[Federal Register Volume 65, Number 181 (Monday, September 18, 2000)]
[Rules and Regulations]
[Pages 56211-56231]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-23354]
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Rules and Regulations
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Federal Register / Vol. 65, No. 181 / Monday, September 18, 2000 /
Rules and Regulations
[[Page 56211]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 70
RIN 3150-AF22
Domestic Licensing of Special Nuclear Material; Possession of a
Critical Mass of Special Nuclear Material
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations governing the domestic licensing of special nuclear
material (SNM) for licensees authorized to possess a critical mass of
SNM, that are engaged in one of the following activities: enriched
uranium processing; fabrication of uranium fuel or fuel assemblies;
uranium enrichment (other than certified existing gaseous diffusion
plants); enriched uranium hexafluoride conversion; plutonium
processing; fabrication of mixed-oxide fuel or fuel assemblies; scrap
recovery of SNM; or any other activity involving a critical mass of SNM
that the Commission determines could significantly affect public health
and safety or the environment. The amendments establish performance
requirements, require affected licensees to perform an integrated
safety analysis (ISA) to identify potential accidents at the facility
and the items relied on for safety necessary to prevent these potential
accidents and/or mitigate their consequences; require the
implementation of measures to ensure that the items relied on for
safety are available and reliable to perform their function when
needed; require the safety bases to be maintained, and changes reported
to NRC; allow for licensees to make certain changes to their safety
program and facilities without prior NRC approval; require reporting of
certain events; and require the NRC to perform a backfit analysis under
specified circumstances.
EFFECTIVE DATE: The final rule, with the exception of Sec. 70.76, is
effective October 18, 2000. Section 70.76 will become effective after
the issuance of staff guidance for the implementation of that
provision. Once such guidance has been developed, the NRC will publish
a Federal Register notice specifying the effective date of Sec. 70.76.
FOR FURTHER INFORMATION CONTACT: Theodore S. Sherr, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, telephone (301) 415-7218; e-mail
[email protected], Heather Astwood, Office of Nuclear Material Safety and
Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, telephone (301) 415-5819; e-mail [email protected], or Andrew Persinko,
Office of Nuclear Material Safety and Safeguards, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-
6522; e-mail [email protected].
SUPPLEMENTARY INFORMATION:
I. Background
II. Public Comments on Proposed Rule
III. Changes from the Proposed Rule
IV. Section-by-Section Analysis of Part 70 Amendments
V. Finding of No Significant Environmental Impact: Availability
VI. Paperwork Reduction Act Statement
VII. Public Protection Notification
VIII. Regulatory Analysis
IX. Regulatory Flexibility Certification
X. Voluntary Consensus Standards
XI. Backfit Statement
XII. Small Business Regulatory Enforcement Fairness Act
XIII. List of Subjects
I. Background
On July 30, 1999 (64 FR 41338), the Commission published a proposed
rule for public comment that would amend its regulations governing the
domestic licensing of SNM for certain licensees authorized to possess a
critical mass of SNM. The Commission's action was in response to a
Petition for Rulemaking, (PRM)-70-7, submitted by the Nuclear Energy
Institute (NEI), which was published on November 26, 1996 (61 FR
60057). The proposed rule was intended to grant the NEI PRM in part and
would modify the petitioner's proposal. The majority of the proposed
modifications to Part 70 were included in a proposed new Subpart H,
``Additional Requirements for Certain Licensees Authorized to Possess a
Critical Mass of Special Nuclear Material.'' These modifications were
proposed in order to increase confidence in the margin of safety at the
facilities affected by the rule.
In developing the proposed rule, the Commission sought to achieve
its objectives through a risk-informed and performance-based regulatory
approach that included: (1) The identification of performance
requirements for prevention of accidents or mitigation of their
consequences; (2) the performance of an ISA to identify potential
accidents at the facility and the items relied on for safety; (3) the
implementation of measures to ensure that the items relied on for
safety are available and reliable to perform their function when
needed; (4) the maintenance of the safety bases, including the
reporting of changes to the NRC; and (5) the allowance for licensees to
make certain changes to their safety program and facilities without
prior NRC approval.
The 75-day public comment period on the proposed rule ended on
October 13, 1999. During and after the public comment period, the NRC
staff posted on the NRC web site revised versions of the draft Standard
Review Plan (SRP) that would implement the proposed requirements (i.e.,
on August 4, 1999, a complete draft SRP was posted, and revised
chapters, taking into account comments received, were posted during the
period March 16-April 3, 2000). In addition, three stakeholder meetings
were held to discuss the SRP (September 14-15, 1999, February 9, 2000,
and April 18-19, 2000).
II. Public Comments on Proposed Rule
In preparing the final rule, the NRC staff carefully reviewed and
considered more than 90 comments on the rule, included in 9 individual
letters filed during the public comment period. In addition, 13
submittals containing more than 200 specific comments were received on
the SRP. To simplify the analysis, the NRC staff grouped all written
comments on the rule into the following major topic areas: Performance
Requirements and Design Criteria; Content of Applications and ISA
Summary; Safety Program; Change Process, License Renewal and Backfit;
Definitions; and Miscellaneous. A more detailed analysis is also
documented as
[[Page 56212]]
an attachment to SECY-00-0111. A review of the comments and NRC staff's
responses follow:
A. Performance Requirements and Design Criteria
Comment A.1: One commenter stated that the proposed rule should
specify dose limits for anticipated occurrences similar to those in
Secs. 72.104 and 72.106. This part of the rule would then cover the
range of likelihood (anticipated, likely, unlikely, and highly
unlikely) of potential accidents that could occur at nuclear fuel cycle
facilities. This could result in an increase in the number of
structures, systems, and components (SSCs) relied on for safety and
would impact the design, operation, and licensing of the mixed-oxide
(MOX) facility.
Response: No change in the rule language has been made. The dose
limits for normal operation are contained in 10 CFR Part 20 [viz., 0.05
Sv (5 rem) Total Effective Dose Equivalent (TEDE)/yr for a trained
worker]. The NRC staff views ``anticipated occurrences'' to be
conditions of normal operations, and believes that the measures
currently used by Part 70 licensees to comply with Part 20 have been
and will continue to be successful in protecting workers and the public
during normal operations. Thus, there is insufficient justification to
require identification of ``items'' to demonstrate compliance with Part
20 during normal operations.
Comment A.2: One commenter proposed that the NRC maintain
consistency with past precedent (i.e., the Commission's rationale in
Part 60) and eliminate the specific worker dose limits in Part 70.
Response: No change in rule language has been made. The regulatory
experience and industry events that initiated the effort to add a
systematic accident analysis to Part 70 primarily involved health
impacts to workers as opposed to the public. The NRC staff believes
that the rule's focus on both the potential impacts on workers and the
public is appropriate. Based on the discussions and correspondence with
the industry and public during development of the proposed rule, and
all other comments on the proposed rule, there appears to be general
consensus on this approach.
Comment A.3: One commenter stated, in response to the Federal
Register notice request for comments on the clarity and effectiveness
of the language used (per June 1, 1998, Presidential Memorandum), that
the language in Sec. 70.61(b) and (c) could be substantially clearer;
the commenter provided an alternative plain language version of this
section.
Response: The language in the proposed rule was written in response
to public comments to focus on risk (i.e., likelihood times
consequence) of accidents. The language has been changed, in response
to the comment, to provide additional clarity. The proposed revisions
provided by the commenter, however, are not merely editorial but
represent substantive changes. They appear to have eliminated the
concept of limiting risk, and instead, focused on the likelihood of
accidents. The revised language in the rule attempts to retain the
emphasis on controlling accident risks within appropriate performance
requirements.
Comment A.4: Three commenters expressed concerns about how the
worker dose limits in Sec. 70.61(f) would be applied to ``co-located
workers.'' One commenter suggested that the performance requirements in
Sec. 70.61 consider the individuals working in the nearby facilities as
public when performing an accident analysis to determine the
consequences of the accidents that may occur at the facility. The
commenter concluded that this would result in a more stringent
application of safety requirements for the protection of workers (e.g.,
additional items relied on for safety) at the MOX Fuel Fabrication
Facility, Pit Disassembly, Conversion Facility, Immobilization
Facility, and any other nearby DOE facilities, and would also have a
substantial impact on the cost of the MOX facility. A second commenter
agreed with this assessment, noting that a worker (as defined in
Sec. 70.4) who leaves the controlled area to perform a work-related
function would have to be treated as a member of the public when
performing an ISA and would be subject to the more stringent public
radiation exposure limits. Outside of the controlled area the TEDE
limit of 1 mSv (0.1 rem) for members of the public would apply [cf. 10
CFR 20.1301(a)(1)] rather than the annual TEDE occupational dose limit
of 50 mSv (5 rems) (10 CFR 20.1201). According to this commenter, this
problem has already arisen at the Hanford Tank Waste Remediation System
where co-located workers are governed by the appreciably lower public
dose limits. A third commenter agreed with the above positions and also
stated that the NRC intends to consider individuals outside of the
controlled boundary as workers if they are subject to Part 20
requirements. The commenter noted, as did the first commenter, that DOE
requirements in 10 CFR Part 835 provide an equivalent level of
protection, such that co-located workers--who are subject to the
requirements of either Part 20 or 10 CFR Part 835--should be considered
``workers,'' provided the licensee can demonstrate the ability to
provide management measures (e.g., notification, evacuation, etc.) in
the event of an emergency.
Response: NRC regulations do not specifically address personnel
designated as ``co-located'' workers. In response to the comments, the
first sentence in Sec. 70.61(f) was changed to read as follows: ``Each
licensee must establish a controlled area, as defined in Sec. 20.1003.
In addition, the licensee must retain the authority to exclude or
remove personnel and property from the area.'' The licensee can set the
controlled area at any location around its facility as long as it
maintains control of that area as specified in Part 20 and retains the
authority to exclude or remove personnel and property from the area. If
the controlled area included the nearby Department of Energy (DOE)
facilities, then NRC would consider the personnel working at those
facilities to be ``workers'' for the purposes of the performance
requirements of Sec. 70.61, provided the conditions of Sec. 70.61(f)(2)
are met. The DOE and its contractors could satisfy these conditions by
documenting their compliance with the requirements of 10 CFR
19.12(a)(1)-(5). To emphasize that the Sec. 70.61(f)(2) requirements,
regarding 10 CFR Part 19 training, can be satisfied in combination with
existing training, rather than separate training solely devoted to 10
CFR Part 19, 10 CFR 70(f)(2) has been changed to read: ``Provides
training that satisfies 10 CFR 19.12(a)(1)-(5)''. To emphasize that the
training provided to satisfy Sec. 70.61(f)(2) requirements includes
making individuals aware of the risks associated with accidents
involving the licensed activities as determined by the ISA, the word
``to'' was changed to ``and,'' so that it now reads ``to these
individuals and ensures that they are aware of the risks associated
with accidents''.
