resolving the need for a change process in
part 71 or whether there may be other regulatory solutions that meet
this need. The NRC staff will then propose a final regulatory solution
to the Commission. The Commission will then determine if subpart I
should be issued as a final rule or if other regulatory solutions to
this issue obviate the need for going forward with a final rule. If a
final rule is not needed, then proposed subpart I will be withdrawn and
the comments received on this issue will be addressed at that time.
Issue 16. Fissile Material Exemptions and General License Provisions
Summary of NRC Final Rule. The final rule adopts various revisions
to the fissile material exemptions and the general license provisions
in part 71 to facilitate effective and efficient regulation of the
transport of small quantities of fissile material. The fissile
exemptions (Sec. 71.15) have been revised to include controls on
fissile package mass limit combined with package fissile-to-nonfissile
mass ratio. The general license for fissile material (Sec. 71.22) has
been revised to consolidate and simplify current fissile general
license provisions from Sec.Sec. 71.18, 71.20, 71.22, and 71.24. Under
the final rule, the general license is based on mass-based limits and
the CSI. In light of comments and applicable DOT requirements, the
final rule removes proposed rule language references to ``storage
incident to transportation.'' Also, the exemptions for low level
materials in Sec. 71.14 were revised to apply only to nonfissile and
fissile-exempt materials.
Affected Sections. Sections 71.4, 71.10, 71.11, 71.18, 71.20,
71.22, 71.24, 71.53, 71.59, and 71.100. (Currently effective Sec. 71.10
was relocated to Sec. 71.14 with additional language. Currently
effective Sec.Sec. 71.18, 71.20, 71.22, 71.24, and 71.53 are replaced
by new Sec.Sec. 71.15 and 71.22.)
Background. The NRC published an emergency final rule amending its
regulations on shipments of small quantities of fissile material (62 FR
5907; February 10, 1997). This rule revised the regulations on fissile
exemptions in Sec. 71.53 and the fissile general licenses in Sec.Sec.
71.18 and 71.22. The NRC determined that good cause existed, under
section 553(b)(B) of the Administrative Procedure Act (APA) (5 U.S.C.
553(b)(B)), to publish this final rule without notice and opportunity
for public comment. Further, the NRC also determined that good cause
existed, under section 553(d)(3) of the APA (5 U.S.C. 553(d)(3)), to
make this final rule immediately effective. Notwithstanding the final
status of the rule, the NRC provided for a 30-day public comment
period. The NRC subsequently published in the Federal Register (64 FR
57769; October 27, 1999) a response to the comments received on the
emergency final rule and a request for information on any unintended
economic impacts caused by the emergency final rule.
The NRC issued this emergency final rule in response to a
regulatory defect in the fissile exemption regulation in Sec. 71.53
which was identified by an NRC licensee. The licensee was evaluating a
proposed shipment of a special fissile material and moderator mixture
(beryllium oxide mixed with a low concentration of high-enriched
uranium). The licensee concluded that while Sec. 71.53 was applicable
to the proposed shipment, applying the requirements of Sec. 71.53
could, in certain circumstances, result in an inadequate level of
criticality safety (i.e., an accidental nuclear criticality was
possible in certain unique circumstances).\3\
---------------------------------------------------------------------------
\3\ For transportation purposes, ``nuclear criticality'' means a
condition in which an uncontrolled, self-sustaining, and neutron-
multiplying fission chain reaction occurs. ``Nuclear criticality''
is generally a concern when sufficient concentrations and masses of
fissile material and neutron moderating material exist together in a
favorable configuration. Neutron moderating material cannot achieve
criticality by itself in any concentration or configuration.
However, it can enhance the ability of fissile material to achieve
criticality by slowing down neutrons or reflecting neutrons.
---------------------------------------------------------------------------
The NRC staff confirmed the licensee's analysis that this beryllium
oxide and high-enriched uranium mixture created the potential for
inadequate criticality safety during transportation. An added factor in
the urgency of the situation was that under the NRC regulations in
Sec.Sec. 71.18, 71.20, 71.22, 71.24, and 71.53, these types of fissile
material shipments could be made without prior approval of NRC. For
many years, NRC allowed these shipments of small quantities of fissile
material based on NRC's understanding of the level of risk involved
with these shipments, as well as industry's historic transportation
practices. This experience base had led NRC (and its predecessor, the
Atomic Energy Commission (AEC)) to conclude that shipments made under
the fissile exemption provisions of part 71 typically required minimal
regulatory oversight (i.e., NRC considered these types of shipments to
be inherently safe).\4\
---------------------------------------------------------------------------
\4\ The NRC's regulations in part 71 ensure protection of public
health and safety by requiring that Type AF, B, or BF packages used
for transportation of large quantities of radioactive materials be
approved by the NRC. This approval is based upon the NRC's review of
applications which contain an evaluation of the package's response
to a specific set of rigorous tests to simulate both normal
conditions of transport (NCT) and hypothetical accident conditions
(HAC). However, certain types of packages are exempted from the
testing and NRC prior approval; these are fissile material packages
that either contain exempt quantities (Sec. 71.53), or are shipped
under the general license provisions of Sec.Sec. 71.18, 71.20,
71.22, or 71.24.
---------------------------------------------------------------------------
All public comments on the emergency final rule supported the need
for limits on special moderators (i.e., moderators with low neutron-
absorption properties such as beryllium, graphite, and deuterium).
However, the commenters stated that the restrictions were far too
limiting (to the point that some inherently safe packages were excluded
from the fissile exemption) and could lead to undue cost burdens with
no benefit to safety. In addition, the commenters believed that the
consignment mass limits set to deter undue accumulation of fissile mass
would be extremely costly. Therefore, the commenters recommended that
further rulemaking was necessary to resolve these excessive
restrictions. Based on the public comments on the emergency final rule,
NRC staff contracted with Oak Ridge National Laboratory (ORNL) to
review the fissile
[[Page 3748]]
material exemptions and general license provisions, study the
regulatory and technical bases associated with these regulations, and
perform criticality model calculations for different mixtures of
fissile materials and moderators. The results of the ORNL study were
documented in NUREG/CR-5342,\5\ and NRC published a notice of the
availability of this document in the Federal Register (63 FR 44477;
August 19, 1998). The ORNL study confirmed that the emergency final
rule was needed to provide safe transportation of packages with special
moderators that are shipped under the general license and fissile
material exemptions, but the regulations may be excessive for shipments
where water moderation is the only concern. The ORNL study recommended
that NRC revise part 71.
---------------------------------------------------------------------------
\5\ NUREG/CR-5342, ``Assessment and Recommendations for Fissile-
Material Packaging Exemptions and General Licenses Within 10 CFR
Part 71,'' July 1998.
---------------------------------------------------------------------------
In the October 27, 1999 (64 FR 57769) final rule, the Commission
requested additional information on the cost impact of the emergency
final rule from the public, industry, and DOE because the NRC staff was
not successful in obtaining this information. Specifically, NRC
requested information on the cost of shipments made under the fissile
material exemptions and general license provisions of part 71, before
the publication of the emergency final rule, and those costs and/or
changes in costs resulting from implementation of the emergency rule.
One commenter agreed with the NRC approach but stated that, ``the
limits for those materials containing no special moderators can and
should be increased, hopefully back to their pre-emergency rule
levels.''
As part of NUREG/CR-5342, ORNL performed computer model
calculations of keff (k-effective) for various combinations
of fissile material and moderating material, including beryllium,
carbon, deuterium, silicon-dioxide, and water, to verify the accuracy
of current minimum critical mass values. These minimum critical mass
values were then applied to the regulatory structure contained in part
71, and revised mass limits for both the general license and exemption
provisions to part 71 were determined. Also, ORNL researched the
historical bases for the fissile material exemption and general license
regulations in part 71 and discussed the impact of the emergency final
rule's restrictions on NRC licensees. ORNL concluded that the
restrictions imposed by the emergency final rule were necessary to
address concerns relative to uncontrolled accumulation of exempt
packages (and thus fissile mass) in a shipment and the potential for
inadequate safety margin for exempt packages with large quantities of
special moderators.
Based on its new keff calculations, ORNL suggested that:
(1) The mass limits in the general license and exemption provisions
could be safely increased and thereby provide greater flexibility to
licensees shipping fissile radioactive material; and (2) additional
revisions to part 71 were appropriate to provide increased
clarification and simplification of the regulations. Copies of NUREG/
CR-5342 may be obtained by writing to the Superintendent of Documents,
U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-
9328. A copy is also available for inspection and copying, for a fee,
at the NRC Public Document Room in the NRC Headquarters at One White
Flint North, Room O-1F21, 11555 Rockville Pike, Rockville, MD 20852-
2738.
The current restrictions on fissile exempt and general license
shipments under Sec.Sec. 71.53, and 71.18 through 71.24, respectively,
are burdensome for a large number of shipments that actually contain no
special moderating materials (i.e., packages that are shipped with
water considered as the potential moderating material). This problem
was clearly expressed in public comments on the emergency final rule.
Another regulatory problem is that the current fissile exempt and
general license provisions are cumbersome and outdated; this was one of
the main conclusions of the ORNL study.
The NRC proposed changes (67 FR 21417) were made on the basis of 17
recommendations contained in NUREG/CR-5342. These changes included: (1)
Revising Sec. 71.10, ``Exemption for low level materials,'' to exclude
fissile material, also redesignate Sec. 71.10 as Sec. 71.14; (2)
redesignating Sec. 71.53 as Sec. 71.15, ``Exemption from classification
as fissile material,'' and revise the fissile exemptions; (3)
consolidation of the existing four general licenses in existing
Sec.Sec. 71.18, 71.20, 71.22, and 71.24 into one general license in new
Sec. 71.22, revise the mass limits, and add Type A package, CSI, and QA
requirements; and (4) consolidation of the existing general license
requirements for plutonium-beryllium sealed sources, which are
contained in existing Sec.Sec. 71.18 and 71.22 into one general license
in new Sec. 71.23 and revise the mass limits. Additionally, changes
were proposed to be made to Sec. 71.4, ``Definitions,'' and Sec.
71.100, ``Criminal penalties.''
The NRC also proposed: (1) To adopt the use of the CSI for general
licensed fissile packages; and (2) to retain the current per package
(CSI) limit of 10, rather than raising the per package limit to 50 (see
Issue 5). TS-R-1 does not address the issue of fissile general
licenses, so no compatibility issues arise with retention of the
current NRC per package limit of 10. NRC staff believes that because
reduced regulatory oversight is imposed on fissile general license
shipments (e.g., the package standards of Sec.Sec. 71.71 and 71.73,
fissile package standards of Sec. 71.55, and fissile array standards of
Sec. 71.59 are not imposed for fissile general license shipments),
retention of the current per package limit of 10 is appropriate.
Furthermore, retention of the current per package limit of 10 would not
impose a new burden on licensees; rather, licensees shipping fissile
material under the general license provisions of Sec.Sec. 71.22 and
71.23 would not be permitted to take advantage of the relaxation of the
per package CSI limit from 10 to 50 that would be permitted for Types
AF and B(F) package shipments.
As a result of stakeholder meetings and public comments, the NRC
has incorporated the following changes to the proposed language for
Sec.Sec. 71.15 and 71.22 in the final rule:
(1) Small quantities of fissile materials such as environmental
samples shipped for testing are judged to be of sufficient low quantity
that, if individually packaged, the risk (probability and consequence)
of accumulating the number and type of packages needed to present a
potential criticality hazard is judged to be inconsequential.
Therefore, a new Sec. 71.15(a) has been added to exempt packages
containing 2 grams or less fissile material.
(2) Proposed Sec. 71.15(a) (Sec. 71.15(b) in the final rule)
specifically referred to iron as the nonfissile material for
calculating limiting ratio of 200:1. Commenters suggested that this
would require a new definition (of iron) and would complicate
implementation. There is no technical reason to require that iron be
identified as the nonfissile materials to be included with a mass ratio
of 200:1. Other nonspecial moderating materials such as stainless
steel, concrete, etc., are appropriate. The mass ratio wording has been
modified. The modification maintains the need for the mass ratio of
200:1, but the required nonfissile material is required to be a solid.
As worded, the nonfissile mass can include the packaging mass. It is
judged that sufficient distribution of fissile material in small
quantities (i.e., 1 g of fissile material per 200 g of solid nonfissile
material) will provide adequate protection against nuclear
[[Page 3749]]
criticality. This specification ensures that large numbers of packages,
containing 15 g of fissile material per package, will remain safely
subcritical because of the fissile material dilution and density
reduction by nonfissile materials which are not special moderators
(e.g., beryllium, graphite, etc.). For example, 1 g of optimally
moderated uranium-235 in a mixture at about 0.05 g Uranium-235/cm\3\
occupies a volume of about 20 cm\3\. Two hundred grams of aluminum
metal at about 2.7 g of aluminum/cm\3\ occupies a volume of about 74
cm\3\. As specified, the 15 g of uranium-235 per package will have a
diluted volume of about 1,410 cm\3\ at a density of about 0.01 g
uranium-235/cm\3\ and a density reduction by a factor of 5. Though
aluminum is a minor absorber of low-energy neutrons, most other common
materials of packaging have moderate neutron-absorbing properties that
further ensure safely subcritical accumulations of such packages. The
increase in the subcritical mass of 620 g of optimally moderated
uranium-235, permitted by the reduction of fissile material density, is
related to the ratio of the densities to the power of 1.8 (see Ref. 1 ,
pp. 19-22). Given the density reduction of 5 in the above example, the
adjusted subcritical mass becomes 11,125 g of uranium-235, requiring in
excess of about 741 packages (containing 15 g of uranium-235 per
package) to exceed the determined equivalent quantity of material.
(3) Proposed Sec. 71.15(b) (Sec. 71.15(c) in the final rule), was
modified by referring to fissile and nonfissile materials as solid
materials instead of using ``noncombustible'' and ``insoluble-in-
water.'' The modification was a pragmatic consideration and was made to
avoid reference to the undefined/specified word, ``noncombustible,''
and the phrase, ``insoluble-in-water,'' while addressing the need to
avoid fissile and nonfissile liquids/gases that easily could be
consolidated or lost (thereby decreasing nuclear criticality safety) in
normal and hypothetical accident transportation circumstances. An
additional modification, Sec. 71.15(c)(2) in the final rule, also
removes the limit of 350 g in a package and instead specifies criteria
for commingling of the material such that, within any selected 360 kg
of nonfissile solid material, there can be no more than 180 g of
fissile material. Thus, a large rail car with a homogenized
distribution of fissile material within a nonfissile waste matrix might
exceed the 180 g limit but would be effectively mixed at low enough
concentration to enable safe shipment.
(4) The basis for Sec. 71.15(c)(1) is that a 2000:1 mass ratio of
nonfissile to fissile material is [sim]60% of the minimum critical
fissile material concentration of 1.33 g uranium-235/L in a 1,600 g
SiO2/L matrix. The 60-percent value is judged to be a
reasonably conservative decrease in g uranium-235/g nonfissile material
(e.g., SiO2) to accommodate other nonfissile materials. The
minimum critical fissile material concentration in SiO2 was
derived from studies to compare ``special'' and ``natural'' neutron
moderators with fissile materials. In those studies various systems
were examined that had different species of fissile material (i.e.,
uranium-235, uranium-233, or plutonium-239) combined with water and
other nonfissile neutron scatterers/moderators (e.g., polyethylene,
beryllium, carbon, deuterium, and SiO2). SiO2 was
selected for consideration in the transport exemptions because it is
judged to be the most representative, arbitrary, and nonspecial
moderator matrix for commingling with fissile material. SiO2
has a very low probability for absorbing neutrons and has a large
abundance in nature (i.e., 33 weight percent, second only to oxygen at
49 weight percent). An independent study compared the relative
importance of other elements to silicon with dilute fissile materials.
Except for the category of special moderators (i.e., deuterium,
beryllium, and graphite) and pure forms of magnesium (i.e., magnesium
carbonate, magnesium fluoride, magnesium oxalate, magnesium oxide,
magnesium peroxide, magnesium silicates) and bismuth (i.e., bismuth
basic carbonate, bismuth tri-or penta-fluorides, bismuth oxide),
silicon or silicon dioxide is the most neutronically reactive diluent
for fissile materials. The 1.6-g SiO2/L is representative of
dry bulk mean world soil density.
(5) Section 71.15(d) (Sec. 71.15(c) in proposed rule) has been
revised to reflect ``mass of beryllium, graphite, and hydrogenous
material enriched in deuterium constitute less than 5 percent of the
uranium mass'' (less than 0.1 percent of the fissile mass being the
proposed phrase). This change was made in response to a comment about
the difficulty that shippers would experience based on the proposed
rule language. The staff reviewed the 0.1 percent of fissile mass
language and determined that limiting the low-neutron-absorbing
materials to the proposed ratio would be impractical to implement. The
final language reflecting 5 percent of the uranium mass assures
subcriticality for all moderators of concern and is less burdensome to
measure and implement as a requirement.
(6) Section 71.15(e) (Sec. 71.15(d) in the proposed rule) states
``total plutonium and uranium-233 content not exceeding 0.002 percent
of the mass of uranium'' while the proposed language stated ``does not
exceed 0.1 percent of the mass of uranium-235.'' This change was made
in response to a public comment that the proposed rule changes should
be consistent with the international regulations. The final language
for this section has been revised to be consistent with the 1996 IAEA
standards.
(7) Section 71.15(f) (proposed Sec. 71.15(e)) was reworded for
clarity but reflects the same requirements and guidance as in the
proposed language.
(8) Proposed Sec. 71.22 (e)(5)(iii), Exemption from classification
as fissile material, was revised to read `` * * * The uranium is of
unknown Uranium-235 enrichment or greater than 24 weight percent
enrichment; or * * * '' The reason for the Sec. 71.22(e)(5)(iii)
modification was that enrichments of U-235 greater than 24 weight
percent were not accommodated in the proposed text. Because the minimum
critical mass transition between 24 and 100 weight percent enrichments
of 235U vary slightly, the text was changed to require the
use of Table 71-1 values for all enrichments greater than 24 weight
percent as well as materials of unknown enrichments. The values in
Table 71-1 were developed for 100 weight percent uranium-235 enriched
uranium and are conservatively applied down to 24 weight percent
uranium-235.
(9) Proposed Sec. 71.22, Table 71-1, was modified in the final rule
to replace uranium-235 (Y) with uranium-233 (Y)--change to uranium-233
(Y). The reason is to correct a typographical error in the table.
In the final rule, the NRC has deleted the phrase ``or stored
incident to transport'' from proposed Sec.Sec. 71.22(d)(3) and
71.23(d)(3). The intent of the storage phrase was to permit segregation
of groups of stored packages, consistent with IAEA and DOT
requirements, but the NRC staff believes that the proposed text did not
accommodate that practice because it did not accommodate storage and
segregation of groups of packages. DOT requirements properly restrict
accumulation of packages during transport, based on summing the
packages' CSI or TI, including during storage incident to transport. In
light of the division of regulatory responsibilities explained in the
NRC-DOT Memorandum of Understanding (44 FR 38690; July 2, 1979), the
NRC
[[Page 3750]]
exemptions for carriers-in-transit in Sec. 70.12, and DOT's revision to
49 CFR 173.457 (67 FR 21384), the NRC staff believes that storage in
transit provisions as proposed in Sec.Sec. 71.22(d)(3) and 71.23(d)(3)
are unnecessary.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. One commenter noted that this is a significant deviation
from the TS-R-1 requirement, which now has a 15-g uranium-235 limit as
well as a mass consignment limit.
Response. On February 10, 1997 (62 FR 5907), the NRC published a
final rule on fissile exemptions. That final rule essentially adopted
the 1996 TS-R-1 requirements, including the 15-g per package limit and
400-g consignment mass limit. Both the consignment mass limit (400 g )
and the package mass limit (15 g) were used to control package
accumulations. In consideration of comments received on the 1997 rule,
the NRC has proposed changes to the fissile exemptions; one of the
principal concerns with the 1997 rule was the practicability of the
350-g consignment mass limit (see 67 FR 21418; April 30, 2002). The
proposed rule suggested a mass ratio system together with the per
package limit to eliminate this consignment mass limit. The IAEA is
currently considering changes to the current international regulations
in the area of the fissile material exemptions.
Comment. Three commenters indicated that this provision would
overly complicate the shipping of fissile material and negatively
impact intermodal and international shipping. One commenter noted that
the three-tiered system would dramatically complicate the shipping of
fissile material because the mass ratio requirement makes it difficult
to determine how to classify UF6 into the three tiers. This
same commenter stated that companies that ship internationally will
have a difficult time complying with the proposed system as well as the
international system and suggested that NRC simplify compliance for
these companies. The other commenter stated that if NRC's proposal is
adopted as written, shippers would need to have detailed information
available regarding the materials in each packaging. The commenter
reasoned that this approach assumes that the detailed information would
be readily available and disseminated to shippers, and further,
shippers making international shipments would likely need to meet both
NRC's domestic requirements for determining fissile exempt quantities
and the international mass consignment limits, thus further
complicating the evaluation of criticality controls for a shipment.
Response. The NRC staff believes that the changes are warranted to
alleviate the unnecessary regulatory burden created by the 1997
emergency final rule, including the consignment mass limit. The changes
implemented by the 1997 rule are essentially the same as TS-R-1. These
amendments permit greater flexibility for domestic transport, in
consideration of the comments received when the U.S. adopted the TS-R-1
approach in 1997. However, NRC recognizes that international transport
will also need to comply with IAEA TS-R-1, and the burden has been
unchanged. The IAEA is currently considering changes to the current
international regulations in the area of the fissile material
exemptions. The NRC staff did review the proposed language for the
proposed Sec. 71.15(c) and determined that the 0.1 percent ratio of the
mass of beryllium, graphite, and hydrogenous material enriched in
deuterium to the total fissile mass was a requirement that was
difficult to implement and therefore the language has been changed as
noted above in the rule language description.
Comment. Several commenters expressed concern about material
definitions, with one commenter noting that the definition of iron is
unclear. One commenter requested clarification of what constitutes iron
with regard to Tier 1 or fissile exempt quantities and specifically
asked if steel is considered iron. Another stated that it is difficult
to obtain information on materials to carry out the calculations under
the proposed regulations.
Response. Many materials have the neutronic properties that would
permit them to be considered as the nonfissile material mass to be
mixed with up to 15 g of fissile material in a ratio of 200:1. Iron,
generic steels, stainless steels, and concrete are good examples of
materials for use. Only lead, beryllium, graphite, and hydrogenous
material enriched in deuterium should be excluded as noted in the
revised text. The wording has been modified and clarified in the final
rule.
Comment. One commenter requested that the NRC explain why NRC
proposes changing the total shipment CSI in cases where there is
storage incident to transport, effectively doing away with an exclusive
use condition. The commenter considered this proposal a significant
change in the method of calculating the CSI per consignment and wanted
to remind us that the proposed rule maintains segregation and storage
requirements.
Response. The ``storage incident to transport'' language has been
deleted. See the comment responses under Issue 5.
Comment. Two commenters said that NRC should clarify how the mass
limits for general license packages (found in Sec. 71.22 (a)(3), Tables
71-1 and 71-2) are used for uranium enriched greater than 24 percent.
Both commenters stated that highly enriched uranium does not meet the
criteria under Sec. 71.22(e)(5). Moreover, if uranium enriched greater
than 24 percent cannot be shipped in a DOT 7A, this provision would
have significant cost and operational impacts on the DOE.
Response. Uranium enriched to greater than 24 percent can be
shipped provided the appropriate X value from Table 71-1 is used in the
equation to determine the CSI. The proposed rule had intended Sec.
71.22(e)(3) to guide the reader to using Table 71-1 for uranium-235
enrichments greater than 24 percent. However, the text for Sec.
71.22(e)(5)(iii) has been revised to clarify the use of Table 71-1 for
uranium-235 enrichments greater than 24 percent.
Comment. Several commenters discussed the economic impact of the
proposed regulation. Two commenters asserted that the regulation will
cause an increase in the number of shipments required with an
associated increase in costs, with one predicting required transports
to increase two-to three-fold. Another warned of significant negative
economic consequences if NRC did not retain the current provision for
15 g per package, at least until it is demonstrated unsafe.
Response. These comments appear to be concerned with the rule's
restrictions on package accumulation based on CSI due to the ``storage
incident to transport'' language in the proposed rule. The ``storage
incident to transport'' language has been deleted. Also see the
response to second comment under Issue 5.
Comment. One commenter stated that ``under no circumstances should
the NRC issue general licenses for shipments of radioactive materials
and wastes (or, for that matter, for other purposes).'' The commenter
then added that NRC shouldn't allow fissile materials to be exempted
from packaging and transportation regulations nor should NRC allow
``transport subject to even remotely possible criticality accidents
during shipment'' under any circumstances. The commenter added that it
is ``an outrage, furthermore, that the NRC had
[[Page 3751]]
approved an ``emergency final rule'' allowing shipments of fissile
materials in 1997 without affording the public full opportunity for
comment * * *'' The commenter cited NRC's footnote (see 67 FR 21418;
April 30, 2002) and stated doubts regarding NRC's process for requiring
NRC's approval for ``all Type AF, B, or BF packages.'' The commenter
concluded by stating that ``NRC approval is virtually guaranteed in
almost all cases, whether or not the decision contributes to public
health and safety, not to mention the environment.''
Response. The NRC staff believes that current regulations and
programs for transporting fissile materials, and in particular the
general licensing approach in part 71, result in a high degree of
safety as evidenced by a long record of safe transport of these
materials. The staff believes that a graded series of requirements for
hazardous materials, including the fissile exemptions and general
licenses, remains appropriate.
Comment. Two commenters expressed concern about the use of the part
110 definitions of ``deuterium'' and ``graphite'' in the proposed rule.
The commenters suggested that NRC reconsider these definitions because
they are inappropriate for the purpose of nuclear criticality safety.
Response. The final rule stipulates that ``Lead, beryllium,
graphite, and hydrogenous material enriched in deuterium may be present
in the package, but must not be included in determining the required
mass of solid nonfissile material.'' Materials enriched in deuterium
and graphite are often termed special moderators because their very low
neutron absorption properties give rise to special consideration for
large systems with low concentration of fissile material and,
therefore, warrant consideration in the criticality control approach.
In the interests of consistency within NRC regulations, the NRC staff
believes that the definitions of graphite and deuterium are sufficient
for purposes of defining the materials that cannot be used in the Sec.
71.15 determination.
Comment. One commenter opposed the fissile material exemptions.
Response. No response is necessary.
Comment. Two commenters expressed general support for the fissile
material exemptions. One of whom expressed support for the graduated
exemptions for fissile material shipments because they would allow
increasing quantities in shipments, provided that the packages also
contained a corresponding increase in the ratio of non-fissile to
fissile material. They also appreciated NRC consolidating four fissile
material general licenses into one and consolidating existing general
license requirements for PuBe sources into one section and updating the
mass limits.
Response. The comments are acknowledged. No further response is
necessary.
Comment. Several commenters requested that NRC include and/or
improve various definitions in the proposed rule. One commenter stated
that improved definitions were necessary to categorize the ratio
calculations.
Three commenters added that NRC should not exclude the definition
of ``shipment'' from the rule. Another suggested that the proposed rule
was ambiguous as to whether iron in the packaging (e.g. internal
structure) can be used to meet the 200:1 ratio requirement in the 15-g
exception.
Two commenters noted that the proposed rule did not include a
definition for ``insoluble in water,'' one of whom stated that the
proposed rule fails to clarify the issue in part because of the
rulemaking's lack of clarity. This same commenter questioned NRC's
decision to omit definitions for ``consignment'' and ``shipment'' and
urged NRC to adopt the TS-R-1 definition for these terms.
Response. The NRC staff believes the terms ``ratio'' and
``calculations'' are sufficiently clear without corresponding
definitions. The terms ``iron in the packaging'' and ``insoluble in
water'' have been deleted from the rule. Because of its bearing upon
the fissile exemptions rule, a definition of ``consignment'' that is
consistent with the definition in DOT's corresponding rulemaking has
been added to the final rule language. The NRC staff does not believe a
definition of the common-usage term shipment is warranted.
Comment. One commenter noted that Sec. 71.15(b) does not identify
what standard is to be used in applying either the term
``noncombustible'' or the term ``insoluble-in-water.'' The commenter
stated that if this section is kept as proposed, there is a need to
clarify the terms and specify an appropriate standard.
Response. The text from the proposed rule has changed. Rather than
clarify the words ``noncombustible'' and ``insoluble-in-water,'' the
new text indicates only the need for the nonfissile material to be a
``solid.'' The NRC believes that new definitions are not necessary.
Comment 13. One commenter requested that NRC delete the proposed
exemptions for plutonium-244 in proposed Sec. 71.14(b)(1) because there
are no special form plutonium-244 sources available.
Response: Section 71.14(b)(1) was changed to provide clarification
and simplification of the language that existed in the current
regulation (Sec. 71.10), while retaining the substance of the
exemption. The current Sec. 71.10 (b)(1) exempts shipments that contain
no more than a Type A quantity of radioactive material from all of the
requirements of part 71, except for Sec.Sec. 71.5 and 71.88. Similarly,
Sec. 71.10(b)(3) exempts domestic shipments that contain less than an
aggregate 20 Curies (Ci) of special form americium or plutonium from
all of the requirements of part 71, except for Sec.Sec. 71.5 and 71.88.
The current Type A (A1) limit for plutonium-244 is 8 Ci. The
rule raises the A1 limit for plutonium-244 to 11 Ci--still
less than the 20-Ci exemption of the current Sec. 71.10(b)(3).
Consequently, for plutonium-244, the two exemption criteria of the
current Sec. 71.10(b)(1) and (b)(3) were in conflict. The NRC's
proposed rule resolved that conflict. The commenter's proposed solution
would retain that conflict. Accordingly, absent a substantive basis for
changing the proposed rule, the NRC is retaining the existing 20-Ci
exemption for domestic shipments of special form americium or plutonium
in Sec. 71.14(b)(1) in this final rule. Furthermore, because the
A1 limits for all other nuclides of plutonium are greater
than 20 Ci, only plutonium-244 is mentioned in paragraph (b)(1).
Comment. Two commenters asserted that the regulations are overly
complex and inconsistent with international regulations. One commenter
agreed with NRC's proposal to change the requirements for fissile
material shipments, but did have several objections. The three primary
objections were that NRC hadn't adequately defined the terms to
categorize the ratio calculations; information on the materials,
necessary to perform calculations, is difficult to obtain; and the
proposal is overly complex and inconsistent with international
regulations. This same commenter stated that the proposed rule does not
adequately account for both packages of large volume and packages of
small volume. The proposed changes do not provide for the ability to
ship large volumes of decommissioning waste in an effective manner and
will complicate international trade of fissile exempt materials.
Furthermore, the proposed ratio control is inadequate, and NRC should
define ``insoluble in water.'' The commenter recommended inclusion of
the TS-R-1 provisions for fissile exempt
[[Page 3752]]
materials. Lastly, the commenter stated that, while NRC should go
forward with the rulemaking, it should work with industry to determine
operational limits that will assure that the mass or concentration
limit is maintained under accident conditions.
Response. The staff has reviewed the proposed rule language and has
determined that section Sec. 71.15(d) was not consistent with the
language in TS-R-1 and has been revised. The commenter should note,
that the intent for this rule change is to provide greater flexibility
in transportation with a concomitant improvement of a shipper's
knowledge about the contents of materials in the package. The rule has
been revised to address the concerns about shipments of very small
quantities of fissile material in small packages and shipment of low
concentrations of fissile material where the large volume of the
container and mass of nonfissile material might enable one to exceed
the fissile limit in the proposed rule. The IAEA is currently
considering changes to the current international regulations in the
area of the fissile material exemptions. The concept put forward in the
current rule is one of those under consideration. The other option
proposed to the IAEA to provide safety in the event of uncontrolled
accumulation of fissile exempt packages is to implement a CSI for all
packages containing fissile material. The NRC considered both options
and chose to implement the option that did not require a CSI on fissile
exempt packages.
Comment. One commenter expressed concern that NRC's proposal to add
atomic ratio criteria to the previously used 15-g \235\U mass criterion
may restrict exemption of fissile materials, not containing special
moderators, that are currently acceptable. Another commenter expressed
support for the concept of exemptions for fissile material shipments
under specific conditions. However, the commenter said that NRC's
proposal in Sec. 71.15 was overly conservative and resulted in a
reduction in the limits of fissile material content without
justification.
Response. The NRC staff agrees, in part, with these comments.
Proposed Sec. 71.15(c)(1) has been modified by removing the limit of
350 g in a package and instead specifies criteria for commingling of
the material such that, within any selected 360 kg of nonfissile solid
material, there can be no more than 180 g of fissile material. Thus, a
large rail car with a homogenized distribution of fissile material
within a nonfissile waste matrix might exceed the 180-g limit but would
be effectively mixed at low enough concentration to enable safe
shipment. In the case of small sample shipments, a limit of 2 g per
package has been added to Sec. 71.15(a) and applies without regard to
any mass ratios.
Comment. One commenter stated that the proposed fissile material
exemptions do not agree with the TS-R-1 exemptions and appear to
contain requirements that are not necessary for nuclear criticality
safety. This commenter also expressed concern about the discontinuance
of the exemption for material containing less than 5 grams of uranium-
235 per 10-liter volume and its impact on shipments related to
decommissioning activities. The commenter also voiced support for the
proposed new limit of 350 g of fissile material with a 2000:1 ratio to
noncombustible and insoluble-in-water material.
Response. The NRC staff acknowledges the comment of support for one
of the proposed changes. Regarding the comment about the exemption
discontinuance, the commenter did not provide any detailed
justification for this concern; thus, no change has been made to the
rule language. As stated above, the NRC has determined for a number of
issues that it does not harmonize completely with all changes made in
the IAEA guidance documents based on safety and other technical
reasons.
Issue 17. Decision on Petition for Rulemaking on Double Containment of
Plutonium (PRM-71-12)
Summary of Decision on PRM-71-12. Currently in 10 CFR 71.63(b),
plutonium in excess of 0.74 TBq (20 Ci) must be packaged in a separate
inner container placed within an outer packaging. This is referred to
as double containment. It is the combination of the inner container and
the outer packaging that is subjected to the normal conditions of
transport (Sec. 71.71) and the hypothetical accident conditions (Sec.
71.73). Upon application of the normal conditions of transport and
hypothetical accident conditions, the acceptance criteria for
shielding, containment, and sub-criticality in Sec. 71.51 must be also
met for the total package (inner container and outer packaging), but
the containment dispersal acceptance (10-6 A2/
hour or 1 A2/week) are applied to each boundary (i.e., the
inner container and the outer packaging). Note however, as a point of
clarification, double containment does not mean two Type B containers
nested into one.
The final rule grants the petitioner's request to remove the double
containment requirement of Sec. 71.63(b). However, the requirement of
Sec. 71.63(a) that shipments whose contents contain greater than 0.74
TBq (20 Ci) of plutonium must be made with the contents in solid form
is retained. Thus, the petitioner's alternative proposal is denied.
This completes action on PRM-71-12.
The NRC has decided to remove the double containment requirement
because this regulation is neither risk-informed nor performance-based.
There are many nuclides with A2 values the same or lower
than plutonium's for which double containment has never been required.
Thus, requiring double containment for plutonium alone is not
consistent with the relative hazard rankings in Table A-1. The Type B
packaging standards, which the outer containment of plutonium shipments
must meet, in and of themselves, provide reasonable assurance that
public health and safety and the environment are protected during the
transportation of radioactive material. This position is supported by
an excellent safety record in which no fatalities or injuries have been
attributed to material transported in a Type B package. The imposition
of an additional packaging requirement (in the form of a separate inner
container) is fundamentally inconsistent with this position and is
technically unnecessary to assure safe transport. Further, removal of
this requirement will reduce an unnecessary regulatory burden on
licensees, will likely result in reduced risk to radiation workers, and
will serve to harmonize part 71 with TS-R-1.
On the other hand, the imposition of the requirement that plutonium
in excess of 0.74 TBq (20 Ci) per package be shipped as a solid does
not create a regulatory inconsistency with the Type B package
standards. The NRC considers the contents of a package when it is
evaluating the adequacy of a packaging's design. The approved content
limits and the approved packaging design together define the CoC for a
package. However, other than criticality controls and the solid form
requirement of Sec. 71.63(a), subparts E and F do not contain any
restrictions on the contents of a package. Thus, while the inner
containment requirement in Sec. 71.63(b) can be seen as conflicting
with the Type B package standard because the inner containment affects
the packaging design, the solid form requirement of Sec. 71.63(a) does
not conflict with the packaging requirements of the Type B package
standard because the solid form requirement affects only the contents
of the package, not the packaging itself.
Affected Sections. Section 71.63.
[[Page 3753]]
Discussion of PRM-71-12: The NRC received a petition for rulemaking
from International Energy Consultants, Inc. (IEC), dated September 25,
1997. The petition was docketed as PRM-71-12 and was published for
public comment (63 FR 8362; February 19, 1998). Based on a request from
General Atomic, the comment period was extended to July 31, 1998 (see
63 FR 34335; June 24, 1998). Nine public comments were received on the
petition. Four commenters supported the petition, and five commenters
opposed the petition.
The petitioner requested that Sec. 71.63(b) be removed. The
petitioner argued that the double containment provisions of Sec.
71.63(b) cannot be supported technically or logically. The petitioner
stated that based on the ``Q-system for the Calculation of
A1 and A2 Values,'' an A2 quantity of
any radionuclide has the same potential for damaging the environment
and the human species as an A2 quantity of any other
radionuclide.
The NRC believes that the Q-values are based upon radiological
exposure hazard models which calculate the allowable quantity limit
(the A1 or A2 value) necessary to produce a known
exposure (i.e., one A2 of plutonium-239 or one A2
of cobalt-60 will both yield the same radiation dose under the Q-system
models, even though the A2 values for these nuclides are
different (e.g., one A2 of plutonium-239 = 2 x
10-4 TBq, and one A2 of cobalt-60 = 1 TBq). The
Q-system models take into account the exposure pathways of the various
radionuclides, typical chemical forms of the radionuclide, methods for
uptake into the body, methods for removal from the body, the type of
radiation the radionuclide emits, and the bodily organs the
radionuclide preferentially affects. The specific A1 and
A2 values for each nuclide are developed using radiation
dosimetry approaches recommended by the World Health Organization and
the ICRP. The models are periodically reviewed by international health
physics experts (including representatives from the United States), and
the A1 and A2 values are updated during the IAEA
revision process, based upon the best available data. (Note that
changes to the A1 and A2 values as a result of
changes to the models in TS-R-1 are also discussed in Issue 3 of this
rule.) These values are then issued by the IAEA in safety standards
such as TS-R-1. When the IAEA has revised the A1 and
A2 values in previous revisions of its transport
regulations, these revised values have been adopted by the NRC and DOT
into the transportation regulations in 10 CFR part 71 and 49 CFR part
173, respectively.
NRC's review of the current A1 and A2 values
in Appendix A to part 71, Table A-1, reveals that 5 radionuclides have
an A2 value lower than plutonium (i.e., plutonium-239), and
11 radionuclides have an A2 value that is equal to
plutonium-239. Because the models used to determine the A1
and A2 values all result in the same radiation exposure
(i.e., hazard), a smaller A1 and A2 value for one
radionuclide would indicate a greater potential hazard to humans than a
radionuclide with a larger A1 and A2 value. Thus,
overall, Table A-1 can also be viewed as a relative hazard ranking (for
transportation purposes) of the listed radionuclides. In that light,
requiring double containment for plutonium alone is not consistent with
the relative hazard rankings in Table A-1.
The petitioner also argued that the Type B package requirements
should be applied consistently for any radionuclide, whenever a
package's contents exceed an A2 limit. However, part 71 is
not consistent by imposing the double containment requirement for
plutonium. The petitioner believes that if Type B package standards are
sufficient for a quantity of a particular radionuclide which exceeds
the A2 limit, then Type B package standards should also be
sufficient for any other radionuclide which also exceeds the
A2 limit. The petitioner stated that:
While, for the most part, part 71 regulations embrace this simple
logical congruence, the congruence fails under 10 CFR 71.63(b) wherein
packages containing plutonium must include a separate inner container
for quantities of plutonium having a radioactivity exceeding 20 curies
(0.74 TBq) (with certain exceptions).
