[Federal Register Volume 65, Number 137 (Monday, July 17, 2000)]
[Proposed Rules]
[Pages 44360-44397]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-18029]
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Part IV
Nuclear Regulatory Commission
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10 CFR Part 71
Major Revision to 10 CFR Part 71: Compatibility With ST-1--The IAEA
Transportation Safety Standards--And Other Transportation Safety
Issues, Issues Paper, and Notice of Public Meetings; Proposed Rule
Federal Register / Vol. 65, No. 137 / Monday, July 17, 2000 /
Proposed Rules
[[Page 44360]]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 71
Major Revision to 10 CFR Part 71: Compatibility With ST-1--The
IAEA Transportation Safety Standards--and Other Transportation Safety
Issues, Issues Paper, and Notice of Public Meetings
AGENCY: Nuclear Regulatory Commission.
ACTION: Request for comment on issues paper, and notice of plans for
public meetings.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is considering a
rulemaking that would revise the Commission's regulations on packaging
and transporting radioactive material to make it compatible with the
International Atomic Energy Agency (IAEA) transportation safety
standards as well as codify other requirements. The NRC is seeking
early public input on the major issues associated with such a
rulemaking. To aid in that process, the NRC is requesting comments on
the issues paper included in this notice. Specifically, the NRC is
interested in public and industry comments related to: Quantitative
information on the costs and benefits resulting from consideration of
the factors described in the issues paper, operational data on
radiation exposures (increased or reduced) that might result from
implementing the contemplated changes; whether the presented factors
are appropriate; and whether other factors should be considered,
including providing quantitative information for these factors. The
Commission believes that the stakeholders' comments will help to
quantify the potential impact of these changes and will assist the NRC,
as the proposed rule is developed, in developing a risk-informed
alternative as its preferred option. NRC also intends to conduct three
public meetings in August and September of this year to discuss those
issues and solicit public comments.
DATES: Submit comments at the public meetings, or in writing by
September 30, 2000. Comments received after this date will be
considered if it is practicable to do so, but the Commission is able to
assure consideration only for comments received on or before this date.
In addition to providing opportunity for written (and electronic)
comments, public meetings on the paper will be held as follows :
August 10, 2000 NRC Headquarters, Washington, DC, 8:30 am-5pm
September 20, 2000 Atlanta, Georgia, J.W. Marriott, 3300 Lenox Road
Northeast, Atlanta, GA 30326, 6-10 pm
September 26, 2000 Oakland, California, Oakland Federal Building,
Edward R. Roybal Auditorium and Conference Center, 1301 Clay Street,
Oakland, CA 94612, 6-10 pm
ADDRESSES: Submit comments to: Secretary, U.S. Nuclear Regulatory
Commission, Washington, D.C. 20555. Attention: Rulemaking and
Adjudications staff.
Deliver comments to 11555 Rockville Pike, Rockville, Maryland,
between 7:30 a.m. and 4:15 p.m. on Federal workdays.
You may also provide comments via the NRC's interactive rulemaking
website at http://ruleforum.llnl.gov). This site provides the
capability to upload comments as files (any format), if your web
browser supports that function. For information about the interactive
rulemaking website, contact Ms. Carol Gallagher, (301) 415-5095 (e-
mail:[email protected]).
Copies of any comments received and documents related to this
action may be examined at the NRC Public Document Room, 2120 L Street
NW (Lower Level), Washington, DC Documents created or received at the
NRC after November 1, 1999 are also available electronically at the
NRC's Public Electronic Reading Room on the Internet at http://www.nrc.gov/NRC/ADAMS/index.html. From this site, the public can gain
entry into the NRC's Agencywide Documents Access and Management System
(ADAMS), which provides text and image files of NRC's public documents.
For more information, contact the NRC Public Document Room (PDR)
Reference staff at 1-800-397-4209, 202-634-3273 or email to
[email protected].
FOR FURTHER INFORMATION CONTACT: Naiem S. Tanious, telephone: (301)
415-6103; e-mail: [email protected], Office of Nuclear Material Safety and
Safeguards, USNRC, Washington, DC 20555-0001. Specific comments on the
public meeting process should be directed to Francis X. Cameron; e-mail
[email protected], telephone: (301) 415-1642; Office of the General Counsel,
USNRC, Washington, DC 20555-0001.
SUPPLEMENTARY INFORMATION:
I. Background
By international agreement and through Commission direction, the
NRC staff is preparing an overall rulemaking effort that addresses the
need to make 10 CFR Part 71 regulations, ``Packaging and Transportation
of Radioactive Material'' compatible with the most current revision of
the IAEA Safety Standards Series No. ST-1. Part 71 is based, in
general, on the safety standards developed by the IAEA. The IAEA has
been revising its transportation standards on approximately a 10-year
cycle, with the last edition, ST-1, published in December 1996.
Further, several additional issues related to other changes to 10 CFR
Part 71 are being considered by NRC. These issues include the fissile
material exemptions, general license provisions, and the current
requirements for double containment of plutonium.
The NRC is supplementing its standard rulemaking process by
conducting enhanced public participatory activities including
facilitated public meetings before the start of any formal rulemaking
process to solicit early and active public input on major issues with
revision of 10 CFR Part 71. The NRC will also utilize its rulemaking
website to make the issues paper available to the public and to solicit
public comments. To facilitate discussion and public comments, the NRC
has prepared an issues paper that describes 18 rulemaking issues (IAEA
and Non-IAEA-related) to be addressed in revisions to Part 71. These
issues are described in more detail in Section III of this notice.
II. Request for Written and Electronic Comments and Plans for
Public Meetings
The NRC is soliciting comments on the items presented in the issues
paper in Section III of this notice. Comments may be submitted either
in writing or electronically as indicated under the ADDRESSES heading.
In addition to providing an opportunity for written comments, the NRC
is holding facilitated public meetings at three different geographical
locations on the issues discussed in Section III (see the DATES heading
of this notice for the dates and locations of these meetings). In
addition to the NRC staff, a representative from the Department of
Transportation (DOT) will be available to answer any questions related
to their concurrent rulemaking efforts.
In addition to inviting public comments on the issues presented in
Section III, NRC is soliciting specific comments related to: (1)
Quantitative information on the costs and benefits resulting from
consideration of the factors described in the issues paper, (2)
operational data on radiation exposures (increased or reduced) that
might result from implementing the Part 71 changes; (3) whether the
presented factors are appropriate; and (4) whether other factors should
be considered, including
[[Page 44361]]
providing quantitative information for these factors. The Commission
believes that the stakeholders' comments will help to quantify the
potential impact of these changes and will assist the NRC, as the
proposed rule is developed, in developing a risk-informed alternative
as its preferred option.
Based on the comments received in written or electronic form, and
at the public meetings, the Commission will then be in a better
position to evaluate options for Part 71 rulemaking, to decide on the
preferred options, and to proceed with development of a proposed rule.
III. Issues Paper on Major Revision to 10 CFR Part 71:
Compatibility with ST-1--the IAEA Transportation Safety Standards--
and Other Transportation Safety Issues
A. Introduction
1. Background
In 1969, the International Atomic Energy Agency (IAEA), recognizing
that its international regulations for the safe transportation of
radioactive material should be revised from time to time because of
scientific and technical advances, and accumulated experience, invited
Member States (the U.S. is a Member State) to submit comments and
suggest changes to its standards. As a result of this initiative, the
IAEA issued revised standards in 1973 (Regulations for the Safe
Transport of Radioactive Material, 1973 Edition, Safety Series (SS) No.
6). The IAEA has periodically reviewed its transportation regulations
(about every ten years) to ensure that the regulations are kept
current. Thus, a review of IAEA regulations was initiated in 1979 and
resulted in the publication of revised regulations in 1985 (1985
Edition, SS No. 6).
The U.S. Nuclear Regulatory Commission (NRC) also periodically
revises its regulations to make them compatible, to the extent
appropriate, with those of the IAEA. On August 5, 1983 (48 FR 35600),
the NRC published, in the Federal Register, a final revision to 10 CFR
Part 71, ``Packaging and Transportation of Radioactive Material.'' That
revision, in combination with a parallel revision of the hazardous
materials transportation regulations of the U.S. Department of
Transportation (DOT), brought U.S. domestic transport regulations into
general accord with the 1973 edition of SS No. 6. The next IAEA
revision of the transportation standards in SS No. 6 resulted in a
revision to Part 71 that was published on September 28, 1995 (60 FR
50248), to make Part 71 compatible with the 1985 edition of SS No. 6.
DOT published its corresponding revision to Title 49 of the Code of
Federal Regulations on the same date.
In each case, the NRC coordinated its Part 71 revisions with the
DOT. DOT is the U.S. Competent Authority for transportation of
hazardous materials. ``Radioactive Materials Regulations'' is a subset
of ``Hazardous Materials Regulations'' in Title 49. The DOT and the NRC
co-regulate transport of radioactive material in the United States and
have a Memorandum of Understanding to that effect.
The last revision to the IAEA SS No. 6 was titled Safety Standards
Series No. ST-1, referred to hereafter as ST-1, and was published in
December 1996.