Regarding the concern about the worker who leaves the controlled
area, the risk levels of Sec. 70.61 for the public pertain to any
individual, including workers, outside the controlled area. On the
other hand, with respect to the applicability of the Part 20
occupational dose limit of 0.05 Sv (5 rem)/yr TEDE, a worker can
receive an occupational dose and be subject to the Part 20 occupational
limit, regardless of his location--including activities outside the
controlled area. The ``assigned duties performed in the course of
employment'' is the distinguishing factor for radiation workers
consistent
[[Page 56213]]
with the definition of ``occupational exposure'' in 10 CFR 20. The
changes to Part 70, including the ``worker'' definition, do not affect
this. In this comment, the relationship between Part 20 annual limits
for radiation exposure and the Sec. 70.61 standards for a forward-
looking severe accident assessment have been misinterpreted. Part 70
revisions do not limit doses outside a controlled area to 1 mSv (0.1)
rem/yr.
Comment A.5: One commenter recommended that baseline criterion (8)
be rewritten as follows: ``the design of items relied on for safety
must provide for adequate inspection, testing, and maintenance, or
adequate training, testing and qualification for personnel whose
activities are relied on for safety, to ensure their availability and
reliability to perform their function when needed.''
Response: No change in rule language has been made. The baseline
design criteria are applied from the outset of new design work and are
primarily focused on physical design and facility features. The intent
is to achieve a conservatively designed facility tolerant of both
upsets and human errors. Adequate training, testing, and qualification,
as noted in the comment, will be required as management measures under
Sec. 70.62, but the NRC does not see a need for the facility physical
design to incorporate such training, testing, and qualification of
personnel.
Comment A.6: One commenter stated that the baseline criterion on
environmental and dynamic effects [Sec. 70.64a(4)] is unclear. For
example, the commenter questioned the need for a formal Equipment
Environmental Qualification Program, similar to that required under 10
CFR 50.49 and Regulatory Guide 1.89. According to the commenter, the
NRC should clarify this requirement and should not impose requirements
that may not be appropriate or necessary because of the nature of the
processes at non-reactor nuclear facilities.
Response: No change in rule language has been made. The baseline
design criterion on environmental and dynamic effects does not require
a formal Equipment Environmental Qualification Program, similar to that
required under 10 CFR 50.49 and Regulatory Guide 1.89. This criterion
applies only to new facilities and new processes and is intended to
ensure that potential ambient conditions are considered during the
design of the facility.
Comment A.7: Two commenters had concerns regarding the defense-in-
depth definition in Sec. 70.64. One commenter stated that the
definition does not reflect the defense-in-depth design philosophy as
defined in WASH-1250, ``The Safety of Power Reactor and Related
facilities,'' which outlined three levels of safety concepts in the
design of a nuclear facility. According to the commenter, the
definition presented in Secs. 70.64(b)(1) and (2) oversimplifies and
does not adequately represent the implementation of the defense-in-
depth philosophy in the design. In particular, the commenter noted that
the preference for engineered controls over administrative controls and
features that reduce challenges to items relied on for safety are only
partially implemented in the concept [of defense-in-depth]. Another
commenter agreed, stating that Sec. 70.64(b)(1) appeared unnecessarily
prescriptive by discouraging a licensee from using anything but an
engineered safety control. According to this commenter, as long as the
licensee can satisfactorily demonstrate that an administrative safety
control or a system of administrative and engineered controls will
enable the performance criteria to be satisfied, the choice of items
relied on for safety and the nature of ``defense-in-depth'' practices
that is applied should be flexible. The commenter's view is that this
flexibility, in the grading of defense-in-depth safety concepts, would
be consistent with the ability granted a licensee to grade all aspects
of its safety program [cf. Sec. 70.62(a)].
Response: With respect to the footnote to Sec. 70.64(b) that
describes defense-in-depth, which applies to new facilities and new
processes at existing facilities, the NRC staff believes that it does
reflect the defense-in-depth design philosophy as defined in WASH-1250.
Further, it reflects the Commission's current guidance on the
relationship between defense-in-depth and risk-informed regulation that
is discussed in the Commission policy white paper, ``Risk-Informed and
Performance-Based Regulation.'' With respect to Secs. 70.64(b)(1) and
(2), the NRC staff did not mean to imply that these provisions
encompassed the defense-in-depth philosophy.
Comment A.8: One commenter recommended that the emergency
capability baseline design criterion in Sec. 70.64(a)(6)(ii) address
on-site personnel (rather than all personnel). The commenter suggested
that the rule language be rewritten as ``Evacuation of on-site
personnel; and * * *.''
Response: The NRC agrees with the comment. The proposed change has
been made and is consistent with the intent of the original rule
language.
Comment A.9: One commenter stated that the criticality performance
objective in Sec. 70.61(d) is not related to Secs. 70.61(b) or
70.61(c); yet, the three conditions are all linked together. The
commenter suggested that subpart (d) should be segregated from (b) and
(c) if (d) is preserved as an independent entry (as would seem
preferable). Otherwise (d) should be subsumed under (b) and/or (c), and
the regulatory basis for criticality prevention should be predicated on
the risks and/or consequences of the accidents, rather than the
presence of initiator precursor, per se.
Response: No change in the rule language has been made. The NRC
believes that a separate performance requirement for nuclear
criticality prevention is appropriate. The NRC staff recognizes that
many (but not all) nuclear criticality accidents would reasonably be
expected to result in worker doses that exceed the high- and
intermediate-consequence standards in Sec. 70.61(b) or (c). However,
regardless of the dose directly resulting from the accident, an
inadvertent nuclear criticality should be avoided. This is consistent
with the Commission's goal to prevent inadvertent criticalities, as
reflected in the NRC Strategic Plan (NUREG-1614).
B. Content of Applications and ISA Summary
Comment B.1: One commenter stated that the rule should not
prescribe an acceptable level of detail required in the application,
but should defer this issue to the SRP. The commenter noted that,
although progress has been made in certain areas (e.g., use of language
such as ``* * * types of accident sequences * * *.''), in
Sec. 70.65(b)(6), which requires the applicant to list all items relied
on for safety for high- and intermediate-consequence accidents, the
required level of descriptive detail for items relied on for safety
(''sufficient detail'') remains vague. The commenter recommends that
information at the ``systems level'' should be required, rather than at
the ``component'' or ``sub-component'' level.
Response: The NRC disagrees with the comment. The current language
permits the description of information at a systems level provided that
there is enough detail to understand the function of the system in
relation to the performance requirements. The degree of detail provided
in the ISA Summary, with the other information available to NRC staff,
must be sufficient for the NRC staff to make the determination
specified in Sec. 70.66 (i.e., that the performance requirements of the
regulation are satisfied).
[[Page 56214]]
Comment B.2: One commenter stated that the list of items relied on
for safety should not include procedures that the personnel must
follow. According to the commenter, since procedures are constantly
being adjusted, revised, and improved, their inclusion in the list of
items relied on for safety would necessitate frequent revisions, to the
ISA Summary, that may have little if any safety significance.
Response: Sec. 70.65(b)(6) requires a list, in the ISA Summary,
briefly describing each item relied on for safety. It does not require
procedures to be listed in the ISA Summary. Therefore, the rule
language permits the approach described in the comment. Typically, the
actual personnel action would be regarded as an item relied on for
safety and this would be expected to be addressed in the ISA Summary.
Comment B.3: Two commenters had concerns about the relationship
among the ISA, the ISA Summary, and the safety program. One commenter
recommended that the NRC clarify the relationship of the ISA Summary to
the license and the safety basis to ensure consistency throughout the
rule with the intent expressed in Sec. 70.65(b). The commenter was
concerned that the wording of Sec. 70.65(a) is inconsistent with the
idea, presented in Sec. 70.65(b), that the ISA Summary will not be
incorporated in the license. The commenter suggested removing the
language in Sec. 70.65(a) that references the inclusion of the ISA
Summary in the license application, since that requirement is
adequately covered in Sec. 70.65(b). The commenter also recommended
that a discussion of management measures be included as part of the ISA
Summary. The second commenter stated that the rule implies that the ISA
Summary, as part of the safety program, is part of the license.
Further, the same commenter stated that the ``Statement of
Considerations'' erroneously states that the results of the ISA must be
submitted for NRC approval.
Response: The NRC generally agrees with the comment. The rule
language in Sec. 70.65(a) has been changed to remove the reference to
the ISA Summary and management measures. This removes the implication
that the ISA Summary is part of the license. With respect to the
relationship of the ISA Summary to the management measures, although
under the proposed rule, the elements of the ISA Summary did not
explicitly include management measures, one of the elements
[70.65(b)(4)] of the ISA Summary required information that demonstrates
compliance with the performance requirements. Such a demonstration
requires information about management measures. As suggested in the
comment, the language in Sec. 70.65(b)(4) has been clarified to
explicitly include a description of the management measures. With
regard to the comment that the ``Statement of Considerations''
erroneously ``states that the results of the ISA must be submitted for
approval'', the assertion that the ``Statement of Considerations'' is
erroneous is incorrect--the ``Statement of Considerations'' is
accurate. In response to this comment, and to clarify the role of the
ISA Summary in licensing determinations, changes have been made to
Sec. 70.62(c)(3)(ii) and Sec. 70.66. In particular,
Sec. 70.62(c)(3)(ii) has been modified to specifically state that the
ISA Summary is submitted for approval consistent with the ``Statement
of Considerations'' for the proposed rule. Section 70.66 states that
this submission will be approved if the Commission determines that
``the applicant has complied with the requirements of Sec. 70.21,
Sec. 70.22, Sec. 70.23, and Sec. 70.60 through Sec. 70.65.'' The degree
of detail provided in the ISA Summary [contents of the ISA Summary are
described in Sec. 70.65(b)] and the other information available, must
be sufficient for the NRC staff to make the determination specified in
Sec. 70.66. To supplement staff understanding of information submitted,
NRC may visit the facility during the licensing review to ensure a
sufficient safety basis for operation.