The petitioner further stated that:
If the NRC allows this failure of congruence to persist, the
regulations will be vulnerable to the following challenges: (1) The
logical foundation of the adequacy of A2 values as a proper
measure of the potential for damaging the environment and the human
species, as set forth under the Q-System, is compromised; (2) the
absence of a limit for every other radionuclide which, if exceeded,
would require a separate inner container, is an inherently inconsistent
safety practice; and (3) the performance requirements for Type B
packages, as called for by 10 CFR part 71, establish containment
conditions under different levels of package trauma. The satisfaction
of these Type B package standards should be a matter of proper design
work by the package designer and proper evaluation of the design
through regulatory review. The imposition of any specific package
design feature such as that contained in 10 CFR 71.63(b) is gratuitous.
The regulations are not formulated as package design specifications,
nor should they be.
The NRC agrees that the part 71 regulations are not formulated as
package design specifications; rather, the part 71 regulations
establish performance standards for a package's design. The NRC reviews
the application to evaluate whether the package's design meets the
performance requirements of part 71. Consequently, the NRC can then
conclude that the design of the package provides reasonable assurance
that public health and safety and the environment are adequately
protected.
The petitioner also believes that the continuing presence of Sec.
71.63(b) engenders excessively high costs in the transport of some
radioactive materials without a clearly measurable net safety benefit.
The petitioner stated that this is so, in part, because the ultimate
release limits allowed under part 71 package performance requirements
are identical with or without a ``separate inner container,'' and
because the presence of a ``separate inner container'' promotes
additional exposures to radiation through the additional handling
required for the ``separate inner container.'' Consequently, the
petitioner asserted that the presence or absence of a separate inner
container barrier does not affect the standard to which the outer
container barrier must perform in protecting public health and safety
and the environment. Therefore, the petitioner concluded that given
that the outer containment barrier provides an acceptable level of
safety, the separate inner container is superfluous and results in
unnecessary cost and radiation exposure. According to the petitioner,
these unnecessary costs involve both the design, review, and
fabrication of a package, as well as the costs of transporting the
package. And the unnecessary radiation exposure involves workers having
to handle (i.e., seal, inspect, or move) the ``separate inner
container.''
As an alternative to the primary petition, the petitioner believes
that an option to eliminate both Sec. 71.63(a) and (b) should also be
considered. Section 71.63(a) requires that plutonium in quantities
greater than 0.74 TBq (20 Ci) be shipped in solid form. This option
would have the effect of removing Sec. 71.63 entirely. The petitioner
believes that the arguments set forth to support the elimination of
Sec. 71.63(b) also support the elimination of Sec. 71.63(a).
[[Page 3754]]
The petitioner did not provide a separate regulatory or cost analysis
supporting the request to remove Sec. 71.63(a).
History of the Double Containment Requirement: On June 17, 1974 (39
FR 20960), the AEC issued a final rule which imposed special
requirements on the shipment of plutonium. These requirements are
located in Sec. 71.63 and apply to shipments of radioactive material
containing quantities of plutonium in excess of 0.74 TBq (20 curies).
Section 71.63 contains two principal requirements. First, the plutonium
contents of the package must be in solid form [Sec. 71.63(a)]. Second,
the packaging containing the plutonium must provide a separate inner
containment (i.e., the ``double containment'' requirement) [Sec.
71.63(b)]. In addition, the AEC specifically excluded from the double
containment requirement of Sec. 71.63(b) plutonium in the form of
reactor fuel elements, metal or metal alloys, and other plutonium-
bearing solids that the Commission (AEC or NRC) may determine, on a
case-by-case basis, do not require double containment. This regulation
remained essentially unchanged from 1974 until 1998, when vitrified
high-level waste in sealed canisters was added to the list of exempt
forms of plutonium in Sec. 71.63(b) (63 FR 32600; June 15, 1998). The
double containment requirement is in addition to the existing 10 CFR
part 71 subparts E and F requirements imposed on Type B packagings
(e.g., the normal conditions of transport and hypothetical accident
conditions of Sec.Sec. 71.71 and 71.73, respectively, and the fissile
package requirements of Sec.Sec. 71.55 and 71.59). Part 71 does not
impose a double containment requirement for any radionuclide other than
plutonium. Additionally, IAEA standard TS-R-1 does not provide for a
double containment requirement (in lieu of the single containment Type
B package standards) for any radionuclide.
The AEC issued this regulation at a time when AEC staff anticipated
widespread reprocessing of commercial spent fuel, and existing
shipments of plutonium were made in the form of liquid plutonium
nitrate. Because of physical changes to the plutonium that was expected
to be reprocessed (i.e., higher levels of burnup in commercial reactors
for spent fuel, which would then be reprocessed), and regulatory
concerns with the possibility of package leakage, the AEC issued a
regulation that imposed the double containment requirement when the
package contained more than 0.74 TBq (20 Ci) of plutonium. This double
containment was in addition to the existing Type B package standards on
packages intended for the shipment of greater than an A1 or
A2 quantity of plutonium.
The NRC staff has reviewed the available regulatory history for
Sec. 71.63, and has provided a recapitulation of the supporting
information which led to the issuance of this regulation. The NRC staff
has extracted the following information from several SECY papers the
AEC staff submitted to the Commission on this regulation. The NRC staff
believes this information is relevant and will provide stakeholders
with perspective in understanding the bases for this regulation, and
thereby assist stakeholders in evaluating the staff's proposed changes
to this regulation.
In SECY-R-702,\6\ the AEC staff identified two considerations that
were the genesis of the rulemaking that led to Sec. 71.63. AEC staff
stated:
---------------------------------------------------------------------------
\6\ SECY-R-702, ``Consideration of Form for Shipping
Plutonium,'' June 1, 1973.
---------------------------------------------------------------------------
First, increasingly larger quantities of plutonium will be
recovered from power reactor spent fuel. Second, the specific activity
of the plutonium will increase with higher reactor fuel burnup
resulting in greater pressure generation potential from plutonium
nitrate solutions in shipping containers, greater heat generation, and
higher gamma and neutron radiation levels. These changes will make the
present nitrate packages obsolete. Thus, from both safety and economic
considerations, the transportation of plutonium as [liquid] nitrate
will soon require substantial redesign of packages to handle larger
quantities as well as to deal with the higher levels of gas evolution
(pressurization), heat generation, and gamma and neutron radiation.
There is little doubt that larger plutonium nitrate packages could
be designed to meet regulatory standards. The increased potential for
human error and the consequences of such error in the shipment of
plutonium nitrate are not so easily controlled by regulation. Even
though such packages may be adequately designed, their loading and
closure requires high operation performance by personnel on a
continuing basis. As the number of packages to be shipped increases,
the probability of leakage through improperly assembled and closed
packages also increases. * * * More refined or stringent regulatory
requirements, such as double containment, would not sufficiently lessen
this concern because of the necessary dependence on people to affect
engineered safeguards.
In SECY-R-74-5,\7\ AEC staff summarized the factors relevant to
consideration of a proposed rule following a June 14, 1973, meeting to
discuss SECY-R-702, between the Regulatory and General Manager's staffs
(i.e., the rulemaking and operational sides of the AEC). The AEC
stated:
---------------------------------------------------------------------------
\7\ SECY-R-74-5, ``Consideration of Form for Shipping
Plutonium,'' dated July 6, 1973.
---------------------------------------------------------------------------
As a result of this meeting (on June 14, 1973), the (Regulatory and
General Manager's) staffs have agreed that the basic factors pertinent
to the consideration of form for shipment of plutonium are:
1. The experience with shipping plutonium as an aqueous nitrate
solution in packages meeting current regulatory criteria has been
satisfactory to date.
2. The changing characteristic of plutonium recovered from power
reactors will make the existing packaging obsolete for plutonium
nitrate solutions and possibly for solid form. Economic factors will
probably dictate considerably larger shipments (and larger packages)
than currently used.
3. It is expected that packages can be designed to meet regulatory
standards for either aqueous solutions or solid plutonium compounds.
Just as in any situation involving the packaging of radioactive
materials, a high level of human performance is necessary to assure
against leakage caused by human error in packaging. As the number of
plutonium shipments increases, as it will, and packages become larger
and more complex in design, the probability of such human error
increases.
4. The probability of human error with the packaging for liquid,
anticipated to be more complex in design, is probably greater than with
the packaging for solid. Furthermore, should a human error occur in
package preparation or closure, the probability of liquid escaping from
the improperly prepared package is greater than for most solids and
particularly for solid plutonium materials expected to be shipped.
5. Staff studies reported in SECY-R-62 and SECY-R-509 \8\ conclude
that the consequences of release of solid or aqueous solutions do not
differ appreciably. Therefore, this paper (SECY-R-702) does not deal
with the consequences of releases.
---------------------------------------------------------------------------
\8\ SECY-R-62, ``Shipment of Plutonium,'' and SECY-R-509,
``Plutonium Handling and Storage,'' dated October 16, 1970. These
papers concluded that there is no scientific or technical reason to
prohibit shipment of plutonium nitrate and recommended that
Commission (AEC) efforts be directed toward providing improved
safety criteria for shipping containers.
---------------------------------------------------------------------------
[[Page 3755]]
6. It is, therefore, concluded that safety would be enhanced if
plutonium were shipped as a solid rather than in solution.
The arguments for requiring a solid form of plutonium for shipment
are largely subjective, in that there is no hard evidence on which to
base statistical probabilities or to assess quantitatively the
incremental increase in safety which is expected. The discussion in the
regulatory paper, SECY-R-702, is not intended to be a technical
argument which incontrovertibly leads to a conclusion. It is, rather, a
presentation of the rationale which has led the Regulatory staff to its
conclusion that a possible problem may develop and that the proposed
action is a step towards increased assurance against the problem
developing. In SECY-R-74-172,\9\ AEC staff submitted a final rule to
the Commission for approval.
---------------------------------------------------------------------------
\9\ SECY-R-74-172, ``Consideration of Form for Shipping
Plutonium,'' April 18, 1974.
---------------------------------------------------------------------------
The proposed rule had contained a requirement that the plutonium be
contained in a special form capsule. However, in response to comments
from the AEC General Manager, the final rule changed this requirement
to a separate inner container (i.e., the double containment
requirement). The AEC staff indicated in a response to a public comment
in Enclosure B (to SECY-R-74-172) that ``[t]he need for the inner
containment is based on the desire to provide a substitute for not
requiring the plutonium to be in a `nonrespirable' form.''
The regulatory history of Sec. 71.63 indicates that the AEC's
decision to require a separate inner container for shipments of
plutonium in excess of 0.74 TBq (20 Ci) was based on existing policy
and regulatory concerns (i.e., ``that a possible problem may develop
and that the proposed action [in SECY-R-702] is a step towards
increased assurance against the problem developing''). Because of the
expectation of a significant increase in the number of liquid plutonium
nitrate shipments, the AEC used a defense-in-depth philosophy (i.e.,
the double containment and solid form requirements), to ensure that
respirable plutonium would not be released to the environment during a
transportation accident. However, the regulatory history does indicate
that the AEC's concerns did not involve the adequacy of existing liquid
plutonium nitrate packages. Rather, the AEC's regulatory concern was on
the increased possibility of human error combined with an expected
increase in the number of shipments that would yield an increased
probability of leakage during shipment. The AEC's policy concern was
based on an economic decision on whether the AEC should require the
reprocessing industry to build new, larger liquid plutonium-nitrate
shipping containers, capable of handling higher burnup reactor spent
fuel, or to build new, dry, powdered plutonium-dioxide shipping
containers. The regulatory history indicates that the AEC staff judged
that new, larger, higher burnup-capacity liquid plutonium-nitrate
packages could be designed, approved, built, and safely used. However,
one of the AEC's principal underlying assumptions for this rule was
obviated in 1979 when the Carter administration decided that
reprocessing of civilian spent fuel and reuse of plutonium was not
desirable. Consequently, the expected plutonium reprocessing economy
and widespread shipments of liquid plutonium nitrate within the U.S.
never materialized.
On June 15, 1998 (63 FR 32600), in response to a petition for
rulemaking submitted by DOE (PRM-71-11) (February 18, 1994; 59 FR
8143), the Commission issued a final rule revising Sec. 71.63(b) to add
vitrified high-level waste (HLW) contained in a sealed canister to the
list of forms of plutonium exempt from the double containment
requirement (June 15, 1998; 63 FR 32600). In its original response to
PRM-71-11, NRC proposed in SECY-96-215 \10\ to make a ``determination''
under Sec. 71.63(b)(3) that vitrified HLW contained in a sealed
canister did not require double containment. However, the Commission in
an SRM on SECY-96-215, dated October 31, 1996, disapproved the staff's
approach and directed that resolution of this petition be addressed
through rulemaking (the June 15, 1998, final rule was the culmination
of this effort). In addition to disapproving the use of a
``determination'' process, the Commission also directed the staff to
``* * * also address whether the technical basis for 10 CFR 71.63
remains valid, or whether a revision or elimination of portions of 10
CFR 71.63 is needed to provide flexibility for current and future
technologies.'' In SECY-97-218,\11\ NRC responded to the SRM's
direction and stated ``[t]he technical basis remains valid and the
provisions provide adequate flexibility for current and future
technologies.''
---------------------------------------------------------------------------
\10\ SECY-96-215, ``Requirements for Shipping Packages Used to
Transport Vitrified Waste Containing Plutonium,'' dated October 8,
1996.
\11\ SECY-97-218, ``Special Provisions for Transport of Large
Quantities of Plutonium (Response to Staff Requirements Memorandum--
SECY-96-215),'' dated September 29, 1997.
---------------------------------------------------------------------------
Summary of Comments Received on the Petition (PRM-71-12): Nine
public comments were received on the petition (petition was published
for public comment in 63 FR 8362; February 19, 1998). Four commenters
supported the petition, and five commenters opposed the petition. The
four commenters supporting the petition essentially stated that the
IAEA's Q-system accurately reflects the dangers of radionuclides,
including plutonium, and that elimination of Sec. 71.63(a) and (b)
would make the regulations more performance based, reduce costs and
personnel exposures, and be consistent with the IAEA standards.
The five commenters opposing the petition essentially stated that:
(1) Plutonium is very dangerous, especially in liquid form, and
therefore additional regulatory requirements are warranted; (2)
existing regulations are not overly burdensome, especially in light of
the total expected transportation cost; (3) TRUPACT-II packages meet
current Sec. 71.63(b) requirements (TRUPACT-II is a package developed
by DOE to transport transuranic wastes (including plutonium) to the
Waste Isolation Pilot Plant (WIPP) and has been issued a part 71 CoC,
No. 9218); (4) a commenter (the Western Governors' Association) has
worked for over 10 years to ensure a safe transportation system for
WIPP, including educating the public about the TRUPACT-II package; (5)
any change now would erode public confidence and be detrimental to the
entire transportation system for WIPP shipments; and (6) additional
personnel exposure due to double containment is insignificant.
Analysis of Public Comments on the Issues Paper: The NRC has
received 48 public comments on this issue in response to the issue
paper, in subsequent public meetings, and the workshop (the issues
paper was published at 65 FR 44360; July 17, 2000). Industry
representatives and some members of the public support the petition.
Public interest organizations, Agreement States and State
representatives, and the Western Governors' Association, and other
members of the public oppose the petition. Several commenters expressed
their belief that Congress, in approving the Waste Isolation Pilot
Plant Land Withdrawal Act (the Act), Pub. L. 102-579 (106 Stat. 4777),
section 16(a), which mandates that the NRC certify the design of
packages used to transport transuranic waste to WIPP, expected those
packages to have a double containment. The NRC researched this
[[Page 3756]]
issue and found that section 16(a) of the Act does not contain any
explicit provisions mandating the use of a double containment in
packages transporting transuranic waste to or from WIPP. Section 16(a)
of the Act states, in part, ``[n]o transuranic waste may be transported
by or for the Secretary [of the DOE] to or from WIPP, except in
packages the design of which has been certified by the Nuclear
Regulatory Commission * * *'' Furthermore, the NRC has reviewed the
legislative history \12\ associated with the Act and has not identified
any discussions on the use of double containment for the shipment of
transuranic waste. The legislative history does mention that the design
of these packages will be certified by the NRC; however, this language
is identical to that contained in the Act itself. Therefore, the NRC
believes the absence of specific language in section 16(a) of the Act
requiring double containment should be interpreted as requiring the NRC
to apply its independent technical judgment in establishing standards
for package designs and in evaluating applications for certification of
package designs, to ensure that such packages would provide reasonable
assurance that public health and safety and the environment would be
adequately protected. In carrying out its mission, the courts have
found that the NRC has broad latitude in establishing, maintaining, and
revising technical performance criteria necessary to provide reasonable
assurance that public health and safety and the environment are
adequately protected. An example of these technical performance
criteria is the Type B package design standards. Accordingly, the NRC
believes that the proposed revision of a technical package standard
(i.e., removal of the double containment requirement for plutonium from
the Type B package standards) is not restricted by the mandate of
section 16(a) of the Act for the NRC to certify the design of packages
intended to transport transuranic material to and from WIPP.
---------------------------------------------------------------------------
\12\ See Congressional Record Vol. 137, November 5, 1991, pages
S15984-15997 (Senate approval of S. 1671); Cong. Rec. Vol. 138, July
21, 1992, pages H6301-6333 (House approval of H.R. 2637); Cong. Rec.
Vol. 138, October 5, 1992, pages H11868-11870 (House approval of
Conference Report on S. 1671); Cong. Rec. Vol. 138, October 8, 1992
(Senate approval of Conference Report on S. 1671); and Cong. Rec.
Vol. 138, October 5, 1992, pages H12221-12226 (Conference Report on
S. 1671-H. Rpt. 102-1037).
---------------------------------------------------------------------------
Other commenters stated that stakeholders' expectations were that
packages intended to transport transuranic material to and from WIPP
would include a double containment provision. Consequently, the
commenters expressed a belief that removal of the double containment
requirement would decrease public confidence in the NRC's
accomplishment of its mission in the approval of the design of packages
for the transportation of transuranic waste to and from WIPP. The
commenters stated that the public would view elimination of the double
containment requirement as a relaxation in safety. The presence of a
separate inner container provides defense-in-depth through an
additional barrier to the release of plutonium during a transportation
accident, according to commenters. In addition, the commenters stated
that plutonium is so inherently deadly, that defense-in-depth is
appropriate. The NRC agrees that a double containment does provide an
additional barrier. However, the NRC believes that, for the reasons
discussed below, double containment is unnecessary to protect public
health and safety. The NRC and AEC have not required an additional
containment barrier for Type B packages transporting any radionuclides
other than plutonium and, before 1974, the AEC did not require double
containment for plutonium.
In response to some of the comments opposed to the petition, the
NRC believes that removal of Sec. 71.63(b) would not invalidate the
design of existing packages intended for the shipment of plutonium.
These packages could continue to be used with a separate inner
container. The NRC agrees with the commenters that a quantitative cost
analysis was not provided by the petitioner.
The NRC has issued part 71 CoC No. 9218 to DOE for the TRUPACT-II
package (Docket No. 71-9218), for the transportation of transuranic
waste (including plutonium) to and from the WIPP. The TRUPACT-II
package complies with the current Sec. 71.63(b) requirements and has a
separate inner container. The TRUPACT-II SAR indicates that the weight
of the inner container and its lid is approximately 2,620 lbs.
Hypothetically, elimination of the separate inner container would
increase the available payload for the TRUPACT-II package from the
current 7,265 to 9,885 lbs. Thus, removal of the double containment
requirement would potentially increase the TRUPACT-II's available
payload by 36 percent. Further, the removal of the inner container from
the TRUPACT-II would also potentially increase the available volume.
The NRC believes that the final rule would not invalidate the existing
TRUPACT-II design (i.e., it would still meet all remaining applicable
requirements of part 71). Thus, DOE could continue to use the TRUPACT-
II to ship transuranic waste to and from WIPP, or DOE could consider an
alternate Type B package.
Additionally, based on comments received in the public meetings,
the NRC believes that a misperception exists with respect to TRUPACT-II
shipments; removal of the Sec. 71.63(b) double containment requirement
would not result in loose plutonium waste being placed inside a
TRUPACT-II package. Based upon information contained in the SAR,
plutonium wastes (i.e., used gloves, anti-Cs, rags, etc.) are placed in
plastic bags, and these bags are sealed inside lined 55-gallon steel
drums. Plutonium residues are placed inside cans which are then sealed
inside a pipe overpack (a 6-inch or 12-inch stainless steel cylinder
with a bolted lid), and the pipe overpack is then sealed inside a lined
55-gallon steel drum. The 55-gallon drums are then sealed inside the
TRUPACT-II inner containment vessel, and finally the inner containment
vessel is sealed inside the TRUPACT-II package. Consequently, the
TRUPACT-II shipping practices employ multiple barriers and would
continue to do so. Removal of the inner containment vessel would not be
expected to produce a significant incremental increase in the
possibility of leakage during normal transportation. The NRC notes that
some NRC regulations have established additional requirements for
plutonium (e.g., the special nuclear material license application
provisions of Sec. 70.22(f)).
The NRC believes that the Type B packaging standards, in and of
themselves, provide reasonable assurance that public health and safety
and the environment would be adequately protected during the
transportation of radioactive material. This belief is supported by an
excellent safety record in which no fatalities or injuries have been
attributed to material transported in a Type B package. Type B
packaging standards have been in existence for approximately 40 years
and have been incorporated into the part 71 regulations by both the NRC
and its predecessor, the AEC. The NRC's Type B package standards are
based on IAEA's Type B package standards. Moreover, IAEA's Type B
package standards have never required a separate inner container for
packages intended to transport plutonium, nor for any other
radionuclide.
Therefore, the NRC believes that imposition of an additional
packaging
[[Page 3757]]
requirement (in the form of a separate inner container) is
fundamentally inconsistent with the position that Type B packaging
standards, in and of themselves, provide reasonable assurance that
public health and safety and the environment would be adequately
protected during the transportation of (any type of) radioactive
material. Thus, the NRC believes that maintaining Sec. 71.63(b) is not
consistent with the other existing Type B packaging standards contained
in part 71.
The NRC also believes that the regulatory history of Sec. 71.63
demonstrates that the AEC's decision to add this section was based on
policy and regulatory concerns. However, the NRC also agrees that the
use of a double containment does provide defense-in-depth and does
decrease the absolute risk of the release of respirable plutonium to
the environment during a transportation accident. Consequently, while
the defense-in-depth afforded by a double containment does reduce risk,
the NRC believes the question which should be focused on is whether the
double containment requirement is risk-informed. The NRC is unaware of
any risk studies that would provide a quantitative indication of the
risk reduction associated with the use of an NRC-certified double
containment packaging in transportation of plutonium. Rather, the NRC
would look to the demonstrated performance record of existing Type B
package standards to conclude that double containment is not necessary.
In summary, the AEC indicated (in SECY-R-702 and SECY-R-74-5) that
liquid plutonium nitrate packages were safe, and new, larger packages
to handle higher burnup reactor spent fuel could also be designed. NRC
believes that the AEC's assumption for initiating this requirement was
that large scale reprocessing of civilian reactor spent fuel and reuse
of plutonium would occur. The decision of former President Carter's
administration to forgo the reprocessing of civilian reactor spent fuel
and reuse of plutonium obviated the AEC's assumption. Consequently, the
AEC's supposition that a human error occurring while sealing a package
of liquid plutonium nitrate was more likely to occur with the expected
increase in shipments of plutonium nitrate was also obviated by the
Government's decision to forgo the reprocessing of civilian reactor
spent fuel. In SECY-97-218, NRC staff indicated that the separate inner
container provided an additional barrier to the release of plutonium in
an accident. NRC continues to believe that a separate inner container
provides an additional barrier to the release of plutonium in an
accident, just as a package with triple containment would provide an
even greater barrier to the release of plutonium in an accident.
However, this type of approach is neither risk informed nor performance
based. Consequently, based upon review of the petition, comments on the
petition, and research into the regulatory history of the double
containment requirement, the NRC agrees that a separate inner container
is not necessary for Type B packages containing solid plutonium. NRC
believes that the worldwide performance record over 40 years of Type B
packages demonstrates that a single containment barrier is adequate.
Therefore, the NRC agrees with the petitioner and believes that Sec.
71.63(b) is not technically necessary to provide a reasonable assurance
that public health and safety and the environment will be adequately
protected during the transportation of plutonium.
While the NRC believes a case can be made for elimination of the
separate inner container requirement in Sec. 71.63(b), elimination of
the solid form requirement in Sec. 71.63(a) is not as clear. While the
same arguments can be made on the obviation of the AEC's basis for
originally issuing Sec. 71.63(a) (i.e., the elimination of reprocessing
of plutonium), the same regulatory inconsistency between Type B package
standards and the inner containment requirement does not exist for the
liquid versus solid form argument. The NRC considers the contents of a
package when it is evaluating the adequacy of a packaging's design. The
approved content limits and the approved packaging design together
define the CoC for a package. However, other than criticality controls
and the liquid form requirement of Sec. 71.63(a), 10 CFR part 71
subparts E and F do not contain any restrictions on the contents of a
package. Thus, while the inner containment requirement in Sec. 71.63(b)
can be seen as conflicting with the Type B package standard because the
inner containment affects the packaging's design, the solid form
requirement of Sec. 71.63(a) does not conflict with the packaging
requirements of the Type B package standard because the solid form
requirement affects only the contents of the package, not the packaging
itself.
The NRC expects that cost and dose savings would accrue from the
removal of Sec. 71.63(b). However, because no shipments of liquid
plutonium nitrate are contemplated in the U.S., NRC would not expect
cost or dose savings to accrue from the removal of Sec. 71.63(a), if
that section were to be also removed. Further, the AEC's original bases
have been obviated by former President Carter's administration's
decision to not pursue a commercial fuel cycle involving the
reprocessing of plutonium.
After weighing this information, the NRC continues to believe that
the Type B package standards, when evaluated against 40 years of use
worldwide, and millions of safe shipments of Type B packages, together
provide reasonable assurance that public health and safety and the
environment would be adequately protected during the transportation of
radioactive material. The NRC believes that, in this case, the
reasonable assurance standard, provided by the Type B package
requirements, provides an adequate basis for the public's confidence in
the NRC's actions.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. Several commenters suggested that all radioactive
materials should require double packaging. Two of these commenters
stated double containment is a security and safety precaution. A third
stated that existing container requirements are the minimum standards
necessary for safety, security, and public acceptance. Another
commenter simply objected to the removal of the requirement for double
containment of plutonium.
Response. The NRC disagrees with these comments. The NRC has made a
finding that single containment of radioactive material provides an
adequate level of safety for all radioactive materials. The
A1 and A2 value summary found at 67 FR 21422;
April 30, 2002, under the heading Issue 3, provides information that
supports the NRC's basis for this decision. The comments provided no
justification for the double containment requirement for shipment of
all nuclear materials.
Comment. Several commenters were concerned with NRC's proposal to
eliminate double containment. The first of these commenters asked if
there is any basis to eliminate the double containment requirement
other than to harmonize our rules with the IAEA regulations. The second
commenter expressed concern that the ``only benefits from eliminating
double containment * * * would accrue to the DOE, to contractors,
licensees, and shippers in the form of cost savings.'' Furthermore, the
commenter stated that the cost of maintaining transportation
[[Page 3758]]
safety standards should be borne by those in the industry and that
costs should not be ``used as an excuse for deregulation or
exemptions.'' A similar argument was made by another commenter who
urged NRC not to remove Sec. 71.63(b) reasoning that, as noted in the
proposed rulemaking, the petitioner did not provide a quantitative cost
analysis; therefore, the contention that ``presence of Sec. 71.63(b)
engenders excessively high costs'' is unsubstantiated. Another
commenter stated that while an 8-13 percent volume reduction due to
weight restrictions caused by double containment is not trivial, the
benefits from reducing this weight penalty needs to be balanced against
the resulting increase in radiation doses, the increased likelihood of
a release in the event of a severe accident, and the increased cost of
certifying a new package.
Response. The primary reason for removing the double containment
requirement is that the NRC has no technical justification or basis for
maintaining double containment for plutonium or any other radionuclide.
The NRC believes the arguments for removing double containment have
been adequately addressed earlier in this notice and in the proposed
rule under this issue.
While NRC acknowledges that there may be monetary benefits
associated with removing double containment, there are other reasons as
well, including reduction in personnel exposure for those individuals
involved in loading packages for transport. Further, while double
containment does provide an additional barrier against release, the NRC
believes that, for reasons previously explained, double containment is
unnecessary to protect public health and safety. Moreover, NRC has been
and remains committed to providing regulations that are not only risk
informed, but also reduce unnecessary regulatory burden.
Comment. One commenter stated that removing the double containment
requirement would reduce costs of packaging and associated hardware.
The commenter asserted that double containment increases costs without
measurable benefit. The commenter then provided cost information and
discussed the design, certification, and fabrication of future
packaging (e.g., TRUPACT III or the DPP-1 and DPP-2) needed to complete
DOE's Accelerated Cleanup strategy for resolution of the legacy wastes
and materials from the Cold War.
Response. NRC acknowledges the comment.
Comment. Many commenters opposed the elimination of the double
containment requirement because of possible public health and safety
consequences.
Response. The commenters provided no basis for their assertions
that removing the double-containment requirement would increase public
exposure risks. The NRC staff believes that the current Type B package
requirements, as applied to all radionuclides, are adequate to protect
public health and safety.
Comment. One commenter stated that the principal benefit of
removing the double containment requirement would be a reduction in
exposure to the workers. The commenter added that it would also result
in lower costs.
Response. NRC acknowledges the comment.
Comment. One commenter expressed concern that the A1 and
A2 values have been used as a justification for single-shell
containers for plutonium.
Response: The NRC does not agree with this unsubstantiated
statement that the A1 and A2 values have been
used as justification for the elimination of the double containment
requirement for plutonium. The justifications for elimination of the
double containment requirement were detailed in the proposed rule on
April 30, 2002 (67 FR 21421 through 21425), and focus more on the fact
that the original AEC requirement for double containment of plutonium
was based on existing policy and regulatory concerns and was not risk
informed. While the A1 and A2 values are
referenced in the discussion, they are referenced from the standpoint
that there are other radionuclides with the same or lower A1
and A2 values than plutonium. Because these radionuclides
have never required double containment, it cannot be argued from a risk
standpoint that the shipment of plutonium should be treated any
differently.
Comment. Three commenters expressed support for the proposed
removal of the requirement for ``double containment'' of plutonium from
Sec. 71.63. One commenter asserted that a single containment barrier is
adequate for Type B packages containing more than 20 curies of solid
form plutonium. The commenter further stated that the former AEC's
rationale for requiring the double containment provision is now moot
because the expectation for liquid plutonium nitrate shipments has
never materialized. The commenter also expressed opposition to the
double containment requirement because it presents continuing costs
without commensurate benefits. The commenter stated that removing the
double containment requirement would result in a small and acceptable
increase in public risk. Furthermore, the requirement removes
flexibility in package designs that might be needed to meet DOE's
mission.
Another commenter expressed concern that the double containment
requirement was implemented in the 1970s without adequate
justification.
The third commenter said that using double containment causes
unnecessary worker radiation exposure. This commenter said this
unnecessary worker radiation is estimated to be 1200 to 1700 person-rem
over a 10-year period. The commenter also said the conditions that
justified double containment during the early 1970s have disappeared.
These include large numbers of shipments of nitrate solutions or other
forms from reprocessing, compounded by crude containment requirements,
and the absence of quality assurance requirements. This position was
justified because France, Germany, and the United Kingdom, as well as
other IAEA Member Nations, no longer require double containment for
plutonium. The commenter believed that harmonization of part 71 with
IAEA TS-R-1 was an important goal of this rulemaking because to do so
would allow for consistent regulation among the principal nations
shipping nuclear materials. Furthermore, it was recommended that NRC
eliminate the special requirements for plutonium shipments in Sec.
71.63 for consistency with the use of prescriptive, performance-based
safety standards.
Response. The comments are generally in line with statements in the
proposed rule on April 30, 2002 (67 FR 21421 through 21425), that
described the NRC's bases for elimination of the double containment
requirement.
Comment. Several commenters stated that double containment provides
more protection to the public than single containment. One of these
commenters stated the belief that the commenter and a majority of the
Western Governors are concerned with the proposal to eliminate the
double containment requirement for plutonium shipments. The commenter
stated that ``the regulatory analysis is defective in its failure to
recognize likely impacts on the agreement among the Western Governors'
Association, the individual Western States, and DOE for a system of
extra regulatory transportation safeguards, which we believe are at the
heart of both government and public acceptance of the WIPP
transportation program.'' One commenter stated that if
[[Page 3759]]
Sec. 71.63(b) is deleted, there will very likely be some use of single-
contained packages for future WIPP shipments.
Response. With respect to the last commenter's statement, the use
of single containment packages for future shipments is one possible
outcome of the change. NRC acknowledges that agreements between DOE and
States may be impacted by the elimination of the double containment
regulatory requirement. However, any change to NRC regulations that
impact how DOE conducts its transportation operations is a DOE
decision. As such, DOE and the States may need to negotiate and resolve
issues related to DOE's operations.
Comment. One commenter stated that the proposed rule is not risk
informed and does not use a common sense approach. Another commenter
stated strong agreement with this first commenter. Another commenter
recommended that both Sec.Sec. 71.63(a) and (b) be retained but that
the limit be expressed as 0.74 TBq (20 Ci) for the total of all
actinides with A2 values equal to or less than 1.0 x
10-3 TBq (2.7 x 10-2 Ci).
Response. The NRC believes the decision to eliminate double
containment is risk informed and reduces an unnecessary regulatory
burden. In this context, there is adequate actual operating experience
with Type B package shipments to support the Commission's decision to
remove the double containment requirement for plutonium packages. There
are many nuclides with A2 values the same or lower than
plutonium's that have never required double containment.
Further, current NRC regulations state that, in certain
circumstances, plutonium in excess of 0.74 TBq (20 Ci) can be shipped
as a normal form solid without requiring double containment. The
shipment of reactor fuel elements containing plutonium is one example.
Using the most conservative A2 value of 0.00541 Ci, 0.74 TBq
(20 Ci) of plutonium (Pu-238, Pu-239, Pu-240) equates to an
A2 multiple of roughly 3700. In contrast, using 19 risk-
significant nuclides (including Am-241) from a typical single boiling
water reactor spent fuel assembly (reference NUREG/CR-6672,
``Reexamination of Spent Fuel Shipment Risk Estimates,'' page 7-17),
one can calculate a curie content of 148,346 Ci with a cumulative
A2 multiple of just under 790,000 (the assembly also would
contain an A2 multiple of 455,000 of plutonium nuclides). If
the A2 multiple is viewed as a measure of potential health
effect, then from a risk-informed standpoint, the shipment of one
particular nuclide in a Type B package should not be treated
differently from any other nuclide of comparable A2 in a
Type B package. It should be noted that for domestic shipments, there
is a well established and excellent safety record associated with the
shipment of spent fuel assemblies in single containment spent fuel
packages.
Comment. Two commenters stated that removing the double containment
requirement would provide health benefits for radiation workers. One
commenter argued that the cost of reducing the exposure to workers to
the required 1 mrem/yr would be very high. One commenter asserted that
we need to balance public safety and the safety of radiation workers.
Response. As discussed in the draft EA, NRC agrees that the removal
of the double containment requirement would result in reduced risk to
radiation workers.
Comment. One commenter stated that worker exposure estimates are
not supported by data. Another commenter stated that the conclusion
that single containment will decrease radiation doses is incorrect for
WIPP shipments. The commenter contends that radiation doses would
increase to both workers and the general public.
Response. The first commenter's remark about lack of data on worker
exposure estimates was true at the time of the public meeting on June
24, 2002, where the comment was made. However, during the comment
period, DOE, one of the major entities affected by the current double
containment rule, submitted the results of a detailed study they
performed to evaluate the impacts for elimination of the current
requirement. In that study, they presented quantifiable data that
indicates that over a 10-year period, they could expect to see a
reduction of 1200 to 1700 person-rem if the double containment
provision is eliminated. The second commenter provided qualitative and
quantitative information (some of which concerned a non-NRC certified
cask) that comes to a contrary conclusion. While the NRC does not
endorse or dispute either study's conclusions, the NRC believes worker
dose would be reduced due to less handling. Further, radiation
protection of transport workers (e.g., drivers, inspectors) and the
public is provided through the package maximum radiation levels set
forth in DOT regulations, which are not a function of double
containment.
Comment. One commenter stated that the NRC has not fully evaluated
the regulatory impact of the proposed change on the use of the TRUPACT
II design.
Response. During the development of the proposed rule, NRC staff
used all available data to evaluate the costs and benefits of the
proposed change. NRC staff requested specific information on costs and
benefits as part of the proposed rule, and the information received was
considered during the development of a final position. NRC received a
study from the commenter and, while the NRC does not endorse or dispute
the study's conclusions, the results are in line with the NRC's
contention that elimination of the double containment requirement will
likely result in a reduction in worker radiation exposure.
Comment. One commenter asked if NRC considers powder a solid form.
Response. Yes, the NRC has always considered powder as a solid form
when implementing Sec. 71.63(a). However, powders, under the eliciting
rule, were not considered as a solid form that was exempt from the
double containment requirements of Sec. 71.63(b).
Comment. One commenter endorsed NRC's proposal to retain the
requirement that shipments whose contents exceed 20 curies of plutonium
must be made in a solid form as provided under Sec. 71.63(a).
Response. The comment is acknowledged.
Comment. One commenter expressed support for the NRC position.
Response. The comment is acknowledged.
Comment. Several commenters expressed concern that removing the
double containment requirement would erode public confidence in the
Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. One of
the commenters noted that NRC's decision is not supported by any
studies to demonstrate that the change is minimal and that NRC should
only relax the double containment provisions when NRC receives
scientific evidence that demonstrates beyond a reasonable doubt that
single containment is as safe as double containment for shipments to
WIPP. Another commenter cited the economic, shipping, and public
confidence aspects of a severe accident release as the primary
arguments in support of retaining double containment.
Response. The comments are acknowledged. With regard to the last
commenter's citation, as is the case with other nuclides, NRC-certified
Type B packagings provide for safety in transportation accidents. With
regard to non-safety focused arguments (economic and public confidence
[[Page 3760]]
issues), as well as the other commenter's concerns, the reader is
referred to a related discussion earlier on this issue, under the
heading: Analysis of Public Comments on the Issues Paper.
Comment. One commenter discussed an incident involving the shipment
of plutonium-containing transuranic waste to DOE's Waste Isolation
Pilot Plant in New Mexico. A truck carrying TRU waste was involved in a
traffic accident. While no radiation was released, the inner container
was discovered to be contaminated with radiation to the extent that it
could not be unloaded. The commenter pointed out that the double-walled
container provided a margin of safety that would not have existed under
the proposed rule. The commenter stated that the incident underscores
the importance of maintaining the double containment requirement, as it
has been a crucial element in the success of the WIPP TRU waste
shipping campaign to date.
Response. In the cited case, NRC staff understands that neither
containment was compromised due to the accident.
Comment. One commenter stated that all shipping requirement
revisions should be more, rather than less, protective of public
health. Two other commenters stated that the AEC's original 1974
reasoning for imposing the double containment requirements was still
valid, including the possibility for human error and expected increases
in the number of shipments. The commenter also responded to the claim
that adopting a single containment requirement would be safer for
personnel who handle the inner container by stating that this may
simply be a shifting of risk from personnel to the public.
Response. The comment that shipping requirement revisions should
all be more, rather than less, protective of public health, is
acknowledged. The NRC's transportation regulations are designed to
provide adequate protection to the public health and safety from
radioactive material transportation activities. In doing so, NRC seeks
to balance its regulations by ensuring public health and safety while
at the same time not creating unnecessary regulatory burden.
Regarding the comment that the AEC's original 1974 reasoning for
imposing double containment is still valid, the NRC notes that the
AEC's original reasoning was based on the fact of transporting liquids;
that is no longer the case. The justifications for elimination of the
double containment requirement detailed in the proposed rule on April
30, 2002 (67 FR 21421 through 21425) is based on technical arguments
and focus on the confidence in Type B packages. While there is an
increase in the number of shipments to WIPP, the vast majority of these
shipments do not involve liquids.
The NRC disagrees with the comment that while the adoption of a
single containment requirement would be safer for personnel who handle
the inner container, this constitutes a shifting of the risk from
personnel to the public. The NRC believes that the risk of shipping
plutonium in a single containment Type B package is no different than
that of shipping other radionuclides with the same or lower
A1 and A2 values than plutonium.
Comment. One commenter stated that although spent fuel that is
damaged to the extent that the rod cladding's integrity is in question
may be subject to the requirements of Sec. 71.63, it is not clear that
all damaged fuel will require double containment.