2. Scope of Part 71 Rulemaking
The Commission has directed the NRC staff to begin rulemaking to
revise Part 71 for compatibility with ST-1. The NRC staff compared ST-1
to SS No. 6 to identify changes made in ST-1, and then identified
affected sections of Part 71. Based on this comparison, the NRC staff
identified eleven Part 71 IAEA-compatibility issues to be addressed
through the rulemaking process. These eleven issues (identified as
issues 1 through 11) are discussed in greater detail in Section B.
Seven additional issues were identified (issues 12 thru 18) for
incorporation in the rulemaking process, through NRC staff
identification and through Commission direction, and are also discussed
in further detail in Section B.
The Part 71 rulemaking and this issues paper are being 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 each
agency. Note that on December 28, 1999 (64 FR 72633), DOT published an
Advance Notice of Proposed Rule regarding adoption of ST-1 in its
regulations, and plans to proceed to develop a proposed rule for public
comments and subsequently a final rule. In order to develop a final
rule concurrent with the timing of the DOT final rule, the NRC staff
developed the following schedule: (1) the NRC staff will submit to the
Commission for approval, a proposed rule to revise Part 71 by March 1,
2001, (2) the proposed rule is expected to be published for public
comment in April 2001, (3) the NRC staff is planning to hold public
meetings during the public comment period, and (4) after the end of the
public comment period, the staff will revise the rule and submit it for
approval as a final rule by June 2002.
The NRC proposed rule will include a cost-benefit (regulatory
analysis). Contrary to the NRC's rulemaking process under the
Administrative Procedure Act, development of the IAEA ST-1 did not
directly involve the public or include a cost-benefit analysis, to our
knowledge. In contrast, NRC is bound to consider costs and benefits in
its regulatory analysis, and is prepared to differ from the ST-1
standards, at least for domestic purposes, to the extent the standards
cannot be justified from a cost-benefit perspective.
B. Issues Format
The following format is used in the presentation of the issues that
follow. Each issue is assigned a tracking number with a short title,
and includes an issue description paragraph and a listing of factors
for consideration. The factors for consideration in this document are
not meant to be a complete or final listing, but are included to help
prompt consideration and discussion of the issue. In August and
September 2000, through a series of public meetings and a summary
workshop, the public and industry will be requested to (1) comment on
and recommend additions, deletions, or modifications to the factors for
consideration; (2) propose implementation options for each issue; and
(3) provide estimated implementation cost information. Other venues for
feedback will be made available through mailings and by internet
through the NRC web site. This public feedback will then be used in
developing implementation options for Commission consideration as the
Part 71 rulemaking process proceeds. Comments received that are outside
the scope of this rulemaking may be addressed in future rulemaking if
warranted.
Factors for consideration that are common to most of the issues are
stated here, rather than repeated in each issue. These include: (1) How
should risk considerations (i.e., what can happen, how likely is it,
what are the consequences) be factored into rulemaking on applicable
issues, (2) costs (i.e., administrative, training, testing) to industry
and/or Government agencies in adopting ST-1 requirements (issues 1-11)
or the NRC-initiated changes (issues 12-18), and (3) potential problems
that may occur as a result of adopting ST-1 requirements, or problems
that may occur from partial or non-adoption of the ST-1 requirements
resulting in dual standards between domestic (10 CFR 71) and
international (ST-1) requirements. For issues 1-11, the ``factors for
consideration'' noted under each issue are generally written
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in the context of adopting the ST-1 requirements into Part 71.
In the case of the eleven IAEA-compatibility issues, portions of
the Safety Standards Series ST-1 are referenced by the corresponding
paragraph number from the original IAEA document. The full text of the
ST-1 references can be found in Appendix A of this issues paper.
Issue 1. Changing Part 71 to SI Units Only
Description
ST-1, Annex II, page 199 states: ``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 ST-1. ST-1 also requires that activity values contained in
shipping papers and displayed on package labels be expressed only in SI
units (paragraphs 543 and 549). SS No. 6, 1985 Edition, used SI units
as the primary controlling units, with subsidiary units in parentheses;
either units were permissible on labels and shipping papers.
The ST-1 requirement regarding only the use of SI units conflicts
with the NRC Metrication Policy issued on June 19, 1996 (61 FR 31169).
This policy allows a dual-unit system to be used; SI units with English
units in parentheses. 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 utilizes the dual unit scheme in accordance with the NRC Metrication
Policy.
Factors for Consideration
What changes would licensees and Certificate of Compliance
holders have to make to relevant documents if NRC revised 10 CFR Part
71 to require SI units only?
What risks and safety impacts might occur in shipments
because of possible confusion or erroneous conversion between the
currently utilized English units and SI units?
What sort of transition period would be needed to allow
for the conversion to exclusive use of SI units?
What other conforming changes would have to be made to
Title 10?
Issue 2. Radionuclide Exemption Values
Description
Exempt materials are those which are of such low potential hazard
that they may not be required to be shipped in accordance with specific
transportation regulations. In ST-1, the IAEA adopted a new approach to
specifying these materials by developing radionuclide-specific activity
concentration values for exempt materials and activity limits for
exempt consignments. These new values are found in ST-1, Tables I and
II, and Section IV. Related information is provided in paragraphs 401
through 406 of ST-1. Exempt materials are those that fall below the
listed activity concentration values. Exempt consignments are packages
or loads that have a total activity less than the listed activity
values.
The exempt materials activity concentration values range from 0.1
to 1,000,000 Bq/g, with most radionuclides in the 1 to 100 Bq/g range.
This IAEA requirement does not currently exist in Part 71. Appendix A
to Part 71--Determination of A1 and A2, does not
contain exemption values for each radionuclide because the exemption
for low-level radioactive material as contained in 10 CFR 71.10(a) is
70 Bq/g (2000 picoCuries per gram) or less.
Some materials, such as ores containing naturally occurring
radionuclides, would be brought into the scope of the regulations for
the first time; however, provisions are included in ST-1 that reduce
the potential impact on natural materials containing radionuclides at
these low levels. The provisions continue to exempt natural material
and ores containing naturally occurring radionuclides, that are not
intended to be processed for the use of these radionuclides, provided
the activity concentration of the material does not exceed 10 times the
values [ST-1 paragraph 107(e)]. Additionally, for materials that may
appear in the scope of the regulations for the first time, but which
have activity concentrations not exceeding 30 times the exempt activity
concentrations, provisions exist in ST-1 to allow them to be
transported as LSA-I materials that may be transported unpackaged (in
bulk). However, there may be unintended consequences in implementing
the ST-1 concentration values where applied to non-transportation
activities. The DOT current exempt material standard of 70 Bq/g (2000
picoCuries per gram), based on previous IAEA transportation standards,
has application by cross reference outside the domain of
transportation.
Factors for Consideration
In some cases, would shippers have to expend resources to:
(1) Identify the radionuclides in a material; (2) measure the activity
concentration of each radionuclide; and, (3) apply the method for
mixtures of radionuclides when determining the basic radionuclide
values for exempt material?
Should the exemption values apply to domestic as well as
export shipments?
If the exemption values only applied to export shipments,
would the resulting standard be practical to implement?
If DOT specifies the exemption values in its regulations
(49 CFR 173), should the NRC incorporate those same exemption values in
Part 71, or simply make reference to the exemption values in the DOT
regulations?
There may be unintended consequences to adoption of
specific exemption values as the current exemption value is used for
non-transportation related activities. To what extent and in what
manner would a change to specific exemption values affect entities
whose non-transportation activities are linked to the current exemption
value?
Issue 3. Revision of A1 and A2
Description
The A1 and A2 values specified in Part 71,
Appendix A, are basic dose-based values used in several areas of the
regulations, including determining the type of package that must be
used for transporting radioactive material. For example, the
A1 values are the maximum activity of special-form materials
allowed in a Type A package, and the A2 values are the
maximum activity of non-special-form material allowed in a Type A
package. The A1 and A2 values are also used for
several other quantitative limits including Type B-package activity
release limits, low-specific activity material specifications, and
excepted package content limits.
The ST-1 revised A1 and A2 values are
primarily based on dosimetric models that use the IAEA's Q system for
dose determination. The Q system includes consideration of a broad
range of specific exposure pathways consisting of: External photon
dose, external beta dose, inhalation dose, skin and ingestion dose
because of contamination, and dose from submersion in gaseous isotopes.
The main changes in the Q system resulted from making the dosimetric
models
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consistent with those used in International Commission on Radiation
Protection (ICRP) Publication 61. The lung model and dose conversion
factors were updated to the latest ICRP models and the radionuclide
values were recalculated. The Q system reference doses and exposure
pathways were not changed.
Factors for Consideration
Is there a practical alternative to adoption of the
A1 and A2 values?
Are there specific values that should be modified for
domestic use only? What would be the justification for doing so?
To what extent should the US partial adoption of ICRP 61
be considered for revising the A1 and A2 values?
Issue 4. Uranium Hexafluoride Package Requirements
Description
ST-1 introduces detailed requirements for uranium hexaflouride
(UF6) packages designed for more than 0.1 kg UF6.