Comment B.4: Two commenters were concerned with the broad nature of
the requirement in Sec. 70.65(b)(3) that seeks information on each
process analyzed in the ISA, regardless of the risk associated with the
process. According to one commenter, the ISA Summary should only
address those processes for which accident sequences have been
identified that would produce consequences that exceed the performance
criteria of Sec. 70.61.
Response: The NRC staff needs some information on each process
analyzed in the ISA to assess completeness and quality of the
licensee's ISA process and to understand and assess the completeness
and functions of the items relied on for safety. The degree of detail
provided in the ISA Summary, together with the other information
available, must be sufficient for the NRC staff to make the
determination specified in Sec. 70.66. In addition the information is
useful in confirming the adequacy of emergency planning.
Comment B.5: According to one commenter, Sec. 70.65(b) implies that
the ISA Summary is a single document. In practice, the commenter noted
that it will be a sequence of documents that cover the facility, and if
multiple documents are submitted, they should all be in the same
format.
Response: The NRC agrees that multiple ISA Summaries are allowed
for each facility resulting in the possibility that the ISA Summary
will be a sequence of documents rather than consisting of a single
document. The definition of ISA Summary in Sec. 70.4 has been changed
to reflect this process. NRC staff approval of individual documents
(e.g., on a process basis) will be conditioned to allow for subsequent
identification of system interaction effects that may be identified
through the review of other ISA summary documents submitted.
Comment B.6: One commenter stated that the requirement, in
Sec. 70.65(b)(7), to provide information on the locations of onsite
chemicals, is unnecessary.
Response: No change in the rule language has been made. Section
70.65(b)(7) does not require information on the locations of onsite
chemicals to be submitted to the NRC. The regulation requires a
description of the proposed quantitative standards used to assess the
consequences to an individual from acute chemical exposure to licensed
material or chemicals produced from licensed material. This information
is necessary to ensure safety and is consistent with NRC's Memorandum
of Understanding (MOU) with the Occupational Safety and Health
Administration (OSHA).
Comment B.7: One commenter objected to the requirement to provide
process descriptions, noting that American Institute of Chemical
Engineers (AIChE) guidelines may result in the process being broken
into ``nodes'' or ``segments.'' The commenter suggested that the rule
should specify the descriptions of segments or nodes that could only be
combined into a process if the boundaries established for the hazard
analysis match.
Response: The NRC agrees with the comment, but does not believe a
change in the rule language is needed. The intent of the
Sec. 70.65(b)(3) requirement is to provide process information so that
the NRC staff can understand: What activities are performed at the site
that involve hazardous materials associated with or produced from
licensed radiological material, including any use, storage,
manufacturing, or handling of those materials; what was analyzed in the
ISA; and the hazards identified in the ISA. The AIChE guidelines use
the term ``process nodes'' with respect to Hazard and Operability
Analysis (HAZOP) and define it as ``sections of equipment with definite
boundaries
[[Page 56215]]
* * * within which process parameters are investigated for deviations *
* *.'' In HAZOP analyses, the term ``node'' designates a pipeline or
vessel that has a common design intent. In meeting the Sec. 70.65(b)(3)
requirement, several nodes may be combined.
C. Safety Program
Comment C.1: Three commenters questioned the narrow definition of
the safety program that is presented in Sec. 70.62(a) and recommended
deleting it from the rule language. According to the commenters, the
safety program is broader than the three elements identified in
Sec. 70.62(a)(1) as process safety information, ISA, and management
measures. The commenters noted that fuel cycle facility safety programs
encompass the three elements identified plus all the other topics
addressed in the license application. This includes, for example,
radiation safety, criticality safety, chemical safety, and fire
protection, in addition to the three elements directly associated with
the ISA.
Response: The NRC staff agrees in principle with the comment. The
term ``safety program,'' as used in Sec. 70.62 (a), is related to the
elements needed to demonstrate compliance with the performance
requirements in Sec. 70.61. This safety program consists of process
safety information, ISA, and management measures. There is no intent to
indicate that these elements represent the total safety program at the
facility. Therefore, the rule language was clarified by changing ``The
three elements of the safety program; namely process safety
information, integrated safety analysis, and management measures, are
described in paragraph (b) through (d) of this section * * *'' to
``Three elements of this safety program; namely process safety
information, integrated safety analysis, and management measures, are
described in paragraph (b) through (d) of this section.''
Comment C.2: One commenter stated that the current proposed rule
offers sufficient flexibility in selecting ISA methodology so that a
broad spectrum of facilities can be addressed and such that licensees
have flexibility to interface with their site processes, procedures and
resources.
Response: The Commission agrees with the comment; therefore, no
change was made to the rule language with respect to ISA methodologies.
The final rule offers sufficient flexibility in selecting an ISA
methodology that can be used to analyze a facility's site, processes
and procedures.
Comment C.3: Two commenters were concerned about the implementation
of the final rule, and, in particular, the time frame for compliance
with those aspects of the rule not related to the completion of the ISA
and the submittal of the ISA Summary. One commenter, citing the
experience when Part 20 was revised, recommended an effective date
sufficiently far into the future so programmatic changes could be
implemented at the operating facilities and any necessary conforming
license amendments could be completed. Regarding the latter issue, both
commenters cited 10 CFR 20.1008 as an example of how potential
contradictions between license applications and regulations could be
addressed. One commenter recommended including an additional provision
of this type, especially in light of license conditions that have been
added to licenses recently renewed by the NRC.
Response: The NRC agrees with the comment. In Sec. 70.76(a), it
states that ``this provision shall apply for subpart H requirements as
soon as the NRC approves that licensee's ISA Summary pursuant to
Sec. 70.66. For requirements other than Subpart H, this provision
applies regardless of the status of the approval of a licensee's ISA
Summary.'' In addition, Appendix A was revised to include the
following: ``Licensees must comply with reporting requirements in this
appendix, except for (a)(1), (a)(2), and (b)(4), after they have
submitted an ISA Summary in accordance with Sec. 70.62(c)(3)(ii).
Licensees must comply with (a)(1), (a)(2), and (b)(4) after October 18,
2000.'' In addition, Sec. 70.62(c)(3)(ii) was revised to further
clarify implementation schedules for existing licensees.
Comment C.4: Two commenters stated that a graded approach should be
used in determining the management measures that need to be applied to
items relied on for safety. One commenter recommended that the language
in Sec. 70.62(d) should be changed as follows: ``The measures applied
to a particular engineered or administrative control or control system
may be graded commensurate with the reduction of the risk attributable
to that control or control system.'' The other commenter recommended
that other factors besides risk including consequences, life cycle, and
magnitude of hazard involved, should be used to determine appropriate
management measures.
Response: The NRC agrees with the comment and has made the
suggested change to the rule language in Sec. 70.62(d). Regarding the
question of considering other factors besides ``risk,'' the NRC notes
that the grading of measures to consequences, life cycle, and magnitude
of hazard, is part of grading the measures to risk. The phrase used in
the rule--``commensurate with the reduction of risk attributable to
that item''--does not imply requiring a quantitative determination of
the risk significance of any particular item relied on for safety. The
rule is non-prescriptive regarding the grading approach and criteria to
be used, allowing applicants to propose such details.
Comment C.5: One commenter stated that the 4-year period for
conducting the ISA and for modifying the facility to address any
identified unacceptable performance deficiencies may be too short and
recommended a 5-year period instead. According to the commenter, a 5-
year time-frame would be consistent with the time allowed for existing
licensees that have committed, by license condition, to perform ISAs.
The commenter also recommended that the period should start on the date
when the NRC approves the plan required in Sec. 70.62(3)(i), noting
that if the clock starts on the effective date of the rule and the NRC
takes one year to approve the ISA plan, the licensee will be unduly
hampered. In addition, the commenter stated that there should be some
incentive for the NRC to complete its approval process in a timely
manner and recommended imposition of a 90-day limit for NRC to issue a
decision on the acceptability of a licensee's ISA approach. The
commenter also recommended that appropriate and sufficient time be
allowed for the licensee to present a plan to the NRC and to implement
the plan to correct any identified unacceptable performance
deficiencies.
Response: Regarding the proposal for a 5-year period for conducting
the ISA and correcting all unacceptable deficiencies, the NRC believes
that the 4-year period proposed in the proposed rule is reasonable.
However, NRC recognizes that there may be some instances where
modifications resulting from the ISA cannot be completed within the 4
years specified and has modified Sec. 70.62(c)(3)(ii) to accommodate
these instances by clarifying that NRC may approve extensions for
reasons that are beyond the control of the licensee. Regarding the
licensee being unduly hampered because of the time required for the NRC
staff to approve the plan required by Sec. 70.62(c)(3)(i), the NRC
staff expects to complete the licensing review within 90 days, assuming
that the information submitted is complete. However, the time it takes
the NRC to approve the
[[Page 56216]]
plan will depend on the quality of the plan submitted by the licensee.
In addition, current industry development of an ISA Summary guidance
document should facilitate the licensing review process.
Comment C.6: One commenter stated that the plan required in
Sec. 70.62(c) which should be submitted within 6 months of the
effective date of the rule, should pertain only if a licensee has not
already completed the actions outlined in Sec. 70.62(c)(3)(ii).
Response: The implementation plan and the ISA must satisfy the
requirements in the final rule. If the actions outlined in
Sec. 70.62(c)(3)(ii) have been completed, then all that would be
required to satisfy Sec. 70.62(c)(3)(i) is submission of a description
of any additional work that must be performed to meet the requirements
in Subpart H of the rule, or a confirmation that the work submitted
meets the requirements in Subpart H of the rule.
Comment C.7: Four commenters disagreed with the requirement in
Sec. 70.62(a)(3) to establish and maintain a log of failures of items
relied on for safety. One commenter stated that the requirement should
be rewritten to be performance-based rather than prescriptive. The
commenter noted that most licensees have an incident reporting and
corrective action system, which is used for all activities at the
facility. As long as these systems meet the performance objective, it
seems unnecessary for the rule language to be prescriptive in how it is
met. Another commenter agreed, stating that it is inappropriate to
impose this extra record-keeping burden on the licensee, because the
licensee already has to generate records of this nature to manage its
business and another different log is unnecessary work. Another
commenter noted that because of the reporting requirements of
Sec. 70.62(a)(2) and Sec. 70.74(a)(1), the NRC will already possess all
of the information sought in the ``log'' of Sec. 70.62(a)(3).
Response: The NRC generally agrees with the comment that
maintenance of the failure log would be unnecessarily prescriptive.