Response. NRC has previously published guidance (ISG-1, Rev. 1,
dated October 25, 2002) on when the double containment provision is
required for damaged spent fuel. Basically, canning (double
containment) is required if the spent fuel contains known or suspected
cladding defects greater than a pinhole leak or hairline crack that
have the potential for release of significant amounts of fuel into the
cask.
Comment. One commenter stated that additional procedures (e.g.,
closures and testing) are required to implement Sec. 71.63, which leads
to added worker exposures. The commenter provided quantitative and
monetized data detailing the extra time and amount of money that the
double containment requirement imposes on TRU Waste, Plutonium Oxides,
and Damaged Spent Nuclear Fuel Operations.
Response. NRC acknowledges this comment.
Comment. One commenter stated that additional containment systems
reduce cask capacities and consequently require more shipments to move
the same material. This commenter also said that the double containment
represents extra weight that must be moved and then provided estimates
of the cost for moving the extra weight in the double-containment
structure in the cases of TRU Waste, Plutonium Oxides, and Damaged
Spent Nuclear Fuel operations.
Response. The comment is acknowledged.
Comment. One commenter stated that design costs and costs for NRC
certification services are incurred by increased design complexity
relating to the provision of the double-containment barrier. The
commenter noted that the alternative to the design and certification
cost penalty is to petition for an exemption under Sec. 71.63(b)(4);
however, preparing this petition is time-consuming and probably similar
in cost to getting a separate containment boundary designed and
certified. The commenter estimated certification and capital cost
penalties for the cases of CH-TRU and RH-TRU Wastes, Plutonium Oxides,
DHLW Glass Exemption, and Damaged Spent Nuclear Fuel.
Response. The comment is acknowledged.
Comment. One commenter stated that while the restrictions of Sec.
71.63 remain in effect, it must continue to expend funds unnecessarily
for double-containment packaging. This commenter provided tables of
monetized breakdowns of these estimates. The commenter estimated that
the net result from all three areas (TRU wastes, plutonium oxides and
residues, and damaged spent nuclear fuel) is that double-containment
requirements will produce an avoidable cost of approximately $12
million in capital cost, $20 million in operational cost, and $26
million to $40 million in shipping and receiving costs. In addition,
the commenter estimated that the double containment requirement will
result in additional worker radiation exposure amounting to 1250 to
1770 person-rem.
Response. The commenter has provided information that appears to
support the NRC's contention that removal of double containment would
provide for cost savings and decreased personnel exposure.
Comment. One commenter stated that double containment provides some
additional protection to the public in both normal and accident
situations. The commenter stated that most of this additional
protection relates to a potential reduction in population exposure.
However, the commenter estimated that the total radiation exposure
reduction in most cases amounts to a maximum of about 30 person-rem/
year distributed among a potentially exposed population of tens of
millions of persons. The commenter stated that such an effect would not
be perceptible.
Response. NRC acknowledges the comment.
Comment. One commenter stated that, although double containment
reduces the risk incurred by the public of exposure to radiation from
the package in incident-free transport, the reduction is likely to be
relatively small. The dose rate is already small enough at distances
[[Page 3761]]
where the public is likely to be exposed that the impact of single-or
double-contained material will not be consequential. This commenter
also noted that one effective containment boundary is sufficient to
meet containment requirements implicit in Type B design approvals, but
the materials shipped are already within one or more inner containers.
The commenter believes the presence of these redundant containers
effectively rules out any problems that might result from human errors
in achieving a required level of leak-tightness for single contained
Type B packages.
Response. NRC acknowledges the comment.
Comment. One commenter stated that doubly contained packages pose
lower risks and is not, by itself, sufficient justification for using
doubly contained packages. The commenter stated that, in general, the
likelihood of achieving an accident sufficient to compromise
containment of a singly contained Type B package has been estimated to
be fewer than 1 in 200 in the event of a severe accident. Achieving
damage to two redundant containments could be expected to be as much as
a factor of 10 lower risk relative to the single containment case. The
commenter stated that this is not as large a benefit as it may seem;
the decrease in absolute risk will be very small because the risk of
shipping singly contained plutonium is exceedingly small to start. The
commenter provided monetized and quantified estimates of the cost/risk
tradeoffs associated with double-containment versus single-containment
for the handling of Contact-Handled TRU Waste, Plutonium Oxide and
Plutonium-Bearing Wastes, Remote-Handled TRU Waste, and Failed Fuel.
Response. NRC acknowledges the comment.
Comment. Two commenters stated that if the NRC continues to pursue
the proposal to relax the plutonium shipment double containment
standards, then it should conduct a series of hearings on the
rulemaking, with at least one of those hearings held in the western
U.S. Another commenter objected to the lack of public education
regarding the ``numerous, confusing, and complicated'' proposed rule
changes, which, when presented as they were, encourage nonengagement.
The commenter requested that an extension be placed on the comment
period and that ``ordinary'' language be used to explain the actual
proposals, how they will impact public health, what agencies and rules
are involved, and how one can easily reply to all agencies involved in
these proposals by mail, email, or fax.
Response. The rulemaking process does not include the opportunity
for formal hearings because the proposed rulemaking is not a licensing
action, which does require hearings. The NRC staff thinks that the
commenter meant holding public meetings to discuss the issue. Hearings
were held in this rulemaking in the form of public meetings. Two
meetings were held in June 2002, in Chicago, IL, and the NRC TWFN
Auditorium, and 3 meetings were held in NRC Headquarters, Atlanta, GA,
and Oakland, CA, during August and September 2000. The NRC did not
extend the 90-day public comment period, because the public had ample
opportunity to comment on this rule during the 1-year period following
March 2001, when the proposed rule was posted on the Secretary of the
Commission Web site.
Issue 18. Contamination Limits as Applied to Spent Fuel and High-Level
Waste (HLW) Packages
Summary of NRC Final Rule. The final rule does not adopt any
changes to part 71 for this issue because experience with regulations
requiring that licensees monitor the external surfaces of labeled
radioactive material packages for contamination upon receipt and
opening indicates the rate of packages exceeding allowable levels en
route is low, and therefore, in transit decontamination of packages is
not warranted. Further, requiring such decontamination of packages
could result in a significant increase in worker doses without a
commensurate increase in public health and safety.
Affected Sections. None (not adopted).
Background. In the period of December 1997 through April 1998, the
French Nuclear Installations Safety Directorate inspected a French
nuclear power plant and railway terminal used by La Hague reprocessing
plant. The inspectors noticed that, since the beginning of the 1990's,
a high percentage of spent fuel packages and/or railcars had a level of
removable surface contamination that exceeded IAEA regulatory limits by
as much as a factor of 1000. Subsequent investigations found that the
contamination incidents involved shipments from other European
countries, and the French transport authorities notified their
counterparts of their findings. Subsequently, French, German, Swiss,
Belgian, and Dutch spent fuel shipments were temporarily suspended.
After estimating the occupational and public doses from the
contamination incidents, the European transport authorities concluded
that these incidents did not have any radiological consequence. The
contamination was believed to be caused by contact of the spent fuel
package surface with contaminated water from the spent fuel storage
pool during package handling operations. The authorities concluded that
there were deficiencies in the contamination measurement procedures and
the distribution of that information.
Media reports on these incidents focused attention on IAEA's
regulations for removable contamination on package surfaces. TS-R-1
contains contamination limits for all packages of 4.0 Bq/cm2
for beta and gamma and low toxicity alpha emitting radionuclides, and
0.4 Bq/cm2 for all other alpha emitting radionuclides.
Although TS-R-1 uses the term ``limit,'' IAEA considers these
``limits'' to be guidance values, or derived values, above which
appropriate action should be considered. In cases of contamination
above the limit, that action is to decontaminate to below the limits.
TS-R-1 further provides that in transport, ``* * * the magnitude of
individual doses, the number of persons exposed, and the likelihood of
incurring exposure shall be kept as low as reasonable, economic and
social factors being taken into account * * *'' The IAEA contamination
regulations have been applied to radioactive material packages in
international commerce for almost 40 years, and practical experience
demonstrates that the regulations can be applied successfully. With
respect to contamination limits, TS-R-1 contains no changes from
previous versions of IAEA's regulations.
Part 71 does not contain contamination limits, but Sec. 71.87(i)
requires that licensees determine that the level of removable
contamination on the external surface of each package offered for
transport is as low as is reasonably achievable, and within the limits
specified in DOT regulations in 49 CFR 173.443.
The IAEA established a Coordinated Research Project (CRP) to review
contamination models, approaches to reduce package contamination,
strategies to address cask-weeping, and possible recommendations for
revisions to the contamination standard that consider risks, costs, and
practical experience. The IAEA CRP facilitates the investigation of
radioactive material transportation issues by key IAEA Member States.
IAEA is considering the CRP report, and any further actions or remedies
that may be warranted are being addressed by the IAEA Transportation
Safety Standards Committee (TRANSSC). NRC supported
[[Page 3762]]
the IAEA initiative to establish the CRP, and NRC would participate in
the IAEA review of surface contamination standards.
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. One commenter expressed support of the NRC position not to
change from current standards.
Response. The NRC acknowledges these comments. No further response
necessary.
Comment. One commenter requested that the NRC keep ``removable
contamination of external ``spent'' fuel shipping packages'' to the
``absolute minimum attainable, even if extra cost is incurred in doing
so.'' The commenter added that ``full data on container surface
contamination must be kept and submitted to the regulatory agency as
part of required manifest records.''
Response. Keeping contamination to an absolute minimum could result
in a significant increase in worker dose, due to the additional
exposures required to achieve that low level of contamination, without
a commensurate increase in public health and safety. Current DOT
regulations require that shippers be able to provide to inspectors upon
request documentation that supports the shipper's certification that
radioactive material shipments were made in compliance with applicable
requirements, including contamination limits. This practice has worked
well, and NRC has no basis to change it.
Comment. One commenter stated that the NRC's measures should allow
for decontamination of nuclear waste shipments during transport if they
begin to exceed allowable radiation levels en route. The commenter
stated that this would reduce exposure to the public and prevent
shipments from having to return to the point of origin.
Response. Current NRC regulations require that licensees monitor
the external surfaces of labeled radioactive material packages for
contamination upon receipt and opening (see details at Sec.
20.1906(b)(1)). Based on its experience with these regulations, the
rate of packages exceeding allowable levels en route is low, and NRC
does not believe that in transit decontamination of packages is
warranted.
Comment. One commenter asserted that there is no reason to seek any
special dose consideration or reduction in the handling and transport
of spent fuel or storage casks. The commenter added that industry has
not attributed any problems with decontamination and dose to the
handling and transport of spent fuel or storage casks. The commenter
did note that although industry did experience some of the weeping
issues in the early 1990's, industry has taken steps to eliminate this
condition.
Response. NRC agrees that incidents of cask weeping have subsided
in recent years. However, NRC notes that considerable occupational dose
is expended to achieve compliance with current regulatory limits that
do not appear to be risk-informed, and that occupational and public
doses associated with spent fuel cask surface contamination limits do
not appear to be optimized.
Comment. One commenter requested that the NRC not relax ``radiation
protection in any shipments, especially high-level wastes and intensely
irradiated ``spent'' fuel,'' the reason being that, in the near future,
shipments of high-level wastes and spent fuel may increase in number,
and this would justify NRC staff's maintaining ``maximum control * * *
as a principal goal of the NRC.'' The commenter also stated that while
``Europeans may dismiss contamination ``incidents'' as having no
radiological consequences * * * that is not convincing, in view of
recent research findings concerning adverse impacts of low-level
radiation at the cellular and molecular levels.''
Response. No change to the contamination limit is being adopted in
the final rule, and no relaxation of radiation protection has been
proposed.
Comment. Two commenters expressed opposition to allowing greater
contamination on surfaces of irradiated fuel and high-level radioactive
waste containers and supported NRC's decision to refuse this. Two other
commenters supported the NRC's proposal to make no changes in the
contamination levels for these packages.
Response. No response is necessary.
Comment. One commenter expressed opposition to allowing greater
contamination on surfaces of irradiated fuel and high level radioactive
waste containers.
Response: The NRC acknowledges these comments. No response is
necessary.
Issue 19. Modifications of Event Reporting Requirements
Summary of NRC Final Rule. The final rule revises, in Sec. 71.95,
the event reporting submission period to provide a written report from
30 to 60 days. Other regulatory requirements to orally notify the NRC
Operations Center promptly of an event and for licensees to report
instances of failure to follow the conditions of the CoC while
packaging was in use remain unchanged. The revision lengthening the
time for submission of the written report is consistent with changes to
similar requirements in Part 50.
Affected Sections. Section 71.95.
Background. The Commission recently issued a final rule to revise
the event reporting requirements in Part 50 (see 65 FR 63769; October
20, 2000). This final rule revised the verbal and written event
notification requirements for power reactor licensees in Sec.Sec. 50.72
and 50.73. In SECY-99-181,\13\ NRC staff informed the Commission that
public comments on the proposed part 50 rule had suggested that
conforming changes also be made to the event notification requirements
in part 72 (Licensing Requirements for the Independent Storage of Spent
Fuel) and part 73 (Physical Protection of Plants and Materials). In
response, the Commission directed the NRC staff to study whether
conforming changes should be made to parts 72 and 73. During this
study, the NRC also reviewed the part 71 event reporting requirements
in Sec. 71.95 and concluded that similar changes could be made to the
part 71 event reporting requirements.
---------------------------------------------------------------------------
\13\ SECY-99-181, ``Proposed Plans and Schedules to Modify
Reporting Requirements Other than 10 CFR 50.72 and 50.73 for Power
Reactors and Material Licensees,'' dated July 9, 1999.
---------------------------------------------------------------------------
Analysis of Public Comments on the Proposed Rule
A review of the comments and the NRC staff's responses for this
issue follows:
Comment. Two commenters expressed support for the proposed
modifications. One commenter stated that the proposed modifications to
event reporting requirements will enhance safety. The other commenter
noted that many States respond to incidents involving radioactive
materials on a regular basis and would not want to wait until the full
60 days for reporting purposes.
Response. The NRC acknowledges the comments supporting the change
to require a 60-day report instead of a 30-day report for a
transportation event. The comment that States would need to respond to
incidents and would need reports sooner than 60 days is not consistent
with the fact that prompt reporting to the National Response Center,
NRC Operations Center, and appropriate State Authorities occurs after
an event. The written report to the NRC will not affect this practice.
Therefore, the change in the time to
[[Page 3763]]
provide a written report would have no effect on the emergency response
and information exchange actions that would still be performed by
licensees or the DOT National Response Center. Therefore, no changes in
the proposed rule language are being made.
Comment. One commenter asked how this proposed change affects other
parts of the proposed rulemaking and urged the NRC to ensure that it
conforms with the rest of the proposed rulemaking.
Response. There are no other impacts on the regulations associated
with adopting this specific change.
Comment. Two commenters opposed the proposed event reporting
requirements. The first commenter stated that there should never be a
30-or 60-day ``delay in filing a report on any event involving
malperformance of a package or container,'' but that a report should be
filed immediately with the NRC when a problem occurs. The second
commenter suggested that ``reporting should serve the needs of the
(NRC) staff-and public safety,'' rather than the licensee. This
commenter also claimed that an extra 30 days may be too long an
extension if there is a serious safety problem.
Response. The NRC notes that if a serious safety problem resulted
from an incident, it would be reported promptly to the NRC Operations
Center. The NRC staff notes that a review of the regulatory analysis
included in the proposed rule stated that: ``In new paragraph (a)(3),
[of section 71.95] the NRC would retain the existing requirement for
licensees to report instances of failure to follow the conditions of
the CoC while a packaging was in use.'' This section was inadvertently
left out of the proposed rule language and was added to the final rule.
Comment. One commenter indicated concern about the lack of data to
support NRC's position on extending the reporting period from 30 to 60
days.
Response. There is sufficient rationale as reflected in other
regulations for reducing the regulatory burden related to the time for
submitting written reports. See the discussion in the proposed rule
(April 30, 2002; 67 FR 21427) for additional detail on the
justification for the change. Therefore, no change to the rule is
proposed.
Comment. One commenter was concerned about difficulties in
compiling a jointly written report by the certificate holder and the
shipper if they are in different countries.
Response. The commenter's concern about coordination of a jointly
written event report is valid; however, the longer time being proposed
for submitting an event report should accommodate delays in the
communication interface and help ensure completion within the 60-day
reporting period. Therefore, no changes have been made to the proposed
rule language.
Comment. One commenter found the event reporting requirements
unclear in two places. The proposed rule would direct the licensee to
request information from certificate holders; however, neither the
supporting discussion nor regulatory text addresses a situation in
which a certificate holder declines to provide comments. The commenter
asked whether the licensee's obligation would be satisfied at the point
that a request is made to CoC holders. The commenter also found it
unclear whether NRC intended to exempt DOT specification and foreign
package designs holding U.S. validations from the reporting
requirements. The commenter asserted that if NRC intends to make a
distinction between NRC-approved packages and other authorized
packages, it may be necessary to develop separate QA procedures and
related instructions. The impacts on resources associated with such
development may require further investigation.
Response. Regarding the first question about what would happen if a
licensee did not receive supporting information in its process to issue
an event report to the NRC to comply with the requirements of Sec.
71.95, the NRC notes that the licensee should make an earnest attempt
to obtain relevant information from the CoC holder. In the case where
the CoC holder refused to provide input to the report, the licensee
would still need to submit the report to the NRC within the 60-day time
period. NRC technical staff would determine if CoC staff input should
have been included in the report and would obtain it directly from the
CoC holder as necessary. Further, if the NRC determined that the CoC
holder's lack of support resulted in a report that was incorrect or
incomplete, then the NRC would pursue appropriate regulatory action
against the CoC holder.
Regarding the second question about the reporting requirement being
applicable to DOT specification and foreign package designs with U.S.
validation, the NRC notes that its regulations only apply directly to
its licensees or CoC holders. NRC will, however, forward this comment
to DOT for appropriate consideration. No change to NRC rule language is
being made.
Comment. One commenter stated that the requirement of the CoC
holder to rely on other licensees or registered users, over whom the
holder has no authority or control, to identify problems or package
deficiencies, is inappropriate and must be modified. Another commenter
stated that the authorized package user should be making the required
report.
Response. Both comments deal with the original language in the
existing Sec. 71.95 which states that licensees are responsible for
providing event reports to the NRC.
IV. Section-by-Section Analysis
Several sections in part 71 are redesignated in this rulemaking to
improve consistency and ease of use. For some sections, only the
section number is changed. However, for other sections, revisions are
being made to the regulatory language. The following table is provided
to aid the public in understanding the numerical changes to sections of
part 71.
Redesignation Table
------------------------------------------------------------------------
New section number Existing section number
------------------------------------------------------------------------
Sec. 71.8................................. Sec. 71.11.
Sec. 71.9................................. New section.
Sec. 71.10................................ New section.
Sec. 71.11 (Reserved)..................... NA.
Sec. 71.12................................ Sec. 71.8.
Sec. 71.13................................ Sec. 71.9.
Sec. 71.14................................ Sec. 71.10.
Sec. 71.15................................ Sec. 71. 53.
Sec. 71.16 (Reserved)..................... NA.
Sec. 71.17................................ Sec. 71.12.
Sec. 71.18 (Reserved)..................... NA.
Sec. 71.19................................ Sec. 71.13.
Sec. 71.20................................ Sec. 71.14.
Sec. 71.21................................ Sec. 71.16.
Sec. 71.22................................ Sec. 71.18.
Sec. 71.23................................ Sec. 71.20.
Sec. 71.24 (Reserved)..................... Sec. 71.22 (Section
removed).
Sec. 71.25 (Reserved)..................... Sec. 71.24 (Section
removed).
Sec. 71.53 (Reserved)..................... Sec. 71.53 (Section
redesignated).
------------------------------------------------------------------------
Subpart A--General Provisions
Section 71.0 Purpose and scope
Paragraph (d) has been reformatted into three paragraphs to
simplify this regulation and to better use plain language. Paragraph
(d)(1) indicates that general licenses, for which no NRC package
approval is required, are issued in new Sec.Sec. 71.20 through 71.23.
This change reflects the removal of existing Sec.Sec. 71.22 and 71.24
(redesignated Sec.Sec. 71.24 and 71.25 (Reserved)). Paragraph (d)(2)
indicates that an application for package approval must be completed in
accordance with subpart D. Paragraph (d)(3) continues to
[[Page 3764]]
require a licensee transporting, or delivering material to a carrier
for transport, to meet the requirements of the applicable portions of
subparts A, G, and H.
New paragraph (e) has been added to indicate that persons who hold,
or apply for, a part 71 CoC for Type AF, Type B, Type BF, Type B(U)F,
or Type B(M)F packages are within the scope of part 71 regulations.
Existing paragraphs (e) and (f) have been redesignated as new
paragraphs (f) and (g), respectively. The rule text in new paragraph
(f) is the same as existing paragraph (e) text. New paragraph (g) has
been revised to reflect the redesignation of existing Sec. 71.11 as new
Sec. 71.8.
Section 71.1 Communications and Records
In Sec. 71.1, paragraph (a) has been revised to indicate that
documents submitted to the NRC should be addressed to the attention of
the ``Document Control Desk,'' not the ``Director of the Office of
Nuclear Material Safety and Safeguards.'' Provisions have also been
added to provide requirements when a due date for a document falls on a
Saturday, Sunday, or Federal holiday. In that case, the document would
be due the next Federal workday. This change is identical to a change
made to Sec. 72.4 in a recent part 72 final rule (see 64 FR 33178; June
22, 1999).
Section 71.2 Interpretations
No changes were made to the text of this section; however, it has
been retained in the revision of this subpart for completeness.
Section 71.3 Requirement for License
No changes were made to the text of this section; however, it has
been retained in the revision of this subpart for completeness.
Section 71.4 Definitions
The existing definitions for ``A1,'' ``Fissile
material,'' ``Low Specific Activity (LSA) material,'' ``Package,'' and
``Transport index (TI)'' are revised as conforming changes. New
definitions for ``A2,'' ``Certificate of Compliance,''
``Consignment,'' ``Criticality Safety Index (CSI),'' ``Deuterium,''
``U.S. Department of Transportation (DOT),'' ``Graphite,'' ``Spent
fuel,'' and ``unirradiated uranium'' have been added as conforming
changes.
The definition of ``A1'' has been revised to split the
previous combined definition for ``A1'' and
``A2'' into two individual definitions. This approach is
consistent with the standard in TS-R-1. Furthermore, no change has been
made to the current technical content of the definition for
``A1''; however, the text is revised to improve readability.
A definition for ``A2'' has been added, because the
previous joint definition for ``A1'' and ``A2''
has been split into two definitions. (See also definition for
``A1.'')
A definition for ``Certificate of Compliance'' has been added. This
definition is similar to the definition for the same term found in Sec.
72.3.
A definition for ``Consignment'' has been added.
A definition of ``Criticality Safety Index (CSI)'' has been added.
A definition of ``Deuterium'' has been added that applies to new
Sec.Sec. 71.15 and 71.22.
A definition of ``U.S. Department of Transportation (DOT)'' has
been added.
The definition of ``Fissile material'' has been revised by removing
238Pu from the list of fissile nuclides; clarifying that
``fissile material'' means the fissile nuclides themselves, not
materials containing fissile nuclides; and redesignating the reference
to exclusions from fissile material controls from Sec. 71.53 to new
Sec. 71.15.
A definition of ``Graphite'' has been added that applies to new
Sec.Sec. 71.15 and 71.22.
The definition of ``Low Specific Activity (LSA)'' material (LSA-I,
LSA-II, and LSA-III) has been revised to be consistent with DOT, and to
reflect the existence of Sec. 71.77 (Sec. 71.77 provides requirements
on the qualification of LSA-III material).
A definition for ``Optimum interspersed hydrogenous moderation''
has been added (the definition itself was included in the proposed rule
Sec. 71.4, but, inadvertently, no mention of that fact was made in this
Section).
The definition of ``Package'' has been revised by clarifying in
paragraph (1) that Fissile material package also means a Type AF, Type
BF, Type B(U)F, or Type B(M)F package. New paragraph (2) has been added
defining Type A packages in accordance with DOT regulations contained
in 49 CFR Part 173. Existing paragraph (2) defining Type B packages has
been redesignated as subparagraph (3). No changes have been made to the
redesignated text.
A definition of ``Spent nuclear fuel'' or ``Spent fuel'' has been
added. This definition is the same as that currently found in Sec.
72.3.
The definition for ``Transport index (TI)'' has been revised to
reflect the new definition of Criticality Safety Index; however, the
method for determining the TI of a package, based on the package's
radiation dose rate, remains unchanged.
A definition for ``unirradiated uranium'' has been added as it is
part of the LSA-I definition.
Section 71.5 Transportation of Licensed Material
No changes were made to the text of this section; however, it has
been included in the revision of this subpart for completeness.
Section 71.6 Information Collection Requirements: OMB Approval
This section has been redesignated from subpart B, Exemptions, to
subpart A, General Provisions. Paragraph (b) of this section has been
revised as a conforming change to reflect the addition of new
information collection requirements. Additionally, the existing
information collection requirement in Appendix A to part 71, paragraph
II, was inadvertently omitted from the list of approved information
collection requirements in a previous rulemaking; consequently, NRC
staff has added Appendix A, paragraph II, to paragraph (b) to correct
this error. Furthermore, the reference to Sec. 71.6a has been removed,
because no such section currently exists in part 71.
Section 71.7 Completeness and Accuracy of Information
This section has been redesignated from subpart B, Exemptions, to
subpart A, General Provisions. Further, paragraphs (a) and (b) have
been revised by adding the terms ``certificate holder'' and ``applicant
for a CoC.''
Section 71.8 Deliberate Misconduct
This section has been redesignated from subpart B, Exemptions, to
subpart A, General Provisions. Further, in subpart A, Sec. 71.11 has
been redesignated as Sec. 71.8. However, the current text of Sec. 71.11
has not changed in the redesignated Sec. 71.8.
Section 71.9 Employee Protection
New Sec. 71.9 has been added to provide requirements on employee
protection. Currently, requirements relating to the protection of
employees against firing or other discrimination when the employee
engages in certain ``protected activities'' are provided under the
parts of title 10 for which a specific license was issued to possess
radioactive material. However, no provisions were provided in part 71
relating to the protection of employees against firing or other
discrimination when employees engage in certain ``protected
activities'' when they are the employees of a certificate holder or
applicant for a CoC.
[[Page 3765]]
The NRC believes these employees should also be afforded the same
rights and protection as are currently afforded employees of licensees.
The new section is identical to the existing Sec. 72.10, ``Employee
protection.'' In including licensees in the new Sec. 71.9, the NRC
recognizes that the potential for duplication occurs for licensees
regulated under multiple title 10 parts. However, the NRC believes that
by including licensees along with certificate holders and applicants
for a CoC, improved regulatory clarity would be achieved, and any
potential confusion would be minimized.
Section 71.10 Public Inspection of Application
A new section has been added indicating that applications and
documents submitted to the Commission, in connection with an
application for a package approval, shall be available for public
review in accordance with the provisions of parts 2 and 9. This new
section is similar to existing Sec. 72.20. Existing Sec. 71.10 has been
redesignated Sec. 71.14 with changes to the text as discussed under
Sec. 71.14, below.
Section 71.11 (Reserved)
This section has been redesignated from subpart B, Exemptions, to
subpart A, General Provisions, and is reserved. Existing Sec. 71.11 has
been redesignated as Sec. 71.8.
Subpart B--Exemptions
Section 71.12 Specific Exemptions
Existing Sec. 71.8 has been redesignated as Sec. 71.12. No changes
have been made to the contents of this section. Existing Sec. 71.12 has
been redesignated as Sec. 71.17, with changes to the text as discussed
under Sec. 71.17, below.
Section 71.13 Exemption of Physicians
Existing Sec. 71.9 has been redesignated as Sec. 71.13. No changes
have been made to the contents of this section. Existing Sec. 71.13 has
been redesignated as Sec. 71.19, with changes to the text as discussed
under Sec. 71.19, below.
Section 71.14 Exemption for Low-Level Materials
Existing Sec. 71.10 has been redesignated as Sec. 71.14. Existing
Sec. 71.14 has been redesignated as Sec. 71.20, with no changes to the
text.
In new Sec. 71.14, paragraph (a) has been revised by removing the
existing single 70 Bq/g (0.002 [mu]Ci/g) specific activity value.
Additionally, paragraph (a) has been reformatted by adding two new
paragraphs. Subparagraph (a)(1) provides an increased exemption for
natural radioactive materials and ores. Subparagraph (a)(2) provides an
exemption for radioactive material based on the ``Activity
Concentration for Exempt Material'' and the ``Activity Limit for Exempt
Consignment'' found in Table A-2 in Appendix A to part 71.
Paragraph (b) has been revised to consolidate the exemption
provisions for LSA and SCO material. The LSA and SCO exemptions
contained in existing paragraphs (b)(2) and (c) of this section have
been consolidated into a revised paragraph (b)(3). The reference to
material exempt from classification as fissile material has been
revised from Sec. 71.53 to Sec. 71.15, because of the redesignation of
the section.
Existing paragraph (b)(3) has been removed. The 0.74-TBq (20-Ci)
exemption for special form americium and special form plutonium has
been removed. However, the 0.74-TBq (20-Ci) exemption for special form
plutonium-244, transported in domestic commerce, has been retained as
new paragraph (b)(2). For international shipments, the A1 quantity
limit for special form plutonium-244 continues to apply.
Section 71.15 Exemption From Classification as Fissile Material
Existing Sec. 71.11 has been redesignated as Sec. 71.8. Existing
Sec. 71.53 has been redesignated as Sec. 71.15, and relocated to
subpart B with the other part 71 exemptions. This section has been
revised by providing mass-ratio based limits in classifying fissile-
exempt material. This approach removes the concentration- and
consignment-based limits of the current Sec. 71.53 and returns to
package-based mass limits, with required minimum ratios of nonfissile-
to-fissile mass.
The title has been changed to ``Exemption from classification as
fissile material.''
New paragraph (a) has been added and allows for small samples of
fissile material to be shipped. In paragraph (b), the fissile mass per
package is limited to 15 grams with a nonfissile-to-fissile mass ratio
of 200:1. In paragraph (c), the allowed provided there is less than 150
g of fissile material per 360 Kg ratio of nonfissile-to-fissile
material is also raised to 2000:1. The mass of any lead, graphite,
beryllium, and deuterium in the package cannot be included in
determining the nonfissile material mass.
In current Sec. 71.53, paragraph (c) has been redesignated as
paragraph (e), and has been reformatted and revised to clarify that the
nitrogen to uranium atomic ratio, for shipments of liquid uranyl
nitrate, must be greater than or equal to 2.0. A new requirement has
been added specifying the use of DOT Type A packaging.
In current Sec. 71.53, paragraph (d) has been redesignated as
paragraph (e), and has been reformatted and revised to clarify the mass
limits for plutonium. No substantive changes have been made to this
paragraph.
Section 71.16 (Reserved)
This section has been redesignated from subpart C, General
Licenses, to subpart B, Exemptions, and is reserved. Further, existing
Sec. 71.16 has been redesignated as Sec. 71.21. However, the current
text of Sec. 71.16 has not been changed in the redesignated Sec. 71.21.
Subpart C--General Licenses
Section 71.17 General License: NRC-Approved Package
Existing Sec. 71.12 has been redesignated as Sec. 71.17. The text
of paragraphs (a) and paragraph (b) has not been changed.
Paragraph (c)(3) has been revised using plain language and to
reflect the NRC's requirement to address information submitted to the
NRC to the attention of the NRC's Document Control Desk, in accordance
with Sec. 71.1.
Paragraph (d) has not been changed.
Paragraph (e) has been revised to reflect the redesignation of Sec.
71.13 to Sec. 71.19. No other change was made for this paragraph.
Section 71.18 Reserved
Section 71.19 Previously Approved Package
Existing Sec. 71.13 has been redesignated as Sec. 71.19. Paragraph
(a) has been revised to reflect the current package designators (e.g.,
B(U)F, B(M)F, AF) and to reflect the redesignation of Sec. 71.12 to
Sec. 71.17. Additionally, the contents of paragraph (a)(2) have been
removed to reflect that these packages are no longer recognized
internationally. Existing paragraph (a)(3) has been redesignated as
(a)(2) with no change to the contents. Also, an expiration date for
grandfathering these packages has been established in new paragraph
(a)(3). Paragraph (b) has been updated to remove the LSA packages, as
these packages no longer exist, and to reflect the redesignation of
Sec. 71.12 to Sec. 71.17. No other changes were made. A new paragraph
(c) has been added to reflect the type B(U) and B(M) packages that have
met the requirements of IAEA Safety Series 6 1985 (as amended 1990) and
to correct a typographical error. Additionally, a date by which
fabrication of these packages must be complete has been added. Existing
paragraph (c) has been redesignated as
[[Page 3766]]
paragraph (d). Existing paragraph (d) has been redesignated as
paragraph (e) and updated to reflect the identification number suffix
of ``-96'' for previously approved package designs that have been
resubmitted for review by the NRC and have been approved, and to remove
the package designated as Type A from this paragraph.
Section 71.20 General License: DOT Specification Container
Existing Sec. 71.14 has been redesignated as Sec. 71.20. No changes
have been made to the contents of paragraphs (a) through (d). New
paragraph (e) has been added to indicate that these types of packages
will be phased out 4 years after the effective date of this final rule.
Section 71.21 General License: Use of Foreign Approved Package
Existing Sec. 71.16 has been redesignated as Sec. 71.21. No changes
have been made to the contents of this section.
Section 71.22 General License: Fissile Material
Existing Sec. 71.18 has been redesignated as Sec. 71.22. The
current Sec. 71.22 has been removed. This section has been amended by
consolidating and simplifying the current fissile general license
provisions contained in existing Sec.Sec. 71.18, 71.20, 71.22, and
71.24 into a new Sec. 71.22. The new Sec. 71.22, while retaining some
of the provisions of the existing general licenses, principally uses
mass-based limits and a Criticality Safety Index (CSI). Concentration-
based limits have been removed. Exceptions relating to plutonium-
beryllium sealed sources in existing Sec.Sec. 71.18 and 71.22 have been
relocated to new Sec. 71.23. The values contained in new Tables 71-1
and 71-2 have been revised from the values contained in the table in
existing Sec. 71.22 and in Table 1 in existing Sec. 71.20,
respectively; and are based on new minimum critical mass calculations
described in NUREG/CR-5342. In some instances, the allowable mass limit
has been increased from the current limits in existing Sec.Sec. 71.18,
71.20, 71.22, and 71.24; in other instances, the allowable mass limit
has been reduced. The values contained in new Tables 71-1 and 71-2 are
used as the variables X, Y, and Z in the equation in paragraph (e).
The title has been revised to indicate that this general license is
not restricted to a specific type of fissile material shipment.
Paragraph (a) has been revised to require that fissile material
shipped under this general license be contained in a DOT Type A
package. Additionally, while the existing exception from subparts E and
F requirements has been maintained, the DOT Type A package regulations
of 49 CFR part 173 has also been specified.
Paragraph (b) remains unchanged.
Paragraph (c) has been revised to remove the specific gram limits
for uranium and plutonium but retains the existing Type A quantity
limit. Revised gram limits have been relocated to new Table 71-1, which
is associated with new paragraphs (d) and (e). A requirement has also
been added to limit the amount of special moderating materials
beryllium, graphite, and hydrogenous material enriched in deuterium
present in a package to less than 500 g.
Existing paragraph (d) has been removed. Revised gram limits for
fissile material mixed with material having a hydrogen density greater
than water (i.e., a moderating effectiveness greater than
H2O) have been placed in new Table 71-1. A note has been
added to new Table 71-1 to indicate that reduced mass limits apply when
more than 15 percent of a mixture of moderating materials contains
moderating material with a hydrogen density greater than
H2O.
New paragraph (d) has been added to require that shipments of
packages containing fissile material be labeled with a CSI, that the
CSI per package be less than or equal to 10.0, and that the sum of the
CSIs in a shipment of multiple fissile material packages be limited to
less than or equal to 50.0 for a nonexclusive use conveyance, and to
less than or equal to 100.0 for an exclusive use conveyance.
Existing Paragraphs (e) and (f) have been removed.
New paragraph (e) has been added to require that the CSI be
calculated via a new equation for any of the fissile nuclides. Guidance
on applying the equation and the mass limit input values of Tables 71-1
and 71-2 is also contained in this paragraph.
Section 71.23 General License: Plutonium-Beryllium Special Form
Material
The existing Sec. 71.20, ``General license: Fissile material,
limited moderator per package,'' has been removed. A new section on the
shipment of plutonium-beryllium (Pu-Be) special-form fissile material
(i.e., sealed sources) has been added as a new Sec. 71.23. New Sec.
71.23 consolidates regulations on shipment of Pu-Be sealed sources
contained in existing Sec.Sec. 71.18 and 71.22 into one location in
part 71. The new Sec. 71.23 reduces the maximum quantity of fissile
plutonium Pu-Be sealed sources that could be shipped on a single
conveyance through changes in the mass limits and calculation of the
CSI. Currently, a Pu-Be sealed source package can contain up to 400 g
of fissile plutonium with a CSI equal to 10.0. Consequently, the
current conveyance limits are 4,000 g per shipment for an exclusive-use
vehicle and 2000 g per shipment for a nonexclusive use vehicle. The new
Sec. 71.23 increases the maximum CSI per package from 10 to 100;
however, the maximum quantity of plutonium per conveyance (i.e.,
shipment) would be reduced to 1000 g. The 1000-g per shipment limit and
240 g of fissile plutonium limit are equivalent to those in new Sec.
71.22(f) (1000 g per shipment and 200 g of fissile plutonium). The 240
g versus 200 g of fissile plutonium per package is due to the increased
confidence that the fissile plutonium, within a sealed source capsule,
would not escape from the capsule during an accident and reconfigure
itself into an unfavorable geometry.
New Sec. 71.23 has been titled: ``General license: Plutonium-
beryllium special form material.'' Paragraph (a) describes the
applicability of this section, exceptions to the requirements of
subparts E and F, and the requirement to ship Pu-Be sealed sources in
DOT Type A packages.
Paragraph (b) requires that shipments of Pu-Be sealed sources be
made under an NRC-approved QA program.
Paragraph (c) requires a 1000 g per package limit. In addition,
plutonium-239 and plutonium-241 constitute only 240 g of the 1000 g
limit.
Paragraph (d) requires that a CSI be calculated per paragraph (e),
and the CSI must be less than or equal to 100.0. For shipments of
multiple packages, the sum of the CSIs is limited to less than or equal
to 50.0 for a nonexclusive use conveyance and to less than or equal to
100.0 for an exclusive use conveyance.
Paragraph (e) provides an equation to calculate the CSI for Pu-Be
sources. This equation is based upon the 240-g mass limit for fissile
nuclide plutonium-239 and plutonium-241 in paragraph (c).
Section 71.24 (Reserved)
Section 71.25 (Reserved)
Existing Sec.Sec. 71.22 and 71.24 have been redesignated as
Sec.Sec. 71.24 and 71.25. New Sec.Sec. 71.24 and 71.25 have been
removed and reserved.
Subpart D--Application for Package Approval
Section 71.41 Demonstration of Compliance
Paragraph (a) has been revised to require that a Type B package
which contains radioactive contents with
[[Page 3767]]
activity greater than 105A2 of any radionuclide
must meet the enhanced deep immersion test found in Sec. 71.61. A new
paragraph (d) has been added to provide special package authorizations.
Section 71.51 Additional Requirements for Type B Packages
Paragraph (a) has been revised to remove the reference to Sec.
71.52, because the requirements of Sec. 71.52 have expired. Paragraph
(d) has been added to require that a package which contains radioactive
contents with activity greater than 105A2 of any
radionuclide must also meet the enhanced deep immersion test found in
Sec. 71.61.
Section 71.53 Fissile Material Exemptions (Reserved)
This section has been removed and reserved; its contents have been
moved to Sec. 71.15.
Section 71.55 General Requirements for Fissile Material Packages
New paragraphs (f) and (g) have been added. Paragraph (f) specifies
design and testing for fissile material package designs for transport
by aircraft, and paragraph (g) addresses UF6 criticality
exception from Sec. 71.55(b). Additionally, as a conforming change,
paragraph (b) has been updated to support new paragraph (g).
Section 71.59 Standards for Arrays of Fissile Material Packages
Paragraphs (b) and (c) have been revised to use the term CSI
(criticality safety index).
Paragraph (b) has been revised to refer to a CSI rather than a TI
for nuclear criticality control. The method for calculating a CSI is
the same as the existing method for a TI for nuclear criticality
control.
Paragraph (c) has been revised to provide direction to licensees
when the CSI is exactly equal to 50 and to use plain language.