NRC certifies Type B and fissile (i.e., enriched uranium)
UF6 packages under 10 CFR Part 71. Although most of these
issues are under DOT in 49 CFR Part 173, the new ST-1 provisions
relevant to 10 CFR Part 71 are summarized as follows (see Appendix A
for a listing of the specific ST-1 provisions):
Para 629: Packages shall be packaged and transported in accordance with
an international standard, ISO 7195, ``Packaging of Uranium
Hexafluoride (UF6) for Transport.'' ST-1 also allows [para
632(a)] for use of equivalent national standards (e.g., ANSI N14.1);
provided that approval by all countries involved in the shipment is
obtained (i.e., multilateral approval).
Para 630: ST-1 requires that packages must withstand: (a) A minimum
internal pressure test to 2.8 MPa (1.4 MPa for multilateral approval),
(b) the ``normal conditions of transport'' drop test, and (c) the
hypothetical accident condition thermal test (except that packages
containing grater than 9000 kg are exempt from this test if given
multilateral approval).
Para 631: ST-1 prohibits packages from utilizing pressure relief
devices.
Para 677(b): ST-1 includes an exception that allows UF6
packages to be evaluated for criticality without considering the in-
leakage of water into the containment system. This provision means that
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 physical contact of the cylinder valve to
any other component of the packaging after the hypothetical accident
tests, the valve remains leak-tight, and when there is a high degree of
quality control in the manufacture, maintenance, and repair of
packaging coupled with tests to demonstrate closure of each package
before each shipment.
Factors for Consideration
NRC practice has been to certify fissile UF6
packages (including the cylinder which is the containment vessel and a
protective overpack) that are shown to be leaktight when subject to the
hypothetical accident tests and to specify that the cylinder meets ANSI
N14.1 (ANSI N14.1 has the domestic pressure test requirement in 630(a),
not the regulations). For this reason, it is believed that NRC-
certified UF6 packages already comply with the above package
performance requirements (para 630 and 677(b)). However, these changes
appear to have significant ramifications for non-fissile UF6
packaging that are under the purview of DOT.
NRC practice has been to reference the ANSI N14.1 standard
in the certification, but not to reference the standard in the rule.
Although the ISO-7195-2000 standard (in draft) has been drafted taking
into account ANSI N14.1, a detailed confirmation of the compatibility
of the two standards has not been performed. NRC has representation on
the ANSI N14.1 revision panel.
Issue 5. Introduction of Criticality Safety Index (CSI)
Requirements
Description
For fissile material packages, ST-1 defines a new term,
``criticality safety index'' (CSI) (paragraph 218), that applies in
addition to the traditional package transport index (TI). In current
domestic regulations and in the previous IAEA regulations, the overall
package TI was determined based upon the more limiting of a ``TI based
upon criticality considerations'' and a ``TI based on package radiation
levels.'' Both NRC and DOT regulations define and rely on the TI to
determine appropriate safety requirements.
The CSI is determined in the same manner as the current TI ``based
upon criticality considerations,'' but it now must be displayed on
shipments of fissile material (paras 544-545) using a new ``fissile
material'' label. A package TI is still determined in the same way as
the ``TI based on package radiation levels'' and continues to be
displayed on the traditional ``radioactive material'' label.
Factors for Consideration
Under the new approach, it is believed that some shipments
of fissile material packages might be made more efficiently (equivalent
safety but more packages allowed in a single shipment), due to avoiding
the situation where separation distance requirements (radiological
safety) restrict package accumulation (criticality safety), or vice
versa.
Are any issues envisioned in the use of two TI values for
shipments?
Issue 6. Type C Packages and Low Dispersible Material
Description
IAEA has adopted the concept of a new category of package, the Type
C package (paragraphs 230, 667-670, 730, 734-737) that could withstand
severe accident conditions in air transport without loss of containment
or significant increase in external radiation levels. At the same time,
ST-1 introduced a new category of material, Low Dispersible Material
(LDM), which due to its limited radiation hazard and low dispersibility
could continue to be transported by aircraft in Type B packages. U.S.
regulations have no Type C package or LDM category, but do have
specific requirements for the air transport of plutonium. These
specific NRC requirements for the air transportation of plutonium (10
CFR 71.64 and 71.74) continue to apply, and will not be addressed in
this rulemaking.
The Type C requirements apply to packages destined for air
transport that contain a total activity above the following thresholds:
for special form material--3,000 A1 or 100,000
A2, whichever is lesser, and for all other radioactive
material--3,000 A2. 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 m/s impact
test is required instead of the 9 m-drop test. A 60-minute fire test is
required instead of the 30-minute Type B requirement. Other additional
tests, such as a puncture/tearing test are also imposed. These tests
are more stringent and are expected to result in package designs that
will survive more severe aircraft accidents than Type B package
designs.
The LDM specification was added to account for materials (package
contents)
[[Page 44364]]
that have inherently limited dispersibility, solubility, and external
radiation levels. The test requirements for LDM 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. 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.
Factors for Consideration
What would be the impact on air transport of currently
certified Type B packages if the activity content is limited to the
activity content thresholds specified above?
What tests and analyses would be a practical method for
demonstrating compliance with the type C package standards?
Issue 7. Deep Immersion Test
Description
The IAEA performance requirement for deep water immersion contained
in ST-1 (para. 657 and 730) is an expansion of the requirement
contained in SS No. 6. Previously, the deep immersion test was only
required for packages of irradiated fuel exceeding 37 PBq (1,000,000
Ci). The ST-1 requirements apply to all Type B(U) and B(M) packages
containing more than 105A2 and to Type C
packages.
10 CFR 71.61 requires a deep immersion test for packages of
irradiated nuclear fuel with activity greater than 106 Ci.
Currently, 10 CFR 71.61 is more conservative than SS No. 6, with
respect to irradiated fuel package design requirements because it
requires that a package for irradiated nuclear fuel must be designed
such that its undamaged containment system can withstand an external
water pressure of 2 MPa for a period of not less than one hour without
collapse, buckling, or in leakage of water. The conservatism lies in
the test criteria of no collapse, buckling, or in leakage as compared
to the ``no rupture'' criteria found in SS No. 6 and ST-1.
To be consistent with ST-1, the NRC would have to revise 10 CFR
Part 71.61 to apply to all packages with activity greater than
105A2 and adopt the ST-1 test criteria.
Factors for Consideration
How should the differences in the acceptance standards be
addressed?
What would be the impact on availability of packages and
shipping costs if all packages with an activity greater than
105A2 are required to pass the immersion test
requirements?
Would US origin package designs have to be specially
reviewed and certified before shippers could export them in accordance
with international regulations if ST-1 requirements were not adopted?
Issue 8. Grandfathering Previously Approved Packages
Description
Historically, 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. Package designs and packagings compliant
with the existing regulations do not become ``unsafe'' when the
regulations are amended (unless a significant safety issue is corrected
in the revision).
Grandfathering typically includes provisions that allow for: (1)
Continued use of existing package designs and packagings already
fabricated, although some additional requirements may be imposed, (2)
completion of packagings 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.
A major change in ST-1 is that ``grandfathering'' should be limited
to only those package designs that have been certified under the last
two major revisions of the regulations. Packages approved under an
earlier revision would either be removed from service or be required to
be re-certified under the revised regulations that result from this
rulemaking.
As revised in 1996, IAEA regulations in ST-1 only recognize the
``grandfathering'' of package designs certified under the 1973 and 1985
editions of IAEA regulations (SS No. 6). Package designs approved under
the 1967 edition of SS No. 6 would be required to be re-certified,
removed from service, or shipped via exemption (i.e., special
arrangement). If this approach to ``grandfathering'' is adopted in DOT
and NRC regulations, package designs approved to earlier versions of
DOT and NRC regulations (i.e., those based on 1967 IAEA regulations)
would be required to be re-certified, removed from service, or shipped
via exemption.
Factors for Consideration
Should the ``grandfathering `` of previously approved
packages be limited to those approved under the last two major
revisions of the regulations? If not, on what basis should the
``grandfathering `` of previously approved packages be allowed?
How long should ``grandfathered'' packages be allowed to
be fabricated or used?
What type and magnitude of package design changes should
be allowed for ``grandfathered'' packages, before re-certification to
the current set of regulations is required?
IAEA has initiated a process to review and update ST-1 on
a two-year frequency and does this new process raise any issues on the
grandfathering limitations to the last two major revisions?
Issue 9. Changes to Various Definitions
Description
The NRC is contemplating changes to various definitions in Part 71
to provide internal consistency and improve correlation with ST-1. 10
CFR 71.4 includes defined terms used throughout Part 71. These terms
require clear definition so that they can be used to accurately
communicate requirements to licensees. The NRC would add the following
definitions from ST-1: (1) Confinement system (paragraph 209), (2)
Criticality safety index (paragraph 218; reference issue 5), (3) Low
dispersible radioactive material (paragraph 225; reference issue 6),
and (4) Quality assurance (paragraph 232). Additionally, the NRC would
propose to revise the definition of ``package'' in 10 CFR 71.4 to be
consistent with ST-1. For reference, the ST-1 definitions are contained
in Appendix A and provided below.