Regarding the concern about prescriptiveness, the rule has been revised
to eliminate the requirement for licensees to establish and maintain a
specific log of information developed and maintained elsewhere.
However, the final rule requires that data be readily retrievable and
available. This information is necessary to evaluate the reliability
and availability of items relied on for safety, the likelihood of
failure of the items, and the effectiveness of management measures
implemented by the licensee. The NRC also anticipates this information
will be reviewed during periodic inspections by NRC as part of the
revised oversight process that is being developed. Regarding the
redundancy of reporting, the rule currently requires the licensee to
report only any loss or degradation of items relied on for safety that
results in failure to meet the performance requirements of Sec. 70.61.
The requirements of Sec. 70.62(a)(3) include a much broader set of
items, including all items relied on for safety or management measures
that have failed to perform their function.
D. Change Process, License Renewal, and Backfit
Comment D.1: Five commenters were concerned about the requirement
in Sec. 70.72(d)(1) to submit changes to the ISA Summary every 90 days.
Two commenters stated that an annual update [similar to the annual
Final Safety Analysis Report updates for reactors per 10 CFR 50.71(e)]
should suffice, considering that the potential consequences of reactor
accidents are significantly greater than those at fuel cycle
facilities. One commenter stated that an annual update to the ISA
Summary would be consistent with the reporting requirements (for
changes to records) of Sec. 70.72(d)(3). Another commenter stated that
the 90-day reporting of changes is entirely too frequent, which would
mean that the facility and the NRC would always have change reporting
in progress. According to the commenter, there is no need for NRC to
have this ``real-time'' knowledge; rather, it is only important that
the licensee have ``real-time'' knowledge. The commenter noted that the
NRC only needs reasonably current knowledge, because the current ISA is
available and accessible at the site. The commenter believes that a 12-
month to 24-month update for reporting, as used in other places, is
satisfactory and more efficient, noting that this seems clearly
justified based on the fact that all the information is available at
the site and accessible to the NRC at any time.
Response: The NRC agrees with the comment that submitting updates
to the ISA Summary to the NRC can be less frequent than required in the
proposed rule. The final rule requires only annual reporting within 30
days after the end of the calender year during which the changes
occurred.
Comment D.2: One commenter noted that, under Sec. 70.72, the NRC
should define ``periodically'' in the context of reporting of changes
made to SSCs etc.
Response: The NRC determined that no change to the reporting
requirements is necessary in response to the comment. The comment
referenced language in the ``Statement of Considerations'' not the
rule. The specific reporting requirements were defined in the proposed
rule and are included, as revised, in the final rule.
Comment D.3: Two commenters were concerned about the footnote in
Sec. 70.72(c) that attempts to explain new types of accident sequences.
Both commenters stated that the language in the footnote would require
nearly all process changes to be approved by NRC through a license
amendment, which would be in conflict with the overall objectives for
the proposed rule. Both commenters recommended that the footnote be
deleted.
Response: The NRC agrees that the footnote did not successfully
clarify the definition of ``new types of accident sequences.'' Thus,
the footnote has been deleted from the final rule. The NRC staff will
develop a guidance document, with input from stakeholders, to describe
an acceptable change process that meets the requirements of the final
rule in more detail. The degree of detail provided in the ISA Summary,
together with the other information available, must be sufficient for
the NRC staff to make the determination specified in Sec. 70.66. In
addition, the NRC staff had added a discussion to Chapter 3 of the SRP
to describe an acceptable level of detail in the identification of the
types of accident sequences.
Comment D.4: Three commenters were concerned about the
requirements, in Sec. 70.72, regarding configuration management and the
overly broad process for making changes at licensed facilities. One
commenter stated that the requirements, as written, apply to all site,
structures, processes, systems, equipment, components, computer
programs, and activities of personnel, regardless of safety
significance. The commenter noted that compliance with these
requirements would appear to require configuration management and
change control applied to everything on the site of the licensed
facility; that could include the wastewater treatment facility, a laser
facility, the administration building, maintenance of the shrubbery,
etc. Every change would require an evaluation and a summary submitted
to the NRC, even though inclusion in the change control process would
make no contribution to the safety of licensed operations and would
impose an undue burden on the licensee. To remedy this, the commenter
recommended that the configuration and change process be limited to any
``changes to the site, processes or items relied on for safety as
described in the
[[Page 56217]]
ISA Summary.'' Another commenter agreed, stating that the requirement
is too broad and all-encompassing and would require configuration
management evaluation of changes having no or absolutely minimal effect
on health and safety (e.g., office remodeling, planting of shrubbery,
changing paint colors). The commenter suggested that rather than
control every change by means of configuration management, the licensee
should first rely on internal procedures to screen any proposed changes
initially for their potential safety significance.
Response: No change in the rule language has been made. The
emphasis of this requirement is clearly on licensed operations and the
associated safety controls. If a licensee has established a
configuration management system in accordance with Sec. 70.72(a), it is
important the licensee use the system to evaluate every change made at
a facility that could affect safety (i.e., generally not shrubbery,
paint color) to ensure that any impacts from those changes on the
safety of operations is identified, considered, documented, before
implementing the change. In some cases, the analysis would be trivial
because no known hazards would be involved in the change (e.g., certain
changes in the administration building, or changes to shrubbery). Often
it is clear that there are no safety implications associated with the
proposed change. However, there may be special cases in which
apparently minor changes could adversely affect safety, such as
installation of a drinking water fountain in a radiological control
area. In addition, every change which is assessed in the configuration
management system does not need to be submitted to the NRC. Section
70.72(d)(2) states that only those changes to records required by
Sec. 70.62(a)(2) need to be submitted. These would include changes to
the process safety information, ISA, and management measures. In
addition, with respect to the use of an ``initial screening''
mechanism, the NRC staff considers an initial screening to assess the
safety impact of a change to be part of an evaluation, as called for in
Sec. 70.72(a). In some cases, this screening will be sufficient.
Comment D.5: One commenter stated that Secs. 70.72(c)(1)(i),
(c)(2), and (c)(3) are wrong to use the ISA Summary as the decision-
making document. The commenter noted that the ISA, the detailed
licensee-generated information and evaluations that the licensee uses
to manage its program, comprises the information base for decisions.
Summaries only provide a general level of information about the more
important elements of the safety system for operations as determined
under the licensed program.
Response: No change in the rule language has been made. The ISA
Summary is prepared based on the ISA, and contains key information that
is directly related to facility safety, such as a list of items relied
on for safety, a description of hazards identified in the ISA, and a
general description of the types of accident sequences. The contents of
the ISA Summary are described in Sec. 70.65(b). The NRC staff could
review the adequacy of changes using the ISA instead of the ISA
Summary, but this approach would require submission of a greater amount
of information to NRC and would pose an unnecessary burden on the
licensee. (Also, see response to Comment B.3.)
Comment D.6: Two commenters are concerned about the annual
requirement in Sec. 70.72(d)(3) to submit a brief summary of all
changes to the records required by Sec. 70.62(a)(2). According to one
commenter, the submittal would cover process safety information
[Sec. 70.62(b)] including procedures, drawings, and detailed equipment
lists. The commenter does not believe the NRC requires a summary of
changes to this type information. A second commenter agreed, stating
that the wording of this section will inadvertently and significantly
expand the information that would have to be reported. In particular,
the view was expressed that Sec. 70.72(d) would require the licensees
to submit voluminous information that could include the update to
process safety information, including drawings, flow process diagrams,
and piping and instrumentation diagrams. The commenter suggested that
this section should be reworded to read: ``a brief summary of all
changes to the integrated safety analysis and ISA summary, that are
made without prior Commission approval, must be submitted to the NRC
every 12 months * * *.''
Response: No change in the rule language has been made. The
regulation currently requires submission of ``* * * a brief summary of
all the changes to the records required by Sec. 70.62(a)(2) * * *''
This does not require the submittal of actual charts and drawings but a
written summary of the changes made. For the reasons cited in the
response to comment D.1, it is important that the NRC be knowledgeable
of changes made to this information.
Comment D.7: One commenter noted that, unlike Sec. 50.59, the
requirements of Sec. 70.72 do not call for the submittal of a brief
description and summary safety evaluation for each change. The
commenter believes that the NRC would benefit from a description of
changes made to the ISA Summary. Accordingly, Sec. 70.72 should require
brief descriptions and summary safety evaluations of each change made
pursuant to Sec. 70.72 and require that an updated ISA Summary be
provided on a biennial basis.
Response: No change in the rule language has been made. The brief
summaries of changes submitted under the requirements of
Sec. 70.72(d)(2) would be expected to include an explanation of each
change, the reasons why the change was made, and why it did not require
pre-approval. This information will be included in a guidance document
to be developed. The NRC staff views this as sufficient and does not
anticipate the need for licensees to submit a summary safety evaluation
for each change, as long as each change has been made in accordance
with the final rule and the approved process.
Comment D.8: Two commenters questioned the current timeframe (10
years) or the need for renewal of licenses, suggesting that the new
rule, in effect, resulted in a ``living license.'' One commenter stated
that if a ``living license'' is truly the outcome as described in the
Supplementary Information, renewal periods as long as 20 years would be
appropriate. The other commenter noted that with updates required every
12 months there is no real need for the NRC to renew the license--it
only becomes a maintenance chore to confirm periodically that the
licensing basis remains intact. The commenter believes that the living
license concept provides advantages for the NRC and the licensee.
Response: Although the NRC generally agrees with those comments, no
change in the rule language has been made. A specific time period for
renewals is not specified in Part 70 and to establish one in the rule
would require consideration of many factors, such as compliance with
the National Environmental Policy Act and the impact of the loss of
commitments linked to license renewal, that were not addressed in the
current rulemaking (e.g., financial assurance for decommissioning).
Establishment of a new term for licenses (e.g., 20 years) in 10 CFR 70
would require an analysis of these factors and an opportunity for
public comment. The NRC staff will evaluate whether a longer term for
fuel cycle licenses is appropriate in light of the new requirements in
Subpart H. In any case, even if NRC ultimately declines to extend the
term of the fuel cycle licenses (nominally 10 years), the burden of
license renewal should be
[[Page 56218]]
significantly reduced because the licensee will be required to maintain
current the ISA Summary, items relied on for safety, and management
measures.