Subparagraph (1) has been revised by replacing the term ``(n)ot in
excess of 10,'' with the term ``(l)ess than or equal to 50.'' New
paragraph (c)(2) has been added to provide for shipment of packages
with a CSI of less than 50 on an exclusive use conveyance. The current
conveyance limit of 100 has been retained. Existing paragraph (c)(2)
has been redesignated as new paragraph (c)(3) and has been revised by
replacing the term ``(i)n excess of 10,'' with the term ``(g)reater
than 50.'' These three changes: (1) Provide greater clarity and
mathematical consistency among paragraphs (c)(1), (c)(2), and (c)(3);
(2) clarify the CSI limits for storage incident to transport; and (3)
increase the CSI limit per package from 10 to 50 for shipments made
with nonexclusive use conveyances.
Section 71.61 Special Requirements for Type B Packages Containing More
Than 105A2
This section has been revised to require an enhanced water
immersion test for packages used for radioactive contents with activity
greater than 105A2. The title of this section has
also been revised to reflect that the scope has been broadened beyond
irradiated nuclear fuel.
Section 71.63 Special Requirement for Plutonium Shipments
The title has been revised to reflect only a single ``requirement''
rather than multiple requirements.
Paragraph (b) has been removed.
The designation of the remaining text as paragraph (a) has been
removed, because only one paragraph remains. The text of former
paragraph (a) has been revised to use plain language. The 0.74-TBq (20-
Ci) limit and solid form requirement have been retained.
Section 71.73 Hypothetical Accident Conditions
A new paragraph (c)(2) has been added to require a crush test for
fissile material packages.
Section 71.88 Air Transport of Plutonium
Paragraph (a)(2) has been revised to remove the 70-Bq/g (0.002-
[mgr]Ci/g) specific activity value and substitute activity
concentration values for plutonium found in Appendix A, Table A-2, of
this part. This revision is a conforming change to the revision to new
Sec. 71.14 to ensure consistent treatment of plutonium between these
two sections.
Subpart G--Operating Controls and Procedures
Section 71.91 Records
As a conforming change to subpart H, paragraphs (b) and (c) have
been redesignated as paragraphs (c) and (d), respectively, and are
revised by adding the terms ``certificate holder'' and ``applicant for
a CoC.'' New paragraph (b) has been added to require a certificate
holder to keep records on the model, serial number, and date of
manufacture of a packaging. These requirements are similar to the
requirements in paragraph (a), though less information is required. No
change has been made to paragraph (a).
Section 71.93 Inspection and Tests
As a conforming change to subpart H, paragraphs (a) and (b) have
been revised by adding the terms ``certificate holder'' and ``applicant
for a CoC.'' Paragraph (c) has been revised to require the certificate
holder to notify the NRC before it begins fabrication of a packaging
that can contain material having a decay heat load in excess of 5 kW or
a maximum normal operating pressure of 103 kPa (kilo Pascals) (15 lbf/
in2) gauge. This notification could be for either
fabricating a single packaging or the beginning of a campaign for
fabricating multiple packagings. This notification is in accordance
with the requirements of Sec. 71.1, rather than an NRC Regional
Administrator. This change in notification location reduces confusion
in identifying the appropriate Regional Administrator when the
certificate holder and fabrication location are overseas. Licensees
have been removed from this paragraph because the NRC believes that
requiring a licensee, who does not own the packaging, to notify the NRC
in advance of a packaging fabrication, when the licensee may not use
the packaging for years, is inappropriate and an unreasonable burden.
The NRC believes that requiring certificate holders and applicants for
a CoC to notify the NRC in advance of fabricating a packaging(s) would
allow the NRC adequate opportunity to inspect these activities. This
change is similar to the current requirement in Sec. 72.232(d) for part
72 certificate holders or applicants for a CoC to notify the NRC 45
days before starting the fabrication of the first storage cask under a
part 72 CoC. This action improves the harmonization between these two
regulations in parts 71 and 72.
Section 71.95 Reports
The existing introductory text and paragraphs (a), (b), and (c)
have been combined into a new paragraph (a) which requires a licensee,
after requesting the certificate holder's input, to submit a written
report to the NRC in certain circumstances. The requirement for the
licensee to request input from the certificate holder during
development of the written event report will ensure that design
deficiency issues have been thoroughly considered. The licensee will
also be required to provide the certificate holder with a copy of the
written event report, after the report is submitted to the NRC. This
will permit the certificate holder to monitor and trend the package
performance information, arising from package use by multiple
licensees. Additionally,
[[Page 3768]]
requirements on timing and submission location for the written reports
have been relocated to new paragraph (c). Furthermore, the 30-day
reporting requirement has been lengthened to a 60-day reporting
requirement.
The existing paragraph (c) has been redesignated as paragraph (b)
and revised for clarity.
New paragraphs (c) and (d) have been added to provide requirements
on the timing, submission location, form, and content of the written
reports.
Section 71.100 Criminal Penalties
Section 223 of the Atomic Energy Act of 1954, as amended, (the Act)
provides for criminal sanctions for willful violation of, attempted
violation of, or conspiracy to violate, any regulation issued under
sections 161b, 161i, or 161o of the Act. The Commission stated in a
final rule on ``Clarification of Statutory Authority for Purposes of
Criminal Enforcement'' (57 FR 55082; November, 24, 1992), that
substantive rules under sections 161b, 161i, or 161o of the Act include
those rules that create ``duties, obligations, conditions,
restrictions, limitations, and prohibitions.'' For the NRC to consider
the possibility of criminal sanctions for willful violation of,
attempted violation of, or conspiracy to violate, any substantive
regulations, the NRC must have clearly identified to affected parties
which regulations in part 71 are substantive rules. Accordingly,
paragraph (b) of this section identifies those part 71 regulations that
the NRC does not consider as substantive regulations. Thus, willful
violation of, attempted violation of, or conspiracy to violate any of
the regulations listed in paragraph (b) is not subject to possible
criminal sanctions.
Paragraph (b) of this section has been revised as a conforming
change. The NRC has reviewed new Sec.Sec. 71.10 and considers that this
regulation is not a substantive rule. Therefore, new Sec.Sec. 71.10 has
been added to the list of sections in paragraph (b). The NRC reviewed
new Sec.Sec. 71.9, 71.18, and 71.23 and considers that these
regulations are substantive rules. Therefore, these sections have not
been added to paragraph (b). Additionally, the NRC has reviewed the
existing Sec.Sec. 71.9, 71.10, and 71.53 and concluded these sections
should be recharacterized as substantive rules. Therefore, new Sec.Sec.
71.13, 71.14, and 71.18 have not been included in paragraph (b).
Additionally, existing Sec.Sec. 71.52 and 71.53 have been removed from
paragraph (b), because these section numbers have been removed from
part 71.
Subpart H--Quality Assurance
Section 71.101 Quality Assurance Requirements
Paragraph (a) has been revised by adding two new sentences to the
end of the paragraph specifying responsibilities for certificate
holders and applicants for a CoC.
Paragraph (b) has been revised to add the terms ``certificate
holder'' and ``applicant for a CoC.'' The second sentence has been
revised to provide greater clarity and consistency within subpart H by
referring to ``the QA requirement's importance to safety.''
Paragraph (c) has been revised by redesignating the existing text
as paragraph (c)(1), and new text has been added on submitting QA
programs in accordance with the requirements of Sec. 71.1. New
paragraph (c)(2) has been added to provide equivalent requirements on
the submission of QA programs for certificate holders and applicants
for a CoC.
Paragraph (f) has been revised to allow the use of existing NRC-
approved part 71 and part 72 QA programs, in lieu of submitting a new
QA program. Additionally, the terms ``certificate holder'' and
``applicant for a CoC'' have been added.
Paragraph (g) has been revised by making a minor change to clarify
that Sec. 34.31(b) is located in chapter I of title 10 of the Code of
Federal Regulations. Additionally, as a conforming change, Sec.
71.12(b) has been redesignated as Sec. 71.17(b).
Section 71.103 Quality Assurance Organization
Paragraph (a) has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.105 Quality Assurance Program
Paragraphs (a) through (d) have been revised by adding the terms
``certificate holder'' and ``applicant for a CoC.''
Section 71.107 Package Design Control
Paragraph (a) has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.'' Further, the last sentence has
been revised to improve clarity and consistency within subpart H by
referring to ``processes that are essential to the functions of the
materials, parts, and components that are important to safety.''
Paragraph (b) has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.'' Additionally, the last sentence
of paragraph (c) has been revised by replacing the text ``(c)hanges in
the conditions specified in the package approval require NRC approval *
* *.'' with ``(c)hanges in the conditions specified in the CoC require
NRC prior approval * * *.''
Section 71.109 Procurement Document Control
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.111 Instructions, Procedures, and Drawings
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.113 Document Control
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.115 Control of Purchased Material, Equipment, and Services
Paragraphs (a) through (c) have been revised by adding the terms
``certificate holder'' and ``applicant for a CoC.''
Section 71.117 Identification and Control of Materials, Parts, and
Components
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.119 Control of Special Processes
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.121 Internal Inspection
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.123 Test Control
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.125 Control of Measuring and Test Equipment
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.127 Handling, Storage, and Shipping Control
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.129 Inspection, Test, and Operating Status
Paragraph (a) has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
[[Page 3769]]
Section 71.131 Nonconforming Materials, Parts, or Components
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.133 Corrective Action
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.135 Quality Assurance Records
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Section 71.137 Audits
This section has been revised by adding the terms ``certificate
holder'' and ``applicant for a CoC.''
Appendix A to Part 71--Determination of A1 and A2
No changes have been made in paragraphs I, III, and V; however,
these paragraphs have been included due to revising Appendix A, in its
entirety.
Paragraph II has been revised to use plain language and has been
redesignated as subparagraph II(a). The intent of existing paragraph II
has not been changed; however, the reference to existing Table A-2 has
been revised as a conforming change to the new Table A-3. New paragraph
II(b) has been added to provide direction on determining exempt
material activity concentration and exempt consignment activity values
when a radionuclide has been identified as a constituent of a proposed
shipment, but the individual radionuclide is not listed in Table A-2.
Consequently, the structure of paragraphs II(a) and II(b) is the same.
New paragraph II(c) has been added to provide direction to licensees on
how to submit requests for Commission prior approval of either
A1 and A2 values or exempt material activity
concentration and exempt consignment activity values, for radionuclides
that are not listed in Tables A-1 and A-2, respectively.
Paragraph IV has been revised by adding new paragraphs (e) and (f)
to provide equations to use in determining a consolidated exempt
material activity concentration and exempt consignment activity value
when a shipment contains multiple radionuclides. The existing text
describing an alternative method for calculating the A1 or
A2 value of a mixture has been redesignated as paragraphs
(c) and (d). No changes have been made from the existing equations.
Appendix A, Table A-1--A1 and A2 Values for
Radionuclides
This Table has been revised to reflect the values from TS-R-1.
Appendix A, Table A-2--Exempt Material Activity Concentrations and
Exempt Consignment Activity Limits for Radionuclides
A new Table A-2 has been added to Appendix A of part 71. This table
contains the values of Exempt Material Activity Concentrations and
Exempt Consignment Activity Limits for selected radionuclides. Table A-
2 is referenced in new Sec. 71.14(a)(2) and is used in Sec. 71.14 to
determine when concentrations of material are not considered
radioactive material, for the purposes of transportation.
Appendix A, Table A-3--General Values for A1 and
A2
The existing Table A-2 has been redesignated as new Table A-3, and
the values have been revised to reflect the changes from TS-R-1.
Appendix A, Table A-4--Activity Mass Relationships for Uranium
The existing Table A-3 has been redesignated as new Table A-4. No
changes have been made to the values contained in new Table A-4.
V. Criminal Penalties
For the purposes of section 223 of the Atomic Energy Act (AEA), the
Commission is amending 10 CFR part 71 under one or more of sections
161b, 161i, or 161o of the AEA. Willful violations of the rule will be
subject to criminal enforcement.
The following is a list of substantive rule sections being revised
or added in this rulemaking: Sec.Sec. 71.1, 71.3, 71.5, 71.8, 71.9,
71.12, 71.13, 71.14, 71.15, 71.17, 71.19, 71.20, 71.21, 71.22, 71.23,
71.61, 71.63, 71.88, 71.91, 71.93, 71.95, 71.101, 71.103, 71.105,
71.107, 71.109, 71.111, 71.113, 71.115, 71.117, 71.119, 71.121, 71.123,
71.125, 71.127, 71.129, 71.131, 71.133, 71.135, 71.137.
VI. Issues of Compatibility for Agreement States
Under the ``Policy Statement on Adequacy and Compatibility of
Agreement State Programs'' which became effective on September 3, 1997
(62 FR 46517), NRC program elements (including regulations) are placed
into four compatibility categories. In addition, NRC program elements
also are identified as having particular health and safety significance
or as being reserved solely to the NRC. Compatibility Category A are
those program elements that are basic radiation protection standards
and scientific terms and definitions that are necessary to understand
radiation protection concepts. An Agreement State should adopt Category
A program elements in an essentially identical manner to provide
uniformity in the regulation of agreement material on a nationwide
basis. Compatibility Category B are those program elements that apply
to activities that have direct and significant effects in multiple
jurisdictions. An Agreement State should adopt Category B program
elements in an essentially identical manner. Compatibility Category C
are those program elements that do not meet the criteria of Category A
or B, but the essential objectives of which an Agreement State should
adopt to avoid conflict, duplication, gaps, or other conditions that
would jeopardize an orderly pattern in the regulation of agreement
material on a nationwide basis. An Agreement State should adopt the
essential objectives of the Category C program elements. Compatibility
Category D are those program elements that do not meet any of the
criteria of Category A, B, or C, and thus do not need to be adopted by
Agreement States for purposes of compatibility. A bracket around a
category means that the section may have been adopted elsewhere, and it
is not necessary to adopt it again. Health and Safety (H&S) are program
elements that are not required for compatibility (i.e., Category D) but
are identified as having a particular health and safety role (i.e.,
adequacy) in the regulation of agreement material within the State.
Although not required for compatibility, the State should adopt program
elements in this category based on those of NRC that embody the
essential objectives of the NRC program elements because of particular
health and safety considerations. Compatibility Category NRC are those
program elements that address areas of regulation that cannot be
relinquished to Agreement States pursuant to the Atomic Energy Act, as
amended, or provisions of title 10 of the Code of Federal Regulations.
These program elements should not be adopted by Agreement States. The
following table lists the part 71 revisions and their corresponding
categorization under the ``Policy Statement on Adequacy and
Compatibility of Agreement State Programs.'' This table has been
revised to incorporate comments received from the States of California
and Wisconsin during the 30-day Agreement States comment period which
began on June 3, 2003.
[[Page 3770]]
Part 71--Packaging and Transportation of Radioactive Material
------------------------------------------------------------------------
Regulation Compatibility
section Section title category Comments
------------------------------------------------------------------------
Sec. 71.0..... Purpose and D, except This requirement is
Scope. paragraph C is designated
[B]. Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this requirement in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this provision
is not necessary.
Sec. 71.1..... Communications D
and Records.
Sec. 71.2..... Interpretations. D
Sec. 71.3..... Requirements for [B]............. This requirement is
license. designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions since
it assures
authorization for
the transport of
licensed material.
An Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this requirement in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this provision
is not necessary.
Sec. 71.4..... Definitions:
A1.............. [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
A2.............. [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in mulitiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Carrier......... [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
[[Page 3771]]
Certificate D--for those This term is used in
holder. States which the sections
have no concerning quality
licensees that assurance programs
us Type B for Type B
packages. or packages. Those
States which have
no licensees that
use Type B packages
are not required to
adopt this
definition. This
definition is
designated
Compatibility
Category B for
those States which
have licensees that
us Type B packages
because it applies
to activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
[B]--for those
States which
have licensees
that use Type B
packages.
Certificate of D--for those This term is used in
compliance. States which the sections
have no concerning quality
licensees that assurance programs
use Type B for Type B
packages. packages. Those
[B]--for those States which have
States which no licensees that
have licensees use Type B packages
that use Type B are not required to
packages. adopt this
definition. This
definition is
designated
Compatibility
Category B for
those States which
have licensees that
use Type B packages
because it applies
to activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Close reflection D............... This definition is
by water. not required for
compatibility since
it defines a term
which pertains to
an area reserved to
NRC. A State may
adopt this
definition for
purposes of clarity
or communication.
This definition can
be adopted by
Agreement States
since it in and of
itself does not
convey any
authority whereby a
State can regulate
in an exclusive NRC
jurisdiction.
However, if a State
chooses to define
the term then the
definition should
be essentially
identical.
Consignment..... [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Containment D............... This term is not
System. used in any section
requiring Agreement
State adoption.
Conveyance...... [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
[[Page 3772]]
Criticality B............... This definition is
safety Index. designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
In addition, this
definition is
needed for a common
understanding
beyond a plain
dictionary meaning
of the term in
order to implement
10 CFR 71.22, 71.23
and 71.59.
Deuterium....... B............... This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
In addition, this
definition is
needed for a common
understanding
beyond a plain
dictionary meaning
of the term in
order to implement
Sec. 71.15.
DOT............. D............... This term does not
meet any of the
criteria of
Category A, B, C,
or H&S because it
is a widely
accepted
abbreviation for
the U. S.
Department of
Transportation.
Exclusive use... [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Fissile material [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Graphite........ B............... This definition is
needed for a common
understanding
beyond a plain
dictionary meaning
of the term in
order to implement
Sec. 71.15, which
has direct and
significant
transboundary
effects.
Licensed [D]............. This term does not
material. meet any of the
criteria of
Category A, B, C,
or H&S because it
is widely accepted
and understood.
This definition
also appears in 10
CFR 20.1003. For
purposes of
compatibility, the
language of the
Part 20 definition
should be used and
is assigned to
Compatibility
Category D.
Low Specific [B]............. This definition is
Activity (LSA) designated
material. Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
[[Page 3773]]
Low toxicity [B]............. This definition is
alpha emitters. designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Maximum normal D............... The definition of
operating the term ``maximum
pressure. normal operating
pressure'' was
changed from a
compatibility
category ``B'' to a
category ``D.''
This term is not
used in any section
requiring Agreement
State adoption; it
relates to the heat
conditions in Sec.
71.71(c)(1), which
is designated a
category ``NRC.''
This definition is
not required for
compatibility since
it defines a term
which pertains to
an area reserved to
the NRC. A State
may adopt this
definition for
purposes of clarity
or communication.
This definition can
be adopted by
Agreement States
since it is and of
itself does not
convey any
authority whereby a
State can regulate
in an exclusive NRC
jurisdiction.
However, if a State
chooses to define
this term, then the
definition should
be essentially
identical.
Natural thorium. [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Normal form [B]............. This definition is
radioactive designated
material. Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Optimum D............... This definition is
interspersed not required for
hydrogenous compatibility since
moderation. it defines a term
which pertains to
an area reserved to
NRC. A State may
adopt this
definition for
purposes of clarity
or communication.
This definition can
be adopted by
Agreement States
since it in and of
itself does not
convey any
authority whereby a
State can regulate
in an exclusive NRC
jurisdiction.
However, if a State
chooses to define
the term, then the
definition should
be essentially
identical.
Package......... [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
[[Page 3774]]
Fissile material [B]............. This definition is
package or Type designated
AF package, Compatibility
Type BF, Type Category B because
B(U)F package, it applies to
or Type B(M)F. activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
then the adoption
of this definition
is not necessary.
Type A package.. [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
then the adoption
of this definition
is not necessary.
Type B package.. [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
then the adoption
of this definition
is not necessary.
Packaging....... [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
then the adoption
of this definition
is not necessary.
Special form [B]............. This definition is
radioactive designated
material. Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
then the adoption
of this definition
is not necessary.
Specific [B]............. This definition is
activity. designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
then the adoption
of this definition
is not necessary.
Spent Nuclear D............... This definition is
Fuel or Spent not required
Fuel. compatibility since
it defines a term
which pertains to
an area reserved to
NRC. A State may
adopt this
definition for
purposes of clarity
or communication.
This definition can
be adopted by
Agreement States
since it in and of
itself does not
convey any
authority whereby a
State can regulate
in an exclusive NRC
jurisdiction.
However, if a State
chooses to define
the term, then the
definition should
be essentially
identical.
State........... D. ....................
[[Page 3775]]
Surface [B]............. This definition is
Contaminated designated
Object (SCO). Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
then the adoption
of this definition
is not necessary.
Transport Index. [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
then the adoption
of this definition
is not necessary.
Type A quantity. [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Type B quantity. [B]............. This definition is
designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Unirradiated [B]............. This definition is
uranium. designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Uranium-- [B]............. This definition is
natural, designated
depleted and Compatibility
enriched. Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this definition in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this definition
is not necessary.
Sec. 71.5..... Transportation [B]............. This requirement is
of Licensed designated
Material. Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this requirement
is not necessary.
Sec. 71.6..... Information D...............
collection
requirements:
OMB approval.
[[Page 3776]]
Sec. 71.7..... Completeness and D...............
accuracy of
Information.
Sec. 71.8..... Deliberate C............... The Commission
misconduct. determined in
response to SECY-97-
156 that Agreement
States should adopt
the essential
objectives of this
provision. The
essential
objectives of this
provision are
provided in
paragraphs (a),
(b), (c), and (d).
If deliberate
misconduct and
wrongdoing issues
involving Agreement
State licensees
were not pursued
and closed by
Agreement States,
then a potential
gap may be created
between NRC and
Agreement State
programs.
Sec. 71.9..... Employee D............... This provision does
Protection. not meet any of the
criteria for
designations
Category A, B, C,
or health and
safety. Thus, it
does not need to be
adopted by
Agreement States.
Sec. 71.10.... Public D............... This provision does
Inspection of not meet any of the
Application. criteria for
designations
Category A, B, C,
or health and
safety. Thus, it
does not need to be
adopted by
Agreement States.
Sec. 71.11.... [RESERVED]......
Sec. 71.12.... Specific D...............
exemptions.
Sec. 71.13.... Exemption for [B]............. This provision is
physicians. designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this requirement
is not necessary.
Sec. 71.14.... Exemptions for [B]-paragraph Paragraph (a) is
low level (a). designated as a
material. NRC--paragraph Compatibility
(b). Category B because
of its significant
transboundary
impacts with
respect to the
establishment of
exempt materials in
the area of
transportation. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this requirement in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this requirement
is not necessary.
Paragraph (b) is
designated
Compatibility
Category ``NRC.''
This provision is
reserved to the NRC
because it
delineates NRC's
authority from that
of DOT's in the
area of
transportation of
radioactive
materials. These
provisions
relinquish to DOT
the control of
types of shipment
that are of low
risk both from
radiation and
criticality
standpoints.
Further, to ensure
that only low
criticality risk
shipments are
included in the
area of DOT
authority, these
provisions restrict
the exemption to
Type A and low-
specific-activity
(LSA) or surface
contaminated
objects (SCOs) that
either contain no
fissile material or
satisfy the fissile
material exemption
requirements in
Sec. 71.11.
Finally, this
provision is
reserved to the NRC
because this
exemption does not
relieve licensees
from DOT
requirements by
reason of NRC's
authority. Thus,
Agreement States
should not adopt
this provision in
order to retain
their ability to
implement all of 49
CFR as directed by
DOT.
[[Page 3777]]
Sec. 71.15.... Exemptions from [B]............. This provision is
classification designated
as fissile Compatibility
material. Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this requirement
is not necessary.
Note: This
provision was
previously
designated ``NRC.''
It was changed to
``B'' to ensure
compatibility
between NRC and
Agreement States in
an area that has
significant and
direct
transboundary
implications.
During further
staff review, it
was noted that the
requirements in
this section
``Fissile material
exemptions'' is the
same as those of
DOT in 49 CFR
173.453, ``Fissile
materials
exceptions.'' Staff
noted that States
adopt these DOT
regulations as a
part of their
transportation
regulations. Staff
also noted that in
accordance with
Sec. 150.11, an
Agreement State can
regulate the
following fissile
materials: U-235 in
quantities not
exceeding 350
grams, U-233 in
quantities not
exceeding 200
grams; plutonium in
quantities not
exceeding 200
grams, or any
combination of
these materials
that would be
sufficient to form
a critical mass.
These requirements
would apply to the
materials Agreement
States regulate.
Thus, the
compatibility of
this requirement
was changed to a
``[B],'' which
indicates that if a
State has adopted
this provision as a
part of the State's
DOT regulations,
then the adoption
of this provision
is not necessary.
Sec. 71.16.... [RESERVED]......
Sec. 71.17.... General license: [B]............. This provision is
NRC--approved designated
package. Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this provision
is not necessary.
Sec. 71.19.... Previously NRC............. This provision is
approved reserved to the NRC
package. because it
addresses packages
intended for both
the storage and
transportation of
spent fuel.
Sec. 71.20.... General license: [B]............. This provision is
DOT designated
specification Compatibility
container Category B because
material. it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this provision
is not necessary.
Sec. 71.21.... General license: [B]............. This provision is
Use of foreign designated
approved Compatibility
package. Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this provision
is not necessary.
[[Page 3778]]
Sec. 71.22.... General license: [B]............. This provision
Fissile designated
material. Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this provision
is not necessary.
Note: A similar
provision was
previously
designated ``NRC.''
It was changed to
``B'' to ensure
compatibility
between NRC and
Agreement States in
an area that has
significant and
direct
transboundary
implications.
During further
staff review, it
was noted that in
accordance with 10
CFR 150.11, an
Agreement State can
regulate the
following fissile
materials: U-235 in
quantities not
exceeding 350
grams, U-233 in
quantities not
exceeding 200
grams; plutonium in
quantities not
exceeding 200
grams, or any
combination of
these materials
that would be
sufficient to form
a critical mass.
These requirements
would apply to the
materials Agreement
States regulate.
Thus, the
compatibility of
this requirement
was changed to a
``[B],'' which
indicates that if a
State has adopted
this provision as a
part of the State's
DOT regulations,
then the adoption
of this provision
is not necessary.
Sec. 71.23.... General license: [B]............. This provision is
Plutonium- designated
beryllium Compatibility
special form Category B because
material. it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this requirement
is not necessary.
Sec. 71.24.... [RESERVED]......
Sec. 71.25.... [RESERVED]......
Sec. 71.31.... Contents of NRC.............
Application.
Sec. 71.33.... Package NRC.............
description.
Sec. 71.35.... Package NRC.............
evaluation.
Sec. 71.37.... Quality NRC.............
Assurance.
Sec. 71.38.... Renewal of a NRC.............
certificate of
compliance or
quality
assurance
program
approval.
Sec. 71.39.... Requirements for NRC.............
additional
information.
Sec. 71.41.... Demonstration of NRC............. This provision is
Compliance. designated NRC
because it
addresses an area
reserved to NRC's
regulatory
authority.
Sec. 71.43.... General NRC.............
Standards for
all packages.
Sec. 71.45.... Lifting and tie- NRC.............
down Standards
for all
packages.
Sec. 71.47.... External [B]............. This requirement was
radiation changed from a
Standards for compatibility
all packages. category ``NRC'' to
``[B].'' This
provision was
changed because it
establishes the
external radiation
standards for all
transportation
packages. It is
essential that the
Agreement States
adopt this
provision in an
essentially
identical manner
because they have
direct and
significant
transboundary
effects. The
bracket,``B,''
indicates that a
State should adopt
this provision in
an essentially
identical manner
because of its
direct and
significant
transboundary
effects; however,
if a State has
adopted this
provision as a part
of its DOT
regulations, then
the adoption of
this section is not
necessary.
[[Page 3779]]
Sec. 71.51.... Additional NRC............. This provision is
Requirements designated NRC
for Type B because it
packages. addresses an area
reserved to NRC's
regulatory
authority.
Sec. 71.53.... [RESERVED]......
Sec. 71.55.... General NRC............. This provision is
Requirements designated NRC
for fissile because it
material addresses an area
packages. reserved to NRC's
regulatory
authority.
Sec. 71.57.... [RESERVED]......
Sec. 71.59.... Standards for NRC............. This provision is
arrays of designated NRC
fissile because it
material addresses an area
packages. reserved to NRC's
regulator
authority.
Sec. 71.61.... Special NRC............. This provision is
requirements designated NRC
for Type B because it
packages addresses an area
containing more reserved to NRC's
than 10\5\A2. regulatory
authority.
Sec. 71.63.... Special NRC............. This provision is
requirements designated NRC
for plutonium because it
shipments. addresses an area
reserved to NRC's
regulatory
authority.
Sec. 71.64.... Special NRC............. This provision is
requirements designated NRC
for plutonium because it
air shipments. addresses an area
reserved to NRC's
regulatory
authority.
Sec. 71.65.... Additional NRC............. This provision is
Requirements. designated NRC
because it
addresses an area
reserved to NRC's
regulatory
authority.
Sec. 71.71.... Normal NRC............. This provision is
conditions of designated NRC
transport. because it
addresses an area
reserved to NRC's
regulatory
authority.
Sec. 71.73.... Hypothetical NRC............. This provision is
accident designated NRC
conditions. because it
addresses an area
reserved to NRC's
regulatory
authority.
Sec. 71.74.... Accident NRC............. This provision is
conditions for designated NRC
air transport because it
of plutonium. addresses an area
reserved to NRC's
regulatory
authority.
Sec. 71.75.... Qualification of NRC............. This provision is
special form designated NRC
radioactive because it
material. addresses an area
reserved to NRC's
regulatory
authority.
Sec. 71.77.... Qualification of NRC............. This provision is
LSA-III designated NRC
material. because it
addresses an area
reserved to NRC's
regulatory
authority.
Sec. 71.81.... Applicability of D............... This requirement was
operating changed from a
controls. compatibility
category ``B'' to
``D.'' This
designation was
changed because it
does not meet any
of the criteria for
designation as
Category A, B, C or
Health and Safety
and is not required
for the purposes of
compatibility.
Sec. 71.83.... Assumptions as [B]............. This requirement was
to unknown changed from a
properties. compatibility
category ``NRC'' to
``[B].'' Agreement
States can regulate
fissile material
below 350g. This
provision is needed
to address fissile
material regulated
by the States and
to assure that a
regulatory gap in
the regulations of
these materials is
not created. The
bracket, ``b,''
indicates that a
State should adopt
this provision in
an essentially
identical manner
because of its
direct and
significant
transboundary
effects; however,
if a State has
adopted this
provision as a part
of its DOT
regulations, then
the adoption of
this section is not
necessary.
Sec. 71.85.... Preliminary [B]............. This provision is
determinations. designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this provision
is not necessary.
[[Page 3780]]
Sec. 71.87.... Routine [B]............. This provision is
determinations. designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this provision
is not necessary.
Sec. 71.88.... Air transport of [B]............. This provision is
plutonium. designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this regulation
is not necessary.
Sec. 71.89.... Opening [B]............. This provision is
instructions. designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this regulation
is not necessary.
Sec. 71.91.... Records......... D............... This provision does
not meet any of the
criteria for
designations
Category A, B, C,
or health and
safety. Thus, it
does not need to be
adopted by
Agreement States.
Sec. 71.93.... Inspection and D............... This provision does
tests. not meet any of the
criteria for
designations
Category A, B, C,
or health and
safety. Thus, it
does not need to be
adopted by
Agreement States.
Sec. 71.95.... Reports......... D............... This provision does
not meet any of the
criteria for
designations
Category A, B, C,
or health and
safety. Thus, it
does not need to be
adopted by
Agreement States.
Sec. 71.97.... Advance B............... This provision is
notification of designated
shipment of Compatibility
irradiated Category B because
reactor fuel it applies to
and nuclear activities that
waste. have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
Sec. 71.99.... Violations...... D...............
Sec. 71.100... Criminal D...............
penalties.
[[Page 3781]]
Sec. 71.101... Quality D--Paragraphs Paragraphs (a), (b),
assurance (a), (b), and and (c)(1) are
requirements. (c)(1) are designated Category
designated D C and the essential
for those objectives of these
States which provisions should
have no users be adopted by those
of Type B Agreement States
packages-other which have
than Industrial licensees who use
Radiography**. Type B packages.
C--Paragraphs These provisions
(a), (b) and are designated
(c)(1) are Category C because
designated C the quality
for those assurance of Type B
States which packages is an
have users of activity that is
Type B packages- needed in order to
other than avoid a nationwide
Industrial gap in the
Radiography.**. regulation of the
D--paragraph (f) transportation of
C--paragraph (g) radioactive
NRC-paragraphs materials. If these
(c)(2), (d) and provisions are not
(e). adopted, this could
**Note: 10 CFR result in
71.101(g) undesirable
indicates that consequences in
QA programs for multiple
industrial jurisdictions. The
radiography essential objective
Type B package of paragraph (a) is
users are that each licensee
covered by 10 who uses a Type B
CFR 34.31(b). package is
It also responsible for the
indicated that quality assurance
this section requirements which
satisfies Sec. apply to the use of
71.12 (b) and a package. The
thus would essential objective
satisfy those of paragraph (b) is
secitons that each licensee
referenced in who uses a Type B
this provision package shall
(Sec.Sec. establish,
71.101 through amintain, and
71.137). execute a quality
assurance program.
The essential
objective of
paragraph (c)(1) is
that each licensee
who uses a Type B
package shall,
prior to the use of
any package for the
shipment of any
material subject to
this part, obtain
approval of its
quality assurance
program by the
regulatory agency.
Paragraph (f) is not
required for
compatibility
because the States
have the
felxibility to
determine whether
they wish to accept
a previously
approved quality
assurance program.
Sec. 71.103... Quality D--for those For paragraph (a),
assurance States which those States which
organization. have no users have licenses that
of Type B use Type B
packages-other packages, and have
than Industrial adopted the
Radiography**. essential
[C]--Paragraph objectives of Sec.
(a) is 71.101(a), it is
designated [C] not necessary for
for those them to adopt this
States which provision again.
have users of Paragraph (b) is
Type B packages- designated as a
other than Category C, and the
Industrial essential
Radiography**. objectives of these
C--Paragraph (b) provisions should
is designated C be adopted by those
for those Agreement States
States which which have
have users of licensees who use
Type B packages- Type B packages.
other than This provision is
Industrial designated Category
Radiography**. C because the
D--paragraphs quality assurance
(d), (e), and of Type B packages
(f). is an activity that
**Note: Sec. is needed in order
71.101 (g) to avoid a
indicates that nationwide gap in
QA programs for the regulation of
industrial the transportation
radiography of radioactive
Type B package materials. If these
users are provisions are not
covered by Sec. adopted, this could
34.31(b). It result in
also indicated undesirable
that this consequences in
section multiple
satisfies Sec. jurisdictions. The
71.12(b) and essential objective
thus would of paragraph (b) is
satisfy those that each licensee
sections who uses a Type B
referenced in package should
this provision verify by
Sec.Sec. 71.101 procedures such as
through 71.137). checking, auditing,
and inspection,
that activities
affecting the
safety-related
functions have been
performed
correctly.
[[Page 3782]]
Sec. 71.105... Quality D--for those Para. (a) is
assurance States which designated [C] and
program. have no users para. (b) is
of Type B designated C for
packages--other those Agreement
than Industrial States with
Radiography. licensees that use
C--Paragraphs Type B packages and
(a), (c), and the essential
(d) and. objectives of these
[C]--paragraph b provisions should
for those be adopted by those
States which Agreement States.
have users of These provisions
Type B packages- are designated
-other than Category C because
Industrial the QA of Type B
Radiography**. packages is an
**Note: 10 CFR activity that is
71.101(g) needed in order to
indicates that avoid a nationwide
QA programs for regulatory gap in
industrial the regulation of
radiography the transportation
Type B package of radioactive
users are materials. If these
covered by 10 provisions are not
CFR 34.31(b). adopted, this could
It also result in
indicated that undesirable
this section consequences in
satisfies Sec. multiple
71.12(b) and jurisdictions. The
thus would essential objective
satisfy those of para. (a) is
sections that each licensee
referenced in who uses a Type B
this provision package shall
(Sec.Sec. document the
71.101 through quality assurance
71.137). program by written
procedures or
instructions and
shall carry out the
program in
accordance with
those procedures
throughout the
period during which
the packaging is
used, and shall
identify the
material and
components covered
by the quality
assurance program.
The essential
objective of para.
(b) is that each
licensee who uses a
Type B package
shall control
activities
affecting the
safety-related
functions of the
Type B package.
Para. (b) is
bracketed ``C'',
because the
essential objective
of this provision
is captured by Sec.
71.103(b); if an
Agreement State
adopts the
essential
objectives of Sec.
71.103(b), it is
not necessary to
adopt this
provision again.
The essential
objective of para.
(c) is that the
licensee and
certificate holder
shall base its QA
program on items
listed in (1)
through (5). The
essential objective
of para. (d) is
that the licensee
and certificate
holder shall
provide training of
personnel
performing
activities
affecting the
quality of the
package to assure
proficiency in
their knowledge of
the QA program;
review the status
and adequacy of the
QA program at
established
intervals; and
regular management
review of the QA
program by all
cognizant
organizations.
Sec. 71.107... Package design NRC............. This provision is
control. reserved to the NRC
because it
addresses the
design,
fabrication,
modification, and
approval of Type B
packages.
Sec. 71.109... Procurement NRC............. This provision is
document reserved to the NRC
control. because it
addresses the
design,
fabrication,
modification, and
approval of Type B
packages.
Sec. 71.111... Instructions, NRC............. This provision is
procedures, and reserved to the NRC
drawings. because it
addresses the
design,
fabrication,
modification, and
approval of Type B
packages.
Sec. 71.113... Document control NRC............. This provision is
reserved to the NRC
because it
addresses the
design,
fabrication,
modification, and
approval of Type B
packages.
Sec. 71.115... Control of NRC............. This provision is
purchased reserved to the NRC
material, because it
equipment, and addresses the
services. design,
fabrication,
modification, and
approval of Type B
packages.
Sec. 71.117... Identification NRC............. This provision is
and control of reserved to the NRC
materials, because it
parts, and addresses the
components. design,
fabrication,
modification, and
approval of Type B
packages.
Sec. 71.119... Control of NRC............. This provision is
special reserved to the NRC
processes. because it
addresses the
design,
fabrication,
modification, and
approval of Type B
packages.
Sec. 71.121... Internal NRC............. This provision is
Inspection. reserved to the NRC
because it
addresses the
design,
fabrication,
modification, and
approval of Type B
packages.
Sec. 71.123... Test control.... NRC............. This provision is
reserved to the NRC
because it
addresses the
design,
fabrication,
modification, and
approval of Type B
packages.
[[Page 3783]]
Sec. 71.125... Control of NRC............. This provision is
measuring and reserved to the NRC
test equipment. because it
addresses the
design,
fabrication,
modification, and
approval of Type B
packages.
Sec. 71.127... Handling, D--for those This provision is
storage, and States which designated Category
shipping have no users C for those States
control. of Type B which have
packages--other licensees that use
than Industrial Type B packages.
Radiography. This provision is
[C]--for those designated Category
States which C because the
have users of quality assurance
Type B packages- of Type B packages
-other than is an activity that
Industrial is needed in order
Radiography**. to avoid nationwide
**Note: 10 CFR gas in the
71.101 (g) regulation of the
indicates that transportation of
QA programs for radioactive
industrial materials. If this
radiography provision is not
Type B package adopted, this could
users are result in
covered by Sec. undesirable
34.31(b). It consequences in
also indicated multiple
that this jurisdications. For
section those States which
satisfies Sec. have licensees that
71.12(b) and use Type B
thus would packages, and have
satisfy those adopted the
sections essential
referenced in objectives of Sec.
this provision 71.105, it is not
(Sec.Sec. necessary for them
71.101 through to adopt this
71.137). provision again.
Sec. 71.129... Inspection, D--for those This provision is
test, and States which designated Category
operating have no users C because the
status. of Type B quality assurance
packages--other of Type B packages
than Industrial is an activity that
Radiography**. is needed in order
[C]--for those to avoid a
States which nationwide gap in
have users of the regulation of
Type B packages- the transportation
-other than of radioactive
Industrial materials. If this
Radiography**. provision is not
**Note: 10 CFR adopted, this could
71.101 (g) result in
indicates that undesirable
QA programs for consequences in
industrial multiple
radiography jurisdictions. For
Type B package those States which
users are have licensees that
covered by Sec. use Type B
34.31(b). It packages, and have
also indicated adopted the
that this essential
section objectives of Sec.
satisfies Sec. 71.105, it is not
71.12(b) and necessary for them
thus would to adopt this
satisfy those provision again.
sections
referenced in
this provision
(Sec.Sec.
71.101 through
71.137).