Para. 209. ``Confinement System shall mean the assembly of fissile
material and packaging components specified by the designer and agreed
to by the competent authority as intended to preserve criticality
safety.''
Para. 218. ``Criticality safety index (CSI) assigned to a package,
overpack or freight container containing fissile material shall mean a
number which is used to provide control over the accumulation of
packages, overpacks or freight containers containing material.''
[[Page 44365]]
Para. 225. ``Low dispersible radioactive material shall mean either
a solid radioactive material or a solid radioactive material in a
sealed capsule, that has limited dispersibility and is not in powdered
form.''
Para. 232. ``Quality assurance shall mean a systematic programme of
controls qand inspections applied by an organization or body involved
in the transport of radioactive material which is aimed at providing
adequate confidence that the standard of safety prescribed in these
Regulations is achieved in practice.''
Factors for Consideration
Do the definitions conflict with existing programs, or
introduce other issues or concerns?
Are there other definitions of terms that are recommended
for incorporation in Part 71?
Issue 10. Crush Test for Fissile Material Package Design
Description
Under requirements for packages containing fissile material, ST-1
682(b) requires tests specified in paragraphs 719-724 followed by
whichever of the following is the more limiting: the drop test onto a
bar as identified in paragraph 727(b) and, either the crush test listed
in paragraph 727(c) for packages having a mass not greater than 500 kg
and an overall density not greater than 1000 kg/m\3\ based on external
dimensions, or the nine meter drop test listed in paragraph 727(a) for
all other packages; or the water immersion test of paragraph 729.
SS No.6 and Part 71 presently require the crush test for fissile
material packages having a mass not greater than 500 kg and an overall
density not greater than 1000 kg/m3 based on external
dimensions, and radioactive contents greater than 1000 A2
not as special form radioactive material. Under ST-1, the crush test is
no longer limited to fissile material packages containing an activity
greater than 1000 A2 because ST-1 has extended the crush
test requirement to include fissile material package designs regardless
of the activity of the contents. This was done 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 designs).
To be consistent with ST-1, the NRC would have to revise 10 CFR
Part 71 wording to recognize removal of the 1000 A2 activity
limit with respect to the crush test requirement for fissile material
package designs. However, full compliance with ST-1 requirements for
fissile material packages would also require changes to the
hypothetical accident conditions test sequencing of 10 CFR 71.73 and
would require performance of the nine-meter free drop test or the crush
test, but not both as presently required by Sec. 71.73.
Factors for Consideration
How should the differences in the test sequencing and
required tests be addressed? Would the test sequencing requirements be
applied to Type B packages as well?
What would be the impact on availability of packages and
shipping costs due to elimination of the 1000 A2 activity
limit for fissile material packages having a mass not greater than 500
kg and an overall density not greater than 1000 kg/m3 based
on external dimensions?
If Part 71 is changed to only eliminate the 1000
A2 activity limit for fissile material packages, but all
other tests and the testing sequence remains unchanged, what
implications would this have for US origin packages for export?
Issue 11. Fissile Material Package Design for Transport by Aircraft
Issue Description
For shipment of fissile material by air, ST-1 requires that
packages with quantities greater than excepted amounts (that would
include all the NRC certified packages) require an additional
criticality evaluation. Specifically, the requirements are:
Para 680(a): Packages must remain subcritical, assuming 20 centimeters
water reflection but not inleakage (i.e., moderation) when subjected to
the tests for Type C packages (see Issue 6). The specification of no
water ingress is given as the objective of this requirement is
protection from criticality events resulting from mechanical or
physical rearrangement of the geometry of the package (i.e., fast
criticality).
Para 680(b) This provision states that if a package takes credit for
``special features,'' this package can only be presented for air
transport if it is shown that these features remain effective even
under the Type C test conditions followed by a water immersion test.
``Special features'' are specified in ST-1 Para 677, and include
features that provide moderator exclusion.
The application of the paragraph 680 requirement to fissile-by-air
packages is in addition to the normal condition tests (and possibly
accident tests) that the package already must meet. Thus:
A Type IF or AF package by air must: 1) Withstand
incident-free conditions of transport with respect to release,
shielding, and maintaining subcriticality (single package and array of
packages), (2) withstand accident condition tests with respect to
maintaining subcriticality (single package and array of packages), and
(3) comply with para 680 with respect to maintaining subcriticality
(single package).
A Type BF package by air must: (1) Withstand incident-free
conditions of transport and Type B tests with respect to release,
shielding, and maintaining subcriticality (single package and array of
packages); and (2) comply with para 680 with respect to maintaining
subcriticality (single package).
A Type C fissile material package must withstand:
incident-free conditions of transport (single package and array of
packages), Type B tests (single package and array of packages), and
Type C tests (single package) with respect to release, shielding, and
maintaining subcriticality.
Factors for Consideration
Certain factors need to be considered in determining the
practical impacts of domestic adoption of ST-1 paragraph 680. First,
all uranium can be shipped in non-Type C package (IF, AF) due to its
A1 and A2 values. The paragraph 680(a)
requirements appear to be readily satisfied by low-enriched uranium,
because low enriched uranium (less than approximately 5% enrichment)
would typically require moderation (e.g., by water) to achieve nuclear
criticality, but the test specifies no water ingress. Secondly, there
are statutory restrictions on air transport of plutonium in the U.S.
Finally, packaging for air transportation may follow International
Civil Aviation Organization Technical Instructions that are also being
revised for compatibility with ST-1.
Issue 12: Special Package Approvals
Description
The transport of large objects that are too large for certified
packagings and cannot satisfy the packaging requirements was not
considered in the development of Part 71. However, as decommissioning
activities increase, the need to transport large objects is rising. For
example, in 1997, Portland General Electric Company (PGE) requested
approval of the Trojan Reactor Vessel Package (TRVP) (including
internals) for transport to the disposal facility
[[Page 44366]]
operated by US Ecology on the Hanford Nuclear Reservation near
Richland, Washington. The TRVP contained approximately 74 petabequerels
(2 million curies) in the form of activated metal and 5.7 terabequerels
(155 curies) in the form of internal surface contamination; was filled
with low-density concrete; and weighed approximately 900 metric tons
(1000 tons).
The Commission approved the Trojan shipment under exemptions issued
through 10 CFR Part 71.8. Also, the U.S. Department of Transportation's
(DOT's) regulations that govern radioactive material shipments do not
recognize packages approved via NRC exemption, so DOT also had to
consider and issue an exemption for the Trojan shipment.
Because it is the Commission's policy to avoid the use of
exemptions for recurring licensing actions, the NRC staff is
considering adding regulatory provisions to Part 71 to address special
package approvals. If adopted, these provisions would provide a
mechanism for review of special packages under the regulations without
the need for exemptions.
Factors for Consideration
Should Part 71 be revised to address reactor vessels
specifically or to address large objects in general?
Should NRC consider adopting an analogue of IAEA's special
arrangement provision modified to address packaging?
What (additional) determinations should be included in an
application for a special package approval?
Should the risk-informed basis used specifically for the
Trojan approval be adopted for other special package approvals?
Issue 13. Expansion of Part 71 Quality Assurance Requirements to
Holders of, and Applicants for, a Certificate of Compliance
Description
The NRC has observed problems with the performance of 10 CFR Part
72 Certificate of Compliance (CoC) holders in implementing the Part 72
quality assurance (QA) requirements. Problems have occurred in design,
design control, fabrication, and corrective action areas. Although CoCs
are legally binding documents, certificate holders or applicants for a
CoC and their contractors and subcontractors have not clearly been
brought within the scope of Part 72 requirements. Therefore, because
the terms ``certificate holder'' and ``applicant for a certificate of
compliance'' do not appear in the Part 72, Subpart G regulations, the
NRC has not had a clear basis to cite these persons for violations of
Part 72 requirements in the same way it treats licensees.
The NRC Enforcement Policy \1\ and its implementing program were
established to support the NRC's overall safety mission in protecting
public health and safety and the environment. Consistent with this
purpose, enforcement actions are used as a deterrent to emphasize the
importance of compliance with requirements and to encourage prompt
identification and comprehensive correction of the violations.
Enforcement sanctions consist of Notices of Violation (NOVs), civil
penalties, and orders of various types. In addition to formal
enforcement actions, the NRC also uses related administrative actions
such as Notices of Nonconformance (NONs), Confirmatory Action Letters,
and Demands for Information to supplement its enforcement program. The
NRC expects licensees, certificate holders, and applicants for a CoC to
adhere to any obligations and commitments that result from these
actions and will not hesitate to issue appropriate orders to ensure
that these obligations and commitments are met. The nature and extent
of the enforcement action are intended to reflect the seriousness of
the violation involved. An NOV is a written notice setting forth one or
more violations of a legally binding requirement.
---------------------------------------------------------------------------
\1\ NUREG-1600, ``General Statement of Policy and Procedures for
NRC Enforcement Actions,'' May 2000.