Comment D.9: Five commenters recommended that a backfit provision
similar to that in 10 CFR 50.109 or 10 CFR 76.76 should be included in
the final rule. One commenter stated that the backfit provision should
apply to current proposed changes at existing facilities. Another
commenter stated that the backfit provision should be immediately
effective for those processes or parts of an existing facility for
which an ISA has been completed. A third commenter favored an
immediately effective backfit provision. However, as an alternative,
the commenter would make the provision effective for facilities or
systems for which the ISA has been completed and the ISA Summary
submitted to the NRC. A fourth commenter stated that deferring
consideration of a backfit provision would be evading an extremely
important issue, expressing the view that it is vital that a formal,
systematic, and disciplined review of new, changed, or differing
positions that could backfit existing facilities be applied to increase
regulatory certainty. According to the commenter, no change to the
backfit language in 10 CFR 50.109, which has been used successfully to
control backfits at power reactors in the past, is needed to allow for
qualitative analysis. 10 CFR 50.109, which the commenter endorses, is
viewed as neither a quantitative nor a qualitative backfit provision.
In contrast to the statement made in the Statement of Considerations of
the proposed rule, the commenter does not believe that a comprehensive
risk baseline is necessary before reasoned judgments can be made on the
benefits and risks of a proposed backfit.
Response: The Commission agrees that regulatory stability and
certainty can be improved by establishing a backfit provision for fuel
cycle facilities covered by Subpart H of the final rule. Consequently,
NRC has included a backfit provision in the rule in Sec. 70.76. The
wording of Sec. 70.76 is similar to the current language in Sec. 76.76.
For requirements other than Subpart H, this provision will apply
immediately after NRC publication of backfit guidance. For Subpart H
requirements, this provision shall apply for a licensee as soon as the
NRC approves that licensee's ISA Summary pursuant to Sec. 70.66. The
NRC will publish guidance that will address the qualitative versus
quantitative analysis issue and consideration of chemical risks. The
NRC staff anticipates completing this guidance within six months of the
publication of the final rule. Under the Sec. 70.76 backfit provision,
a backfit analysis is not required for modifications necessary to bring
the facility into compliance with the rule, including the performance
requirements in Subpart H. The subject of backfit is discussed in more
detail in an attachment to SECY-00-0111.
E. Definitions
Comment E.1: One commenter recommended a change (from 4 percent to
5 percent enrichment) in the definition of a critical mass of SNM to
reflect the higher enrichments that are currently in use.
Response: The definition of critical mass of SNM in Part 70 is used
solely to determine when Subpart H applies. To emphasize this point,
the definition was changed to include the phrase, ``for purposes of
subpart H.'' The definition, including the 4 percent figure, is
identical to that used in Sec. 70.24, which requires criticality
accident alarms and other related measures.
Comment E.2: Regarding the issue of ``reasonable assurance,'' two
commenters stated that, in the definition of available and reliable to
perform their function when needed, the use of the term ``ensure''
implies a level of certainty that is unrealistically high. Both
commenters recommended replacing the term ``ensure'' with the term
``provide reasonable assurance.'' One commenter also recommended
removing the word ``continuous'' from the definition, which would now
read ``* * * means that * * * items relied on for safety will perform
their intended safety function when needed and management measures will
be implemented to provide reasonable assurance of compliance with the
performance requirements of Sec. 70.61.''
Response: The definition was revised to remove the word
``continuous,'' but no change was made regarding ``ensure.'' With
respect to ``ensure,'' the proposed rule language does not indicate a
level of certainty that is unrealistic. The term ``ensure'' is used
extensively throughout NRC's regulations in the context of a licensee's
obligations to connote ``make sure'' or ``make certain.'' Specifically,
elsewhere in Part 70 alone, the term is used in this context eight
times: Secs. 70.24(a)(3), 70.32(j), 70.38(g)(4)(iii), 70.51(a)(10),
70.52(c), 70.57(b)(3), 70.57(b)(4), and 70.57(b)(6). Whereas, the term
``reasonable assurance'' is used just once in Part 70, in
Sec. 70.23(b), to describe the level of assurance that the Commission
must find in order to approve construction. The use of ``ensure'' in
the definition of ``available and reliable to perform their function
when needed''' in Sec. 70.4 is appropriate. In short, licensees
``ensure'' and the Commission determines ``with reasonable assurance.''
Regarding the issue of ``continuous compliance,'' the definition of
``available and reliable'' in Sec. 70.4 has been modified to delete the
word ``continuous.'' This change recognizes the concept that a failure
of an item relied on for safety does not automatically infer a failure
to meet the performance requirements of Sec. 70.61. In addition, the
NRC recognizes that items relied on for safety may temporarily not be
available (i.e., not continuous) when taken out of service for
maintenance or functional testing; however, the performance
requirements must still be met. A discussion has been added to Chapter
3 in the SRP to address the relationship of failures of items relied on
for safety to meeting the performance requirements.
Comment E.3: One commenter stated that there is a ``disconnect''
regarding the definition of the term items relied on for safety and
recommends that the term be replaced by the term Measures relied on for
safety.
Response: The reason for the comment is not clear, but perhaps the
commenter objects to the use of the term ``item'' to refer to a
personnel action. Part 70 does, in fact, allow human actions to be
items relied on for safety and permits flexibility in determining how
the items and measures are defined. Consequently, the Commission has
retained the original text in the final rule. (See related Comments
B.2. and E.4.)
Comment E.4: One commenter was concerned that the term items relied
on for safety includes ``activities of personnel,'' and proposed
changing the definition in 70.4 to limit items relied on for safety to
``structures, systems, equipment, and components.'' According to the
commenter, it is reasonably straightforward to classify physical items
as being relied upon for safety, and to apply graded quality assurance
controls, including management measures, to design, construction,
operation, and maintenance, etc., of those physical items, based on
their respective safety functions. The commenter stated that it can be
confusing to try and classify and grade items when they include
``personnel activities,'' since an activity has little importance
absent the context of its influence on a physical item's safety
function. Removing ``personnel activities'' from the definition of
items relied on for safety would not limit their
[[Page 56219]]
importance but rather, would put activities in context with the
structures, systems, equipment, or components to which they are
related, without necessitating a change in the balance of the proposed
rule. The commenter stated that removing personnel activities from the
definition of items relied on for safety will also help address the
concern raised (in comment B.2) regarding the treatment of procedures
as items relied on for safety.
Response: No change was made to the rule language. Human actions
that are relied on to prevent an accident (i.e., administrative
controls) are as important as the ``physical items'' needed to prevent
an accident. Just as there are measures (e.g., maintenance,
configuration management) needed to ensure the availability and
reliability of physical controls, there are analogous measures (e.g.,
training, procedures) needed to ensure the availability and reliability
of human actions. Graded approaches that can be applied to the
maintenance of a physical control depending on the risk significance of
the control could also be applied to the training of workers who
perform safety functions, depending on the risk significance of the
human's actions. Although the reliability of engineered controls may be
higher than administrative controls, the final rule allows licensees
the flexibility to employ both engineered as well as administrative
controls.
Comment E.5: One commenter stated that the NRC should define the
terms likely, unlikely, highly unlikely, and credible in the rule so
that there will be one set of definitions applied to all nuclear fuel
facilities. The commenter stated that this will minimize the
interpretation and application of these terms in the ISA.
Response: No change in the rule language has been made. Part 70
applies to different types of fuel cycle facilities, some of which are
more complex and have more accident sequences than others. Accordingly,
since the application of the terms in the rule will be necessarily
specific to the individual context in which they are applied, the
development of a definition for these terms in the rule language is
impracticable. The Commission, however, will provide general guidance
on the application of the terms unlikely and highly unlikely in the SRP
to aid licensees in implementing the provisions of the rule.
Comment E.6: One commenter recommended a change in the definition
of worker. In particular, the language ``* * * exposure to radiation
and /or radioactive material from licensed and unlicensed sources of
radiation'' would be replaced with ``* * * exposure to radiation and /
or radioactive material from licensed sources of radiation, and
radiation from man-made non-regulated sources (e.g., an individual).''
As originally defined, persons who are subject to occupational doses
from natural sources of radiation, (e.g., airline pilots and astronauts
subject to high cosmic background might be included, whereas workers
involved with the possession or use of unlicensed radioactive materials
might not be). The commenter stated that the proposed change removes
this source of confusion.
Response: The NRC staff agrees in principle with the comment.
However, the commenter's proposed change does not eliminate the
confusion (e.g., some man-made unlicensed sources of radiation are part
of background or otherwise not included in occupational doses as
defined in NRC's radiation protection standards in 10 CFR 20). Instead,
in response to the comment, the definition in Sec. 70.4 was changed to:
Worker, as used in Subpart H, means an individual who receives an
occupational dose as defined in 10 CFR 20.1003.
F. Miscellaneous
Comment F.1: One commenter recommended that the criticality
requirements of Sec. 70.24 be revised to permit alternate criticality
control provisions to be accepted for DOE facilities without requiring
an exemption.
Response: Comments on Sec. 70.24 are outside the scope of the
rulemaking.
Comment F.2: Two commenters recommended changes in the
decommissioning requirements of Secs. 70.22(a)(9) and 70.38. In
particular, one commenter recommended that the timeliness and schedule
provisions in the decommissioning requirements of Sec. 70.38 be revised
to include separate requirements for DOE facilities.
Response: Comments on Secs. 70.22(a)(9) and 70.38 are outside the
scope of the rulemaking.
Comment F.3: One commenter expressed concern with the language in
Sec. 70.23(b), which states that the Commission will approve
construction of a plutonium processing and fuel fabrication facility
only after determining that the design bases of SSCs, and the attendant
quality assurance program are adequate to protect against natural
phenomena and the consequences of potential accidents. In particular,
the commenter stated that this provision, as written, seems contrary to
other changes being proposed under the draft rule, because it addresses
consequences of potential accidents, as opposed to the risk associated
with credible accidents.
Response: Section 70.23(b) has not been modified in this
rulemaking. The reference to ``consequences'' in the rule language does
not preclude a risk-informed approach in satisfying this requirement.
The NRC will need to consider the risk, and thus the likelihood of
consequences of potential accidents occurring, in order to determine
whether there is reasonable assurance of protection against such
consequences. This consideration of risk will be important in
determining the need for (and the ability of) the applicant to reduce
the likelihood of accidents and to mitigate their consequences.
Comment F.4: One commenter recommended that Sec. 70.11 be revised
to reflect the applicability of NRC authority over a MOX fuel
fabrication facility owned by the DOE, pursuant to changes in law last
year.