Sec. 71.131... Nonconforming D--for those This provision is
materials, States which designated Category
parts, or have no users C because the
components. of Type B quality assurance
packages-other of Type B packages
than Industrial is an activity that
Radiography**. is needed in order
[C]--for those to avoid a
States which nationwide gap in
have users of the regulation of
Type B packages- the transportation
-other than of radioactive
Industrial materials. If this
Radiography**. provision is not
**Note: 10 CFR adopted, this could
71.101 (g) result in
indicates that undesirable
QA programs for consequences in
industrial multiple
radiography jurisdictions. For
Type B package those States which
users are have licensees that
covered by Sec. use Type B
34.31(b). It packages, and have
also indicated adopted the
that this essential
section objectives of Sec.
satisfies Sec. 71.105, it is not
71.12(b) and necessary for them
thus would to adopt this
satisfy those provision again.
sections
referenced in
this provision
(Sec.Sec.
71.101 through
71.137).
Sec. 71.133... Corrective D--for those This provision is
action. States which designated Category
have no users C for those States
of Type B which have
packages--other licensees that use
than Industrial Type B packages.
Radiography**. This provision is
C--for those designated Category
States which C because the
have users of quality assurance
Type B packages- of Type B packages
-other than is an activity that
Industrial is needed in order
Radiography**. to avoid a
**Note: 10 CFR nationwide gap in
71.101 (g) the regulation of
indicates that the transportation
QA programs for of radioactive
industrial materials. If this
radiography provision is not
Type B package adopted, this could
users are result in
covered by Sec. undesirable
34.31(b). It consequences in
also indicated multiple
that this jurisdictions. The
section essential objective
satisfies Sec. of this provision
71.12(b) and is that each
thus would licensee who uses a
satisfy those Type B package
sections shall establish
referenced in measures to assure
this provision that conditions
(Sec.Sec. adverse to quality,
71.101 through such as
71.137). deficiencies,
deviations,
defective material
and equipment, and
nonconformances,
are promptly
identified and
corrected.
[[Page 3784]]
Sec. 71.135... Quality D--for those This provision is
assurance States which designated a
records. have no users Category C for
of Type B those States which
packages--other have licensees that
than industrial use Type B
Radiography**. packages. This
C--for those provision is
States which designated Category
have users of C because the
Type B packages- quality assurance
-other than of Type B packages
industrial is an activity that
radiography**. is needed in order
**Note: 10 CFR to avoid a
71.101(g) nationwide gap in
indicates that the regulation of
QA programs for the transportation
industrial of radioactive
radiography materials. If this
Type B package provision is not
users are adopted, this could
covered by Sec. result in
34.31(b). It undesirable
also indicated consequences in
that this multiple
section jurisdictions. The
satisfies Sec. essential objective
71.12(b) and of this provision
thus would is that each
satisfy those licensee who uses a
sections Type B package
referenced in shall maintain
this provision sufficient written
(Sec.Sec. records to
71.101 through demonstrate
71.137). compliance with the
quality assurance
program.
Sec. 71.137... Audits.......... D--for those This provision is
States which designated a
have no users Category C for
of Type B those States which
packages--other have licensees that
than Industrial use Type B
Radiography**. packages. This
C--for those provision is
States which designated Category
have users of C because the
Type B packages- quality assurance
-other than of Type B packages
Industrial is an activity that
Radiography**. is needed in order
**Note: 10 CFR to avoid a
71.101(g) nationwide gap in
indicates that the regulation of
QA program for the transportation
industrial of radioactive
radiography materials. If this
Type B package provision is not
users are adopted, this could
covered by Sec. result in
34.31(b). It undesirable
also indicated consequences in
that this multiple
section jurisdictions. The
satisfies Sec. essential
71.12(b) and objectives of this
thus would provision are that
satisfy those each licensee who
sections uses a Type B
referenced in package shall carry
this provision out a system of
Sec.Sec. 71.101 planned and
through 71.137). periodic audits to:
(1) verify
compliance with all
aspects of the
quality assurance
program, (2)
determine the
effectiveness of
the program, (3)
verify that the
audits are
performed by
appropriately
trained personnel,
(4) audits
performed in
accordance with
procedures; (5)
audit results
documented and
reviewed by
appropriate
management; and (6)
follow-up actions
are taken as
necessary.
Appendix A.... Determination of [B]............. This definition is
A1 and A2. designated
Compatibility
Category B because
it applies to
activities that
have direct and
significant effects
in multiple
jurisdictions. An
Agreement State
should adopt
Category B program
elements in an
essentially
identical manner.
The bracket, ``B,''
indicates that if a
State has adopted
this provision in
another portion of
its regulations,
such as the State's
DOT regulations,
then the adoption
of this requirement
is not necessary.
------------------------------------------------------------------------
VII. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995, Pub.
L. 104-113, requires that Federal agencies use technical standards that
are developed or adopted by voluntary consensus standard bodies unless
the use of such a standard is inconsistent with applicable law or
otherwise impractical. In this rule, the NRC considered but decided not
to adopt the ASME Code, Section III, Division 3, as described in Issue
14. However, NRC has amended its transportation regulations to make
them compatible with the IAEA transportation standards. This action
does not constitute the establishment of a standard that establishes
generally applicable requirements.
VIII. Environmental Assessment: Finding of No Significant Environmental
Impact
The Commission has prepared an environmental assessment entitled
Final Environmental Assessment (EA) of Major Revision of 10 CFR part 71
(NUREG/CR-6711, December 2003), on this regulation. The EA is available
on the NRC rulemaking Web site (http://ruleforum.llnl.gov) and is also
available for inspection in the NRC Public Document Room, 11555
Rockville Pike, Room O-1F21, Rockville, MD. The following is a brief
summary of the EA.
The EA grouped the proposed action into 19 different changes to
part 71, which could be adopted either all together as one list or
independently in a partial list. Of these 19 changes, the following 4
meet the NRC's categorical exclusion criteria:
Changes to Various Definitions (Issue 9);
Expansion of Part 71 Quality Assurance
Requirements to Certificate of Compliance (CoC) Holders (Issue 13);
Change Authority for Dual-Purpose Package
Certificate Holders (Issue 15); and
Modifications of Event Reporting Requirements
(Issue 19).
None of the remaining 15 changes are expected to cause a
significant impact to human health, safety, or the environment, whether
issued altogether or individually. In fact, most of the
[[Page 3785]]
changes would have negligible effects or result in slight improvements
in health, safety, and environmental protection. In particular, the
following changes are primarily administrative in nature, would not
cause any new negative impacts, and would result in the beneficial
effect of simplifying and/or harmonizing the NRC's regulations with TS-
R-1:
Changing Part 71 to the International System of
Units (SI) Only (Issue 1);
Revision of A1 and A2
(Issue 3);
A new requirement to display the Criticality
Safety Index on shipping packages of fissile material (Issue 5);
A provision to ``grandfather'' older shipping
packages under the part 71 requirements in existence when their
Certificates of Compliance were issued (Issue 8); and
Procedures for approval of special arrangements
for shipment of special packages (Issue 12).
The following changes would result in slight net improvements in
health, safety, and environmental protection:
Addition of uranium hexafluoride package
requirements (Issue 4);
Strengthening the requirements in Sec. 71.61 to
ensure package containment in deep submersion scenarios (Issue 7);
Adoption of the crush test for fissile material
package design (Issue 10);
Adoption of fissile material package design
requirements for transport by aircraft (Issue 11); and
Adoption of the ASME Code for spent fuel
transportation casks (Issue 14).
The proposal to change the existing 70-Bq/g (0.002-[mu]Ci/g) level
to radionuclide-specific activity limits (Issue 2) is expected to have
mixed, although overall minor, effects. For radionuclides with new
exemption values that are lower than the current limit, there could be
a decrease in the number of exempted shipments and a commensurate
slight increase in the level of protection. For radionuclides with new
exemption values that are higher than the current limit, there could be
an increase in the number of exempted shipments and a commensurate
slight increase in associated radiation exposures. However, IAEA and
the NRC have determined that this change would not significantly
increase the risk to individuals.
The addition of the Type C package and low level dispersible
material concepts (Issue 6) would result in mixed, although overall
minor, effects. If the same number of packages are handled, the
radiation doses to workers loading and unloading Type C packages
shipped by air will be slightly higher than the doses to workers
loading and unloading other kinds of packages shipped by other means.
At the same time, ``incident-free'' doses during the shipping of Type C
packages are expected to be slightly reduced compared to baseline
conditions, while the risks associated with accidents during shipping
could be slightly increased or decreased depending on the shipping
scenario.
Changes to transportation regulations for fissile materials
actually consist of 17 individual recommendations for revisions to part
71 (Issue 16). Ten of these recommendations are expected to result in
no impact, as they simply clarify definitions, consolidate related
requirements into single sections, or streamline the regulations. Four
of the recommendations will result in small improvements to health,
safety, and environmental protection by eliminating confusion among
licensees and/or providing added assurance for critical safety. The
last two recommendations, which would revise exemptions for low-level
material and remove or modify provisions related to the shipment of Pu-
Be neutron sources, are expected to significantly improve criticality
safety.
Changes to the requirements for plutonium shipments in Sec. 71.63
(PRM-71-12) could result in a slight increase in the probability and
consequences of accidental releases, primarily when and if plutonium is
shipped in liquid form. However, most plutonium shipments are either
related to the disposition of plutonium wastes or to the production of
mixed oxides, neither of which involve the shipment of a liquid
solution of plutonium.
No changes have been identified for the issue related to surface
contamination limits as applied to spent fuel and high level waste
(Issue 18). The issue was included in the proposed rule in response to
Commission direction in SRM-SECY-00-0117. NRC is seeking input on
whether the NRC should address this issue in future rulemaking
activities. As a result, no regulatory options were developed, and
therefore no environmental assessment conducted.
The Commission has determined, under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
subpart A of 10 CFR part 51, that this rule is not a major Federal
action significantly affecting the quality of the human environment,
and therefore an environmental impact statement (EIS) is not required.
The Commission's ``Final Environmental Statement on the
Transportation of Radioactive Material by Air and Other Modes,'' NUREG-
0170 \14\, dated December 1977, is NRC's generic EIS, covering all
types of radioactive material transportation by all modes (road, rail,
air, and water). From the Commission's latest survey of radioactive
material shipments and their characteristics, ``Transport of
Radioactive Material in the United States,'' SAND 84-7174, April 1985,
the NRC concluded that current radioactive material shipments are not
so different from those evaluated in NUREG-0170 as to invalidate the
results or conclusions of that EIS. The environmental assessment of the
impacts associated with this rulemaking is evaluated in Final
Environmental Assessment (EA) of Major Revision of 10 CFR part 71
(NUREG/CR-6711, December 2003).
---------------------------------------------------------------------------
\14\ Copies of NUREG-0170 may be purchased from the
Superintendent of Documents, U.S. Government Printing Office, P.O.
Box 37082, Washington, DC 20013-7082. Copies are also available from
the National Technical Information Service, 5285 Port Royal Road,
Springfield, VA 22161. A copy is also available for inspection and
copying for a fee in the NRC Public Document Room, 11555 Rockville
Pike, Room O-1F21, Rockville, MD.
---------------------------------------------------------------------------
NUREG-0170 established the nonaccident related radiation exposures
associated with transportation of radioactive material in the United
States as 98 person-Sv (9800 person-rem) which, based on the
conservative linear radiation dose hypothesis, resulted in a maximum of
1.7 genetic effects and 1.2 latent cancer effects per year. More than
half this impact resulted from shipment of medical-use radioactive
materials. Accident related impacts were established at a maximum of
one genetic effect and one latent cancer fatality for 200 years of
transporting radioactive materials. The principal nonradiological
impacts were found to be two injuries per year and less than one
accidental death per 4 years. In contrast, nonaccident related
radiation exposures and accident related impacts associated with this
rulemaking would not change from the impact of the current part 71
requirements (i.e., no increase or decrease). Nonradiological traffic
injuries and nonradiological traffic deaths would not change. These
impacts are judged to be insignificant compared with the baseline
impacts established in NUREG-0170.
IX. Paperwork Reduction Act Statement
This final rule amends information collection requirements that are
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). These
[[Page 3786]]
requirements were approved by the Office of Management and Budget,
approval number 3150-0008.
The burden to the public for these information collections is
estimated to average 19.2 hours per licensee, including the time for
reviewing instructions, searching existing data sources, gathering and
maintaining the data needed, and completing and reviewing the
information collection. Send comments on any aspect of these
information collections, including suggestions for reducing the burden,
to the Records Management Branch (T-5F52), U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, or by Internet electronic mail
to [email protected]; and to the Desk Officer, Office of Information
and Regulatory Affairs, NEOB-10202,(3150-0008), Office of Management
and Budget, Washington, DC 20503.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement unless the requesting document displays a currently valid
OMB control number.
X. Regulatory Analysis
The Commission has prepared a regulatory analysis entitled ``Final
Regulatory Analysis of Major Revision of 10 CFR part 71--NUREG/CR-6713,
December 2003. `` To support the discussions of the proposed changes,
selected material from this regulatory analysis has been included
earlier under each issue. The analysis examines the costs and benefits
of the alternatives considered by the Commission. The regulatory
analysis is available on the NRC rulemaking Web site, and is also
available for inspection at the NRC Public Document Room, 11555
Rockville Pike, Room O-1F21, Rockville, MD.
XI. Regulatory Flexibility Act Certification
In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C.
605(b)), the Commission certifies that this rule will not have a
significant economic impact on a substantial number of small entities.
This rule affects NRC licensees, including operators of nuclear power
plants, who transport or deliver to a carrier for transport, relatively
large quantities of radioactive material in a single package. These
companies do not generally fall within the scope of the definition of
``small entities'' set forth in the Regulatory Flexibility Act or the
size standards adopted by the NRC (10 CFR 2.810).
Only one small entity commented on the proposed changes suggesting
that small entities would be negatively affected by the rule. Reviewing
records of licensed QA programs, NRC found that only 15 of the 127 NRC-
licensed QA progams were small entities. Furthermore, of these 15
companies, NRC staff expects that only two or three would be negatively
affected by the final rule, given these companies' lines of business
and day-to-day operations. Based on these data, it is believed there
will not be significant economic impacts for a substantial number of
small entities.
XII. Backfit Analysis
The NRC has determined that the backfit rule does not apply to this
rule; therefore, a backfit analysis is not required for this rule
because these amendments do not involve any provisions that would
require backfits as defined in 10 CFR chapter I.
List of Subjects in 10 CFR Part 71
Criminal penalties, Hazardous materials transportation, Nuclear
materials, Packaging and containers, Reporting and recordkeeping
requirements.
0
For the reasons set out in the preamble and under the authority of the
Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of
1974, as amended; and 5 U.S.C. 552 and 553, the Commission is adopting
the following amendments to 10 CFR part 71.
PART 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL
0
1. The authority citation for part 71 continues to read as follows:
Authority: Secs. 53, 57, 62, 63, 81, 161, 182, 183, 234, 68
Stat. 930, 932, 933, 935, 948, 953, 954, as amended, sec. 1701, 106
Stat. 2951, 2952, 2953 (42 U.S.C. 2073, 2077, 2092, 2093, 2111,
2201, 2232, 2233, 2297f); secs. 201, as amended, 202, 206, 88 Stat.
1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); sec.
1704, 112 Stat. 2750 (44 U.S.C. 3504 note).
Section 71.97 also issued under sec. 301, Pub. L. 96-295, 94
Stat. 789-790.
0
2. Subparts A, B, and C to part 71 are revised to read as follows:
Subpart A--General Provisions
Sec.
71.0 Purpose and scope.
71.1 Communications and records.
71.2 Interpretations.
71.3 Requirement for license.
71.4 Definitions.
71.5 Transportation of licensed material.
71.6 Information collection requirements: OMB approval.
71.7 Completeness and accuracy of information.
71.8 Deliberate misconduct.
71.9 Employee protection.
71.10 Public inspection of application.
71.11 [Reserved]
Subpart B--Exemptions
71.12 Specific exemptions.
71.13 Exemption of physicians.
71.14 Exemption for low-level materials.
71.15 Exemption from classification as fissile material.
71.16 [Reserved]
Subpart C--General Licenses
71.17 General license: NRC-approved package.
71.18 [Reserved]
71.19 Previously approved package.
71.20 General license: DOT specification container.
71.21 General license: Use of foreign approved package.
71.22 General license: Fissile material.
71.23 General license: Plutonium-beryllium special form material.
71.24 [Reserved]
71.25 [Reserved]
Subpart A--General Provisions
Sec. 71.0 Purpose and scope.
(a) This part establishes--
(1) Requirements for packaging, preparation for shipment, and
transportation of licensed material; and
(2) Procedures and standards for NRC approval of packaging and
shipping procedures for fissile material and for a quantity of other
licensed material in excess of a Type A quantity.
(b) The packaging and transport of licensed material are also
subject to other parts of this chapter (e.g., 10 CFR parts 20, 21, 30,
40, 70, and 73) and to the regulations of other agencies (e.g., the
U.S. Department of Transportation (DOT) and the U.S. Postal Service)
\1\ having jurisdiction over means of transport. The requirements of
this part are in addition to, and not in substitution for, other
requirements.
---------------------------------------------------------------------------
\1\ Postal Service manual (Domestic Mail Manual), Section 124,
which is incorporated by reference at 39 CFR 111.1.
---------------------------------------------------------------------------
(c) The regulations in this part apply to any licensee authorized
by specific or general license issued by the Commission to receive,
possess, use, or transfer licensed material, if the licensee delivers
that material to a carrier for transport, transports the material
outside the site of usage as specified in the NRC license, or
transports that material on public highways. No provision of this part
authorizes possession of licensed material.
(d)(1) Exemptions from the requirement for license in Sec. 71.3 are
specified in Sec. 71.14. General licenses for which no NRC package
approval is
[[Page 3787]]
required are issued in Sec.Sec. 71.20 through 71.23. The general
license in Sec. 71.17 requires that an NRC certificate of compliance or
other package approval be issued for the package to be used under this
general license.
(2) Application for package approval must be completed in
accordance with subpart D of this part, demonstrating that the design
of the package to be used satisfies the package approval standards
contained in subpart E of this part, as related to the tests of subpart
F of this part.
(3) A licensee transporting licensed material, or delivering
licensed material to a carrier for transport, shall comply with the
operating control requirements of subpart G of this part; the quality
assurance requirements of subpart H of this part; and the general
provisions of subpart A of this part, including DOT regulations
referenced in Sec. 71.5.
(e) The regulations of this part apply to any person holding, or
applying for, a certificate of compliance, issued pursuant to this
part, for a package intended for the transportation of radioactive
material, outside the confines of a licensee's facility or authorized
place of use.
(f) The regulations in this part apply to any person required to
obtain a certificate of compliance, or an approved compliance plan,
pursuant to part 76 of this chapter, if the person delivers radioactive
material to a common or contract carrier for transport or transports
the material outside the confines of the person's plant or other
authorized place of use.
(g) This part also gives notice to all persons who knowingly
provide to any licensee, certificate holder, quality assurance program
approval holder, applicant for a license, certificate, or quality
assurance program approval, or to a contractor, or subcontractor of any
of them, components, equipment, materials, or other goods or services,
that relate to a licensee's, certificate holder's, quality assurance
program approval holder's, or applicant's activities subject to this
part, that they may be individually subject to NRC enforcement action
for violation of Sec. 71.8.
Sec. 71.1 Communications and records.
(a) Except where otherwise specified, all communications and
reports concerning the regulations in this part and applications filed
under them should be sent by mail addressed: ATTN: Document Control
Desk, Director, Spent Fuel Project Office, Office of Nuclear Material
Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001, by hand delivery to the NRC's offices at 11555 Rockville
Pike, Rockville, Maryland; or, where practicable, by electronic
submission, for example, via Electronic Information Exchange, or CD-
ROM. Electronic submissions must be made in a manner that enables the
NRC to receive, read, authenticate, distribute, and archive the
submission, and process and retrieve it a single page at a time.
Detailed guidance on making electronic submissions can be obtained by
visiting the NRC's Web site at http://www.nrc.gov/site-help/eie.html,
by calling (301) 415-6030, by e-mail to [email protected], or by writing the
Office of the Chief Information Officer, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001. The guidance discusses, among
other topics, the formats the NRC can accept, the use of electronic
signatures, and the treatment of nonpublic information.
(b) Each record required by this part must be legible throughout
the retention period specified by each Commission regulation. The
record may be the original or a reproduced copy or a microform provided
that the copy or microform is authenticated by authorized personnel and
that the microform is capable of producing a clear copy throughout the
required retention period. The record may also be stored in electronic
media with the capability for producing legible, accurate, and complete
records during the required retention period. Records such as letters,
drawings, and specifications must include all pertinent information
such as stamps, initials, and signatures. The licensee shall maintain
adequate safeguards against tampering with and loss of records.
Sec. 71.2 Interpretations.
Except as specifically authorized by the Commission in writing, no
interpretation of the meaning of the regulations in this part by any
officer or employee of the Commission, other than a written
interpretation by the General Counsel, will be recognized to be binding
upon the Commission.
Sec. 71.3 Requirement for license.
Except as authorized in a general license or a specific license
issued by the Commission, or as exempted in this part, no licensee may-
-
(a) Deliver licensed material to a carrier for transport; or
(b) Transport licensed material.
Sec. 71.4 Definitions.
The following terms are as defined here for the purpose of this
part. To ensure compatibility with international transportation
standards, all limits in this part are given in terms of dual units:
The International System of Units (SI) followed or preceded by U.S.
standard or customary units. The U.S. customary units are not exact
equivalents but are rounded to a convenient value, providing a
functionally equivalent unit. For the purpose of this part, either unit
may be used.
A1 means the maximum activity of special form radioactive material
permitted in a Type A package. This value is either listed in Appendix
A, Table A-1, of this part, or may be derived in accordance with the
procedures prescribed in Appendix A of this part.
A2 means the maximum activity of radioactive material, other than
special form material, LSA, and SCO material, permitted in a Type A
package. This value is either listed in Appendix A, Table A-1, of this
part, or may be derived in accordance with the procedures prescribed in
Appendix A of this part.
Carrier means a person engaged in the transportation of passengers
or property by land or water as a common, contract, or private carrier,
or by civil aircraft.
Certificate holder means a person who has been issued a certificate
of compliance or other package approval by the Commission.
Certificate of Compliance (CoC) means the certificate issued by the
Commission under subpart D of this part which approves the design of a
package for the transportation of radioactive material.
Close reflection by water means immediate contact by water of
sufficient thickness for maximum reflection of neutrons.
Consignment means each shipment of a package or groups of packages
or load of radioactive material offered by a shipper for transport.
Containment system means the assembly of components of the
packaging intended to retain the radioactive material during transport.
Conveyance means:
(1) For transport by public highway or rail any transport vehicle
or large freight container;
(2) For transport by water any vessel, or any hold, compartment, or
defined deck area of a vessel including any transport vehicle on board
the vessel; and
(3) For transport by any aircraft.
Criticality Safety Index (CSI) means the dimensionless number
(rounded up to the next tenth) assigned to and placed on the label of a
fissile material package, to designate the degree of control of
[[Page 3788]]
accumulation of packages containing fissile material during
transportation. Determination of the criticality safety index is
described in Sec.Sec. 71.22, 71.23, and 71.59.
Deuterium means, for the purposes of Sec.Sec. 71.15 and 71.22,
deuterium and any deuterium compounds, including heavy water, in which
the ratio of deuterium atoms to hydrogen atoms exceeds 1:5000.
DOT means the U.S. Department of Transportation.
Exclusive use means the sole use by a single consignor of a
conveyance for which all initial, intermediate, and final loading and
unloading are carried out in accordance with the direction of the
consignor or consignee. The consignor and the carrier must ensure that
any loading or unloading is performed by personnel having radiological
training and resources appropriate for safe handling of the
consignment. The consignor must issue specific instructions, in
writing, for maintenance of exclusive use shipment controls, and
include them with the shipping paper information provided to the
carrier by the consignor.
Fissile material means the radionuclides uranium-233, uranium-235,
plutonium-239, and plutonium-241, or any combination of these
radionuclides. Fissile material means the fissile nuclides themselves,
not material containing fissile nuclides. Unirradiated natural uranium
and depleted uranium and natural uranium or depleted uranium, that has
been irradiated in thermal reactors only, are not included in this
definition. Certain exclusions from fissile material controls are
provided in Sec. 71.15.
Graphite means, for the purposes of Sec.Sec. 71.15 and 71.22,
graphite with a boron equivalent content less than 5 parts per million
and density greater than 1.5 grams per cubic centimeter.
Licensed material means byproduct, source, or special nuclear
material received, possessed, used, or transferred under a general or
specific license issued by the Commission pursuant to the regulations
in this chapter.
Low Specific Activity (LSA) material means radioactive material
with limited specific activity which is nonfissile or is excepted under
Sec. 71.15, and which satisfies the descriptions and limits set forth
below. Shielding materials surrounding the LSA material may not be
considered in determining the estimated average specific activity of
the package contents. LSA material must be in one of three groups:
(1) LSA--I.
(i) Uranium and thorium ores, concentrates of uranium and thorium
ores, and other ores containing naturally occurring radioactive
radionuclides which are not intended to be processed for the use of
these radionuclides;
(ii) Solid unirradiated natural uranium or depleted uranium or
natural thorium or their solid or liquid compounds or mixtures;
(iii) Radioactive material for which the A2 value is
unlimited; or
(iv) Other radioactive material in which the activity is
distributed throughout and the estimated average specific activity does
not exceed 30 times the value for exempt material activity
concentration determined in accordance with Appendix A.
(2) LSA--II.
(i) Water with tritium concentration up to 0.8 TBq/liter (20.0 Ci/
liter); or
(ii) Other material in which the activity is distributed throughout
and the average specific activity does not exceed 10-\4\
A2/g for solids and gases, and
10-\5\A2/g for liquids.
(3) LSA--III. Solids (e.g., consolidated wastes, activated
materials), excluding powders, that satisfy the requirements of Sec.
71.77, in which:
(i) The radioactive material is distributed throughout a solid or a
collection of solid objects, or is essentially uniformly distributed in
a solid compact binding agent (such as concrete, bitumen, ceramic,
etc.);
(ii) The radioactive material is relatively insoluble, or it is
intrinsically contained in a relatively insoluble material, so that
even under loss of packaging, the loss of radioactive material per
package by leaching, when placed in water for 7 days, would not exceed
0.1 A2; and
(iii) The estimated average specific activity of the solid does not
exceed 2 x 10-3 A2/g.
Low toxicity alpha emitters means natural uranium, depleted
uranium, natural thorium; uranium-235, uranium-238, thorium-232,
thorium-228 or thorium-230 when contained in ores or physical or
chemical concentrates or tailings; or alpha emitters with a half-life
of less than 10 days.
Maximum normal operating pressure means the maximum gauge pressure
that would develop in the containment system in a period of 1 year
under the heat condition specified in Sec. 71.71(c)(1), in the absence
of venting, external cooling by an ancillary system, or operational
controls during transport.
Natural thorium means thorium with the naturally occurring
distribution of thorium isotopes (essentially 100 weight percent
thorium-232).
Normal form radioactive material means radioactive material that
has not been demonstrated to qualify as ``special form radioactive
material.''
Optimum interspersed hydrogenous moderation means the presence of
hydrogenous material between packages to such an extent that the
maximum nuclear reactivity results.
Package means the packaging together with its radioactive contents
as presented for transport.
(1) Fissile material package or Type AF package, Type BF package,
Type B(U)F package, or Type B(M)F package means a fissile material
packaging together with its fissile material contents.
(2) Type A package means a Type A packaging together with its
radioactive contents. A Type A package is defined and must comply with
the DOT regulations in 49 CFR part 173.
(3) Type B package means a Type B packaging together with its
radioactive contents. On approval, a Type B package design is
designated by NRC as B(U) unless the package has a maximum normal
operating pressure of more than 700 kPa (100 lbs/in2) gauge
or a pressure relief device that would allow the release of radioactive
material to the environment under the tests specified in Sec. 71.73
(hypothetical accident conditions), in which case it will receive a
designation B(M). B(U) refers to the need for unilateral approval of
international shipments; B(M) refers to the need for multilateral
approval of international shipments. There is no distinction made in
how packages with these designations may be used in domestic
transportation. To determine their distinction for international
transportation, see DOT regulations in 49 CFR Part 173. A Type B
package approved before September 6, 1983, was designated only as Type
B. Limitations on its use are specified in Sec. 71.19.
Packaging means the assembly of components necessary to ensure
compliance with the packaging requirements of this part. It may consist
of one or more receptacles, absorbent materials, spacing structures,
thermal insulation, radiation shielding, and devices for cooling or
absorbing mechanical shocks. The vehicle, tie-down system, and
auxiliary equipment may be designated as part of the packaging.
Special form radioactive material means radioactive material that
satisfies the following conditions:
(1) It is either a single solid piece or is contained in a sealed
capsule that can be opened only by destroying the capsule;
(2) The piece or capsule has at least one dimension not less than 5
mm (0.2 in); and
[[Page 3789]]
(3) It satisfies the requirements of Sec. 71.75. A special form
encapsulation designed in accordance with the requirements of Sec. 71.4
in effect on June 30, 1983 (see 10 CFR part 71, revised as of January
1, 1983), and constructed before July 1, 1985, and a special form
encapsulation designed in accordance with the requirements of Sec. 71.4
in effect on March 31, 1996 (see 10 CFR part 71, revised as of January
1, 1983), and constructed before April 1, 1998, may continue to be
used. Any other special form encapsulation must meet the specifications
of this definition.
Specific activity of a radionuclide means the radioactivity of the
radionuclide per unit mass of that nuclide. The specific activity of a
material in which the radionuclide is essentially uniformly distributed
is the radioactivity per unit mass of the material.
Spent nuclear fuel or Spent fuel means fuel that has been withdrawn
from a nuclear reactor following irradiation, has undergone at least 1
year's decay since being used as a source of energy in a power reactor,
and has not been chemically separated into its constituent elements by
reprocessing. Spent fuel includes the special nuclear material,
byproduct material, source material, and other radioactive materials
associated with fuel assemblies.
State means a State of the United States, the District of Columbia,
the Commonwealth of Puerto Rico, the Virgin Islands, Guam, American
Samoa, and the Commonwealth of the Northern Mariana Islands.
Surface Contaminated Object (SCO) means a solid object that is not
itself classed as radioactive material, but which has radioactive
material distributed on any of its surfaces. SCO must be in one of two
groups with surface activity not exceeding the following limits:
(1) SCO-I: A solid object on which:
(i) The nonfixed contamination on the accessible surface averaged
over 300 cm2 (or the area of the surface if less than 300
cm2) does not exceed 4 Bq/cm2 (10-4
microcurie/cm2) for beta and gamma and low toxicity alpha
emitters, or 0.4 Bq/cm2 (10-5 microcurie/
cm2) for all other alpha emitters;
(ii) The fixed contamination on the accessible surface averaged
over 300 cm2 (or the area of the surface if less than 300
cm2) does not exceed 4 x 10-4 Bq/cm2
(1.0 microcurie/cm2) for beta and gamma and low toxicity
alpha emitters, or 4 x 103 Bq/cm2 (0.1
microcurie/cm2) for all other alpha emitters; and
(iii) The nonfixed contamination plus the fixed contamination on
the inaccessible surface averaged over 300 cm2 (or the area
of the surface if less than 300 cm2) does not exceed 4 x
104 Bq/cm2 (1 microcurie/cm2) for beta
and gamma and low toxicity alpha emitters, or 4 x 103 Bq/cm2
(0.1 microcurie/cm2) for all other alpha emitters.
(2) SCO-II: A solid object on which the limits for SCO-I are
exceeded and on which:
(i) The nonfixed contamination on the accessible surface averaged
over 300 cm2 (or the area of the surface if less than 300
2) does not exceed 400 Bq/cm2 (10-2
microcurie/cm2) for beta and gamma and low toxicity alpha
emitters or 40 Bq/cm2 (10-3 microcurie/
cm2) for all other alpha emitters;
(ii) The fixed contamination on the accessible surface averaged
over 300 cm2 (or the area of the surface if less than 300
cm2) does not exceed 8 x 105 Bq/cm2
(20 microcuries/cm2) for beta and gamma and low toxicity
alpha emitters, or 8 x 104 Bq/cm2 (2 microcuries/
cm2) for all other alpha emitters; and
(iii) The nonfixed contamination plus the fixed contamination on
the inaccessible surface averaged over 300 cm2 (or the area
of the surface if less than 300 2) does not exceed 8 x
105 Bq/cm2 (20 microcuries/cm2) for
beta and gamma and low toxicity alpha emitters, or 8 x 104
Bq/cm2 (2 microcuries/cm2) for all other alpha
emitters.
Transport index (TI) means the dimensionless number (rounded up to
the next tenth) placed on the label of a package, to designate the
degree of control to be exercised by the carrier during transportation.
The transport index is the number determined by multiplying the maximum
radiation level in millisievert (mSv) per hour at 1 meter (3.3 ft) from
the external surface of the package by 100 (equivalent to the maximum
radiation level in millirem per hour at 1 meter (3.3 ft)).
Type A quantity means a quantity of radioactive material, the
aggregate radioactivity of which does not exceed A1 for
special form radioactive material, or A2, for normal form
radioactive material, where A1 and A2 are given
in Table A-1 of this part, or may be determined by procedures described
in Appendix A of this part.
Type B quantity means a quantity of radioactive material greater
than a Type A quantity.
Unirradiated uranium means uranium containing not more than 2 x
103 Bq of plutonium per gram of uranium-235, not more than 9
x 106 Bq of fission products per gram of uranium-235, and
not more than 5 x 10-3 g of uranium-236 per gram of uranium-
235.
Uranium--natural, depleted, enriched:
(1) Natural uranium means uranium with the naturally occurring
distribution of uranium isotopes (approximately 0.711 weight percent
uranium-235, and the remainder by weight essentially uranium-238).
(2) Depleted uranium means uranium containing less uranium-235 than
the naturally occurring distribution of uranium isotopes.
(3) Enriched uranium means uranium containing more uranium-235 than
the naturally occurring distribution of uranium isotopes.
Sec. 71.5 Transportation of licensed material.
(a) Each licensee who transports licensed material outside the site
of usage, as specified in the NRC license, or where transport is on
public highways, or who delivers licensed material to a carrier for
transport, shall comply with the applicable requirements of the DOT
regulations in 49 CFR parts 170 through 189 appropriate to the mode of
transport.
(1) The licensee shall particularly note DOT regulations in the
following areas:
(i) Packaging--49 CFR part 173: subparts A, B, and I.
(ii) Marking and labeling--49 CFR part 172: subpart D, Sec.Sec.
172.400 through 172.407, Sec.Sec. 172.436 through 172.440, and subpart
E.
(iii) Placarding--49 CFR part 172: subpart F, especially Sec.Sec.
172.500 through 172.519, 172.556, and appendices B and C.
(iv) Accident reporting--49 CFR part 171: Sec.Sec. 171.15 and
171.16.
(v) Shipping papers and emergency information--49 CFR part 172:
subparts C and G.
(vi) Hazardous material employee training--49 CFR part 172: subpart
H.
(vii) Hazardous material shipper/carrier registration--49 CFR part
107: subpart G.
(2) The licensee shall also note DOT regulations pertaining to the
following modes of transportation:
(i) Rail--49 CFR part 174: subparts A through D and K.
(ii) Air--49 CFR part 175.
(iii) Vessel--49 CFR part 176: subparts A through F and M.
(iv) Public Highway--49 CFR part 177 and parts 390 through 397.
(b) If DOT regulations are not applicable to a shipment of licensed
material, the licensee shall conform to the standards and requirements
of the DOT specified in paragraph (a) of this section to the same
extent as if the shipment or transportation were subject to DOT
regulations. A request for
[[Page 3790]]
modification, waiver, or exemption from those requirements, and any
notification referred to in those requirements, must be filed with, or
made to, the Director, Office of Nuclear Material Safety and
Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001.
Sec. 71.6 Information collection requirements: OMB approval.
(a) The Nuclear Regulatory Commission has submitted the information
collection requirements contained in this part to the Office of
Management and Budget (OMB) for approval as required by the Paperwork
Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or
sponsor, and a person is not required to respond to, a collection of
information unless it displays a currently valid OMB control number.
OMB has approved the information collection requirements contained in
this part under control number 3150-0008.
(b) The approved information collection requirements contained in
this part appear in Sec.Sec. 71.5, 71.7, 71.9, 71.12, 71.17, 71.19,
71.20, 71.22, 71.23, 71.31, 71.33, 71.35, 71.37, 71.38, 71.39, 71.41,
71.47, 71.85, 71.87, 71.89, 71.91, 71.93, 71.95, 71.97, 71.101, 71.103,
71.105, 71.107, 71.109, 71.111, 71.113, 71.115, 71.117, 71.119, 71.121,
71.123, 71.125, 71.127, 71.129, 71.131, 71.133, 71.135, 71.137, and
Appendix A, Paragraph II.
Sec. 71.7 Completeness and accuracy of information.
(a) Information provided to the Commission by a licensee,
certificate holder, or an applicant for a license or CoC; or
information required by statute or by the Commission's regulations,
orders, license or CoC conditions, to be maintained by the licensee or
certificate holder, must be complete and accurate in all material
respects.
(b) Each licensee, certificate holder, or applicant for a license
or CoC must notify the Commission of information identified by the
licensee, certificate holder, or applicant for a license or CoC as
having, for the regulated activity, a significant implication for
public health and safety or common defense and security. A licensee,
certificate holder, or an applicant for a license or CoC violates this
paragraph only if the licensee, certificate holder, or applicant for a
license or CoC fails to notify the Commission of information that the
licensee, certificate holder, or applicant for a license or CoC has
identified as having a significant implication for public health and
safety or common defense and security. Notification must be provided to
the Administrator of the appropriate Regional Office within 2 working
days of identifying the information. This requirement is not applicable
to information which is already required to be provided to the
Commission by other reporting or updating requirements.
Sec. 71.8 Deliberate misconduct.
(a) This section applies to any--
(1) Licensee;
(2) Certificate holder;
(3) Quality assurance program approval holder;
(4) Applicant for a license, certificate, or quality assurance
program approval;
(5) Contractor (including a supplier or consultant) or
subcontractor, to any person identified in paragraph (a)(4) of this
section; or
(6) Employees of any person identified in paragraphs (a)(1) through
(a)(5) of this section.
(b) A person identified in paragraph (a) of this section who
knowingly provides to any entity, listed in paragraphs (a)(1) through
(a)(5) of this section, any components, materials, or other goods or
services that relate to a licensee's, certificate holder's, quality
assurance program approval holder's, or applicant's activities subject
to this part may not:
(1) Engage in deliberate misconduct that causes or would have
caused, if not detected, a licensee, certificate holder, quality
assurance program approval holder, or any applicant to be in violation
of any rule, regulation, or order; or any term, condition or limitation
of any license, certificate, or approval issued by the Commission; or
(2) Deliberately submit to the NRC, a licensee, a certificate
holder, quality assurance program approval holder, an applicant for a
license, certificate or quality assurance program approval, or a
licensee's, applicant's, certificate holder's, or quality assurance
program approval holder's contractor or subcontractor, information that
the person submitting the information knows to be incomplete or
inaccurate in some respect material to the NRC.
(c) A person who violates paragraph (b)(1) or (b)(2) of this
section may be subject to enforcement action in accordance with the
procedures in 10 CFR part 2, subpart B.
(d) For the purposes of paragraph (b)(1) of this section,
deliberate misconduct by a person means an intentional act or omission
that the person knows:
(1) Would cause a licensee, certificate holder, quality assurance
program approval holder, or applicant for a license, certificate, or
quality assurance program approval to be in violation of any rule,
regulation, or order; or any term, condition, or limitation of any
license or certificate issued by the Commission; or
(2) Constitutes a violation of a requirement, procedure,
instruction, contract, purchase order, or policy of a licensee,
certificate holder, quality assurance program approval holder,
applicant, or the contractor or subcontractor of any of them.
Sec. 71.9 Employee protection.
(a) Discrimination by a Commission licensee, certificate holder, an
applicant for a Commission license or a CoC, or a contractor or
subcontractor of any of these, against an employee for engaging in
certain protected activities, is prohibited. Discrimination includes
discharge and other actions that relate to compensation, terms,
conditions, or privileges of employment. The protected activities are
established in section 211 of the Energy Reorganization Act of 1974, as
amended, and in general are related to the administration or
enforcement of a requirement imposed under the Atomic Energy Act of
1954, as amended, or the Energy Reorganization Act of 1974, as amended.