---------------------------------------------------------------------------
However, when the NRC has identified a failure to comply with Part
72 QA requirements by certificate holders or applicants for a CoC, it
has issued an NON rather than an 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 believed that limiting the available enforcement sanctions to
administrative actions was insufficient to address the performance
problems observed in industry.
In response to this problem, the NRC staff submitted a rulemaking
plan to revise Part 72 to the Commission in SECY-97-214.\2\ In a Staff
Requirements Memorandum (SRM) to SECY-97-214, the Commission approved
the staff's rulemaking plan and directed the staff to also consider
whether conforming changes to the quality assurance (QA) regulations in
Part 71 would be necessary, because of dual purpose cask designs. Dual
purpose cask designs are intended for both the storage of spent fuel
under Part 72 and the transportation of spent fuel under Part 71. In a
memorandum from the EDO to the Commission, dated December 3, 1997, the
NRC staff indicated that expansion of the Part 71 QA provisions to
include certificate holders and applicants for a Certificate of
Compliance (CoC) would be made as part of the rulemaking to conform
Part 71 to IAEA standard ST-1.
---------------------------------------------------------------------------
\2\ SECY-97-214, ``Changes to 10 CFR Part 72, Expand
Applicability to Include Certificate Holders and Applicants and
Their Contractors and Subcontractors,'' dated September 24, 1997.
This rulemaking plan expanded the applicability of the QA provision
of Part 72, Subpart G, to specifically include Part 72 certificate
holders and applicants for a Certificate of Compliance.
---------------------------------------------------------------------------
The Commission recently issued a final rule expanding QA
regulations in Part 72, Subpart G, to specifically include certificate
holders and applicants for a CoC. Consequently, the NRC is now
considering similarly expanding the QA regulations in Part 71, Subpart
H, to specifically include certificate holders and applicants for a
CoC. The NRC believes that this change is necessary to ensure
consistency between the QA provisions of Parts 71 and 72, particularly
in light of NRC approval of dual purpose cask designs. As with the Part
72 final rule, this issue would provide explicit notice to certificate
holders and applicants for a CoC of their QA responsibilities; and
would provide the NRC staff with additional enforcement sanction--
should violations of the Part 71 QA requirements occur.
Factors for Consideration
Should consistency be maintained between the QA provisions
of Parts 71 and 72, in light of the existence of dual purpose cask
designs?
Issue 14. Adoption of ASME Code
Description
The NRC staff proposes that the ASME (American Society of
Mechanical Engineers) Code, Section III, Division 3, be incorporated by
reference in 10 CFR Part 71 via rulemaking. This rule will ensure
implementation of the ASME
[[Page 44367]]
Code in cask fabrication, including all QA aspects of the code, such as
the presence of an authorized nuclear inspector (ANI) during the
fabrication to ensure that the code requirements are met, and stamping
of components after fabrication is complete. This approach would be
similar to how the ASME Code is endorsed for power reactors under 10
CFR 50.55(a) and would make the fabrication process for transportation
cask containments commensurate with that used for nuclear power plant
components.
NRC inspections of vendors'/fabricators' shops (for fabrication of
spent fuel storage canisters and transportation casks) have identified,
over the past several years, quality control (QC) and quality assurance
(QA) problems in these fabricated systems. A major reason for these
problems is that these fabricators/vendors do not fully use a code for
QA in the fabrication process of these systems. These QA problems have
in some instances continued in spite of repeated adverse NRC and
licensee findings.
The NRC staff intends to incorporate two recent developments.
First, ASME issued a consensus code 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, that would
require stamping of components constructed to it (i.e., the
transportation cask's containment). Second, Public Law 104-113
``National Technology Transfer and Advancement Act'' was enacted in
1996 to require that Federal agencies use consensus standards (e.g.,
the ASME B&PV Code), except when there are justified reasons for not
doing so. These two developments support efforts to initiate rulemaking
in this area.
Factors for Consideration
Can other regulatory vehicles for NRC endorsement of Code
be used or should this only be done by rulemaking?
Are there other voluntary consensus standards that should
be considered in addition to, or in lieu of, ASME code?
Issue 15. Adoption of Changes, Tests, and Experiments Authority
Description
The Commission recently approved a final rule to expand the
provisions of 10 CFR 72.48, ``Changes, Tests, and Experiments,'' to
include Part 72 certificate holders (October 4, 1999; 64 FR 53582). 10
CFR Part 72 Certificate holders are allowed to make changes to a spent
fuel storage cask design or conduct tests and experiments, without
prior NRC review and approval, if certain requirements are met.
However, Part 71 contains no similar provisions to permit a certificate
holder to change the design of a Part 71 transportation package. The
NRC has issued Certificates of Compliance (CoC) under Parts 71 and 72
for dual purpose casks [packages] (i.e., containers intended for both
the storage and transportation of spent fuel). This has created the
situation where a 10 CFR Part 72 certificate holder is authorized to
change a storage design feature of a dual-purpose storage/
transportation cask without obtaining NRC prior approval; however, the
10 CFR Part 71 certificate holder is not authorized to modify
transportation package design without obtaining NRC prior approval,
even when the same physical component and change is involved.
In SECY-99-130 \3\ and SECY-99-054.\4\ the staff indicated that
comments had been received on the proposed rule that requested that
authority similar to 10 CFR 72.48 be created in Part 71, particularly
with respect to dual purpose casks. Staff indicated that this issue
would be addressed in the subsequent rulemaking to conform Part 71 with
IAEA standard ST-1. The Commission adopted the staff's recommendations
in a Staff Requirements Memorandum (SRM) dated June 22, 1999.
---------------------------------------------------------------------------
\3\ SECY-99-130, ``Final Rule--Revisions to Requirements of 10
CFR Parts 50 and 72 Concerning Changes, Tests, and Experiments,''
dated May 12, 1999.
\4\ SECY-99-054, ``Plans for Final Rule--Revisions to
Requirements of 10 CFR Parts 50, 52, and 72 Concerning Changes,
Tests, and Experiments,'' dated February 22, 1999.
---------------------------------------------------------------------------
In SECY-99-054 staff recommended that a similar authority to 10 CFR
72.48 be created for spent fuel transportation packages intended for
domestic use only. Staff also recommended that this authority be
limited to Part 50 and 72 licensees shipping spent fuel and the Part 71
certificate holder. Furthermore, other supporting changes to Part 71
would be required to ensure consistency with the process contained in
10 CFR 72.48. These changes would include using common terminology such
as ``changes to the cask design, as described in the final safety
analysis report'' (FSAR) and a process for requesting amendments to a
CoC. Requirements for periodically updating a transportation package
FSAR would also be required to ensure an accurate ``licensing'' basis
is available for evaluating future proposed changes, and requirements
for package users to have a copy of the FSAR, and the updated FSAR.
The current IAEA standard ST-1 does not contain any equivalent
provisions for changing a transportation package's design, without
prior review by the competent authority.
Factors for Consideration
Should this change authority apply to spent fuel packages
involved in domestic commerce only?
Should this change authority be expanded to include all
types of transportation packages, licensees, or users?
Should the change authority apply to all domestic
transportation packages?
Should the change authority apply to dual purpose spent
fuel packages?
Issue 16. Fissile Material Exemptions and General License
Provisions
Discussion
The NRC published an emergency final rule on February 10, 1997 (62
FR 5907), amending Part 71 regulations that deal with shipments of
exempt quantities of fissile material and shipments of fissile material
under a general license. An NRC licensee had identified that a shipment
of waste material (beryllium oxide containing a low concentration of
high-enriched uranium) that met the fissile exemption provisions of 10
CFR 71.53 had the potential for an accidental criticality in certain
specific circumstances. Packages shipped under the provisions of 10 CFR
71.53 were considered inherently safe for criticality-safety purposes.
These regulations assumed that only ordinary water (H2O)
could be present as a moderating material. The regulations did not
contemplate the presence of special moderating materials (e.g.,
beryllium, graphite, or deuterium). Because of this criticality safety
issue, the NRC published a rule that was immediately effective with no
opportunity for pre-promulgation public comment. The NRC did solicit
comments after the rule was effective. All public comments supported
the need for the emergency final rule when the shipments contained
special moderators (moderators other than water); however, the
commenters stated that the rule had gone too far for water moderated
shipments, that it was excessively restrictive and costly to licensees,
and that further rulemaking was necessary.
Based on these comments, NRC staff contracted with Oak Ridge
National Laboratory (ORNL) to thoroughly review fissile material
exemptions and general license provisions. ORNL performed
[[Page 44368]]
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 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. The ORNL study
was issued as NUREG/CR-5342 in July 1998 (available via the following
NRC website: http://www.nrc.gov/NRC/NUREGS/CR5342/index.html). The ORNL
study confirmed that the emergency rule was needed to provide safe
transportation of packages with special moderators that are shipped
under the general license and fissile material exemptions, but may be
excessive for water-moderated shipments.