Response: The NRC agrees with the comment, but believes a separate
rulemaking is required. Since October 17, 1998, when the amendment to
Section 202 of the Energy Reorganization Act of 1974 was enacted,
Sec. 70.11, as well as several other subsections of the regulations,
need to be updated to reflect this legislative change. However, to
address this subsection and all the other instances, and to avoid the
necessity for potential future revisions of this type, the NRC intends
to institute an administrative-type rule amendment to conform all of
the references to Section 202 in the regulations, including Sec. 70.11,
to merely cite Section 202, rather than repeat the text of that
section. Because this rule change affects various parts of the
regulations, it will be conducted independently of the current Part 70
amendments.
Comment F.5: One commenter stated that as additional DOE facilities
are licensed by the NRC under Part 70, the NRC should ensure that the
requirements address the full range of fissionable and fissile
materials at these facilities.
Response: This issue is beyond the scope of the rulemaking. It will
be addressed, if necessary, in the future.
Comment F.6: One commenter agreed that the proposed rule is
entirely consistent with the U.S. Environmental Protection Agency's
Risk Management Program regulations and the general duty clause of the
Clean Air Act, and contains appropriate complementary safety measures
for facilities possessing a critical mass of SNM.
[[Page 56220]]
Response: No response necessary.
Comment F.7: One commenter strongly recommended that the NRC adopt,
by reference, the 1998 edition of National Fire Protection Association
(NFPA) 801, ``Facilities Handling Radioactive Materials.'' NFPA 801
would apply to Sec. 70.62, ``Safety program and integrated safety
analysis,'' that addresses protection from all relevant hazards,
including radiological, criticality, fire, and chemical. The NFPA
standard would also apply to Sec. 70.64, ``Requirements for new
facilities or new processes at existing facilities,'' that addresses
fire protection. The reference to NFPA 801 is in keeping with the
requirements of Public Law 104-113, ``National Technology Transfer and
Advancement Act,'' that requires Federal agencies to use private
sector-developed national consensus technical standards in carrying out
public policy, wherever appropriate.
Response: The suggested change would be an unnecessarily
prescriptive rule requirement. Instead, the NRC identifies the
standards in NFPA 801 and 600 as an acceptable approach for
demonstrating compliance with 10 CFR Part 70 in the SRP.
Comment F.8: One commenter noted that the proposed rule
incorporates the current terms of the MOU between the NRC and OSHA.
This should avoid misunderstanding and result in more effective
implementation for all concerned parties.
Response: Although the rule is consistent with the NRC-OSHA MOU,
the rule itself does not incorporate the terms of the MOU.
Nevertheless, the NRC agrees with the spirit of the comment.
Comment F.9: Two commenters expressed concern over those portions
of Secs. 70.22 and 70.23 of the existing rule that address the
regulation of plutonium processing and fuel fabrication facilities. One
commenter asked if Sec. 70.22 (f) should be coordinated with
Sec. 70.65. The commenter noted that it is not clear if the
requirements are collateral, complementary, or redundant. The same
commenter stated that Sec. 70.23(b) should be examined to clarify the
need for this requirement in light of similar information being
submitted pursuant to Sec. 70.65. The second commenter agreed, stating
that Sec. 70.22(f) requires plutonium-related applicants to provide
information on the facility site and design basis of principal SSCs,
etc., as part of the license application. The commenter believed that
this information is also required in other sections of the revised
rule, and thus is redundant.
Response: No change in the rule language has been made. The
requirements are not viewed as redundant, considering: the timeframe
for submittal of information required by the two sections could be
different; and Sec. 70.23(b) contains a requirement for NRC
construction approval before the start of construction.
Comment F.10: Two commenters were concerned about the construction
authorization provisions in Secs. 70.23(b) and 70.23(a)(7). According
to one commenter, irrespective of Sec. 70.65, the construction
authorization provision in Sec. 70.23(b) appears to be an unnecessary
step and should be considered for deletion by the NRC. If the NRC
chooses to retain Sec. 70.23(b), the NRC should clarify how the
authorization process would be conducted, given that the procedural
step has never been exercised. Furthermore, the NRC should identify how
the ``design basis'' authorization is defined, why it is necessary, and
how it relates to the ISA. The second commenter noted that
Sec. 70.23(a)(7), which applies to other Part 70 licensees, allows
construction to commence based on a conclusion by the Director, NRC
Office of Nuclear Material Safety and Safeguards, that environmental
impacts have been appropriately addressed. The commenter stated that
this discretion afforded the NRC under Sec. 70.23(a)(7)--i.e., NRC's
authority over construction associated with ``any * * * activity which
the Commission determines will significantly affect the quality of the
environment'' is adequate to ensure the sufficiency of information
provided to the NRC to authorize or disallow construction. The
commenter proposes that Sec. 70.23(a)(7) be clarified for applicability
to plutonium facilities, and that Secs. 70.22(f), 70.23(a)(7),and
70.23(b) be eliminated. Doing so would avoid the preconception that,
irrespective of design features and material composition, plutonium is
``more special'' than other SNM.
Response: No change in the rule language has been made. The Atomic
Energy Commission specifically established these requirements (see 36
FR 9786; May 28, 1971 and 36 FR 17573; September 2, 1971) for plutonium
facilities in recognition of the potential exposures and ground-
contamination levels that may result if only a small fraction of the
dispersible plutonium in process were released (see SECY-R 188, March
17, 1971). The current revisions to Part 70 do not impact this section
and therefore, the suggested change is outside the scope of the
rulemaking. Regarding the authorization process, the NRC staff has
clarified this process in a letter to Duke, Cogema, Stone & Webster,
dated September 10, 1999. The design basis was also identified in this
letter. NRC provided additional guidance on this process in the draft
Standard Review Plan for a Mixed Oxide Fuel Fabrication Facility. In
addition, the NRC staff is currently assessing the opportunities for
hearings associated with the review of a license application for a
plutonium processing facility and may offer additional guidance on this
topic later in 2000.
Comment F.11: One commenter noted that, under Sec. 70.23(a)(8), the
NRC will approve a plutonium facility's license application only after
construction of principal SSCs has been completed in accordance with
the application. Certainly this is not a requirement unique to
plutonium facilities. The NRC already has the authority to grant
licenses conditional on successful completion of certain actions (such
as successful start-up testing, training, etc.). Completion of
construction in accordance with the license application seems such an
obvious condition that this specific provision seems redundant and
therefore unnecessary.
Response: The NRC established Sec. 70.23(a)(8) specifically for
plutonium processing facilities. Because the current revisions to Part
70 do not impact this section, comments regarding this section are
outside the scope of the rulemaking. See response to Comment F.10.
Comment F.12: One commenter noted that the terminology in Appendix
A (b)(1) clearly ties the failure to the performance requirements. The
phrase, ``and which results in failure to meet the performance
requirements of Sec. 70.61'' is very clear. This phrase should be
consistently included in Appendix A (b)(2)-(5) using the exact same
wording.
Response: No change in the rule language has been made. The linkage
to the failure to meet the performance requirements is already included
in Appendix A (b)(2) and (b)(3). For the events described in Appendix A
(b)(4) and (5), the NRC staff desires to be informed when such events
occur, regardless of the licensee's determination with respect to the
performance requirements. In these cases, the NRC staff will
independently confirm the licensee's assessment of whether the
performance requirements were met, on the basis of the information
reported.
Comment F.13: One commenter stated that the reporting requirements
of Sec. 70.50 continue to misrepresent the principles of the 1988 NRC-
OSHA MOU. Section 70.50(c)(1)(iii)(A)
[[Page 56221]]
requires the reporting of chemical hazards and Sec. 70.50(c)(1)(iii)(B)
requires the reporting of personnel exposures to chemicals. According
to the commenter, although the MOU principles have been correctly
incorporated into other proposed revisions to Part 70 (e.g.,
Secs. 70.4, 70.61(b), 70.62(c), 70.64(a), 70.74, and Appendix A), they
were incorrectly referenced in Sec. 70.50. MOU principle (2) limits NRC
jurisdiction to regulation of chemical hazards of licensed material and
hazardous chemicals produced from licensed material. The two
aforementioned sections of Sec. 70.50 should be corrected to properly
incorporate the MOU principles.
Response: The rule was revised in response to the comment to
reflect more precisely the language in the NRC-OSHA MOU.
Comment F.14: One commenter noted that applicants for licenses to
operate new facilities or new processes at existing facilities would be
expected (``Statements of Consideration,'' 64 FR 41346) to update their
ISAs, based on as-built conditions, and submit the results to the NRC
before operation. The process for uranium enrichment facilities that
must comply with Sec. 70.23a would differ from this description.
Uranium enrichment facilities would submit a complete license
application, including an ISA Summary, for construction and operation.
This application would be the basis for NRC review, and culminate in
issuance of a license for construction and operation. After issuance of
the license, the licensee would institute change control under
Sec. 70.72. The licensee would then be required to submit summaries of
changes and ISA Summary updates as required by Sec. 70.72. An
inspection would verify that the facility has been constructed in
accordance with the license, before operation, as required by
Sec. 70.32(k). No pre-operational submittal and review of an updated
ISA Summary are anticipated for uranium enrichment facilities because
their configuration would be controlled after issuance of the
construction and operation license.
Response: As recognized by the commenter, no changes to the rule
are necessary. The differences in the licensing process for enrichment
facilities other than the gaseous diffusion plants (regulated under 10
CFR Part 76) reflect the process mandated in Section 193 of the Atomic
Energy Act of 1954, et. seq.
III. Changes from the Proposed Rule
Subpart A--General Provisions
Authority
This section has been changed to reflect the redesignations of
Secs. 70.61 and 70.62 as Secs. 70.81 and 70.82, respectively.
Section 70.4 Definitions
The definition of ``available and reliable to perform their
function when needed'' has been modified to eliminate the need to
maintain ``continuous'' compliance with the performance requirements of
Sec. 70.61. The definition of ``configuration management'' has been
modified to clarify its role as a ``management measure.'' The
definition of ``critical mass of special nuclear material'' has been
modified to emphasize that the definition is only for the purposes of
Subpart H. The definition of ``double contingency'' has been changed to
provide minor clarification. The definition for ``worker'' has been
clarified and has been revised to emphasize that the definition is only
for purposes of Subpart H. The definitions of ``ISA'' and ``ISA
Summary'' have been changed to indicate that the ISA can be performed
on a process by process basis and the ISA Summary can be submitted in
multiple documents that cover all the portions of the facility.