(1) The protected activities include, but are not limited to:
(i) Providing the Commission or his or her employer information
about alleged violations of either of the statutes named in paragraph
(a) of this section or possible violations of requirements imposed
under either of those statutes;
(ii) Refusing to engage in any practice made unlawful under either
of the statutes named in paragraph (a) of this section or under these
requirements if the employee has identified the alleged illegality to
the employer;
(iii) Requesting the Commission to institute action against his or
her employer for the administration or enforcement of these
requirements;
(iv) Testifying in any Commission proceeding, or before Congress,
or at any Federal or State proceeding regarding any provision (or
proposed provision) of either of the statutes named in paragraph (a) of
this section; and
(v) Assisting or participating in, or is about to assist or
participate in, these activities.
(2) These activities are protected even if no formal proceeding is
actually initiated as a result of the employee's assistance or
participation.
(3) This section has no application to any employee alleging
discrimination prohibited by this section who, acting without direction
from his or her employer (or the employer's agent),
[[Page 3791]]
deliberately causes a violation of any requirement of the Energy
Reorganization Act of 1974, as amended, or the Atomic Energy Act of
1954, as amended.
(b) Any employee who believes that he or she has been discharged or
otherwise discriminated against by any person for engaging in protected
activities specified in paragraph (a)(1) of this section may seek a
remedy for the discharge or discrimination through an administrative
proceeding in the Department of Labor. The administrative proceeding
must be initiated within 180 days after an alleged violation occurs.
The employee may do this by filing a complaint alleging the violation
with the Department of Labor, Employment Standards Administration, Wage
and Hour Division. The Department of Labor may order reinstatement,
back pay, and compensatory damages.
(c) A violation of paragraph (a), (e), or (f) of this section by a
Commission licensee, certificate holder, applicant for a Commission
license or a CoC, or a contractor or subcontractor of any of these may
be grounds for:
(1) Denial, revocation, or suspension of the license or the CoC;
(2) Imposition of a civil penalty on the licensee or applicant; or
(3) Other enforcement action.
(d) Actions taken by an employer, or others, which adversely affect
an employee may be predicated upon nondiscriminatory grounds. The
prohibition applies when the adverse action occurs because the employee
has engaged in protected activities. An employee's engagement in
protected activities does not automatically render him or her immune
from discharge or discipline for legitimate reasons or from adverse
action dictated by nonprohibited considerations.
(e)(1) Each licensee, certificate holder, and applicant for a
license or CoC must prominently post the current revision of NRC Form
3, ``Notice to Employees,'' referenced in Sec. 19.11(c) of this
chapter. This form must be posted at locations sufficient to permit
employees protected by this section to observe a copy on the way to or
from their place of work. The premises must be posted not later than 30
days after an application is docketed and remain posted while the
application is pending before the Commission, during the term of the
license or CoC, and for 30 days following license or CoC termination.
(2) Copies of NRC Form 3 may be obtained by writing to the Regional
Administrator of the appropriate U.S. Nuclear Regulatory Commission
Regional Office listed in Appendix D to part 20 of this chapter or by
calling the NRC Publishing Services Branch at 301-415-5877.
(f) No agreement affecting the compensation, terms, conditions, or
privileges of employment, including an agreement to settle a complaint
filed by an employee with the Department of Labor pursuant to section
211 of the Energy Reorganization Act of 1974, as amended, may contain
any provision which would prohibit, restrict, or otherwise discourage
an employee from participating in a protected activity as defined in
paragraph (a)(1) of this section including, but not limited to,
providing information to the NRC or to his or her employer on potential
violations or other matters within NRC's regulatory responsibilities.
Sec. 71.10 Public inspection of application.
Applications for approval of a package design under this part,
which are submitted to the Commission, may be made available for public
inspection, in accordance with provisions of parts 2 and 9 of this
chapter. This includes an application to amend or revise an existing
package design, any associated documents and drawings submitted with
the application, and any responses to NRC requests for additional
information.
Sec. 71.11 [Reserved]
Subpart B--Exemptions
Sec. 71.12 Specific exemptions.
On application of any interested person or on its own initiative,
the Commission may grant any exemption from the requirements of the
regulations in this part that it determines is authorized by law and
will not endanger life or property nor the common defense and security.
Sec. 71.13 Exemption of physicians.
Any physician licensed by a State to dispense drugs in the practice
of medicine is exempt from Sec. 71.5 with respect to transport by the
physician of licensed material for use in the practice of medicine.
However, any physician operating under this exemption must be licensed
under 10 CFR part 35 or the equivalent Agreement State regulations.
Sec. 71.14 Exemption for low-level materials.
(a) A licensee is exempt from all the requirements of this part
with respect to shipment or carriage of the following low-level
materials:
(1) Natural material and ores containing naturally occurring
radionuclides that are not intended to be processed for use of these
radionuclides, provided the activity concentration of the material does
not exceed 10 times the values specified in Appendix A, Table A-2, of
this part.
(2) Materials for which the activity concentration is not greater
than the activity concentration values specified in Appendix A, Table
A-2 of this part, or for which the consignment activity is not greater
than the limit for an exempt consignment found in Appendix A, Table A-
2, of this part.
(b) A licensee is exempt from all the requirements of this part,
other than Sec.Sec. 71.5 and 71.88, with respect to shipment or
carriage of the following packages, provided the packages do not
contain any fissile material, or the material is exempt from
classification as fissile material under Sec. 71.15:
(1) A package that contains no more than a Type A quantity of
radioactive material;
(2) A package transported within the United States that contains no
more than 0.74 TBq (20 Ci) of special form plutonium-244; or
(3) The package contains only LSA or SCO radioactive material,
provided--
(i) That the LSA or SCO material has an external radiation dose of
less than or equal to 10 mSv/h (1 rem/h), at a distance of 3 m from the
unshielded material; or
(ii) That the package contains only LSA-I or SCO-I material.
Sec. 71.15 Exemption from classification as fissile material.
Fissile material meeting the requirements of at least one of the
paragraphs (a) through (f) of this section are exempt from
classification as fissile material and from the fissile material
package standards of Sec.Sec. 71.55 and 71.59, but are subject to all
other requirements of this part, except as noted.
(a) Individual package containing 2 grams or less fissile material.
(b) Individual or bulk packaging containing 15 grams or less of
fissile material provided the package has at least 200 grams of solid
nonfissile material for every gram of fissile material. Lead,
beryllium, graphite, and hydrogenous material enriched in deuterium may
be present in the package but must not be included in determining the
required mass for solid nonfissile material.
(c)(1) Low concentrations of solid fissile material commingled with
solid nonfissile material, provided that:
(i) There is at least 2000 grams of solid nonfissile material for
every gram of fissile material, and
(ii) There is no more than 180 grams of fissile material
distributed within 360 kg of contiguous nonfissile material.
(2) Lead, beryllium, graphite, and hydrogenous material enriched in
[[Page 3792]]
deuterium may be present in the package but must not be included in
determining the required mass of solid nonfissile material.
(d) Uranium enriched in uranium-235 to a maximum of 1 percent by
weight, and with total plutonium and uranium-233 content of up to 1
percent of the mass of uranium-235, provided that the mass of any
beryllium, graphite, and hydrogenous material enriched in deuterium
constitutes less than 5 percent of the uranium mass.
(e) Liquid solutions of uranyl nitrate enriched in uranium-235 to a
maximum of 2 percent by mass, with a total plutonium and uranium-233
content not exceeding 0.002 percent of the mass of uranium, and with a
minimum nitrogen to uranium atomic ratio (N/U) of 2. The material must
be contained in at least a DOT Type A package.
(f) Packages containing, individually, a total plutonium mass of
not more than 1000 grams, of which not more than 20 percent by mass may
consist of plutonium-239, plutonium-241, or any combination of these
radionuclides.
Sec. 71.16 [Reserved]
Subpart C--General Licenses
Sec. 71.17 General license: NRC-approved package.
(a) A general license is issued to any licensee of the Commission
to transport, or to deliver to a carrier for transport, licensed
material in a package for which a license, certificate of compliance
(CoC), or other approval has been issued by the NRC.
(b) This general license applies only to a licensee who has a
quality assurance program approved by the Commission as satisfying the
provisions of subpart H of this part.
(c) This general license applies only to a licensee who--
(1) Has a copy of the CoC, or other approval of the package, and
has the drawings and other documents referenced in the approval
relating to the use and maintenance of the packaging and to the actions
to be taken before shipment;
(2) Complies with the terms and conditions of the license,
certificate, or other approval, as applicable, and the applicable
requirements of subparts A, G, and H of this part; and
(3) Before the licensee's first use of the package, submits in
writing to: ATTN: Document Control Desk, Director, Spent Fuel Project
Office, Office of Nuclear Material Safety and Safeguards, using an
appropriate method listed in Sec. 71.1(a), the licensee's name and
license number and the package identification number specified in the
package approval.
(d) This general license applies only when the package approval
authorizes use of the package under this general license.
(e) For a Type B or fissile material package, the design of which
was approved by NRC before April 1, 1996, the general license is
subject to the additional restrictions of Sec. 71.19.
Sec. 71.18 [Reserved]
Sec. 71.19 Previously approved package.
(a) A Type B package previously approved by NRC, but not designated
as B(U), B(M), B(U)F, or B(M)F in the identification number of the NRC
CoC, or Type AF packages approved by the NRC prior to September 6,
1983, may be used under the general license of Sec. 71.17 with the
following additional conditions:
(1) Fabrication of the packaging was satisfactorily completed by
August 31, 1986, as demonstrated by application of its model number in
accordance with Sec. 71.85(c);
(2) A serial number that uniquely identifies each packaging which
conforms to the approved design is assigned to, and legibly and durably
marked on, the outside of each packaging; and
(3) Paragraph (a) of this section expires (insert date 4 years
after the effective date of this final rule). The effective date of
this final rule is October 1, 2004.
(b) A Type B(U) package, a Type B(M) package, or a fissile material
package, previously approved by the NRC but without the designation ``-
85'' in the identification number of the NRC CoC, may be used under the
general license of Sec. 71.17 with the following additional conditions:
(1) Fabrication of the package is satisfactorily completed by April
1, 1999, as demonstrated by application of its model number in
accordance with Sec. 71.85(c);
(2) A package used for a shipment to a location outside the United
States is subject to multilateral approval as defined in DOT
regulations at 49 CFR 173.403; and
(3) A serial number which uniquely identifies each packaging which
conforms to the approved design is assigned to and legibly and durably
marked on the outside of each packaging.
(c) A Type B(U) package, a Type B(M) package, or a fissile material
package previously approved by the NRC with the designation ``-85'' in
the identification number of the NRC CoC, may be used under the general
license of Sec. 71.17 with the following additional conditions:
(1) Fabrication of the package must be satisfactorily completed by
December 31, 2006, as demonstrated by application of its model number
in accordance with Sec. 71.85(c); and
(2) After December 31, 2003, a package used for a shipment to a
location outside the United States is subject to multilateral approval
as defined in DOT regulations at 49 CFR 173.403.
(d) NRC will approve modifications to the design and authorized
contents of a Type B package, or a fissile material package, previously
approved by NRC, provided--
(1) The modifications of a Type B package are not significant with
respect to the design, operating characteristics, or safe performance
of the containment system, when the package is subjected to the tests
specified in Sec.Sec. 71.71 and 71.73;
(2) The modifications of a fissile material package are not
significant, with respect to the prevention of criticality, when the
package is subjected to the tests specified in Sec.Sec. 71.71 and
71.73; and
(3) The modifications to the package satisfy the requirements of
this part.
(e) NRC will revise the package identification number to designate
previously approved package designs as B, BF, AF, B(U), B(M), B(U)F,
B(M)F, B(U)-85, B(U)F-85, B(M)-85, B(M)F-85, or AF-85 as appropriate,
and with the identification number suffix ``-96'' after receipt of an
application demonstrating that the design meets the requirements of
this part.
Sec. 71.20 General license: DOT specification container.
(a) A general license is issued to any licensee of the Commission
to transport, or to deliver to a carrier for transport, licensed
material in a specification container for fissile material or for a
Type B quantity of radioactive material as specified in DOT regulations
at 49 CFR parts 173 and 178.
(b) This general license applies only to a licensee who has a
quality assurance program approved by the Commission as satisfying the
provisions of subpart H of this part.
(c) This general license applies only to a licensee who--
(1) Has a copy of the specification; and
(2) Complies with the terms and conditions of the specification and
the applicable requirements of subparts A, G, and H of this part.
(d) This general license is subject to the limitation that the
specification
[[Page 3793]]
container may not be used for a shipment to a location outside the
United States, except by multilateral approval, as defined in DOT
regulations at 49 CFR 173.403.
(e) This section expires October 1, 2008.
Sec. 71.21 General license: Use of foreign approved package.
(a) A general license is issued to any licensee of the Commission
to transport, or to deliver to a carrier for transport, licensed
material in a package, the design of which has been approved in a
foreign national competent authority certificate, that has been
revalidated by DOT as meeting the applicable requirements of 49 CFR
171.12.
(b) Except as otherwise provided in this section, the general
license applies only to a licensee who has a quality assurance program
approved by the Commission as satisfying the applicable provisions of
subpart H of this part.
(c) This general license applies only to shipments made to or from
locations outside the United States.
(d) This general license applies only to a licensee who--
(1) Has a copy of the applicable certificate, the revalidation, and
the drawings and other documents referenced in the certificate,
relating to the use and maintenance of the packaging and to the actions
to be taken before shipment; and
(2) Complies with the terms and conditions of the certificate and
revalidation, and with the applicable requirements of subparts A, G,
and H of this part. With respect to the quality assurance provisions of
subpart H of this part, the licensee is exempt from design,
construction, and fabrication considerations.
Sec. 71.22 General license: Fissile material.
(a) A general license is issued to any licensee of the Commission
to transport fissile material, or to deliver fissile material to a
carrier for transport, if the material is shipped in accordance with
this section. The fissile material need not be contained in a package
which meets the standards of subparts E and F of this part; however,
the material must be contained in a Type A package. The Type A package
must also meet the DOT requirements of 49 CFR 173.417(a).
(b) The general license applies only to a licensee who has a
quality assurance program approved by the Commission as satisfying the
provisions of subpart H of this part.
(c) The general license applies only when a package's contents:
(1) Contain less than a Type A quantity of fissile material; and
(2) Contain less than 500 total grams of beryllium, graphite, or
hydrogenous material enriched in deuterium.
(d) The general license applies only to packages containing fissile
material that are labeled with a CSI which:
(1) Has been determined in accordance with paragraph (e) of this
section;
(2) Has a value less than or equal to 10; and
(3) For a shipment of multiple packages containing fissile
material, the sum of the CSIs must be less than or equal to 50 (for
shipment on a nonexclusive use conveyance) and less than or equal to
100 (for shipment on an exclusive use conveyance).
(e)(1) The value for the CSI must be greater than or equal to the
number calculated by the following equation:
[GRAPHIC] [TIFF OMITTED] TR26JA04.012
(2) The calculated CSI must be rounded up to the first decimal
place;
(3) The values of X, Y, and Z used in the CSI equation must be
taken from Tables 71-1 or 71-2, as appropriate;
(4) If Table 71-2 is used to obtain the value of X, then the values
for the terms in the equation for uranium-233 and plutonium must be
assumed to be zero; and
(5) Table 71-1 values for X, Y, and Z must be used to determine the
CSI if:
(i) Uranium-233 is present in the package;
(ii) The mass of plutonium exceeds 1 percent of the mass of
uranium-235;
(iii) The uranium is of unknown uranium-235 enrichment or greater
than 24 weight percent enrichment; or
(iv) Substances having a moderating effectiveness (i.e., an average
hydrogen density greater than H2O) (e.g., certain
hydrocarbon oils or plastics) are present in any form, except as
polyethylene used for packing or wrapping.
Table 71-1.--Mass Limits for General License Packages Containing Mixed
Quantities of Fissile Material or Uranium-235 of Unknown Enrichment per
Sec. 71.22(e)
------------------------------------------------------------------------
Fissile material Fissile material
mass mixed with mass mixed with
moderating moderating
substances having substances having
Fissile material an average an average
hydrogen density hydrogen density
less than or equal greater than H2Oa
to H2O (grams) (grams)
------------------------------------------------------------------------
\235\ U (X)..................... 60 38
\233\ U (Y)..................... 43 27
\239\ Pu or \241\ Pu (Z)........ 37 24
------------------------------------------------------------------------
a When mixtures of moderating substances are present, the lower mass
limits shall be used if more than 15 percent of the moderating
substance has an average hydrogen density greater than H2O.
[[Page 3794]]
Table 71-2.--Mass Limits for General License Packages Containing Uranium-
235 of Known Enrichment per Sec. 71.22(e)
------------------------------------------------------------------------
Fissile
material
Uranium enrichment in weight percent of \235\ U not mass of
exceeding \235\ U (X)
(grams)
------------------------------------------------------------------------
24........................................................ 60
20........................................................ 63
15........................................................ 67
11........................................................ 72
10........................................................ 76
9.5....................................................... 78
9......................................................... 81
8.5....................................................... 82
8......................................................... 85
7.5....................................................... 88
7......................................................... 90
6.5....................................................... 93
6......................................................... 97
5.5....................................................... 102
5......................................................... 108
4.5....................................................... 114
4......................................................... 120
3.5....................................................... 132
3......................................................... 150
2.5....................................................... 180
2......................................................... 246
1.5....................................................... 408
1.35...................................................... 480
1......................................................... 1,020
0.92...................................................... 1,800
------------------------------------------------------------------------
Sec. 71.23 General license: Plutonium-beryllium special form material.
(a) A general license is issued to any licensee of the Commission
to transport fissile material in the form of plutonium-beryllium (Pu-
Be) special form sealed sources, or to deliver Pu-Be sealed sources to
a carrier for transport, if the material is shipped in accordance with
this section. This material need not be contained in a package which
meets the standards of subparts E and F of this part; however, the
material must be contained in a Type A package. The Type A package must
also meet the DOT requirements of 49 CFR 173.417(a).
(b) The general license applies only to a licensee who has a
quality assurance program approved by the Commission as satisfying the
provisions of subpart H of this part.
(c) The general license applies only when a package's contents:
(1) Contain less than a Type A quantity of material; and
(2) Contain less than 1000 g of plutonium, provided that:
plutonium-239, plutonium-241, or any combination of these
radionuclides, constitutes less than 240 g of the total quantity of
plutonium in the package.
(d) The general license applies only to packages labeled with a CSI
which:
(1) Has been determined in accordance with paragraph (e) of this
section;
(2) Has a value less than or equal to 100; and
(3) For a shipment of multiple packages containing Pu-Be sealed
sources, the sum of the CSIs must be less than or equal to 50 (for
shipment on a nonexclusive use conveyance) and less than or equal to
100 (for shipment on an exclusive use conveyance).
(e)(1) The value for the CSI must be greater than or equal to the
number calculated by the following equation:
[GRAPHIC] [TIFF OMITTED] TR26JA04.013
(2) The calculated CSI must be rounded up to the first decimal
place.
Sec. 71.24 [Reserved]
Sec. 71.25 [Reserved]
0
3. In Sec. 71.41, paragraph (a) is revised, and a new paragraph (d) is
added to read as follows:
Sec. 71.41 Demonstration of compliance.
(a) The effects on a package of the tests specified in Sec. 71.71
(``Normal conditions of transport''), and the tests specified in Sec.
71.73 (``Hypothetical accident conditions''), and Sec. 71.61 (``Special
requirements for Type B packages containing more than 105
A2''), must be evaluated by subjecting a specimen or scale
model to a specific test, or by another method of demonstration
acceptable to the Commission, as appropriate for the particular feature
being considered.
* * * * *
(d) Packages for which compliance with the other provisions of
these regulations is impracticable shall not be transported except
under special package authorization. Provided the applicant
demonstrates that compliance with the other provisions of the
regulations is impracticable and that the requisite standards of safety
established by these regulations have been demonstrated through means
alternative to the other provisions, a special package authorization
may be approved for one-time shipments. The applicant shall demonstrate
that the overall level of safety in transport for these shipments is at
least equivalent to that which would be provided if all the applicable
requirements had been met.
0
4. In Sec. 71.51, the introductory text of paragraph (a) is revised,
and a new paragraph (d) is added to read as follows:
Sec. 71.51 Additional requirements for Type B packages.
(a) A Type B package, in addition to satisfying the requirements of
Sec.Sec. 71.41 through 71.47, must be designed, constructed, and
prepared for shipment so that under the tests specified in:
* * * * *
(d) For packages which contain radioactive contents with activity
greater than 105 A2, the requirements of Sec. 71.61 must be
met.
Sec. 71.53 [Reserved)
0
5. Section 71.53 is removed and reserved.
0
6. In Sec. 71.55, the introductory text of paragraph (b) is revised,
and new paragraphs (f) and (g) are added to read as follows:
Sec. 71.55 General requirements for fissile material packages.
* * * * *
[[Page 3795]]
(b) Except as provided in paragraph (c) or (g) of this section, a
package used for the shipment of fissile material must be so designed
and constructed and its contents so limited that it would be
subcritical if water were to leak into the containment system, or
liquid contents were to leak out of the containment system so that,
under the following conditions, maximum reactivity of the fissile
material would be attained:
* * * * *
(f) For fissile material package designs to be transported by air:
(1) The package must be designed and constructed, and its contents
limited so that it would be subcritical, assuming reflection by 20 cm
(7.9 in) of water but no water inleakage, when subjected to sequential
application of:
(i) The free drop test in Sec. 71.73(c)(1);
(ii) The crush test in Sec. 71.73(c)(2);
(iii) A puncture test, for packages of 250 kg or more, consisting
of a free drop of the specimen through a distance of 3 m (120 in) in a
position for which maximum damage is expected at the conclusion of the
test sequence, onto the upper end of a solid, vertical, cylindrical,
mild steel probe mounted on an essentially unyielding, horizontal
surface. The probe must be 20 cm (7.9 in) in diameter, with the
striking end forming the frustum of a right circular cone with the
dimensions of 30 cm height, 2.5 cm top diameter, and a top edge rounded
to a radius of not more than 6 mm (0.25 in). For packages less than 250
kg, the puncture test must be the same, except that a 250 kg probe must
be dropped onto the specimen which must be placed on the surface; and
(iv) The thermal test in Sec. 71.73(c)(4), except that the duration
of the test must be 60 minutes.
(2) The package must be designed and constructed, and its contents
limited, so that it would be subcritical, assuming reflection by 20 cm
(7.9 in) of water but no water inleakage, when subjected to an impact
on an unyielding surface at a velocity of 90 m/s normal to the surface,
at such orientation so as to result in maximum damage. A separate,
undamaged specimen can be used for this evaluation.
(3) Allowance may not be made for the special design features in
paragraph (c) of this section, unless water leakage into or out of void
spaces is prevented following application of the tests in paragraphs
(f)(1) and (f)(2) of this section, and subsequent application of the
immersion test in Sec. 71.73(c)(5).
(g) Packages containing uranium hexafluoride only are excepted from
the requirements of paragraph (b) of this section provided that:
(1) Following the tests specified in Sec. 71.73 (``Hypothetical
accident conditions''), there is no physical contact between the valve
body and any other component of the packaging, other than at its
original point of attachment, and the valve remains leak tight;
(2) There is an adequate quality control in the manufacture,
maintenance, and repair of packagings;
(3) Each package is tested to demonstrate closure before each
shipment; and
(4) The uranium is enriched to not more than 5 weight percent
uranium-235.
0
7. In Sec. 71.59, paragraphs (b) and (c) are revised to read as
follows:
Sec. 71.59 Standards for arrays of fissile material packages.
* * * * *
(b) The CSI must be determined by dividing the number 50 by the
value of ``N'' derived using the procedures specified in paragraph (a)
of this section. The value of the CSI may be zero provided that an
unlimited number of packages are subcritical, such that the value of
``N'' is effectively equal to infinity under the procedures specified
in paragraph (a) of this section. Any CSI greater than zero must be
rounded up to the first decimal place.
(c) For a fissile material package which is assigned a CSI value--
(1) Less than or equal to 50, that package may be shipped by a
carrier in a nonexclusive use conveyance, provided the sum of the CSIs
is limited to less than or equal to 50.
(2) Less than or equal to 50, that package may be shipped by a
carrier in an exclusive use conveyance, provided the sum of the CSIs is
limited to less than or equal to 100.
(3) Greater than 50, that package must be shipped by a carrier in
an exclusive use conveyance, provided the sum of the CSIs is limited to
less than or equal to 100.
0
8. Section 71.61 is revised to read as follows:
Sec. 71.61 Special requirements for Type B packages containing more
than 10\5\A[bdi2].
A Type B package containing more than 10\5\A2 must be
designed so that its undamaged containment system can withstand an
external water pressure of 2 MPa (290 psi) for a period of not less
than 1 hour without collapse, buckling, or inleakage of water.
0
9. Section 71.63 is revised to read as follows:
Sec. 71.63 Special requirement for plutonium shipments.
Shipments containing plutonium must be made with the contents in
solid form, if the contents contain greater than 0.74 TBq (20 Ci) of
plutonium.
0
10. In Sec. 71.73, paragraph (c)(2) is revised to read as follows:
Sec. 71.73 Hypothetical accident conditions.
* * * * *
(c) * * *
(2) Crush. Subjection of the specimen to a dynamic crush test by
positioning the specimen on a flat, essentially unyielding horizontal
surface so as to suffer maximum damage by the drop of a 500-kg (1100-
lb) mass from 9 m (30 ft) onto the specimen. The mass must consist of a
solid mild steel plate 1 m (40 in) by 1 m (40 in) and must fall in a
horizontal attitude. The crush test is required only when the specimen
has a mass not greater than 500 kg (1100 lb), an overall density not
greater than 1000 kg/m \3\ (62.4 lb/ft \3\) based on external
dimension, and radioactive contents greater than 1000 A2 not
as special form radioactive material. For packages containing fissile
material, the radioactive contents greater than 1000 A2
criterion does not apply.
* * * * *
0
11. In Sec. 71.88, paragraph (a)(2) is revised to read as follows:
Sec. 71.88 Air transport of plutonium.
(a) * * *
(2) The plutonium is contained in a material in which the specific
activity is less than or equal to the activity concentration values for
plutonium specified in Appendix A, Table A-2, of this part, and in
which the radioactivity is essentially uniformly distributed; or
* * * * *
0
12. In Sec. 71.91, paragraphs (b) and (c) are revised, and a new
paragraph (d) is added to read as follows:
Sec. 71.91 Records.
* * * * *
(b) Each certificate holder shall maintain, for a period of 3 years
after the life of the packaging to which they apply, records
identifying the packaging by model number, serial number, and date of
manufacture.
(c) The licensee, certificate holder, and an applicant for a CoC,
shall make available to the Commission for inspection, upon reasonable
notice, all records required by this part. Records are only valid if
stamped, initialed, or signed and dated by authorized personnel, or
otherwise authenticated.
(d) The licensee, certificate holder, and an applicant for a CoC
shall maintain sufficient written records to furnish evidence of the
quality of packaging. The records to be maintained include results of
the determinations required by Sec. 71.85; design, fabrication,
[[Page 3796]]
and assembly records; results of reviews, inspections, tests, and
audits; results of monitoring work performance and materials analyses;
and results of maintenance, modification, and repair activities.
Inspection, test, and audit records must identify the inspector or data
recorder, the type of observation, the results, the acceptability, and
the action taken in connection with any deficiencies noted. These
records must be retained for 3 years after the life of the packaging to
which they apply.
0
13. Section 71.93 is revised to read as follows:
Sec. 71.93 Inspection and tests.
(a) The licensee, certificate holder, and applicant for a CoC shall
permit the Commission, at all reasonable times, to inspect the licensed
material, packaging, premises, and facilities in which the licensed
material or packaging is used, provided, constructed, fabricated,
tested, stored, or shipped.
(b) The licensee, certificate holder, and applicant for a CoC shall
perform, and permit the Commission to perform, any tests the Commission
deems necessary or appropriate for the administration of the
regulations in this chapter.
(c) The certificate holder and applicant for a CoC shall notify the
NRC, in accordance with Sec. 71.1, 45 days in advance of starting
fabrication of the first packaging under a CoC. This paragraph applies
to any packaging used for the shipment of licensed material which has
either--
(1) A decay heat load in excess of 5 kW; or
(2) A maximum normal operating pressure in excess of 103 kPa (15
lbf/in \2\) gauge.
0
14. Section 71.95 is revised to read as follows:
Sec. 71.95 Reports.
(a) The licensee, after requesting the certificate holder's input,
shall submit a written report to the Commission of--
(1) Instances in which there is a significant reduction in the
effectiveness of any NRC-approved Type B or Type AF packaging during
use; or
(2) Details of any defects with safety significance in any NRC-
approved Type B or fissile material packaging, after first use.
(3) Instances in which the conditions of approval in the
Certificate of Compliance were not observed in making a shipment.
(b) The licensee shall submit a written report to the Commission of
instances in which the conditions in the certificate of compliance were
not followed during a shipment.
(c) Each licensee shall submit, in accordance with Sec. 71.1, a
written report required by paragraph (a) or (b) of this section within
60 days of the event or discovery of the event. The licensee shall also
provide a copy of each report submitted to the NRC to the applicable
certificate holder. Written reports prepared under other regulations
may be submitted to fulfill this requirement if the reports contain all
the necessary information, and the appropriate distribution is made.
Using an appropriate method listed in Sec. 71.1(a), the licensee shall
report to: ATTN: Document Control Desk, Director, Spent Fuel Project
Office, Office of Nuclear Material Safety and Safeguards. These written
reports must include the following:
(1) A brief abstract describing the major occurrences during the
event, including all component or system failures that contributed to
the event and significant corrective action taken or planned to prevent
recurrence.
(2) A clear, specific, narrative description of the event that
occurred so that knowledgeable readers conversant with the requirements
of part 71, but not familiar with the design of the packaging, can
understand the complete event. The narrative description must include
the following specific information as appropriate for the particular
event.
(i) Status of components or systems that were inoperable at the
start of the event and that contributed to the event;
(ii) Dates and approximate times of occurrences;
(iii) The cause of each component or system failure or personnel
error, if known;
(iv) The failure mode, mechanism, and effect of each failed
component, if known;
(v) A list of systems or secondary functions that were also
affected for failures of components with multiple functions;
(vi) The method of discovery of each component or system failure or
procedural error;
(vii) For each human performance-related root cause, a discussion
of the cause(s) and circumstances;
(viii) The manufacturer and model number (or other identification)
of each component that failed during the event; and
(ix) For events occurring during use of a packaging, the quantities
and chemical and physical form(s) of the package contents.
(3) An assessment of the safety consequences and implications of
the event. This assessment must include the availability of other
systems or components that could have performed the same function as
the components and systems that failed during the event.
(4) A description of any corrective actions planned as a result of
the event, including the means employed to repair any defects, and
actions taken to reduce the probability of similar events occurring in
the future.
(5) Reference to any previous similar events involving the same
packaging that are known to the licensee or certificate holder.
(6) The name and telephone number of a person within the licensee's
organization who is knowledgeable about the event and can provide
additional information.
(7) The extent of exposure of individuals to radiation or to
radioactive materials without identification of individuals by name.
(d) Report legibility. The reports submitted by licensees and/or
certificate holders under this section must be of sufficient quality to
permit reproduction and micrographic processing.
0
15. In Sec. 71.100, paragraph (b) is revised to read as follows:
Sec. 71.100 Criminal penalties.
* * * * *
(b) The regulations in part 71 that are not issued under sections
161b, 161i, or 161o for the purposes of section 223 are as follows:
Sec.Sec. 71.0, 71.2, 71.4, 71.6, 71.7, 71.10, 71.31, 71.33, 71.35,
71.37, 71.38, 71.39, 71.40, 71.41, 71.43, 71.45, 71.47, 71.51, 71.55,
71.59, 71.65, 71.71, 71.73, 71.74, 71.75, 71.77, 71.99, and 71.100.
0
16. Subpart H to part 71 is revised to read as follows:
Subpart H--Quality Assurance
Sec.
71.101 Quality assurance requirements.
71.103 Quality assurance organization.
71.105 Quality assurance program.
71.107 Package design control.
71.109 Procurement document control.
71.111 Instructions, procedures, and drawings.
71.113 Document control.
71.115 Control of purchased material, equipment, and services.
71.117 Identification and control of materials, parts, and
components.
71.119 Control of special processes.
71.121 Internal inspection.
71.123 Test control.
71.125 Control of measuring and test equipment.
71.127 Handling, storage, and shipping control.
71.129 Inspection, test, and operating status.
71.131 Nonconforming materials, parts, or components.
71.133 Corrective action.
71.135 Quality assurance records.
71.137 Audits.
[[Page 3797]]
Subpart H--Quality Assurance
Sec. 71.101 Quality assurance requirements.
(a) Purpose. This subpart describes quality assurance requirements
applying to design, purchase, fabrication, handling, shipping, storing,
cleaning, assembly, inspection, testing, operation, maintenance,
repair, and modification of components of packaging that are important
to safety. As used in this subpart, ``quality assurance'' comprises all
those planned and systematic actions necessary to provide adequate
confidence that a system or component will perform satisfactorily in
service. Quality assurance includes quality control, which comprises
those quality assurance actions related to control of the physical
characteristics and quality of the material or component to
predetermined requirements. The licensee, certificate holder, and
applicant for a CoC are responsible for the quality assurance
requirements as they apply to design, fabrication, testing, and
modification of packaging. Each licensee is responsible for the quality
assurance provision which applies to its use of a packaging for the
shipment of licensed material subject to this subpart.
(b) Establishment of program. Each licensee, certificate holder,
and applicant for a CoC shall establish, maintain, and execute a
quality assurance program satisfying each of the applicable criteria of
Sec.Sec. 71.101 through 71.137 and satisfying any specific provisions
that are applicable to the licensee's activities including procurement
of packaging. The licensee, certificate holder, and applicant for a CoC
shall execute the applicable criteria in a graded approach to an extent
that is commensurate with the quality assurance requirement's
importance to safety.
(c) Approval of program. (1) Before the use of any package for the
shipment of licensed material subject to this subpart, each licensee
shall obtain Commission approval of its quality assurance program.
Using an appropriate method listed in Sec. 71.1(a), each licensee shall
file a description of its quality assurance program, including a
discussion of which requirements of this subpart are applicable and how
they will be satisfied, by submitting the description to: ATTN:
Document Control Desk, Director, Spent Fuel Project Office, Office of
Nuclear Material Safety and Safeguards.
(2) Before the fabrication, testing, or modification of any package
for the shipment of licensed material subject to this subpart, each
licensee, certificate holder, or applicant for a CoC shall obtain
Commission approval of its quality assurance program. Each certificate
holder or applicant for a CoC shall, in accordance with Sec. 71.1, file
a description of its quality assurance program, including a discussion
of which requirements of this subpart are applicable and how they will
be satisfied.
(d) Existing package designs. The provisions of this paragraph deal
with packages that have been approved for use in accordance with this
part before January 1, 1979, and which have been designed in accordance
with the provisions of this part in effect at the time of application
for package approval. Those packages will be accepted as having been
designed in accordance with a quality assurance program that satisfies
the provisions of paragraph (b) of this section.
(e) Existing packages. The provisions of this paragraph deal with
packages that have been approved for use in accordance with this part
before January 1, 1979, have been at least partially fabricated before
that date, and for which the fabrication is in accordance with the
provisions of this part in effect at the time of application for
approval of package design. These packages will be accepted as having
been fabricated and assembled in accordance with a quality assurance
program that satisfies the provisions of paragraph (b) of this section.
(f) Previously approved programs. A Commission-approved quality
assurance program that satisfies the applicable criteria of subpart H
of this part, Appendix B of part 50 of this chapter, or subpart G of
part 72 of this chapter, and that is established, maintained, and
executed regarding transport packages, will be accepted as satisfying
the requirements of paragraph (b) of this section. Before first use,
the licensee, certificate holder, and applicant for a CoC shall notify
the NRC, in accordance with Sec. 71.1, of its intent to apply its
previously approved subpart H, Appendix B, or subpart G quality
assurance program to transportation activities. The licensee,
certificate holder, and applicant for a CoC shall identify the program
by date of submittal to the Commission, Docket Number, and date of
Commission approval.
(g) Radiography containers. A program for transport container
inspection and maintenance limited to radiographic exposure devices,
source changers, or packages transporting these devices and meeting the
requirements of Sec. 34.31(b) of this chapter or equivalent Agreement
State requirement, is deemed to satisfy the requirements of Sec.Sec.
71.17(b) and 71.101(b).
Sec. 71.103 Quality assurance organization.
(a) The licensee,\2\ certificate holder, and applicant for a CoC
shall be responsible for the establishment and execution of the quality
assurance program. The licensee, certificate holder, and applicant for
a CoC may delegate to others, such as contractors, agents, or
consultants, the work of establishing and executing the quality
assurance program, or any part of the quality assurance program, but
shall retain responsibility for the program. These activities include
performing the functions associated with attaining quality objectives
and the quality assurance functions.
---------------------------------------------------------------------------
\2\ While the term ``licensee'' is used in these criteria, the
requirements are applicable to whatever design, fabrication,
assembly, and testing of the package is accomplished with respect to
a package before the time a package approval is issued.
---------------------------------------------------------------------------
(b) The quality assurance functions are--
(1) Assuring that an appropriate quality assurance program is
established and effectively executed; and
(2) Verifying, by procedures such as checking, auditing, and
inspection, that activities affecting the functions that are important
to safety have been correctly performed.
(c) The persons and organizations performing quality assurance
functions must have sufficient authority and organizational freedom to-
-
(1) Identify quality problems;
(2) Initiate, recommend, or provide solutions; and
(3) Verify implementation of solutions.
(d) The persons and organizations performing quality assurance
functions shall report to a management level that assures that the
required authority and organizational freedom, including sufficient
independence from cost and schedule, when opposed to safety
considerations, are provided.
(e) Because of the many variables involved, such as the number of
personnel, the type of activity being performed, and the location or
locations where activities are performed, the organizational structure
for executing the quality assurance program may take various forms,
provided that the persons and organizations assigned the quality
assurance functions have the required authority and organizational
freedom.
(f) Irrespective of the organizational structure, the individual(s)
assigned the responsibility for assuring effective execution of any
portion of the quality assurance program, at any location where
activities subject to this section
[[Page 3798]]
are being performed, must have direct access to the levels of
management necessary to perform this function.
Sec. 71.105 Quality assurance program.
(a) The licensee, certificate holder, and applicant for a CoC shall
establish, at the earliest practicable time consistent with the
schedule for accomplishing the activities, a quality assurance program
that complies with the requirements of Sec.Sec. 71.101 through 71.137.
The licensee, certificate holder, and applicant for a CoC shall
document the quality assurance program by written procedures or
instructions and shall carry out the program in accordance with those
procedures throughout the period during which the packaging is used.
The licensee, certificate holder, and applicant for a CoC shall
identify the material and components to be covered by the quality
assurance program, the major organizations participating in the
program, and the designated functions of these organizations.
(b) The licensee, certificate holder, and applicant for a CoC,
through its quality assurance program, shall provide control over
activities affecting the quality of the identified materials and
components to an extent consistent with their importance to safety, and
as necessary to assure conformance to the approved design of each
individual package used for the shipment of radioactive material. The
licensee, certificate holder, and applicant for a CoC shall assure that
activities affecting quality are accomplished under suitably controlled
conditions. Controlled conditions include the use of appropriate
equipment; suitable environmental conditions for accomplishing the
activity, such as adequate cleanliness; and assurance that all
prerequisites for the given activity have been satisfied. The licensee,
certificate holder, and applicant for a CoC shall take into account the
need for special controls, processes, test equipment, tools, and skills
to attain the required quality, and the need for verification of
quality by inspection and test.
(c) The licensee, certificate holder, and applicant for a CoC shall
base the requirements and procedures of its quality assurance program
on the following considerations concerning the complexity and proposed
use of the package and its components:
(1) The impact of malfunction or failure of the item to safety;
(2) The design and fabrication complexity or uniqueness of the
item;
(3) The need for special controls and surveillance over processes
and equipment;
(4) The degree to which functional compliance can be demonstrated
by inspection or test; and
(5) The quality history and degree of standardization of the item.
(d) The licensee, certificate holder, and applicant for a CoC shall
provide for indoctrination and training of personnel performing
activities affecting quality, as necessary to assure that suitable
proficiency is achieved and maintained. The licensee, certificate
holder, and applicant for a CoC shall review the status and adequacy of
the quality assurance program at established intervals. Management of
other organizations participating in the quality assurance program
shall review regularly the status and adequacy of that part of the
quality assurance program they are executing.