NUREG/CR-5342 identified 16 recommended actions for additional
rulemaking. Additionally, the Commission's SRM on SECY-96-268 approving
the emergency final rule directed the staff to issue guidance for
instances where fissile materials may be mixed in the same shipping
container with different moderators. The staff indicated that this
issue would be addressed in a forthcoming rulemaking (memorandum from
the EDO to the Commission, dated September 8, 1998). On October 27,
1999, the NRC published Federal Register Notice 64 FR 57769 responding
to public comments on the emergency final rule, and also requesting
information on the cost impact of the final rule from the public,
industry, and the DOE, because the NRC staff had not been successful in
obtaining this information. The requirements for the fissile material
general licenses are provided in 10 CFR 71.18, 71.20, 71.22, and 71.24,
and the fissile material exemptions are provided in 71.53.
IAEA standard ST-1 contains language on fissile exemptions and
restrictions on the use of special moderators. However, ST-1 does not
presently contain provisions on general licenses for shipment of
fissile material; previous version did contain general license
conditions.
Factors for Consideration
Should all, or only some, of the 16 sub-issues (i.e., the
recommendations contained in NUREG/CR-5342) be included in this
rulemaking on this issue?
Should additional issues or alternative approaches on the
fissile exemptions or general license provisions be included in this
rulemaking?
Is there available cost data that may help to understand
the cost impact of the implemented emergency rule; or help to better
understand the possible cost impact of the ORNL recommendations?
Issue 17. Double Containment of Plutonium (PRM-71-12)
Description
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 on
February 19, 1998. The comment period was extended to July 31, 1998.
The petitioner requested that regulations in 10 CFR 71.63 be
eliminated. The petitioner argued that the double containment
requirement in 71.63(b) was not consistent with the basis for other
packaging standards (i.e., the Q-value system for identifying the
A1 and A2 values for each nuclide). The
petitioner also argued that the use of double containment for shipments
of plutonium imposed unnecessary costs (i.e., fabrication of shipping
packages and a weight penalty). As an option, the petitioner requested
that 71.63 be entirely eliminated.
In 1974, the Atomic Energy Commission (AEC) issued 10 CFR 71.63
which imposed special requirements on the shipment of plutonium in
excess of 0.74 terabecquerels (20 curies). These requirements specify
that plutonium must be in solid form (71.63(a)) and that packages used
to ship plutonium must provide a separate inner containment (i.e., the
``double containment'' requirement) (71.63(b)). In adopting these
requirements, the AEC specifically excluded plutonium in the form of
reactor fuel elements, metal or metal alloys, and other plutonium-
bearing solids that the Commission determines, on a case-by-case basis,
do not require double containment. These regulations have remained
essentially unchanged since 1974, except for the addition in 1998 of
vitrified high-level waste in sealed canisters to the list of exempt
forms of plutonium. Double containment is in addition to Type B
packaging standards and is not required for any other nuclides that are
listed in Part 71. Additionally, IAEA standard ST-1 does not contain a
double containment requirement for any nuclide.
The AEC issued this regulation at a time when wide-spread
reprocessing of commercial spent fuel was anticipated. The AEC expected
increases in the quantities of plutonium to be shipped and the number
of shipments of plutonium. In addition, the specific activity of the
plutonium was expected to increase with increased burnup, resulting in
higher gamma and neutron radiation levels, greater heat generation, and
greater pressure generation potential from plutonium nitrate solutions
in shipping containers. Because of these expected changes and because
of the susceptibility of liquids to leakage, the AEC believed that
safety would be significantly enhanced if the basic form for shipments
of plutonium were changed from liquid to solid, and if the solid form
of plutonium were required to be shipped in a package providing double
containment of the contents.
The AEC indicated that ``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.''
\5\ The AEC also indicated that the double containment provision
compensates for the fact that the plutonium may not be in a
``nonrespirable'' form. Notwithstanding these rationales, some of the
underlying assumptions for this rule were altered in 1979 when the U.S.
government decided that reprocessing of civilian spent fuel and reuse
of plutonium was not desirable. Consequently, the expected plutonium
reprocessing economy and wide-spread shipments never materialized.
---------------------------------------------------------------------------
\5 \ SECY-R-74-5, dated July 6, 1973.
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With respect to PRM-71-12, eight public comments were received on
the petition; of those, three supported the petition and five opposed
the petition. The supporting comments essentially stated that the
IAEA's Q-System accurately reflects the dangers of nuclides, including
plutonium, and that elimination of 10 CFR 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 opposing comments essentially stated that plutonium is
very dangerous, especially in liquid form, and therefore additional
regulatory requirements are warranted, that existing regulations are
not overly burdensome, especially in light of the
[[Page 44369]]
total expected transportation cost, that TRUPACT-II package meets
71.63(b) requirement, that a commenter (i.e., 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, and that any change now would erode public
confidence and be detrimental to the entire transportation system for
WIPP shipments, and that additional personnel exposure due to double
containment is insignificant.
Factors for Consideration
Should NRC change any of the special requirements for the
transportation of plutonium?
Should the double containment requirement in 71.63(b) be
eliminated?
Should both the solid form and the double containment
requirements of 71.63(a) and (b) be eliminated?
Is consistency with IAEA standard ST-1 important on this
issue?
Issue 18. Contamination Limits as Applied to Spent Fuel and High
Level Waste (HLW) Packages
Description
As part of the NRC's upcoming public meetings on proposed changes
to 10 CFR Part 71, the Commission will consider the issue of removable
package contamination limits for transportation (i.e., radioactive
material that can be removed from the surface of a package prior to
shipment). This issue involves contamination limits for all
transportation packages, including spent fuel and HLW packages,
contained in DOT regulations which are based on the international
transportation standards for contamination limits. The NRC staff
requests public and stakeholder views on whether different
contamination limits should be considered for spent fuel and HLW
packages, and recommendations for future interactions that NRC has with
DOT and IAEA on this issue. NRC staff is aware that the IAEA is
starting a review of contamination models and limits, and this review
will be conducted over the next few years.
The removable contamination limit of 4 Becquerels per square
centimeter (4Bq/cm2) is contained in IAEA Safety Series 6, in ST-1, in
U.S. DOT regulations (49 CFR 173.443), and by reference to DOT's
regulations in NRC's 10 CFR Part 71. The limit applies to the
transportation of all packages, regardless of size. Thus, the 4 Bq/cm2
contamination limit applies to shipment of spent fuel and HLW packages,
even though the unique aspects of these packages were not explicitly
considered in the modeling assumptions used in developing the
contamination limit. Specifically, the contamination limit was designed
to reduce delivery worker exposure from external contamination on small
packages during frequent manual handling of these packages in freight
facilities; however, unlike small packages moved by delivery workers,
handling of spent fuel and HLW packages is done by cranes and other
manipulation equipment, due to the large weights involved, and does not
involve extensive personnel contact, thereby reducing worker exposure
from external package contamination.
Irrespective of remote handling, workers must obtain contamination
readings on a spent fuel or HLW package's external surfaces to ensure
compliance with the 4 Bq/cm2 limit prior to release for shipment. Due
to the large surface areas involved in the contamination checks, and
the prolonged time that workers are in the vicinity of a loaded package
while performing these checks, they receive exposure from radiation
emanating through the package walls. Further, should the contamination
checks reveal contamination above 4 Bq/cm2, then additional worker
exposure occurs during decontamination activities and subsequent checks
of contamination levels to achieve the 4 Bq/cm2 limit. It should be
noted that if the contamination limit for spent fuel and HLW packages
was changed, workers would still be required to check the packages for
contamination (under the changed limit) and thus receive exposure while
performing this activity and any required decontamination activities.
Factors for Consideration
Should the 4 Bq/cm2 limit continue to apply to spent fuel
and HLW packages or should an alternative limit be developed? Is there
an alternate contamination limit or alternative approach that will
result in lowered exposure to workers, yet ensure that the rail and
truck workers as well as the public are adequately protected from
external package contamination?
If alternative contamination limits are established for
spent fuel and HLW packages, is there any concern with the possible
resulting difference in US domestic regulations and international
standards?
Appendix A--Paragraphs Referenced from IAEA ST-1
Appendix A contains the full text of specific paragraphs from
ST-1 referenced in the eleven IAEA-compatibility issues. Paragraphs
are listed numerically in ascending order, with the corresponding
issue identified in bold text at the end of the reference.
107. The Regulations do not apply to:
(e) 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 paras 401-
406. (Issue 2)
209. Confinement system shall mean the assembly of fissile
material and packaging components specified by the designer and
agreed to by the competent authority as intended to preserve
criticality safety. (Issue 9)
218. Criticality safety index (CSI) assigned to a package,
overpack or freight container containing fissile material shall mean
a number which is used to provide control over the accumulation of
packages, overpacks or freight containers containing fissile
material. (Issue 9)
225. Low dispersible radioactive material shall mean either a
solid radioactive material or a solid radioactive material in a
sealed capsule, that has limited dispersibility and is not in powder
form. (Issue 9)
230. Package shall mean the packaging with its radioactive
contents as presented for transport. The types of packages covered
by these Regulations, which are subject to the activity limits and
material restrictions of Section IV and meet the corresponding
requirements, are:
(a) Excepted package;
(b) Industrial package Type 1 (Type IP-1);
(c) Industrial package Type 2 (Type IP-2);
(d) Industrial package Type 3 (Type IP-3);
(e) Type A package;
(f) Type B(U) package;
(g) Type B(M) package;
(h) Type C package.