Subpart G--Special Nuclear Material Control Records, Reports, and
Inspections
Section 70.50 Reporting Requirements
The reporting requirements for hazardous chemicals have been
revised to be consistent with the language of the 1988 NRC-OSHA MOU (53
FR 43950; November 22, 1988).
Subpart H--Additional Requirements for Certain Licensees Authorized To
Possess a Critical Mass of Special Nuclear Material
Section 70.60 Applicability
The applicability of the Subpart H requirements has been revised to
clarify when the requirements will take effect.
Section 70.61 Performance Requirements
The performance requirements in Secs. 70.61(b) and (c) have been
revised to provide clarification. The requirement to establish a
controlled area in Sec. 70.61(f) has been revised to clarify the
conditions for establishing the controlled area, and to clarify the
applicability of the performance requirements to individuals within the
controlled area. The requirement in Sec. 70.61(f)(2) to provide
training ``in accordance with'' 10 CFR 19.12(a)(1)-(5) has been revised
to clarify that equivalent training is acceptable. The new language
specifies that this training must ``satisfy'' 10 CFR 19.12(a)(1)-(5).
Section 70.62 Safety program and integrated safety analysis
The requirement to establish and maintain a safety program in
Sec. 70.62(a)(1) has been revised to clarify that the safety program
referred to in this section is focused on the safety program for
satisfying the new Subpart H requirements, and that the application of
management measures may be graded according to risk. Section
70.62(a)(3) has been modified to make it performance-based by
eliminating the prescriptive requirement to maintain a log of failures
for items relied on for safety. Section 70.62(c)(3) has been revised to
clarify the schedule for planning and performing an ISA, correcting all
performance deficiencies, and submitting the ISA Summary to the NRC for
approval. The ISA Summary can be submitted as a single document, or as
a sequence of documents (e.g., on a process basis). The approval
process for the ISA Summary will require the issuance of a license
amendment; however, a license condition will be established that allows
the licensee to make changes in accordance with Sec. 70.72, including
certain changes that do not require prior NRC approval. In addition, a
provision has been added to the schedule for complying with the
requirements in subpart H for factors beyond control of the licensee.
This would allow additional time for correcting a performance
deficiency if the NRC approves. Also Sec. 70.62(d) has been modified to
reflect that management measures may be graded commensurate with the
reduction in risk.
Section 70.64 Requirements for New Facilities or New Processes at
Existing Facilities
Section 70.64(a) has been revised to provide the correct reference,
Sec. 70.62(c), to the performance of an ISA. Section 70.64(a)(6)(ii)
has been modified to specify that the required emergency capability is
concerned with the evacuation of only on-site personnel.
Section 70.65 Additional Content of Applications
Section 70.65(a) has been revised to clarify that the ISA Summary
is not part of the safety program description required for inclusion in
the license application. Rather, the ISA Summary, that contains a
description of
[[Page 56222]]
management measures, is submitted with the license application.
Section 70.66 Additional Requirements for Approval of License
Application
Section 70.66(b) has been added to clarify, for existing licensees,
the basis for Commission approval of the ISA plan, submitted under
Sec. 70.62(c)(3)(i), and the ISA Summary, submitted under
Sec. 70.62(c)(3)(ii).
Section 70.72 Facility Changes and Change Process
Section 70.72(c)(1) has been revised to eliminate the footnote,
which did not adequately clarify the meaning of ``new types of accident
sequences.'' The revised section 70.72(d)(3) replaces proposed rule
section 70.72(d)(1) to reflect a modified schedule for submission of
revised ISA Summary pages.
Section 70.76 Backfitting
Section 70.76 was added to include requirements for performing a
backfit analysis. The wording in Sec. 70.76 is similar to the current
language in Sec. 76.76 for gaseous diffusion plants with one exception,
70.76(a)(4)(i). The exception in Sec. 70.76(a)(4)(i) relates to the
backfit requirements being inapplicable to changes associated with
bringing the facility in compliance with the requirements of the new
subpart H. The backfit section includes a provision stating that it
shall apply for subpart H requirements as soon as NRC approves that
licensee's ISA Summary (contents of ISA Summary described in
Sec. 70.65(b)) pursuant to Sec. 70.66 and, for requirements other than
Subpart H, it shall apply immediately after NRC publication of backfit
guidance.
Section 70.92 Criminal Penalties
This section has been changed to reflect the redesignations of
Secs. 70.13a, 70.14, 70.61, 70.62, 70.71, and 70.72 as Secs. 70.14,
70.17, 70.81, 70.82, 70.91, and 70.92, respectively, and the addition
of Secs. 70.66, 70.73, and 70.76.
IV. Section-by-Section Analysis of Part 70 Amendments
Authority
This section has been changed to reflect the redesignations of
Secs. 70.61 and 70.62 as Secs. 70.81 and 70.82, respectively.
Subpart A--General Provisions
Section 70.4 Definitions
Definitions of the following 12 terms have been added to this
section to clarify the meaning of certain terms and phrases used in the
new Subpart H: ``Acute,'' ``Available and reliable to perform their
function when needed,'' ``Configuration management,'' ``Critical mass
of SNM,'' ``Double contingency principle,'' ``Hazardous chemicals
produced from licensed materials,'' ``Integrated safety analysis,''
``Integrated safety analysis summary,'' ``Items relied on for safety,''
``Management measures,'' ``Unacceptable performance deficiencies,'' and
``Worker.''
Section 70.14 Foreign Military Aircraft
This section reflects an administrative change to redesignate this
section, formerly Sec. 70.13a.
Section 70.17 Specific Exemptions
This section reflects an administrative change to redesignate this
section, formerly Sec. 70.14.
Subpart G--Special Nuclear Material Control Records, Reports, and
Inspections
Section 70.50 Reporting Requirements
Paragraph (c) has been reworded to include information to be
transmitted when making verbal or written reports to the NRC. The new
information derives from the specifics of the new Subpart H, such as
sequence of events and whether the event was evaluated in the ISA. To
the extent the new information is also applicable to licensees not
subject to Subpart H, the information was added with no differentiation
noted. The new information that would only apply to Subpart H licensees
is noted.
Subpart H--Additional Requirements for Certain Licensees Authorized To
Possess a Critical Mass of Special Nuclear Material
Section 70.60 Applicability
This section lists the types of NRC licensees or applicants that
are subject to the new Part 70, Subpart H, and describes when the new
requirements will be effective.
Section 70.61 Performance Requirements
This section identifies the performance requirements that licensees
subject to Part 70, Subpart H must satisfy. These performance
requirements explicitly address the risks to workers or members of the
public and the environmental releases caused by accidents. Because
accidents are unanticipated events that usually occur over a relatively
short period of time, the Part 70 changes seek to ensure adequate
protection of workers, members of the public, and the environment by
limiting the risk (product of likelihood and consequence) of such
accidents. If, without the implementation of controls, a high
consequence event under Sec. 70.61(b) is highly unlikely, then it is
not necessary for the licensee to apply the engineered or
administrative controls mentioned in the rule. Similarly, if, without
the implementation of controls an intermediate consequence event under
Sec. 70.61(c) is unlikely, then it is not necessary for the licensee to
apply the engineered or administrative controls mentioned in the rule.
Section 70.62 Safety Program and Integrated Safety Analysis
This section describes requirements for establishing and
maintaining a safety program that demonstrates compliance with the
performance requirements of Sec. 70.61. The elements of this safety
program include the compilation of process safety information, the
performance of an ISA, and the application of management measures to
ensure the availability and reliability of items relied on for safety.
Section 70.64 Requirements for New Facilities or New Processes at
Existing Facilities
This section describes baseline design criteria for new facilities
or new processes at existing facilities. The application of these
criteria, which are similar to the general design criteria in Part 50,
Appendix A; Part 72, Subpart F; and 10 CFR 60.131, are consistent with
good engineering practice, which dictates that certain minimum
requirements be applied as design and safety considerations for any new
nuclear process or facility. The baseline design criteria do not
provide relief from compliance with the performance requirements of
Sec. 70.61.
Section 70.65 Additional Content of Applications
In addition to the information that currently must be submitted to
the NRC under Sec. 70.22, for a license application, this section
requires additional information to be submitted to demonstrate
compliance with the new subpart H requirements. In particular, this
additional information includes a description of the applicant's safety
program established under Sec. 70.62. This information will be
incorporated in the license, as appropriate. The ISA Summary must be
submitted with the license application or in accordance with
70.62(c)(3)(ii), but will not be incorporated in the license.
[[Page 56223]]
Section 70.66 Additional Requirements for Approval of License
Application
This section contains the provision that the applicant must comply
with the requirements of Secs. 70.60 through 70.65 (in addition to
Secs. 70.21 through 70.23, in the existing regulation) before a license
will be granted. It also contains the requirements for approving the
ISA plan and the ISA Summary for existing licensees. If the ISA Summary
is submitted as a sequence of documents (e.g., on a process basis), NRC
staff approval of an individual document will be conditioned for system
interaction effects that may be identified through the review of other
ISA Summary documents submitted.
Section 70.72 Facility Changes and Change Process
This section contains requirements that govern changes to site,
structures, systems, equipment, components, and activities of personnel
after a license application has been approved. It requires the licensee
to establish and use a configuration management system to evaluate
changes and the potential impacts of those changes before implementing
them. The regulation permits the licensee to make certain changes
without NRC pre-approval, but requires the licensee to submit a brief
summary of the changes plus updated ISA Summary pages annually within
30 days after the end of the calender year during which the changes
occurred.
Section 70.73 Renewal of Licenses
This section contains the requirements for renewing licenses. It
references the existing renewal requirements and the additional
contents of application in Sec. 70.65.
Section 70.74 Additional Reporting Requirements
This section contains new requirements, in addition to those in
existing Parts 20 and 70, for reporting events to the NRC. The new
approach, based on consideration of the performance requirements
established in 10 CFR 70.61(b), is intended to eventually replace and
modify the approach licensees have currently been using for reporting
criticality events under NRC Bulletin 91-01. The new approach would
cover all types of events, not just criticality events, and establish a
timeframe for reporting that is scaled according to risk.
Section 70.76 Backfitting
This section contains requirements for performing a backfit
analysis that are based on those in 10 CFR 76.76. It will become
effective after NRC publication of backfit guidance. NRC staff will
work with stakeholders to develop backfit guidance which will include
making clear that an adequate demonstration can be based on
quantitative or qualitative evaluations of the nature of the increase
in the overall health and safety protection of the public. The NRC will
publish a separate Federal Register notice upon publication of the
guidance to indicate the effectiveness of Sec. 70.76. After that notice
is published, this provision will not be applied to Subpart H
requirements until the NRC approves the licensee's ISA Summary pursuant
to Sec. 70.66. If the approved ISA Summary is one of a sequence of
approvals (e.g., on a process basis), the backfit provision will apply
to the portion of the facility covered by that ISA Summary document.