Sec. 71.107 Package design control.
(a) The licensee, certificate holder, and applicant for a CoC shall
establish measures to assure that applicable regulatory requirements
and the package design, as specified in the license or CoC for those
materials and components to which this section applies, are correctly
translated into specifications, drawings, procedures, and instructions.
These measures must include provisions to assure that appropriate
quality standards are specified and included in design documents and
that deviations from standards are controlled. Measures must be
established for the selection and review for suitability of application
of materials, parts, equipment, and processes that are essential to the
functions of the materials, parts, and components of the packaging that
are important to safety.
(b) The licensee, certificate holder, and applicant for a CoC shall
establish measures for the identification and control of design
interfaces and for coordination among participating design
organizations. These measures must include the establishment of written
procedures, among participating design organizations, for the review,
approval, release, distribution, and revision of documents involving
design interfaces. The design control measures must provide for
verifying or checking the adequacy of design, by methods such as design
reviews, alternate or simplified calculational methods, or by a
suitable testing program. For the verifying or checking process, the
licensee shall designate individuals or groups other than those who
were responsible for the original design, but who may be from the same
organization. Where a test program is used to verify the adequacy of a
specific design feature in lieu of other verifying or checking
processes, the licensee, certificate holder, and applicant for a CoC
shall include suitable qualification testing of a prototype or sample
unit under the most adverse design conditions. The licensee,
certificate holder, and applicant for a CoC shall apply design control
measures to the following:
(1) Criticality physics, radiation shielding, stress, thermal,
hydraulic, and accident analyses;
(2) Compatibility of materials;
(3) Accessibility for inservice inspection, maintenance, and
repair;
(4) Features to facilitate decontamination; and
(5) Delineation of acceptance criteria for inspections and
tests.
(c) The licensee, certificate holder, and applicant for a CoC shall
subject design changes, including field changes, to design control
measures commensurate with those applied to the original design.
Changes in the conditions specified in the CoC require prior NRC
approval.
Sec. 71.109 Procurement document control.
The licensee, certificate holder, and applicant for a CoC shall
establish measures to assure that adequate quality is required in the
documents for procurement of material, equipment, and services, whether
purchased by the licensee, certificate holder, and applicant for a CoC
or by its contractors or subcontractors. To the extent necessary, the
licensee, certificate holder, and applicant for a CoC shall require
contractors or subcontractors to provide a quality assurance program
consistent with the applicable provisions of this part.
Sec. 71.111 Instructions, procedures, and drawings.
The licensee, certificate holder, and applicant for a CoC shall
prescribe activities affecting quality by documented instructions,
procedures, or drawings of a type appropriate to the circumstances and
shall require that these instructions, procedures, and drawings be
followed. The instructions, procedures, and drawings must include
appropriate quantitative or qualitative acceptance criteria for
determining that important activities have been satisfactorily
accomplished.
Sec. 71.113 Document control.
The licensee, certificate holder, and applicant for a CoC shall
establish measures to control the issuance of documents such as
instructions, procedures, and drawings, including
[[Page 3799]]
changes, that prescribe all activities affecting quality. These
measures must assure that documents, including changes, are reviewed
for adequacy, approved for release by authorized personnel, and
distributed and used at the location where the prescribed activity is
performed.
Sec. 71.115 Control of purchased material, equipment, and services.
(a) The licensee, certificate holder, and applicant for a CoC shall
establish measures to assure that purchased material, equipment, and
services, whether purchased directly or through contractors and
subcontractors, conform to the procurement documents. These measures
must include provisions, as appropriate, for source evaluation and
selection, objective evidence of quality furnished by the contractor or
subcontractor, inspection at the contractor or subcontractor source,
and examination of products on delivery.
(b) The licensee, certificate holder, and applicant for a CoC shall
have available documentary evidence that material and equipment conform
to the procurement specifications before installation or use of the
material and equipment. The licensee, certificate holder, and applicant
for a CoC shall retain, or have available, this documentary evidence
for the life of the package to which it applies. The licensee,
certificate holder, and applicant for a CoC shall assure that the
evidence is sufficient to identify the specific requirements met by the
purchased material and equipment.
(c) The licensee, certificate holder, and applicant for a CoC shall
assess the effectiveness of the control of quality by contractors and
subcontractors at intervals consistent with the importance, complexity,
and quantity of the product or services.
Sec. 71.117 Identification and control of materials, parts, and
components.
The licensee, certificate holder, and applicant for a CoC shall
establish measures for the identification and control of materials,
parts, and components. These measures must assure that identification
of the item is maintained by heat number, part number, or other
appropriate means, either on the item or on records traceable to the
item, as required throughout fabrication, installation, and use of the
item. These identification and control measures must be designed to
prevent the use of incorrect or defective materials, parts, and
components.
Sec. 71.119 Control of special processes.
The licensee, certificate holder, and applicant for a CoC shall
establish measures to assure that special processes, including welding,
heat treating, and nondestructive testing are controlled and
accomplished by qualified personnel using qualified procedures in
accordance with applicable codes, standards, specifications, criteria,
and other special requirements.
Sec. 71.121 Internal inspection.
The licensee, certificate holder, and applicant for a CoC shall
establish and execute a program for inspection of activities affecting
quality by or for the organization performing the activity, to verify
conformance with the documented instructions, procedures, and drawings
for accomplishing the activity. The inspection must be performed by
individuals other than those who performed the activity being
inspected. Examination, measurements, or tests of material or products
processed must be performed for each work operation where necessary to
assure quality. If direct inspection of processed material or products
is not carried out, indirect control by monitoring processing methods,
equipment, and personnel must be provided. Both inspection and process
monitoring must be provided when quality control is inadequate without
both. If mandatory inspection hold points, which require witnessing or
inspecting by the licensee's designated representative and beyond which
work should not proceed without the consent of its designated
representative, are required, the specific hold points must be
indicated in appropriate documents.
Sec. 71.123 Test control.
The licensee, certificate holder, and applicant for a CoC shall
establish a test program to assure that all testing required to
demonstrate that the packaging components will perform satisfactorily
in service is identified and performed in accordance with written test
procedures that incorporate the requirements of this part and the
requirements and acceptance limits contained in the package approval.
The test procedures must include provisions for assuring that all
prerequisites for the given test are met, that adequate test
instrumentation is available and used, and that the test is performed
under suitable environmental conditions. The licensee, certificate
holder, and applicant for a CoC shall document and evaluate the test
results to assure that test requirements have been satisfied.
Sec. 71.125 Control of measuring and test equipment.
The licensee, certificate holder, and applicant for a CoC shall
establish measures to assure that tools, gauges, instruments, and other
measuring and testing devices used in activities affecting quality are
properly controlled, calibrated, and adjusted at specified times to
maintain accuracy within necessary limits.
Sec. 71.127 Handling, storage, and shipping control.
The licensee, certificate holder, and applicant for a CoC shall
establish measures to control, in accordance with instructions, the
handling, storage, shipping, cleaning, and preservation of materials
and equipment to be used in packaging to prevent damage or
deterioration. When necessary for particular products, special
protective environments, such as inert gas atmosphere, and specific
moisture content and temperature levels must be specified and provided.
Sec. 71.129 Inspection, test, and operating status.
(a) The licensee, certificate holder, and applicant for a CoC shall
establish measures to indicate, by the use of markings such as stamps,
tags, labels, routing cards, or other suitable means, the status of
inspections and tests performed upon individual items of the packaging.
These measures must provide for the identification of items that have
satisfactorily passed required inspections and tests, where necessary
to preclude inadvertent bypassing of the inspections and tests.
(b) The licensee shall establish measures to identify the operating
status of components of the packaging, such as tagging valves and
switches, to prevent inadvertent operation.
Sec. 71.131 Nonconforming materials, parts, or components.
The licensee, certificate holder, and applicant for a CoC shall
establish measures to control materials, parts, or components that do
not conform to the licensee's requirements to prevent their inadvertent
use or installation. These measures must include, as appropriate,
procedures for identification, documentation, segregation, disposition,
and notification to affected organizations. Nonconforming items must be
reviewed and accepted, rejected, repaired, or reworked in accordance
with documented procedures.
Sec. 71.133 Corrective action.
The licensee, certificate holder, and applicant for a CoC shall
establish
[[Page 3800]]
measures to assure that conditions adverse to quality, such as
deficiencies, deviations, defective material and equipment, and
nonconformances, are promptly identified and corrected. In the case of
a significant condition adverse to quality, the measures must assure
that the cause of the condition is determined and corrective action
taken to preclude repetition. The identification of the significant
condition adverse to quality, the cause of the condition, and the
corrective action taken must be documented and reported to appropriate
levels of management.
Sec. 71.135 Quality assurance records.
The licensee, certificate holder, and applicant for a CoC shall
maintain sufficient written records to describe the activities
affecting quality. The records must include the instructions,
procedures, and drawings required by Sec. 71.111 to prescribe quality
assurance activities and must include closely related specifications
such as required qualifications of personnel, procedures, and
equipment. The records must include the instructions or procedures
which establish a records retention program that is consistent with
applicable regulations and designates factors such as duration,
location, and assigned responsibility. The licensee, certificate
holder, and applicant for a CoC shall retain these records for 3 years
beyond the date when the licensee, certificate holder, and applicant
for a CoC last engage in the activity for which the quality assurance
program was developed. If any portion of the written procedures or
instructions is superseded, the licensee, certificate holder, and
applicant for a CoC shall retain the superseded material for 3 years
after it is superseded.
Sec. 71.137 Audits.
The licensee, certificate holder, and applicant for a CoC shall
carry out a comprehensive system of planned and periodic audits to
verify compliance with all aspects of the quality assurance program and
to determine the effectiveness of the program. The audits must be
performed in accordance with written procedures or checklists by
appropriately trained personnel not having direct responsibilities in
the areas being audited. Audited results must be documented and
reviewed by management having responsibility in the area audited.
Followup action, including reaudit of deficient areas, must be taken
where indicated.
0
17. Appendix A to part 71 is revised to read as follows:
Appendix A to Part 71--Determination of A1 and A2
I. Values of A1 and A2 for individual radionuclides, which are
the bases for many activity limits elsewhere in these regulations,
are given in Table A-1. The curie (Ci) values specified are obtained
by converting from the Terabecquerel (TBq) figure. The curie values
are expressed to three significant figures to assure that the
difference in the TBq and Ci quantities is one tenth of one percent
or less. Where values of A1 and A2 are unlimited, it is for
radiation control purposes only. For nuclear criticality safety,
some materials are subject to controls placed on fissile material.
II. a. For individual radionuclides whose identities are known,
but which are not listed in Table A-1, the A1 and A2 values
contained in Table A-3 may be used. Otherwise, the licensee shall
obtain prior Commission approval of the A1 and A2 values for
radionuclides not listed in Table A-1, before shipping the material.
b. For individual radionuclides whose identities are known, but
which are not listed in Table A-2, the exempt material activity
concentration and exempt consignment activity values contained in
Table A-3 may be used. Otherwise, the licensee shall obtain prior
Commission approval of the exempt material activity concentration
and exempt consignment activity values for radionuclides not listed
in Table A-2, before shipping the material.
c. The licensee shall submit requests for prior approval,
described under paragraphs II.a. and II.b. of this Appendix, to the
Commission, in accordance with Sec. 71.1 of this part.
III. In the calculations of A1 and A2 for
a radionuclide not in Table A-1, a single radioactive decay chain,
in which radionuclides are present in their naturally occurring
proportions, and in which no daughter radionuclide has a half-life
either longer than 10 days, or longer than that of the parent
radionuclide, shall be considered as a single radionuclide, and the
activity to be taken into account, and the A1 and A2 value to be
applied, shall be those corresponding to the parent radionuclide of
that chain. In the case of radioactive decay chains in which any
daughter radionuclide has a half-life either longer than 10 days, or
greater than that of the parent radionuclide, the parent and those
daughter radionuclides shall be considered as mixtures of different
radionuclides.
IV. For mixtures of radionuclides whose identities and
respective activities are known, the following conditions apply:
a. For special form radioactive material, the maximum quantity
transported in a Type A package is as follows:
[GRAPHIC] [TIFF OMITTED] TR26JA04.014
where B(i) is the activity of radionuclide I, and A1(i) is the A1
value for radionuclide I.
b. For normal form radioactive material, the maximum quantity
transported in a Type A package is as follows:
[GRAPHIC] [TIFF OMITTED] TR26JA04.015
where B(i) is the activity of radionuclide I, and A2(i) is the A2(i)
value for radionuclide I.
c. Alternatively, the A1 value for mixtures of
special form material may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TR26JA04.016
where f(i) is the fraction of activity for radionuclide I in the
mixture, and A1(i) is the appropriate A1 value for radionuclide I.
d. Alternatively, the A2 value for mixtures of normal form
material may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TR26JA04.017
where f(i) is the fraction of activity for radionuclide I in the
mixture, and A2(i) is the appropriate A2 value for
radionuclide I.
e. The exempt activity concentration for mixtures of nuclides
may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TR26JA04.018
where f(i) is the fraction of activity concentration of radionuclide
I in the mixture, and [A] is the activity concentration for exempt
material containing radionuclide I.
f. The activity limit for an exempt consignment for mixtures of
radionuclides may be determined as follows:
[[Page 3801]]
[GRAPHIC] [TIFF OMITTED] TR26JA04.019
where f(i) is the fraction of activity of radionuclide I in the
mixture, and A is the activity limit for exempt consignments for
radionuclide I.
V. When the identity of each radionuclide is known, but the
individual activities of some of the radionuclides are not known,
the radionuclides may be grouped, and the lowest A1 or A2 value, as
appropriate, for the radionuclides in each group may be used in
applying the formulas in paragraph IV. Groups may be based on the
total alpha activity and the total beta/gamma activity when these
are known, using the lowest A1 or A2 values
for the alpha emitters and beta/gamma emitters.
Table A-1.--A1 and A2 Values For Radionuclides
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
Specific activity
Symbol of radionuclide Element and atomic A1 (TBq) A1 (Ci) A2 (TBq) A2 (Ci) --------------------------------------------
number (TBq/g) (Ci/g)
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
Ac-225 (a)....................... Actinium (89)....... 8.0x10-1 2.2x101 6.0x10-3 1.6x10-1 2.1x103 5.8x104
Ac-227 (a)....................... .................... 9.0x10-1 2.4x101 9.0x10-5 2.4x10-3 2.7 7.2x101
Ac-228........................... .................... 6.0x10-1 1.6x101 5.0x10-1 1.4x101 8.4x104 2.2x106
Ag-105........................... Silver (47)......... 2.0 5.4x101 2.0 5.4x101 1.1x103 3.0x104
Ag-108m (a)...................... .................... 7.0x10-1 1.9x101 7.0x10-1 1.9x101 9.7x10-1 2.6x101
Ag-110m (a)...................... .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 1.8x102 4.7x103
Ag-111........................... .................... 2.0 5.4x101 6.0x10-1 1.6x101 5.8x103 1.6x105
Al-26............................ Aluminum (13)....... 1.0x10-1 2.7 1.0x10-1 2.7 7.0x10-4 1.9x10-2
Am-241........................... Americium (95)...... 1.0x101 2.7x102 1.0x10-3 2.7x10-2 1.3x10-1 3.4
Am-242m (a)...................... .................... 1.0x101 2.7x102 1.0x10-3 2.7x10-2 3.6x10-1 1.0x101
Am-243 (a)....................... .................... 5.0 1.4x102 1.0x10-3 2.7x10-2 7.4x10-3 2.0x10-1
Ar-37............................ Argon (18).......... 4.0x101 1.1x103 4.0x101 1.1x103 3.7x103 9.9x104
Ar-39............................ .................... 4.0x101 1.1x103 2.0x101 5.4x102 1.3 3.4x101
Ar-41............................ .................... 3.0x10-1 8.1 3.0x10-1 8.1 1.5x106 4.2x107
As-72............................ Arsenic (33)........ 3.0x10-1 8.1 3.0x10-1 8.1 6.2x104 1.7x106
As-73............................ .................... 4.0x101 1.1x103 4.0x101 1.1x103 8.2x102 2.2x104
As-74............................ .................... 1.0 2.7x101 9.0x10-1 2.4x101 3.7x103 9.9x104
As-76............................ .................... 3.0x10-1 8.1 3.0x10-1 8.1 5.8x104 1.6x106
As-77............................ .................... 2.0x101 5.4x102 7.0x10-1 1.9x101 3.9x104 1.0x106
At-211 (a)....................... Astatine (85)....... 2.0x101 5.4x102 5.0x10-1 1.4x101 7.6x104 2.1x106
Au-193........................... Gold (79)........... 7.0 1.9x102 2.0 5.4x101 3.4x104 9.2x105
Au-194........................... .................... 1.0 2.7x101 1.0 2.7x101 1.5x104 4.1x105
Au-195........................... .................... 1.0x101 2.7x102 6.0 1.6x102 1.4x102 3.7x103
Au-198........................... .................... 1.0 2.7x101 6.0x10-1 1.6x101 9.0x103 2.4x105
Au-199........................... .................... 1.0x101 2.7x102 6.0x10-1 1.6x101 7.7x103 2.1x105
Ba-131 (a)....................... Barium (56)......... 2.0 5.4x101 2.0 5.4x101 3.1x103 8.4x104
Ba-133........................... .................... 3.0 8.1x101 3.0 8.1x101 9.4 2.6x102
Ba-133m.......................... .................... 2.0x101 5.4x102 6.0x10-1 1.6x101 2.2x104 6.1x105
Ba-140 (a)....................... .................... 5.0x10-1 1.4x101 3.0x10-1 8.1 2.7x103 7.3x104
Be-7............................. Beryllium (4)....... 2.0x101 5.4x102 2.0x101 5.4x102 1.3x104 3.5x105
Be-10............................ .................... 4.0x101 1.1x103 6.0x10-1 1.6x101 8.3x10-4 2.2x10-2
Bi-205........................... Bismuth (83)........ 7.0x10-1 1.9x101 7.0x10-1 1.9x101 1.5x10-3 4.2x104
Bi-206........................... .................... 3.0x10-1 8.1 3.0x10-1 8.1 3.8x103 1.0x105
Bi-207........................... .................... 7.0x10-1 1.9x101 7.0x10-1 1.9x101 1.9 5.2x101
Bi-210........................... .................... 1.0 2.7x101 6.0x10-1 1.6x101 4.6x103 1.2x105
Bi-210m (a)...................... .................... 6.0x10-1 1.6x101 2.0x10-2 5.4x10-1 2.1x10-5 5.7x10-4
Bi-212 (a)....................... .................... 7.0x10-1 1.9x101 6.0x10-1 1.6x101 5.4x105 1.5x107
Bk-247........................... Berkelium (97)...... 8.0 2.2x102 8.0x10-4 2.2x10-2 3.8x10-2 1.0
Bk-249 (a)....................... .................... 4.0x101 1.1x103 3.0x10-1 8.1 6.1x101 1.6x103
Br-76............................ Bromine (35)........ 4.0x10-1 1.1x101 4.0x10-1 1.1x101 9.4x104 2.5x106
Br-77............................ .................... 3.0 8.1x101 3.0 8.1x101 2.6x104 7.1x105
Br-82............................ .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 4.0x104 1.1x106
C-11............................. Carbon (6).......... 1.0 2.7x101 6.0x10-1 1.6x101 3.1x107 8.4x108
C-14............................. .................... 4.0x101 1.1x103 3.0 8.1x101 1.6x10-1 4.5
Ca-41............................ Calcium (20)........ Unlimited Unlimited Unlimited Unlimited 3.1x10-3 8.5x10-2
Ca-45............................ .................... 4.0x101 1.1x103 1.0 2.7x101 6.6x102 1.8x104
Ca-47 (a)........................ .................... 3.0 8.1x101 3.0x10-1 8.1 2.3x104 6.1x105
Cd-109........................... Cadmium (48)........ 3.0x101 8.1x102 2.0 5.4x101 9.6x101 2.6x103
Cd-113m.......................... .................... 4.0x101 1.1x103 5.0x10-1 1.4x101 8.3 2.2x102
Cd-115 (a)....................... .................... 3.0 8.1x101 4.0x10-1 1.1x101 1.9x104 5.1x105
Cd-115m.......................... .................... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 9.4x102 2.5x104
Ce-139........................... Cerium (58)......... 7.0 1.9x102 2.0 5.4x101 2.5x102 6.8x103
Ce-141........................... .................... 2.0x101 5.4x102 6.0x10-1 1.6x101 1.1x103 2.8x104
Ce-143........................... .................... 9.0x10-1 2.4x101 6.0x10-1 1.6x101 2.5x104 6.6x105
Ce-144 (a)....................... .................... 2.0x10-1 5.4 2.0x10-1 5.4 1.2x102 3.2x103
Cf-248........................... Californium (98).... 4.0x101 1.1x103 6.0x10-3 1.6x10-1 5.8x101 1.6x103
Cf-249........................... .................... 3.0 8.1x101 8.0x10-4 2.2x10-2 1.5x10-1 4.1
[[Page 3802]]
Cf-250........................... .................... 2.0x101 5.4x102 2.0x10-3 5.4x10-2 4.0 1.1x102
Cf-251........................... .................... 7.0 1.9x102 7.0x10-4 1.9x10-2 5.9x10-2 1.6
Cf-252 (h)....................... .................... 5.0x10-2 1.4 3.0x10-3 8.1x10-2 2.0x101 5.4x102
Cf-253 (a)....................... .................... 4.0x101 1.1x103 4.0x10-2 1.1 1.1x103 2.9x104
Cf-254........................... .................... 1.0x10-3 2.7x10-2 1.0x10-3 2.7x10-2 3.1x102 8.5x103
Cl-36............................ Chlorine (17)....... 1.0x101 2.7x102 6.0x10-1 1.6x101 1.2x10-3 3.3x10-2
Cl-38............................ .................... 2.0x10-1 5.4 2.0x10-1 5.4 4.9x106 1.3x108
Cm-240........................... Curium (96)......... 4.0x101 1.1x103 2.0x10-2 5.4x10-1 7.5x102 2.0x104
Cm-241........................... .................... 2.0 5.4x101 1.0 2.7x101 6.1x102 1.7x104
Cm-242........................... .................... 4.0x101 1.1x103 1.0x10-2 2.7x10-1 1.2x102 3.3x103
Cm-243........................... .................... 9.0 2.4x102 1.0x10-3 2.7x10-2 1.9x10-3 5.2x101
Cm-244........................... .................... 2.0x101 5.4x102 2.0x10-3 5.4x10-2 3.0 8.1x101
Cm-245........................... .................... 9.0 2.4x102 9.0x10-4 2.4x10-2 6.4x10-3 1.7x10-1
Cm-246........................... .................... 9.0 2.4x102 9.0x10-4 2.4x10-2 1.1x10-2 3.1x10-1
Cm-247 (a)....................... .................... 3.0 8.1x101 1.0x10-3 2.7x10-2 3.4x10-6 9.3x10-5
Cm-248........................... .................... 2.0x10-2 5.4x10-1 3.0x10-4 8.1x10-3 1.6x10-5 4.2x10-3
Co-55............................ Cobalt (27)......... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 1.1x105 3.1x106
Co-56............................ .................... 3.0x10-1 8.1 3.0x10-1 8.1 1.1x103 3.0x104
Co-57............................ .................... 1.0x101 2.7x102 1.0x101 2.7x102 3.1x102 8.4x103
Co-58............................ .................... 1.0 2.7x101 1.0 2.7x101 1.2x103 3.2x104
Co-58m........................... .................... 4.0x101 1.1x103 4.0x101 1.1x103 2.2x105 5.9x106
Co-60............................ .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 4.2x101 1.1x103
Cr-51............................ Chromium (24)....... 3.0x101 8.1x102 3.0x101 8.1x102 3.4x103 9.2x104
Cs-129........................... Cesium (55)......... 4.0 1.1x102 4.0 1.1x102 2.8x104 7.6x105
Cs-131........................... .................... 3.0x101 8.1x102 3.0x101 8.1x102 3.8x103 1.0x105
Cs-132........................... .................... 1.0 2.7x101 1.0 2.7x101 5.7x103 1.5x105
Cs-134........................... .................... 7.0x10-1 1.9x101 7.0x10-1 1.9x101 4.8x101 1.3x103
Cs-134m.......................... .................... 4.0x101 1.1x103 6.0x10-1 1.6x101 3.0x105 8.0x106
Cs-135........................... .................... 4.0x101 1.1x103 1.0 2.7x101 4.3x10-5 1.2x10-3
Cs-136........................... .................... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 2.7x103 7.3x104
Cs-137 (a)....................... .................... 2.0 5.4x101 6.0x10-1 1.6x101 3.2 8.7x101
Cu-64............................ Copper (29)......... 6.0 1.6x102 1.0 2.7x101 1.4x105 3.9x106
Cu-67............................ .................... 1.0x101 2.7x102 7.0x10-1 1.9x101 2.8x104 7.6x105
Dy-159........................... Dysprosium (66)..... 2.0x101 5.4x102 2.0x101 5.4x102 2.1x102 5.7x103
Dy-165........................... .................... 9.0x10-1 2.4x101 6.0x10-1 1.6x101 3.0x105 8.2x106
Dy-166 (a)....................... .................... 9.0x10-1 2.4x101 3.0x10-1 8.1 8.6x103 2.3x105
Er-169........................... Erbium (68)......... 4.0x101 1.1x103 1.0 2.7x101 3.1x103 8.3x104
Er-171........................... .................... 8.0x10-1 2.2x101 5.0x10-1 1.4x101 9.0x104 2.4x106
Eu-147........................... Europium (63)....... 2.0 5.4x101 2.0 5.4x101 1.4x103 3.7x104
Eu-148........................... .................... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 6.0x102 1.6x104
Eu-149........................... .................... 2.0x101 5.4x102 2.0x101 5.4x102 3.5x102 9.4x103
Eu-150 (short lived)............. .................... 2.0 5.4x101 7.0x10-1 1.9x101 6.1x104 1.6x106
Eu-150 (long lived).............. .................... 7 x 10-1 1.9x101 7.0x10-1 1.9x101 6.1x104 1.6x106
Eu-152........................... .................... 1.0 2.7x101 1.0 2.7x101 6.5 1.8x102
Eu-152m.......................... .................... 8.0x10-1 2.2x101 8.0x10-1 2.2x101 8.2x104 2.2x106
Eu-154........................... .................... 9.0x10-1 2.4x101 6.0x10-1 1.6x101 9.8 2.6x102
Eu-155........................... .................... 2.0x101 5.4x102 3.0 8.1x101 1.8x101 4.9x102
Eu-156........................... .................... 7.0x10-1 1.9x101 7.0x10-1 1.9x101 2.0x103 5.5x104
F-18............................. Fluorine (9)........ 1.0 2.7x101 6.0x10-1 1.6x101 3.5x106 9.5x107
Fe-52 (a)........................ Iron (26)........... 3.0x10-1 8.1 3.0x10-1 8.1 2.7x105 7.3x106
Fe-55............................ .................... 4.0x101 1.1x103 4.0x101 1.1x103 8.8x101 2.4x103
Fe-59............................ .................... 9.0x10-1 2.4x101 9.0x10-1 2.4x101 1.8x103 5.0x104
Fe-60 (a)........................ .................... 4.0x101 1.1x103 2.0x10-1 5.4 7.4x10-4 2.0x10-2
Ga-67............................ Gallium (31)........ 7.0 1.9x102 3.0 8.1x101 2.2x104 6.0x105
Ga-68............................ .................... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 1.5x106 4.1x107
Ga-72............................ .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 1.1x105 3.1x106
Gd-146 (a)....................... Gadolinium (64)..... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 6.9x102 1.9x104
Gd-148........................... .................... 2.0x101 5.4x102 2.0x10-3 5.4x10-2 1.2 3.2x101
Gd-153........................... .................... 1.0x101 2.7x102 9.0 2.4x102 1.3x102 3.5x103
Gd-159........................... .................... 3.0 8.1x101 6.0x10-1 1.6x101 3.9x104 1.1x106
Ge-68 (a)........................ Germanium (32)...... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 2.6x102 7.1x103
Ge-71............................ .................... 4.0x101 1.1x103 4.0x101 1.1x103 5.8x103 1.6x105
Ge-77............................ .................... 3.0x10-1 8.1 3.0x10-1 8.1 1.3x105 3.6x106
Hf-172 (a)....................... Hafnium (72)........ 6.0x10-1 1.6x101 6.0x10-1 1.6x101 4.1x101 1.1x103
Hf-175........................... .................... 3.0 8.1x101 3.0 8.1x101 3.9x102 1.1x104
Hf-181........................... .................... 2.0 5.4x101 5.0x10-1 1.4x101 6.3x102 1.7x104
Hf-182........................... .................... Unlimited Unlimited Unlimited Unlimited 8.1x10-6 2.2x10-4
Hg-194 (a)....................... Mercury (80)........ 1.0 2.7x101 1.0 2.7x101 1.3x10-1 3.5
[[Page 3803]]
Hg-195m (a)...................... .................... 3.0 8.1x101 7.0x10-1 1.9x101 1.5x104 4.0x105
Hg-197........................... .................... 2.0x101 5.4x102 1.0x101 2.7x102 9.2x103 2.5x105
Hg-197m.......................... .................... 1.0x101 2.7x102 4.0x10-1 1.1x101 2.5x104 6.7x105
Hg-203........................... .................... 5.0 1.4x102 1.0 2.7x101 5.1x102 1.4x104
Ho-166........................... Holmium (67)........ 4.0x10-1 1.1x101 4.0x10-1 1.1x101 2.6x104 7.0x105
Ho-166m.......................... .................... 6.0x10-1 1.6x101 5.0x10-1 1.4x101 6.6x10-2 1.8
I-123............................ Iodine (53)......... 6.0 1.6x102 3.0 8.1x101 7.1x104 1.9x106
I-124............................ .................... 1.0 2.7x101 1.0 2.7x101 9.3x103 2.5x105
I-125............................ .................... 2.0x101 5.4x102 3.0 8.1x101 6.4x102 1.7x104
I-126............................ .................... 2.0 5.4x101 1.0 2.7x101 2.9x103 8.0x104
I-129............................ .................... Unlimited Unlimited Unlimited Unlimited 6.5x10-6 1.8x10-4
I-131............................ .................... 3.0 8.1x101 7.0x10-1 1.9x101 4.6x103 1.2x105
I-132............................ .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 3.8x105 1.0x107
I-133............................ .................... 7.0x10-1 1.9x101 6.0x10-1 1.6x101 4.2x104 1.1x106
I-134............................ .................... 3.0x10-1 8.1 3.0x10-1 8.1 9.9x105 2.7x107
I-135 (a)........................ .................... 6.0x10-1 1.6x101 6.0x10-1 1.6x101 1.3x105 3.5x106
In-111........................... Indium (49)......... 3.0 8.1x101 3.0 8.1x101 1.5x104 4.2x105
In-113m.......................... .................... 4.0 1.1x102 2.0 5.4x101 6.2x105 1.7x107
In-114m (a)...................... .................... 1.0x101 2.7x102 5.0x10-1 1.4x101 8.6x102 2.3x104
In-115m.......................... .................... 7.0 1.9x102 1.0 2.7x101 2.2x105 6.1x106
Ir-189 (a)....................... Iridium (77)........ 1.0x101 2.7x102 1.0x101 2.7x102 1.9x103 5.2x104
Ir-190........................... .................... 7.0x10-1 1.9x101 7.0x10-1 1.9x101 2.3x103 6.2x104
Ir-192 (c)....................... .................... 1.0 2.7x101 6.0x10-1 1.6x101 3.4x102 9.2x103
Ir-194........................... .................... 3.0x10-1 8.1 3.0x10-1 8.1 3.1x104 8.4x105
K-40............................. Potassium (19)...... 9.0x10-1 2.4x101 9.0x10-1 2.4x101 2.4x10-7 6.4x10-6
K-42............................. .................... 2.0x10-1 5.4 2.0x10-1 5.4 2.2x105 6.0x106
K-43............................. .................... 7.0x10-1 1.9x101 6.0x10-1 1.6x101 1.2x105 3.3x106
Kr-81............................ Krypton (36)........ 4.0x101 1.1x103 4.0x101 1.1x103 7.8x10-4 2.1x10-2
Kr-85............................ .................... 1.0x101 2.7x102 1.0x101 2.7x102 1.5x101 3.9x102
Kr-85m........................... .................... 8.0 2.2x102 3.0 8.1x101 3.0x105 8.2x106
Kr-87............................ .................... 2.0x10-1 5.4 2.0x10-1 5.4 1.0x106 2.8x107
La-137........................... Lanthanum (57)...... 3.0x101 8.1x102 6.0 1.6x102 1.6x10-3 4.4x10-2
La-140........................... .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 2.1x104 5.6x105
Lu-172........................... Lutetium (71)....... 6.0x10-1 1.6x101 6.0x10-1 1.6x101 4.2x103 1.1x105
Lu-173........................... .................... 8.0 2.2x102 8.0 2.2x102 5.6x101 1.5x103
Lu-174........................... .................... 9.0 2.4x102 9.0 2.4x102 2.3x101 6.2x102
Lu-174m.......................... .................... 2.0x101 5.4x102 1.0x101 2.7x102 2.0x102 5.3x103
Lu-177........................... .................... 3.0x101 8.1x102 7.0x10-1 1.9x101 4.1x103 1.1x105
Mg-28 (a)........................ Magnesium (12)...... 3.0x10-1 8.1 3.0x10-1 8.1 2.0x105 5.4x106
Mn-52............................ Manganese (25)...... 3.0x10-1 8.1 3.0x10-1 8.1 1.6x104 4.4x105
Mn-53............................ .................... Unlimited Unlimited Unlimited Unlimited 6.8x10-5 1.8x10-3
Mn-54............................ .................... 1.0 2.7x101 1.0 2.7x101 2.9x102 7.7x103
Mn-56............................ .................... 3.0x10-1 8.1 3.0x10-1 8.1 8.0x105 2.2x107
Mo-93............................ Molybdenum (42)..... 4.0x101 1.1x103 2.0x101 5.4x102 4.1x10-2 1.1
Mo-99 (a) (i).................... .................... 1.0 2.7x101 6.0x10-1 1.6x101 1.8x104 4.8x105
N-13............................. Nitrogen (7)........ 9.0x10-1 2.4x101 6.0x10-1 1.6x101 5.4x107 1.5x109
Na-22............................ Sodium (11)......... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 2.3x102 6.3x103
Na-24............................ .................... 2.0x10-1 5.4 2.0x10-1 5.4 3.2x105 8.7x106
Nb-93m........................... Niobium (41)........ 4.0x101 1.1x103 3.0x101 8.1x102 8.8 2.4x102
Nb-94............................ .................... 7.0x10-1 1.9x101 7.0x10-1 1.9x101 6.9x10-3 1.9x10-1
Nb-95............................ .................... 1.0 2.7x101 1.0 2.7x101 1.5x103 3.9x104
Nb-97............................ .................... 9.0x10-1 2.4x101 6.0x10-1 1.6x101 9.9x105 2.7x107
Nd-147........................... Neodymium (60)...... 6.0 1.6x102 6.0x10-1 1.6x101 3.0x103 8.1x104
Nd-149........................... .................... 6.0x10-1 1.6x101 5.0x10-1 1.4x101 4.5x105 1.2x107
Ni-59............................ Nickel (28)......... Unlimited Unlimited Unlimited Unlimited 3.0x10-3 8.0x10-2
Ni-63............................ .................... 4.0x101 1.1x103 3.0x101 8.1x102 2.1 5.7x101
Ni-65............................ .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 7.1x105 1.9x107
Np-235........................... Neptunium (93)...... 4.0x101 1.1x103 4.0x101 1.1x103 5.2x101 1.4x103
Np-236 (short-lived)............. .................... 2.0x101 5.4x102 2.0 5.4x101 4.7x10-4 1.3x10-2
Np-236 (long-lived).............. .................... 9.0x100 2.4x102 2.0x10-2 5.4x10-1 4.7x10-4 1.3x10-2
Np-237........................... .................... 2.0x101 5.4x102 2.0x10-3 5.4x10-2 2.6x10-5 7.1x10-4
Np-239........................... .................... 7.0 1.9x102 4.0x10-1 1.1x101 8.6x103 2.3x105
Os-185........................... Osmium (76)......... 1.0 2.7x101 1.0 2.7x101 2.8x102 7.5x103
Os-191........................... .................... 1.0x101 2.7x102 2.0 5.4x101 1.6x103 4.4x104
Os-191m.......................... .................... 4.0x101 1.1x103 3.0x101 8.1x102 4.6x104 1.3x106
Os-193........................... .................... 2.0 5.4x101 6.0x10-1 1.6x101 2.0x104 5.3x105
Os-194 (a)....................... .................... 3.0x10-1 8.1 3.0x10-1 8.1 1.1x101 3.1x102
P-32............................. Phosphorus (15)..... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 1.1x104 2.9x105
[[Page 3804]]
P-33............................. .................... 4.0x101 1.1x103 1.0 2.7x101 5.8x103 1.6x105
Pa-230 (a)....................... Protactinium (91)... 2.0 5.4x101 7.0x10-2 1.9 1.2x103 3.3x104
Pa-231........................... .................... 4.0 1.1x102 4.0x10-4 1.1x10-2 1.7x10-3 4.7x10-2
Pa-233........................... .................... 5.0 1.4x102 7.0x10-1 1.9x101 7.7x102 2.1x104
Pb-201........................... Lead (82)........... 1.0 2.7x101 1.0 2.7x101 6.2x104 1.7x106
Pb-202........................... .................... 4.0x101 1.1x103 2.0x101 5.4x102 1.2x10-4 3.4x10-3
Pb-203........................... .................... 4.0 1.1x102 3.0 8.1x101 1.1x104 3.0x105
Pb-205........................... .................... Unlimited Unlimited Unlimited Unlimited 4.5x10-6 1.2x10-4
Pb-210 (a)....................... .................... 1.0 2.7x101 5.0x10-2 1.4 2.8 7.6x101
Pb-212 (a)....................... .................... 7.0x10-1 1.9x101 2.0x10-1 5.4 5.1x104 1.4x106
Pd-103 (a)....................... Palladium (46)...... 4.0x101 1.1x103 4.0x101 1.1x103 2.8x103 7.5x104
Pd-107........................... .................... Unlimited Unlimited Unlimited Unlimited 1.9x10-5 5.1x10-4
Pd-109........................... .................... 2.0 5.4x101 5.0x10-1 1.4x101 7.9x104 2.1x106
Pm-143........................... Promethium (61)..... 3.0 8.1x101 3.0 8.1x101 1.3x102 3.4x103
Pm-144........................... .................... 7.0x10-1 1.9x101 7.0x10-1 1.9x101 9.2x101 2.5x103
Pm-145........................... .................... 3.0x101 8.1x102 1.0x101 2.7x102 5.2 1.4x102
Pm-147........................... .................... 4.0x101 1.1x103 2.0 5.4x101 3.4x101 9.3x102
Pm-148m (a)...................... .................... 8.0x10-1 2.2x101 7.0x10-1 1.9x101 7.9x102 2.1x104
Pm-149........................... .................... 2.0 5.4x101 6.0x10-1 1.6x101 1.5x104 4.0x105
Pm-151........................... .................... 2.0 5.4x101 6.0x10-1 1.6x101 2.7x104 7.3x105
Po-210........................... Polonium (84)....... 4.0x101 1.1x103 2.0x10-2 5.4x10-1 1.7x102 4.5x103
Pr-142........................... Praseodymium (59)... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 4.3x104 1.2x106
Pr-143........................... .................... 3.0 8.1x101 6.0x10-1 1.6x101 2.5x103 6.7x104
Pt-188 (a)....................... Platinum (78)....... 1.0 2.7x101 8.0x10-1 2.2x101 2.5x103 6.8x104
Pt-191........................... .................... 4.0 1.1x102 3.0 8.1x101 8.7x103 2.4x105
Pt-193........................... .................... 4.0x101 1.1x103 4.0x101 1.1x103 1.4 3.7x101
Pt-193m.......................... .................... 4.0x101 1.1x103 5.0x10-1 1.4x101 5.8x103 1.6x105
Pt-195m.......................... .................... 1.0x101 2.7x102 5.0x10-1 1.4x101 6.2x103 1.7x105
Pt-197........................... .................... 2.0x101 5.4x102 6.0x10-1 1.6x101 3.2x104 8.7x105
Pt-197m.......................... .................... 1.0x101 2.7x102 6.0x10-1 1.6x101 3.7x105 1.0x107
Pu-236........................... Plutonium (94)...... 3.0x101 8.1x102 3.0x10-3 8.1x10-2 2.0x101 5.3x102
Pu-237........................... .................... 2.0x101 5.4x102 2.0x101 5.4x102 4.5x102 1.2x104
Pu-238........................... .................... 1.0x101 2.7x102 1.0x10-3 2.7x10-2 6.3x10-1 1.7x101
Pu-239........................... .................... 1.0x101 2.7x102 1.0x10-3 2.7x10-2 2.3x10-3 6.2x10-2
Pu-240........................... .................... 1.0x101 2.7x102 1.0x10-3 2.7x10-2 8.4x10-3 2.3x10-1
Pu-241 (a)....................... .................... 4.0x101 1.1x103 6.0x10-2 1.6 3.8 1.0x102
Pu-242........................... .................... 1.0x101 2.7x102 1.0x10-3 2.7x10-2 1.5x10-4 3.9x10-3
Pu-244 (a)....................... .................... 4.0x10-1 1.1x101 1.0x10-3 2.7x10-2 6.7x10-7 1.8x10-5
Ra-223 (a)....................... Radium (88)......... 4.0x10-1 1.1x101 7.0x10-3 1.9x10-1 1.9x103 5.1x104
Ra-224 (a)....................... .................... 4.0x10-1 1.1x101 2.0x10-2 5.4x10-1 5.9x103 1.6x105
Ra-225 (a)....................... .................... 2.0x10-1 5.4 4.0x10-3 1.1x10-1 1.5x103 3.9x104
Ra-226 (a)....................... .................... 2.0x10-1 5.4 3.0x10-3 8.1x10-2 3.7x10-2 1.0
Ra-228 (a)....................... .................... 6.0x10-1 1.6x101 2.0x10-2 5.4x10-1 1.0x101 2.7x102
Rb-81............................ Rubidium (37)....... 2.0 5.4x101 8.0x10-1 2.2x101 3.1x105 8.4x106
Rb-83 (a)........................ .................... 2.0 5.4x101 2.0 5.4x101 6.8x102 1.8x104
Rb-84............................ .................... 1.0 2.7x101 1.0 2.7x101 1.8x103 4.7x104
Rb-86............................ .................... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 3.0x103 8.1x104
Rb-87............................ .................... Unlimited Unlimited Unlimited Unlimited 3.2x10-9 8.6x10-8
Rb(nat).......................... .................... Unlimited Unlimited Unlimited Unlimited 6.7x106 1.8x108
Re-184........................... Rhenium (75)........ 1.0 2.7x101 1.0 2.7x101 6.9x102 1.9x104
Re-184m.......................... .................... 3.0 8.1x101 1.0 2.7x101 1.6x102 4.3x103
Re-186........................... .................... 2.0 5.4x101 6.0x10-1 1.6x101 6.9x103 1.9x105
Re-187........................... .................... Unlimited Unlimited Unlimited Unlimited 1.4x10-9 3.8x10-8
Re-188........................... .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 3.6x104 9.8x105
Re-189 (a)....................... .................... 3.0 8.1x101 6.0x10-1 1.6x101 2.5x104 6.8x105
Re(nat).......................... .................... Unlimited Unlimited Unlimited Unlimited 0.0 2.4x10-8
Rh-99............................ Rhodium (45)........ 2.0 5.4x101 2.0 5.4x101 3.0x103 8.2x104
Rh-101........................... .................... 4.0 1.1x102 3.0 8.1x101 4.1x101 1.1x103
Rh-102........................... .................... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 4.5x101 1.2x103
Rh-102m.......................... .................... 2.0 5.4x101 2.0 5.4x101 2.3x102 6.2x103
Rh-103m.......................... .................... 4.0x101 1.1x103 4.0x101 1.1x103 1.2x106 3.3x107
Rh-105........................... .................... 1.0x101 2.7x102 8.0x10-1 2.2x101 3.1x104 8.4x105
Rn-222 (a)....................... Radon (86).......... 3.0x10-1 8.1 4.0x10-3 1.1x10-1 5.7x103 1.5x105
Ru-97............................ Ruthenium (44)...... 5.0 1.4x102 5.0 1.4x102 1.7x104 4.6x105
Ru-103 (a)....................... .................... 2.0 5.4x101 2.0 5.4x101 1.2x103 3.2x104
Ru-105........................... .................... 1.0 2.7x101 6.0x10-1 1.6x101 2.5x105 6.7x106
Ru-106 (a)....................... .................... 2.0x10-1 5.4 2.0x10-1 5.4 1.2x102 3.3x103
S-35............................. Sulphur (16)........ 4.0x101 1.1x103 3.0 8.1x101 1.6x103 4.3x104
Sb-122........................... Antimony (51)....... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 1.5x104 4.0x105
Sb-124........................... .................... 6.0x10-1 1.6x101 6.0x10-1 1.6x101 6.5x102 1.7x104
[[Page 3805]]
Sb-125........................... .................... 2.0 5.4x101 1.0 2.7x101 3.9x101 1.0x103
Sb-126........................... .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 3.1x103 8.4x104
Sc-44............................ Scandium (21)....... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 6.7x105 1.8x107
Sc-46............................ .................... 5.0x10-1 1.4x101 5.0x10-1 1.4x101 1.3x103 3.4x104
Sc-47............................ .................... 1.0x101 2.7x102 7.0x10-1 1.9x101 3.1x104 8.3x105
Sc-48............................ .................... 3.0x10-1 8.1 3.0x10-1 8.1 5.5x104 1.5x106
Se-75............................ Selenium (34)....... 3.0 8.1x101 3.0 8.1x101 5.4x102 1.5x104
Se-79............................ .................... 4.0x101 1.1x103 2.0 5.4x101 2.6x10-3 7.0x10-2
Si-31............................ Silicon (14)........ 6.0x10-1 1.6x101 6.0x10-1 1.6x101 1.4x106 3.9x107
Si-32............................ .................... 4.0x101 1.1x103 5.0x10-1 1.4x101 3.9 1.1x102
Sm-145........................... Samarium (62)....... 1.0x101 2.7x102 1.0x101 2.7x102 9.8x101 2.6x103
Sm-147........................... .................... Unlimited Unlimited Unlimited Unlimited 8.5x10-1 2.3x10-8
Sm-151........................... .................... 4.0x101 1.1x103 1.0x101 2.7x102 9.7x10-1 2.6x101
Sm-153........................... .................... 9.0 2.4x102 6.0x10-1 1.6x101 1.6x104 4.4x105
Sn-113 (a)....................... Tin (50)............ 4.0 1.1x102 2.0 5.4x101 3.7x102 1.0x104
Sn-117m.......................... .................... 7.0 1.9x102 4.0x10-1 1.1x101 3.0x103 8.2x104
Sn-119m.......................... .................... 4.0x101 1.1x103 3.0x101 8.1x102 1.4x102 3.7x103
Sn-121m (a)...................... .................... 4.0x101 1.1x103 9.0x10-1 2.4x101 2.0 5.4x101
Sn-123........................... .................... 8.0x10-1 2.2x101 6.0x10-1 1.6x101 3.0x102 8.2x103
Sn-125........................... .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 4.0x103 1.1x105
Sn-126 (a)....................... .................... 6.0x10-1 1.6x101 4.0x10-1 1.1x101 1.0x10-3 2.8x10-2
Sr-82 (a)........................ Strontium (38)...... 2.0x10-1 5.4 2.0x10-1 5.4 2.3x103 6.2x104
Sr-85............................ .................... 2.0 5.4x101 2.0 5.4x101 8.8x102 2.4x104
Sr-85m........................... .................... 5.0 1.4x102 5.0 1.4x102 1.2x106 3.3x107
Sr-87m........................... .................... 3.0 8.1x101 3.0 8.1x101 4.8x105 1.3x107
Sr-89............................ .................... 6.0x10-1 1.6x101 6.0x10-1 1.6x101 1.1x103 2.9x104
Sr-90 (a)........................ .................... 3.0x10-1 8.1 3.0x10-1 8.1 5.1 1.4x102
Sr-91 (a)........................ .................... 3.0x10-1 8.1 3.0x10-1 8.1 1.3x105 3.6x106
Sr-92 (a)........................ .................... 1.0 2.7x101 3.0x10-1 8.1 4.7x105 1.3x107