Packages containing fissile material or uranium hexafluoride are
subject to additional requirements. (Issue 6)
232. Quality assurance shall mean a systematic programme of
controls and inspections applied by any organization or body
involved in the transport of radioactive material which is aimed at
providing adequate confidence that the standard of safety prescribed
in these Regulations is achieved in practice. (Issue 9)
401. The following basic values for individual radionuclides are
given in Table I:
(a) A1 and A2 in TBq;
(b) activity concentration for exempt material in Bq/g; and
(c) activity limits for exempt consignments in Bq. (Issue 2)
402. For individual radionuclides which are not listed in Table
I the determination of the basic radionuclide values referred to in
para. 401 shall require competent authority approval or, for
international transport, multilateral approval. Where the chemical
form of each radionuclide is known, it is permissible to use the
A2 value related to its solubility class as recommended
by the International Commission on Radiological Protection, if the
chemical forms under both normal and accident conditions of
transport
[[Page 44370]]
are taken into consideration. Alternatively, the radionuclide values
in Table II may be used without obtaining competent authority
approval. (Issue 2)
403. In the calculations of A1 and A2 for
a radionuclide not in Table I, a single radioactive decay chain in
which the radionuclides are present in their naturally occurring
proportions, and in which no daughter nuclide has a half-life either
longer than 10 days or longer than that of the parent nuclide, shall
be considered as a single radionuclide; and the activity to be taken
into account and the A1 or A2 value to be
applied shall be those corresponding to the parent nuclide of that
chain. In the case of radioactive decay chains in which any daughter
nuclide has a half-life either longer than 10 days or greater than
that of the parent nuclide, the parent and such daughter nuclides
shall be considered as mixtures of different nuclides. (Issue 2)
404. For mixtures of radionuclides, the determination of the
basic radionuclide values referred to in para. 401 may be determined
as follows:
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BILLING CODE 7590-01-C
FIG. 4. Category III-YELLOW label. The background colour of the upper
half of the label shall be yellow and the lower half white, the colour
of the trefoil and the printing shall be black, and the clour of the
category bars shall be red.
(a) Contents:
(i) Except for LSA-I material, the name(s) of the
radionuclide(s) as taken from Table I, using the symbols prescribed
therein. For mixtures of radionuclides, the most restrictive
nuclides must be listed to the extent the space on the line permits.
The group of LSA or SCO shall be shown following the name(s) of the
radionuclide(s). The terms ``LSA-II'',''LSA-III'', ``SCO-I'' and
``SCO-II'' shall be used for this purpose.
(ii) For LSA-I material, the term ``LSA-I'' is all that is
necessary; the name of the radionuclide is not necessary.
(b) Activity: The maximum activity of the radioactive contents
during transport expressed in units of becquerels (Bq) with the
appropriate SI prefix (see Annex II). For fissile material, the mass
of fissile material in units of grams (g), or multiples thereof, may
be used in place of activity.
(c) For overpacks and freight containers the ``contents'' and
``activity'' entries on the label shall bear the information
required in subparas 543(a) and 543(b), respectively, totalled
together for the entire contents of the overpack or freight
container except that on labels for overpacks or freight containers
containing mixed loads of packages containing different
radionuclides, such entries may read ``See Transport Documents''.
(d) Transport index: See paras 526 and 527. (No transport index
entry is required for category I-WHITE.) (Issue 1)
544. Each label conforming to the model in Fig. 5 shall be
completed with the criticality safety index (CSI) as stated in the
certificate of approval for special arrangement or the certificate
of approval for the package design issued by the competent
authority. (Issue 5)
545. For overpacks and freight containers, the criticality
safety index (CSI) on the label shall bear the information required
in para. 544 totalled together for the fissile contents of the
overpack or freight container. (Issue 5)
549. The consignor shall include in the transport documents with
each consignment the following information, as applicable in the
order given:
(a) The proper shipping name, as specified in Table VIII;
(b) The United Nations Class number ``7'';
(c) The United Nations number assigned to the material as
specified in Table VIII, preceded by the letters ``UN'';
(d) The name or symbol of each radionuclide or, for mixtures of
radionuclides, an appropriate general description or a list of the
most restrictive nuclides;
(e) A description of the physical and chemical form of the
material, or a notation that the material is special form
radioactive material or low dispersible radioactive material. A
generic chemical description is acceptable for chemical form;
(f) The maximum activity of the radioactive contents during
transport expressed in units
[[Page 44396]]
of becquerels (Bq) with an appropriate SI prefix (see Annex II). For
fissile material, the mass of fissile material in units of grams
(g), or appropriate multiples thereof, may be used in place of
activity.
(g) The category of the package, i.e. I-WHITE, II-YELLOW, III-
YELLOW;
(h) The transport index (categories II-YELLOW and III-YELLOW
only);
(i) For consignments including fissile material other than
consignments excepted under para. 672, the criticality safety index;
(j) The identification mark for each competent authority
approval certificate (special form radioactive material, low
dispersible radioactive material, special arrangement, package
design, or shipment) applicable to the consignment;
(k) For consignments of packages in an overpack or freight
container, a detailed statement of the contents of each package
within the overpack or freight container and, where appropriate, of
each overpack or freight container in the consignment. If packages
are to be removed from the overpack or freight container at a point
of intermediate unloading, appropriate transport documents shall be
made available;
(l) Where a consignment is required to be shipped under
exclusive use, the statement ``EXCLUSIVE USE SHIPMENT''; and
(m) For LSA-II, LSA-III, SCO-I and SCO-II, the total activity of
the consignment as a multiple of A2. (Issue 1)
629. Except as allowed in para. 632, uranium hexafluoride shall
be packaged and transported in accordance with the provisions of the
International Organization for Standardization document ISO 7195:,
``Packaging of uranium hexafluoride (UF6) for transport''
1, and the requirements of paras 630-631. The package
shall also meet the requirements prescribed elsewhere in these
Regulations which pertain to the radioactive and fissile properties
of the material. (Issue 4)
630. Each package designed to contain 0.1 kg or more of uranium
hexafluoride shall be designed so that it would meet the following
requirements:
(a) withstand without leakage and without unacceptable stress,
as specified in the International Organization for Standardization
document ISO 7195\10\, the structural test as specified in para.
718;
(b) withstand without loss or dispersal of the uranium
hexafluoride the test specified in para. 722; and
(c) withstand without rupture of the containment system the test
specified in para. 728. (Issue 4)
631. Packages designed to contain 0.1 kg or more of uranium
hexafluoride shall not be provided with pressure relief devices.
(Issue 4)
632. Subject to the approval of the competent authority,
packages designed to contain 0.1 kg or more of uranium hexafluoride
may be transported if:
(a) the packages are designed to requirements other than those
given in ISO 7195\10\ and paras 630-631 but, notwithstanding, the
requirements of paras 630-631 are met as far as practicable. (Issue
4)
657. A package for radioactive contents with activity greater
than 10\5\ A2 shall be so designed that if it were
subjected to the enhanced water immersion test specified in para.
730, there would be no rupture of the containment system. (Issue 7)
667. Type C packages shall be designed to meet the requirements
specified in paras 606-619, and of paras 634-647, except as
specified in para. 646(a), and of the requirements specified in
paras 651-654, paras 658-664, and, in addition, of paras 668-670.
(Issue 6)
668. A package shall be capable of meeting the assessment
criteria prescribed for tests in paras 656(b) and 660 after burial
in an environment defined by a thermal conductivity of 0.33 W/m.K
and a temperature of 38 deg.C in the steady state. Initial
conditions for the assessment shall assume that any thermal
insulation of the package remains intact, the package is at the
maximum normal operating pressure and the ambient temperature is
38 deg.C. (Issue 6)
669. A package shall be so designed that, if it were at the
maximum normal operating pressure and subjected to:
(a) the tests specified in paras 719-724, it would restrict the
loss of radioactive contents to not more than 10-\6\
A2 per hour; and
(b) the test sequences in para. 734, it would meet the following
requirements:
(i) retain sufficient shielding to ensure that the radiation
level at 1 m from the surface of the package would not exceed 10
mSv/h with the maximum radioactive contents which the package is
designed to contain; and
(ii) restrict the accumulated loss of radioactive contents in a
period of 1 week to not more than 10 A2 for krypton-85
and not more than A2 for all other radionuclides.