T(H-3)........................... Tritium (1)......... 4.0x101 1.1x103 4.0x101 1.1x103 3.6x102 9.7x103
Ta-178 (long-lived).............. Tantalum (73)....... 1.0 2.7x101 8.0x10-1 2.2x101 4.2x106 1.1x108
Ta-179........................... .................... 3.0x101 8.1x102 3.0x101 8.1x102 4.1x101 1.1x103
Ta-182........................... .................... 9.0x10-1 2.4x101 5.0x10-1 1.4x101 2.3x102 6.2x103
Tb-157........................... Terbium (65)........ 4.0x101 1.1x103 4.0x101 1.1x103 5.6x10-1 1.5x101
Tb-158........................... .................... 1.0 2.7x101 1.0 2.7x101 5.6x10-1 1.5x101
Tb-160........................... .................... 1.0 2.7x101 6.0x10-1 1.6x101 4.2x102 1.1x104
Tc-95m (a)....................... Technetium (43)..... 2.0 5.4x101 2.0 5.4x101 8.3x102 2.2x104
Tc-96............................ .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 1.2x104 3.2x105
Tc-96m (a)....................... .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 1.4x106 3.8x107
Tc-97............................ .................... Unlimited Unlimited Unlimited Unlimited 5.2x10-5 1.4x10-3
Tc-97m........................... .................... 4.0x101 1.1x103 1.0 2.7x101 5.6x102 1.5x104
Tc-98............................ .................... 8.0x10-1 2.2x101 7.0x10-1 1.9x101 3.2x10-5 8.7x10-4
Tc-99............................ .................... 4.0x101 1.1x103 9.0x10-1 2.4x101 6.3x10-4 1.7x10-2
Tc-99m........................... .................... 1.0x101 2.7x102 4.0 1.1x102 1.9x105 5.3x106
Te-121........................... Tellurium (52)...... 2.0 5.4x101 2.0 5.4x101 2.4x103 6.4x104
Te-121m.......................... .................... 5.0 1.4x102 3.0 8.1x101 2.6x102 7.0x103
Te-123m.......................... .................... 8.0 2.2x102 1.0 2.7x101 3.3x102 8.9x103
Te-125m.......................... .................... 2.0x101 5.4x102 9.0x10-1 2.4x101 6.7x102 1.8x104
Te-127........................... .................... 2.0x101 5.4x102 7.0x10-1 1.9x101 9.8x104 2.6x106
Te-127m (a)...................... .................... 2.0x101 5.4x102 5.0x10-1 1.4x101 3.5x102 9.4x103
Te-129........................... .................... 7.0x10-1 1.9x101 6.0x10-1 1.6x101 7.7x105 2.1x107
Te-129m (a)...................... .................... 8.0x10-1 2.2x101 4.0x10-1 1.1x101 1.1x103 3.0x104
Te-131m (a)...................... .................... 7.0x10-1 1.9x101 5.0x10-1 1.4x101 3.0x104 8.0x105
Te-132 (a)....................... .................... 5.0x10-1 1.4x101 4.0x10-1 1.1x101 1.1x104 8.0x105
Th-227........................... Thorium (90)........ 1.0x101 2.7x102 5.0x10-3 1.4x10-1 1.1x103 3.1x104
Th-228 (a)....................... .................... 5.0x10-1 1.4x101 1.0x10-3 2.7x10-2 3.0x101 8.2x102
Th-229........................... .................... 5.0 1.4x102 5.0x10-4 1.4x10-2 7.9x10-3 2.1x10-1
Th-230........................... .................... 1.0x101 2.7x102 1.0x10-3 2.7x10-2 7.6x10-4 2.1x10-2
Th-231........................... .................... 4.0x101 1.1x103 2.0x10-2 5.4x10-1 2.0x104 5.3x105
Th-232........................... .................... Unlimited Unlimited Unlimited Unlimited 4.0x10-9 1.1x10-7
Th-234 (a)....................... .................... 3.0x10-1 8.1 3.0x10-1 8.1 8.6x102 2.3x104
Th(nat).......................... .................... Unlimited Unlimited Unlimited Unlimited 8.1x10-9 2.2x10-7
Ti-44 (a)........................ Titanium (22)....... 5.0x10-1 1.4x101 4.0x10-1 1.1x101 6.4 1.7x102
Tl-200........................... Thallium (81)....... 9.0x10-1 2.4x101 9.0x10-1 2.4x101 2.2x104 6.0x105
Tl-201........................... .................... 1.0x101 2.7x102 4.0 1.1x102 7.9x103 2.1x105
Tl-202........................... .................... 2.0 5.4x101 2.0 5.4x101 2.0x103 5.3x104
Tl-204........................... .................... 1.0x101 2.7x102 7.0x10-1 1.9x101 1.7x101 4.6x102
Tm-167........................... Thulium (69)........ 7.0 1.9x102 8.0x10-1 2.2x101 3.1x103 8.5x104
Tm-170........................... .................... 3.0 8.1x101 6.0x10-1 1.6x101 2.2x102 6.0x103
[[Page 3806]]
Tm-171........................... .................... 4.0x101 1.1x103 4.0x101 1.1x103 4.0x101 1.1x103
U-230 (fast lung absorption) Uranium (92)........ 4.0x101 1.1x103 1.0x10-1 2.7 1.0x103 2.7x104
(a)(d).
U-230 (medium lung absorption) .................... 4.0x101 1.1x103 4.0x10-3 1.1x10-1 1.0x103 2.7x104
(a)(e).
U-230 (slow lung absorption) .................... 3.0x101 8.1x102 3.0x10-3 8.1x10-2 1.0x103 2.7x104
(a)(f).
U-232 (fast lung absorption) (d). .................... 4.0x101 1.1x103 1.0x10-2 2.7x10-1 8.3x10-1 2.2x101
U-232 (medium lung absorption) .................... 4.0x101 1.1x103 7.0x10-3 1.9x10-1 8.3x10-1 2.2x101
(e).
U-232 (slow lung absorption) (f). .................... 1.0x101 2.7x102 1.0x10-3 2.7x10-2 8.3x10-1 2.2x101
U-233 (fast lung absorption) (d). .................... 4.0x101 1.1x103 9.0x10-2 2.4 3.6x10-4 9.7x10-3
U-233 (medium lung absorption) .................... 4.0x101 1.1x103 2.0x10-2 5.4x10-1 3.6x10-4 9.7x10-3
(e).
U-233 (slow lung absorption) (f). .................... 4.0x101 1.1x103 6.0x10-3 1.6x10-1 3.6x10-4 9.7x10-3
U-234 (fast lung absorption) (d). .................... 4.0x101 1.1x103 9.0x10-2 2.4 2.3x10-4 6.2x10-3
U-234 (medium lung absorption) .................... 4.0x101 1.1x103 2.0x10-2 5.4x10-1 2.3x10-4 6.2x10-3
(e).
U-234 (slow lung absorption) (f). .................... 4.0x101 1.1x103 6.0x10-3 1.6x10-1 2.3x10-4 6.2x10-3
U-235 (all lung absorption types) .................... Unlimited Unlimited Unlimited Unlimited 8.0x10-8 2.2x10-6
(a),(d),(e),(f).
U-236 (fast lung absorption) (d). .................... Unlimited Unlimited Unlimited Unlimited 2.4x10-6 6.5x10-5
U-236 (medium lung absorption) .................... 4.0x101 1.1x103 2.0x10-2 5.4x10-1 2.4x10-6 6.5x10-5
(e).
U-236 (slow lung absorption) (f). .................... 4.0x101 1.1x103 6.0x10-3 1.6x10-1 2.4x10-6 6.5x10-5
U-238 (all lung absorption types) .................... Unlimited Unlimited Unlimited Unlimited 1.2x10-8 3.4x10-7
(d),(e),(f).
U (nat).......................... .................... Unlimited Unlimited Unlimited Unlimited 2.6x10-8 7.1x10-7
U (enriched to 20% or less)(g)... .................... Unlimited Unlimited Unlimited Unlimited See Table A-4 See Table A-4
U (dep).......................... .................... Unlimited Unlimited Unlimited Unlimited See Table A-4 See Table A-4
V-48............................. Vanadium (23)....... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 6.3x103 1.7x105
V-49............................. .................... 4.0x101 1.1x103 4.0x101 1.1x103 3.0x102 8.1x103
W-178 (a)........................ Tungsten (74)....... 9.0 2.4x102 5.0 1.4x102 1.3x103 3.4x104
W-181............................ .................... 3.0x101 8.1x102 3.0x101 8.1x102 2.2x102 6.0x103
W-185............................ .................... 4.0x101 1.1x103 8.0x10-1 2.2x101 3.5x102 9.4x103
W-187............................ .................... 2.0 5.4x101 6.0x10-1 1.6x101 2.6x104 7.0x105
W-188 (a)........................ .................... 4.0x10-1 1.1x101 3.0x10-1 8.1 3.7x102 1.0x104
Xe-122 (a)....................... Xenon (54).......... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 4.8x104 1.3x106
Xe-123........................... .................... 2.0 5.4x101 7.0x10-1 1.9x101 4.4x105 1.2x107
Xe-127........................... .................... 4.0 1.1x102 2.0 5.4x101 1.0x103 2.8x104
Xe-131m.......................... .................... 4.0x101 1.1x103 4.0x101 1.1x103 3.1x103 8.4x104
Xe-133........................... .................... 2.0x101 5.4x102 1.0x101 2.7x102 6.9x103 1.9x105
Xe-135........................... .................... 3.0 8.1x101 2.0 5.4x101 9.5x104 2.6x106
Y-87 (a)......................... Yttrium (39)........ 1.0 2.7x101 1.0 2.7x101 1.7x104 4.5x105
[[Page 3807]]
Y-88............................. .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 5.2x102 1.4x104
Y-90............................. .................... 3.0x10-1 8.1 3.0x10-1 8.1 2.0x104 5.4x105
Y-91............................. .................... 6.0x10-1 1.6x101 6.0x10-1 1.6x101 9.1x102 2.5x104
Y-91m............................ .................... 2.0 5.4x101 2.0 5.4x101 1.5x106 4.2x107
Y-92............................. .................... 2.0x10-1 5.4 2.0x10-1 5.4 3.6x105 9.6x106
Y-93............................. .................... 3.0x10-1 8.1 3.0x10-1 8.1 1.2x105 3.3x106
Yb-169........................... Ytterbium (70)...... 4.0 1.1x102 1.0 2.7x101 8.9x102 2.4x104
Yb-175........................... .................... 3.0x101 8.1x102 9.0x10-1 2.4x101 6.6x103 1.8x105
Zn-65............................ Zinc (30)........... 2.0 5.4x101 2.0 5.4x101 3.0x102 8.2x103
Zn-69............................ .................... 3.0 8.1x101 6.0x10-1 1.6x101 1.8x106 4.9x107
Zn-69m (a)....................... .................... 3.0 8.1x101 6.0x10-1 1.6x101 1.2x105 3.3x106
Zr-88............................ Zirconium (40)...... 3.0 8.1x101 3.0 8.1x101 6.6x102 1.8x104
Zr-93............................ .................... Unlimited Unlimited Unlimited Unlimited 9.3x10-5 2.5x10-3
Zr-95 (a)........................ .................... 2.0 5.4x101 8.0x10-1 2.2x101 7.9x102 2.1x104
Zr-97 (a)........................ .................... 4.0x10-1 1.1x101 4.0x10-1 1.1x101 7.1x104 1.9x106
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
a A1 and/or A2 values include contributions from daughter nuclides with half-lives less than 10 days.
b [Reserved]
c The quantity may be determined from a measurement of the rate of decay or a measurement of the radiation level at a prescribed distance from the source.
d These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of transport.
e These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and hexavalent compounds in both normal and accident conditions of transport.
f These values apply to all compounds of uranium other than those specified in notes (d) and (e) of this table.
g These values apply to unirradiated uranium only.
h A1 = 0.1 TBq (2.7 Ci) and A2 = 0.001 TBq (0.027 Ci) for Cf-252 for domestic use.
i A2 = 0.74 TBq (20 Ci) for Mo-99 for domestic use.
Table A-2.--Exempt Material Activity Concentrations and Exempt Consignment Activity Limits for Radionuclides
--------------------------------------------------------------------------------------------------------------------------------------------------------
Activity concentration Activity concentration Activity limit for Activity limit for
Symbol of radionuclide Element and atomic for exempt material for exempt material exempt consignment exempt consignment
number (Bq/g) (Ci/g) (Bq) (Ci)
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ac-225............................ Actinium (89)........ 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Ac-227............................ ..................... 1.0x10-1 2.7x10-12 1.0x103 2.7x10-8
Ac-228............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Ag-105............................ Silver (47).......... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Ag-108m (b)....................... ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Ag-110m........................... ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Ag-111............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Al-26............................. Aluminum (13)........ 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Am-241............................ Americium (95)....... 1.0 2.7x10-11 1.0x104 2.7x10-7
Am-242m (b)....................... ..................... 1.0 2.7x10-11 1.0x104 2.7x10-7
Am-243 (b)........................ ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Ar-37............................. Argon (18)........... 1.0x106 2.7x10-5 1.0x108 2.7x10-3
Ar-39............................. ..................... 1.0x107 2.7x10-4 1.0x104 2.7x10-7
Ar-41............................. ..................... 1.0x102 2.7x10-9 1.0x109 2.7x10-2
As-72............................. Arsenic (33)......... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
As-73............................. ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
As-74............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
As-76............................. ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
As-77............................. ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
At-211............................ Astatine (85)........ 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Au-193............................ Gold (79)............ 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Au-194............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Au-195............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Au-198............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Au-199............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Ba-131............................ Barium (56).......... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Ba-133............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Ba-133m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Ba-140 (b)........................ ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Be-7.............................. Beryllium (4)........ 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Be-10............................. ..................... 1.0x104 2.7x10-7 1.0x106 2.7x10-5
Bi-205............................ Bismuth (83)......... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Bi-206............................ ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Bi-207............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
[[Page 3808]]
Bi-210............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Bi-210m........................... ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Bi-212 (b)........................ ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Bk-247............................ Berkelium (97)....... 1.0 2.7x10-11 1.0x104 2.7x10-7
Bk-249............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Br-76............................. Bromine (35)......... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Br-77............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Br-82............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
C-11.............................. Carbon (6)........... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
C-14.............................. ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Ca-41............................. Calcium (20)......... 1.0x105 2.7x10-6 1.0x107 2.7x10-4
Ca-45............................. ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Ca-47............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Cd-109............................ Cadmium (48)......... 1.0x104 2.7x10-7 1.0x106 2.7x10-5
Cd-113m........................... ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Cd-115............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Cd-115m........................... ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Ce-139............................ Cerium (58).......... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Ce-141............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Ce-143............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Ce-144 (b)........................ ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Cf-248............................ Californium (98)..... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Cf-249............................ ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Cf-250............................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Cf-251............................ ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Cf-252............................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Cf-253............................ ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Cf-254............................ ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Cl-36............................. Chlorine (17)........ 1.0x104 2.7x10-7 1.0x106 2.7x10-5
Cl-38............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Cm-240............................ Curium (96).......... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Cm-241............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Cm-242............................ ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Cm-243............................ ..................... 1.0 2.7x10-11 1.0x104 2.7x10-7
Cm-244............................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Cm-245............................ ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Cm-246............................ ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Cm-247............................ ..................... 1.0 2.7x10-11 1.0x104 2.7x10-7
Cm-248............................ ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Co-55............................. Cobalt (27).......... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Co-56............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Co-57............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Co-58............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Co-58m............................ ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Co-60............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Cr-51............................. Chromium (24)........ 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Cs-129............................ Cesium (55).......... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Cs-131............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Cs-132............................ ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Cs-134............................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Cs-134m........................... ..................... 1.0x103 2.7x10-8 1.0x105 2.7x10-6
Cs-135............................ ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Cs-136............................ ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Cs-137 (b)........................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Cu-64............................. Copper (29).......... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Cu-67............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Dy-159............................ Dysprosium (66)...... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Dy-165............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Dy-166............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Er-169............................ Erbium (68).......... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Er-171............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Eu-147............................ Europium (63)........ 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Eu-148............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Eu-149............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Eu-150 (short lived).............. ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Eu-150 (long lived)............... ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Eu-152............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
[[Page 3809]]
Eu-152m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Eu-154............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Eu-155............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Eu-156............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
F-18.............................. Fluorine (9)......... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Fe-52............................. Iron (26)............ 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Fe-55............................. ..................... 1.0x104 2.7x10-7 1.0x106 2.7x10-5
Fe-59............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Fe-60............................. ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Ga-67............................. Gallium (31)......... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Ga-68............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Ga-72............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Gd-146............................ Gadolinium (64)...... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Gd-148............................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Gd-153............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Gd-159............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Ge-68............................. Germanium (32)....... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Ge-71............................. ..................... 1.0x104 2.7x10-7 1.0x108 2.7x10-3
Ge-77............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Hf-172............................ Hafnium (72)......... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Hf-175............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Hf-181............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Hf-182............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Hg-194............................ Mercury (80)......... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Hg-195m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Hg-197............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Hg-197m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Hg-203............................ ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Ho-166............................ Holmium (67)......... 1.0x103 2.7x10-8 1.0x105 2.7x10-6
Ho-166m........................... ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
I-123............................. Iodine (53).......... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
I-124............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
I-125............................. ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
I-126............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
I-129............................. ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
I-131............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
I-132............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
I-133............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
I-134............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
I-135............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
In-111............................ Indium (49).......... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
In-113m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
In-114m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
In-115m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Ir-189............................ Iridium (77)......... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Ir-190............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Ir-192............................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Ir-194............................ ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
K-40.............................. Potassium (19)....... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
K-42.............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
K-43.............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Kr-81............................. Krypton (36)......... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Kr-85............................. ..................... 1.0x105 2.7x10-6 1.0x104 2.7x10-7
Kr-85m............................ ..................... 1.0x103 2.7x10-8 1.0x1010 2.7x10-1
Kr-87............................. ..................... 1.0x102 2.7x10-9 1.0x109 2.7x10-2
La-137............................ Lanthanum (57)....... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
La-140............................ ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Lu-172............................ Lutetium (71)........ 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Lu-173............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Lu-174............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Lu-174m........................... ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Lu-177............................ ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Mg-28............................. Magnesium (12)....... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Mn-52............................. Manganese (25)....... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Mn-53............................. ..................... 1.0x104 2.7x10-7 1.0x109 2.7x10-2
Mn-54............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Mn-56............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
[[Page 3810]]
Mo-93............................. Molybdenum (42)...... 1.0x103 2.7x10-8 1.0x108 2.7x10-3
Mo-99............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
N-13.............................. Nitrogen (7)......... 1.0x102 2.7x10-9 1.0x109 2.7x10-2
Na-22............................. Sodium (11).......... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Na-24............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Nb-93m............................ Niobium (41)......... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Nb-94............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Nb-95............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Nb-97............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Nd-147............................ Neodymium (60)....... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Nd-149............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Ni-59............................. Nickel (28).......... 1.0x104 2.7x10-7 1.0x108 2.7x10-3
Ni-63............................. ..................... 1.0x105 2.7x10-6 1.0x108 2.7x10-3
Ni-65............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Np-235............................ Neptunium (93)....... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Np-236 (short-lived).............. ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Np-236 (long-lived)............... ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Np-237 (b)........................ ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Np-239............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Os-185............................ Osmium (76).......... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Os-191............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Os-191m........................... ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Os-193............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Os-194............................ ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
P-32.............................. Phosphorus (15)...... 1.0x103 2.7x10-8 1.0x105 2.7x10-6
P-33.............................. ..................... 1.0x105 2.7x10-6 1.0x108 2.7x10-3
Pa-230............................ Protactinium (91).... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Pa-231............................ ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Pa-233............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Pb-201............................ Lead (82)............ 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Pb-202............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Pb-203............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Pb-205............................ ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Pb-210 (b)........................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Pb-212 (b)........................ ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Pd-103............................ Palladium (46)....... 1.0x103 2.7x10-8 1.0x108 2.7x10-3
Pd-107............................ ..................... 1.0x105 2.7x10-6 1.0x108 2.7x10-3
Pd-109............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Pm-143............................ Promethium (61)...... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Pm-144............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Pm-145............................ ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Pm-147............................ ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Pm-148m........................... ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Pm-149............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Pm-151............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Po-210............................ Polonium (84)........ 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Pr-142............................ Praseodymium (59).... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Pr-143............................ ..................... 1.0x104 2.7x10-7 1.0x106 2.7x10-5
Pt-188............................ Platinum (78)........ 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Pt-191............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Pt-193............................ ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Pt-193m........................... ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Pt-195m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Pt-197............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Pt-197m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Pu-236............................ Plutonium (94)....... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Pu-237............................ ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Pu-238............................ ..................... 1.0 2.7x10-11 1.0x104 2.7x10-7
Pu-239............................ ..................... 1.0 2.7x10-11 1.0x104 2.7x10-7
Pu-240............................ ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Pu-241............................ ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Pu-242............................ ..................... 1.0 2.7x10-11 1.0x104 2.7x10-7
Pu-244............................ ..................... 1.0 2.7x10-11 1.0x104 2.7x10-7
Ra-223 (b)........................ Radium (88).......... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Ra-224 (b)........................ ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Ra-225............................ ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Ra-226 (b)........................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
[[Page 3811]]
Ra-228 (b)........................ ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Rb-81............................. Rubidium (37)........ 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Rb-83............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Rb-84............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Rb-86............................. ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Rb-87............................. ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Rb(nat)........................... ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Re-184............................ Rhenium (75)......... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Re-184m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Re-186............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Re-187............................ ..................... 1.0x106 2.7x10-5 1.0x109 2.7x10-2
Re-188............................ ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Re-189............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Re(nat)........................... ..................... 1.0x106 2.7x10-5 1.0x109 2.7x10-2
Rh-99............................. Rhodium (45)......... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Rh-101............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Rh-102............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Rh-102m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Rh-103m........................... ..................... 1.0x104 2.7x10-7 1.0x108 2.7x10-3
Rh-105............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Rn-222 (b)........................ Radon (86)........... 1.0x101 2.7x10-10 1.0x108 2.7x10-3
Ru-97............................. Ruthenium (44)....... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Ru-103............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Ru-105............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Ru-106 (b)........................ ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
S-35.............................. Sulphur (16)......... 1.0x105 2.7x10-6 1.0x108 2.7x10-3
Sb-122............................ Antimony (51)........ 1.0x102 2.7x10-9 1.0x104 2.7x10-7
Sb-124............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Sb-125............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Sb-126............................ ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Sc-44............................. Scandium (21)........ 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Sc-46............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Sc-47............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Sc-48............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Se-75............................. Selenium (34)........ 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Se-79............................. ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Si-31............................. Silicon (14)......... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Si-32............................. ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Sm-145............................ Samarium (62)........ 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Sm-147............................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Sm-151............................ ..................... 1.0x104 2.7x10-7 1.0x108 2.7x10-3
Sm-153............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Sn-113............................ Tin (50)............. 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Sn-117m........................... ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Sn-119m........................... ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Sn-121m........................... ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Sn-123............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Sn-125............................ ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Sn-126............................ ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Sr-82............................. Strontium (38)....... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Sr-85............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Sr-85m............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Sr-87m............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Sr-89............................. ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Sr-90 (b)......................... ..................... 1.0x102 2.7x10-9 1.0x104 2.7x10-7
Sr-91............................. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Sr-92............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
T(H-3)............................ Tritium (1).......... 1.0x106 2.7x10-5 1.0x109 2.7x10-2
Ta-178 (long-lived)............... Tantalum (73)........ 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Ta-179............................ ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Ta-182............................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Tb-157............................ Terbium (65)......... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Tb-158............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Tb-160............................ ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Tc-95m............................ Technetium (43)...... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Tc-96............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Tc-96m............................ ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
[[Page 3812]]
Tc-97............................. ..................... 1.0x103 2.7x10-8 1.0x108 2.7x10-3
Tc-97m............................ ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Tc-98............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Tc-99............................. ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
Tc-99m............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Te-121............................ Tellurium (52)....... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Te-121m........................... ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Te-123m........................... ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Te-125m........................... ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Te-127............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Te-127m........................... ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Te-129............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Te-129m........................... ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Te-131m........................... ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Te-132............................ ..................... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Th-227............................ Thorium (90)......... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Th-228 (b)........................ ..................... 1.0 2.7x10-11 1.0x104 2.7x10-7
Th-229 (b)........................ ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Th-230............................ ..................... 1.0 2.7x10-11 1.0x104 2.7x10-7
Th-231............................ ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Th-232............................ ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
Th-234 (b)........................ ..................... 1.0x103 2.7x10-8 1.0x105 2.7x10-6
Th (nat) (b)...................... ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
Ti-44............................. Titanium (22)........ 1.0x101 2.7x10-10 1.0x105 2.7x10-6
Tl-200............................ Thallium (81)........ 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Tl-201............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Tl-202............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Tl-204............................ ..................... 1.0x104 2.7x10-7 1.0x104 2.7x10-7
Tm-167............................ Thulium (69)......... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Tm-170............................ ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Tm-171............................ ..................... 1.0x104 2.7x10-7 1.0x108 2.7x10-3
U-230 (fast lung absorption) Uranium (92)......... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
(b),(d).
U-230 (medium lung absorption) (e) ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
U-230 (slow lung absorption) (f).. ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
U-232 (fast lung absorption) ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
(b),(d).
U-232 (medium lung absorption) (e) ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
U-232 (slow lung absorption) (f).. ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
U-233 (fast lung absorption) (d).. ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
U-233 (medium lung absorption) (e) ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
U-233 (slow lung absorption) (f).. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
U-234 (fast lung absorption) (d).. ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
U-234 (medium lung absorption) (e) ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
U-234 (slow lung absorption) (f).. ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
U-235 (all lung absorption types) ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
(b),(d),(e),(f).
U-236 (fast lung absorption) (d).. ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
U-236 (medium lung absorption) (e) ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
U-236 (slow lung absorption) (f).. ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
U-238 (all lung absorption types) ..................... 1.0x101 2.7x10-10 1.0x104 2.7x10-7
(b),(d),(e),(f).
U (nat) (b)....................... ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
[[Page 3813]]
U (enriched to 20% or less)(g).... ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
U (dep)........................... ..................... 1.0 2.7x10-11 1.0x103 2.7x10-8
V-48.............................. Vanadium (23)........ 1.0x101 2.7x10-10 1.0x105 2.7x10-6
V-49.............................. ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
W-178............................. Tungsten (74)........ 1.0x101 2.7x10-10 1.0x106 2.7x10-5
W-181............................. ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
W-185............................. ..................... 1.0x104 2.7x10-7 1.0x107 2.7x10-4
W-187............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
W-188............................. ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Xe-122............................ Xenon (54)........... 1.0x102 2.7x10-9 1.0x109 2.7x10-2
Xe-123............................ ..................... 1.0x102 2.7x10-9 1.0x109 2.7x10-2
Xe-127............................ ..................... 1.0x103 2.7x10-8 1.0x105 2.7x10-6
Xe-131m........................... ..................... 1.0x104 2.7x10-7 1.0x104 2.7x10-7
Xe-133............................ ..................... 1.0x103 2.7x10-8 1.0x104 2.7x10-7
Xe-135............................ ..................... 1.0x103 2.7x10-8 1.0x1010 2.7x10-1
Y-87.............................. Yttrium (39)......... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Y-88.............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Y-90.............................. ..................... 1.0x103 2.7x10-8 1.0x105 2.7x10-6
Y-91.............................. ..................... 1.0x103 2.7x10-8 1.0x106 2.7x10-5
Y-91m............................. ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Y-92.............................. ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Y-93.............................. ..................... 1.0x102 2.7x10-9 1.0x105 2.7x10-6
Yb-169............................ Ytterbium (70)....... 1.0x102 2.7x10-9 1.0x107 2.7x10-4
Yb-175............................ ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Zn-65............................. Zinc (30)............ 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Zn-69............................. ..................... 1.0x104 2.7x10-7 1.0x106 2.7x10-5
Zn-69m............................ ..................... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Zr-88............................. Zirconium (40)....... 1.0x102 2.7x10-9 1.0x106 2.7x10-5
Zr-93 (b)......................... ..................... 1.0x103 2.7x10-8 1.0x107 2.7x10-4
Zr-95............................. ..................... 1.0x101 2.7x10-10 1.0x106 2.7x10-5
Zr-97 (b)......................... ..................... 1.0x101 2.7x10-10 1.0x105 2.7x10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
a [Reserved]
b Parent nuclides and their progeny included in secular equilibrium are listed in the following:
Sr-90 Y-90
Zr-93 Nb-93m
Zr-97 Nb-97
Ru-106 Rh-106
Cs-137 Ba-137m
Ce-134 La-134
Ce-144 Pr-144
Ba-140 La-140
Bi-212 Tl-208 (0.36), Po-212 (0.64)
Pb-210 Bi-210, Po-210
Pb-212 Bi-212, Tl-208 (0.36), Po-212 (0.64)
Rn-220 Po-216
Rn-222 Po-218, Pb-214, Bi-214, Po-214
Ra-223 Rn-219, Po-215, Pb-211, Bi-211, Tl-207
Ra-224 Rn-220, Po-216, Pb-212, Bi-212, Tl-208(0.36), Po-212 (0.64)
Ra-226 Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210
Ra-228 Ac-228
Th-226 Ra-222, Rn-218, Po-214
Th-228 Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
Th-229 Ra-225, Ac-225, Fr-221, At-217, Bi-213, Po-213, Pb-209
Th-nat Ra-228, Ac-228, Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
Th-234 Pa-234m
U-230 Th-226, Ra-222, Rn-218, Po-214
U-232 Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
U-235 Th-231
U-238 Th-234, Pa-234m
U-nat Th-234, Pa-234m, U-234, Th-230, Ra-226, Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210
U-240 Np-240m
Np-237 Pa-233
Am-242m Am-242
Am-243 Np-239
c [Reserved]
d These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of
transport.
e These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and hexavalent compounds in both normal and accident
conditions of transport.
f These values apply to all compounds of uranium other than those specified in notes (d) and (e) of this table.
[[Page 3814]]
g These values apply to unirradiated uranium only.
Table A-3.--General Values for A1 and A2
--------------------------------------------------------------------------------------------------------------------------------------------------------
A1 A2 Activity Activity Activity Activity
-------------------------------------------------------------- concentration concentration limits for limits for
Contents for exempt for exempt exempt exempt
(TBq) (Ci) (TBq) (Ci) material (Bq/ material (Ci/ consignments consignments
g) g) (Bq) (Ci)
--------------------------------------------------------------------------------------------------------------------------------------------------------
Only beta or gamma emitting 1 x 10 -1 2.7 x 100 2 x 10 -2 5.4 x 10 -1 1 x 10 -1 2.7 x 10 -10 1 x 10 -4 2.7 x 10 -7
radionuclides are known to
be present.
Only alpha emitting 2 x 10 -1 5.4 x 10 0 9 x 10 -5 2.4 x 10 -3 1 x 10 -1 2.7 x 10 -12 1 x 10 3 2.7 x 10 -8
radionuclides are known to
be present.
No relevant data are 1 x 10 -3 2.7 x 10 -2 9 x 10 -5 2.4 x 10 -3 1 x 10 -1 2.7 x 10 -12 1 x 10 3 2.7 x 10 -8
available.
--------------------------------------------------------------------------------------------------------------------------------------------------------
Table A-4.--Activity-mass Relationships for Uranium
------------------------------------------------------------------------
Specific Activity
Uranium Enrichment \1\ wt % U-235 --------------------------------------
present TBq/g Ci/g
------------------------------------------------------------------------
0.45............................. 1.8 x 10 -8 5.0 x 10 -7
0.72............................. 2.6 x 10 -8 7.1 x 10 -7
1................................ 2.8 x 10 -8 7.6 x 10 -7
1.5.............................. 3.7 x 10 -8 1.0 x 10 -6
5................................ 1.0 x 10 -7 2.7 x 10 -6
10............................... 1.8 x 10 -7 4.8 x 10 -6
20............................... 3.7 x 10 -7 1.0 x 10 -5
35............................... 7.4 x 10 -7 2.0 x 10 -5
50............................... 9.3 x 10 -7 2.5 x 10 -5
90............................... 2.2 x 10 -6 2.8 x 10 -5
93............................... 2.6 x 10 -6 7.0 x 10 -5
95............................... 3.4 x 10 -6 9.1 x 10 -5
------------------------------------------------------------------------
\1\ The figures for uranium include representative values for the
activity of the uranium-234 that is concentrated during the enrichment
process.
Dated in Rockville, Maryland, this 29th day of December, 2003.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 04-35 Filed 1-23-04; 8:45 am]
BILLING CODE 7590-01-P