Where mixtures of different radionuclides are present, the
provisions of paras 404-406 shall apply except that for krypton-85
an effective A2(i) value equal to 10 A2 may be
used. For case (a) above, the assessment shall take into account the
external contamination limits of para. 508. (Issue 6)
670. A package shall be so designed that there will be no
rupture of the containment system following performance of the
enhanced water immersion test specified in para. 730. (Issue 6)
677. For a package in isolation, it shall be assumed that water
can leak into or out of all void spaces of the package, including
those within the containment system. However, if the design
incorporates special features to prevent such leakage of water into
or out of certain void spaces, even as a result of error, absence of
leakage may be assumed in respect of those void spaces. Special
features shall include the following:
(a) Multiple high standard water barriers, each of which would
remain watertight if the package were subject to the tests
prescribed in para. 682(b), a high degree of quality control in the
manufacture, maintenance and repair of packagings and tests to
demonstrate the closure of each package before each shipment; or
(b) For packages containing uranium hexafluoride only:
(i) packages where, following the tests prescribed in para.
682(b), there is no physical contact between the valve and any other
component of the packaging other than at its original point of
attachment and where, in addition, following the test prescribed in
para. 728 the valves remain leaktight; and
(ii) a high degree of quality control in the manufacture,
maintenance and repair of packagings coupled with tests to
demonstrate closure of each package before each shipment. (Issue 4
and issue 11)
680. For packages to be transported by air:
(a) the package shall be subcritical under conditions consistent
with the tests prescribed in para. 734 assuming reflection by at
least 20cm of water but no water inleakage; and
(b) allowance shall not be made for special features of para.
677 unless, following the tests specified in para. 734 and,
subsequently, para. 733, leakage of water into or out of the void
spaces is prevented. (Issue 11)
682. A number ``N'' shall be derived, such that two times ``N''
shall be subcritical for the arrangement and package conditions that
provide the maximum neutron multiplication consistent with the
following:
(a) Hydrogenous moderation between packages, and the package
arrangement reflected on all sides by at least 20 cm of water; and
(b) The tests specified in paras 719-724 followed by whichever
of the following is the more limiting:
(i) the tests specified in para. 727(b) and, either para. 727(c)
for packages having a mass not greater than 500 kg and an overall
density not greater than 1000 kg/m3 based on the external
dimensions, or para. 727(a) for all other packages; followed by the
test specified in para. 728 and completed by the tests specified in
paras 731-733; or
(ii) the test specified in para. 729; and
(c) Where any part of the fissile material escapes from the
containment system following the tests specified in para. 682(b), it
shall be assumed that fissile material escapes from each package in
the array and all of the fissile material shall be arranged in the
configuration and moderation that results in the maximum neutron
multiplication with close reflection by at least 20 cm of water.
(Issue 10)
719. The tests are: the water spray test, the free drop test,
the stacking test and the penetration test. Specimens of the package
shall be subjected to the free drop test, the stacking test and the
penetration test, preceded in each case by the water spray test. One
specimen may be used for all the tests, provided that the
requirements of para. 720 are fulfilled. (Issue 10)
720. The time interval between the conclusion of the water spray
test and the succeeding test shall be such that the water has soaked
in to the maximum extent, without appreciable drying of the exterior
of the specimen. In the absence of any evidence to the contrary,
this interval shall be taken to be two hours if the water spray is
applied from four directions simultaneously. No time interval shall
elapse, however, if the water spray is applied from each of the four
directions consecutively. (Issue 10)
721. Water spray test: The specimen shall be subjected to a
water spray test that simulates exposure to rainfall of
approximately 5 cm per hour for at least one hour. (Issue 10).
[[Page 44397]]
722. Free drop test: The specimen shall drop onto the target so
as to suffer maximum damage in respect of the safety features to be
tested.
(a) The height of drop measured from the lowest point of the
specimen to the upper surface of the target shall be not less than
the distance specified in Table XIII for the applicable mass. The
target shall be as defined in para. 717.
(b) For rectangular fibreboard or wood packages not exceeding a
mass of 50 kg, a separate specimen shall be subjected to a free drop
onto each corner from a height of 0.3 m.
(c) For cylindrical fibreboard packages not exceeding a mass of
100 kg, a separate specimen shall be subjected to a free drop onto
each of the quarters of each rim from a height of 0.3 m. (Issue 10)
723. Stacking test: Unless the shape of the packaging
effectively prevents stacking, the specimen shall be subjected, for
a period of 24 h, to a compressive load equal to the greater of the
following:
(a) The equivalent of 5 times the mass of the actual package;
and
(b) The equivalent of 13 kPa multiplied by the vertically
projected area of the package.
The load shall be applied uniformly to two opposite sides of the
specimen, one of which shall be the base on which the package would
typically rest. (Issue 10)
724. Penetration test: The specimen shall be placed on a rigid,
flat, horizontal surface which will not move significantly while the
test is being carried out.
(a) A bar of 3.2 cm in diameter with a hemispherical end and a
mass of 6 kg shall be dropped and directed to fall, with its
longitudinal axis vertical, onto the centre of the weakest part of
the specimen, so that, if it penetrates sufficiently far, it will
hit the containment system. The bar shall not be significantly
deformed by the test performance.
(b) The height of drop of the bar measured from its lower end to
the intended point of impact on the upper surface of the specimen
shall be 1 m. (Issue 10)
727. Mechanical test: The mechanical test consists of three
different drop tests. Each specimen shall be subjected to the
applicable drops as specified in para. 656 or para. 682. The order
in which the specimen is subjected to the drops shall be such that,
on completion of the mechanical test, the specimen shall have
suffered such damage as will lead to the maximum damage in the
thermal test which follows.
(a) For drop I, the specimen shall drop onto the target so as to
suffer the maximum damage, and the height of the drop measured from
the lowest point of the specimen to the upper surface of the target
shall be 9 m. The target shall be as defined in para. 717.
(b) For drop II, the specimen shall drop so as to suffer the
maximum damage onto a bar rigidly mounted perpendicularly on the
target. The height of the drop measured from the intended point of
impact of the specimen to the upper surface of the bar shall be 1 m.
The bar shall be of solid mild steel of circular section, (15.0
0.5) cm in diameter and 20 cm long unless a longer bar
would cause greater damage, in which case a bar of sufficient length
to cause maximum damage shall be used. The upper end of the bar
shall be flat and horizontal with its edges rounded off to a radius
of not more than 6 mm. The target on which the bar is mounted shall
be as described in para. 717.
(c) For drop III, the specimen shall be subjected to a dynamic
crush test by positioning the specimen on the target so as to suffer
maximum damage by the drop of a 500 kg mass from 9 m onto the
specimen. The mass shall consist of a solid mild steel plate 1 m by
1 m and shall fall in a horizontal attitude. The height of the drop
shall be measured from the underside of the plate to the highest
point of the specimen. The target on which the specimen rests shall
be as defined in para. 717. (Issue 10)
729. Water immersion test: The specimen shall be immersed under
a head of water of at least 15 m for a period of not less than eight
hours in the attitude which will lead to maximum damage. For
demonstration purposes, an external gauge pressure of at least 150
kPa shall be considered to meet these conditions. (Issue 10)
730. Enhanced water immersion test: The specimen shall be
immersed under a head of water of at least 200 m for a period of not
less than one hour. For demonstration purposes, an external gauge
pressure of at least 2 MPa shall be considered to meet these
conditions. (Issue 7)
734. Specimens shall be subjected to the effects of each of the
following test sequences in the orders specified:
(a) the tests specified in paras 727(a), 727(c), 735 and 736;
and
(b) the test specified in para. 737.
Separate specimens are allowed to be used for each of the
sequences (a) and (b). (Issue 6)
735. Puncture/tearing test: The specimen shall be subjected to
the damaging effects of a solid probe made of mild steel. The
orientation of the probe to the surface of the specimen shall be as
to cause maximum damage at the conclusion of the test sequence
specified in para. 734(a).
(a) The specimen, representing a package having a mass less than
250 kg, shall be placed on a target and subjected to a probe having
a mass of 250 kg falling from a height of 3 m above the intended
impact point. For this test the probe shall be a 20 cm diameter
cylindrical bar with the striking end forming a frustum of a right
circular cone with the following dimensions: 30 cm height and 2.5 cm
in diameter at the top. The target on which the specimen is placed
shall be as specified in para. 717.
(b) For packages having a mass of 250 kg or more, the base of
the probe shall be placed on a target and the specimen dropped onto
the probe. The height of the drop , measured from the point of
impact with the specimen to the upper surface of the probe shall be
3 m. For this test the probe shall have the same properties and
dimensions as specified in (a) above, except that the length and
mass of the probe shall be such as to incur maximum damage to the
specimen. The target on which the base of the probe is placed shall
be as specified in para. 717. (Issue 6)
736. Enhanced thermal test: The conditions for this test shall
be as specified in para. 728, except that the exposure to the
thermal environment shall be for a period of 60 minutes. (Issue 6)
737. Impact test: The specimen shall be subject to an impact on
a target at a velocity of not less than 90 m/s, at such an
orientation as to suffer maximum damage. The target shall be as
defined in para. 717. (Issue 6)
Dated at Rockville, Maryland, this 11th day of July, 2000.
For the Nuclear Regulatory Commission.
William F. Kane,
Director, Office of Nuclear Material Safety and Safeguards.
[FR Doc. 00-18029 Filed 7-14-00; 8:45 am]
BILLING CODE 7590-01-P