However, the backfit provision does not apply to changes to those
portions of the facility that are required by the NRC staff to address
system interaction effects identified through the review of other ISA
Summary submissions for that facility.
Subpart J--Enforcement
Section 70.92 Criminal Penalties
This section has been changed to reflect the redesignations of
Secs. 70.13a, 70.14, 70.61, 70.62, 70.71, and 70.72 as Secs. 70.14,
70.17, 70.81, 70.82, 70.91, and 70.92, respectively, and the addition
of Secs. 70.66, 70.73, and 70.76.
Appendix A--Reportable Safety Events
This appendix contains a list of events that licensees must report
to the NRC. These events are categorized according to their
consequences (or potential consequences) and fall into two classes: a
1-hour or 24-hour reporting timeframe. The emphasis on consequences,
rather than risk, is appropriate in this case because the event has
already occurred. Appendix A also requires concurrent reporting of
events when a news release is made or if other Government agencies are
notified, as is done under 10 CFR 50.72, to enhance coordination and
support NRC's ability to respond to questions concerning the safety of
NRC-licensed facilities.
V. Finding of No Significant Environmental Impact: Availability
The Commission has determined, under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
Subpart A of 10 CFR Part 51, that this rule is not a major Federal
action significantly affecting the quality of the human environment and
therefore, an environmental impact statement is not required.
The amendments to 10 CFR Part 70 are intended to provide increased
confidence in the margin of safety at certain facilities that possess a
critical mass of SNM. To accomplish this objective, the amendments: (1)
Identify appropriate performance requirements and the level of
protection needed to prevent or mitigate accidents that exceed such
requirements; (2) require affected licensees to perform an ISA to
identify potential accidents at the facility and the items relied on
for safety; (3) require the implementation of measures to ensure that
the items relied on for safety are available and reliable to perform
their functions when needed; (4) require the safety bases to be
maintained, and changes reported to the NRC; (5) allow for licensees to
make certain changes to their safety program and facilities without
prior NRC approval; (6) require reporting of certain events; and (7)
require a backfit analysis under specified conditions.
The rule language that defines the performance requirements is
relevant to the question of environmental impact. Licensees are
required to protect against the occurrence of or to mitigate the
consequences of accidents that could adversely affect workers, the
public, or the environment. For example, licensees are required to
provide an adequate level of protection against a ``release of
radioactive material to the environment outside the restricted area in
concentrations that, if averaged over 24 hours, exceed 5000 times the
values specified in Table 2 of Appendix B to 10 CFR Part 20.''
Implementation of the new amendments, including the requirement to
protect against events that could damage the environment, is expected
to result in a significant improvement in licensees', NRC's, other
governmental agencies', and the public's understanding of the risks at
these facilities and licensees' ability to ensure that those risks are
appropriately controlled. For existing licensees, any deficiencies
identified in the ISA (which must be completed within 4 years) will
need to be promptly addressed. For new licensees, operations will not
begin unless licensees demonstrate an adequate level of protection
against potential accidents identified in the ISA. As a result, the
safety and environmental impact of the new amendments is positive.
There will be less potential adverse impact on the environment from
licensed operations carried out under the final rule than if those
operations were carried out under the existing Part 70 regulation.
Thus, the Commission has determined, based on
[[Page 56224]]
the Environmental Assessment that supports the rule, that there will be
no significant impact on the human environment from this action.
The NRC requested public comments on any environmental justice
considerations that may be related to this rule. No comments were
received in response to this request.
The NRC also requested the States' views on the environmental
assessment for this rule. No comments were received in response to this
request.
The Environmental Assessment is available for inspection at the NRC
Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC.
Single copies of the Environmental Assessment are available from Barry
Mendelsohn, Office of Nuclear Material Safety and Safeguards, U.S.
Nuclear Regulatory Commission, Washington, DC, 20555-0001, telephone
(301) 415-7262; e-mail: [email protected].
VI. Paperwork Reduction Act Statement
This final rule amends information collection requirements that are
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). These requirements were approved by the Office of Management and
Budget, approval number 3150-0009.
The public reporting burden for this information collection is
estimated to average 92 hours per response and the recordkeeping burden
is estimated to average 548 hours, including the time for reviewing
instructions, searching existing data sources, gathering and
maintaining the data needed, and completing and reviewing the
information collection. Send comments on any aspect of this information
collection, including suggestions for reducing the burden, to the
Records Management Branch (T-6 E6), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, or by Internet electronic mail at
[email protected]; and to the Desk Officer, Office of Information and
Regulatory Affairs, NEOB-10202, (3150-0009), Office of Management and
Budget, Washington, DC 20503.
VII. Public Protection Notification
If a means used to impose an information collection does not
display a currently valid OMB control number, the NRC may not conduct
or sponsor, and a person is not required to respond to, the information
collection.
VIII. Regulatory Analysis
The Commission has prepared a regulatory analysis on this final
regulation. The analysis examines the costs and benefits of the
alternatives considered by the Commission. The analysis is available
for inspection in the NRC Public Document Room, 2120 L Street NW.
(Lower Level), Washington, DC. Single copies of the environmental
assessment are available from Barry Mendelsohn, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, DC, 20555-0001, telephone (301) 415-7262; e-mail:
[email protected].
IX. Regulatory Flexibility Certification
As required by the Regulatory Flexibility Act, as amended, 5 U.S.C.
605(b), the Commission certifies that this rule does not have a
significant economic impact on a substantial number of small entities.
The regulation affects facilities that are authorized to possess a
critical mass of SNM and that are engaged in one of the following
activities: enriched uranium processing; fabrication of uranium fuel or
fuel assemblies; uranium enrichment; enriched uranium hexafluoride
conversion; plutonium processing; fabrication of mixed-oxide fuel or
fuel assemblies; scrap recovery of SNM or any other activity involving
a critical mass of SNM that the Commission determines could
significantly affect public health and safety or the environment. These
licensees do not fall within the scope of the definition of ``small
entities'' set forth in the Regulatory Flexibility Act, nor the size
standards published by the NRC (10 CFR 2.810).
X. Voluntary Consensus Standards
The National Technology Transfer Act of 1995, Pub. L. 104-113,
requires that Federal agencies use technical standards developed or
adopted by voluntary consensus standards bodies unless the use of such
a standard is inconsistent with applicable law or otherwise
impractical. In this regulation, the NRC will use the following
voluntary consensus standard--ANSI/ANS Standard 8.1-1983, ``Nuclear
Criticality Safety in Operations with Fissionable Material Outside
Reactors'' developed by the American Nuclear Society. Portions of the
standard were used in the definition of double contingency and in
Sec. 70.61(d). A consensus standard with the complete scope of the
requirements established in this rulemaking does not exist. The
Commission will reference ANS 8.1 and other consensus standards, as
appropriate, as acceptable approaches to demonstrate compliance with
specific portions of the final rule. This will be addressed in the
Standard Review Plan that is being established with the rule.
XI. Backfit Statement
The NRC has determined that the backfit rule does not apply to this
final rule; therefore, a backfit analysis is not required for this
final rule because these amendments do not involve any provisions that
would impose backfits as defined in 10 CFR Chapter I. However, future
changes to the requirements in subpart H or NRC requirements that apply
to facilities covered by subpart H will be subject to the backfit
requirements in Sec. 70.76 established in this rule.
XII. Small Business Regulatory Enforcement Fairness Act
In accordance with the Small Business Regulatory Enforcement
Fairness Act of 1996, the NRC has determined that this action is
not a major rule and has verified this determination with the Office of
Information and Regulatory Affairs of OMB.
List of Subjects in 10 CFR Part 70
Hazardous materials transportation, Nuclear materials, Packaging
and containers, Radiation protection, Reporting and recordkeeping
requirements, Scientific equipment, Security measures.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553; the NRC is adopting
the following amendments to Part 70.
PART 70--DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL
1. The authority citation for part 70 is revised to read as
follows:
Authority: Secs. 51, 53, 161, 182, 183, 68 Stat. 929, 930, 948,
953, 954, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C.
2071, 2073, 2201, 2232, 2233, 2282, 2297f); secs. 201, as amended,
202, 204, 206, 88 Stat. 1242, as amended, 1244, 1245, 1246 (42
U.S.C. 5841, 5842, 5845, 5846). Sec. 193, 104 Stat. 2835, as amended
by Pub. L. 104-134, 110 Stat. 1321, 1321-349 (42 U.S.C. 2243).
Sections 70.1(c) and 70.20a(b) also issued under secs. 135, 141,
Pub. L. 97-425, 96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).
Section 70.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951 (42 U.S.C. 5851). Section 70.21(g) also issued under sec. 122,
68 Stat. 939 (42 U.S.C. 2152). Section 70.31 also issued under sec.
57d, Pub. L. 93-377, 88 Stat. 475 (42 U.S.C. 2077). Sections 70.36
and 70.44 also issued under sec. 184, 68 Stat. 954, as amended (42
U.S.C. 2234). Section 70.81 also issued under secs. 186, 187, 68
Stat. 955 (42 U.S.C. 2236, 2237). Section 70.82 also issued under
sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138).
[[Page 56225]]
2. The undesignated center heading ``GENERAL PROVISIONS'' is
redesignated as ``Subpart A--General Provisions.''
3. In Sec. 70.4, the definitions of Acute, Available and reliable
to perform their function when needed, Configuration management,
Critical mass of special nuclear material, Double contingency
principle, Hazardous chemicals produced from licensed materials,
Integrated safety analysis, Integrated safety analysis summary, Items
relied on for safety, Management measures, Unacceptable performance
deficiencies, and Worker are added, in alphabetical order, as follows:
Sec. 70.4 Definitions.
* * * * *
Acute, as used in this part, means a single radiation dose or
chemical exposure event or multiple radiation dose or chemical exposure
events occurring within a short time (24 hours or less).
* * * * *
Available and reliable to perform their function when needed, as
used in subpart H of this part, means that, based on the analyzed,
credible conditions in the integrated safety analysis, items relied on
for safety will perform their intended safety function when needed, and
management measures will be implemented that ensure compliance with the
performance requirements of Sec. 70.61 of this part, considering
factors such as necessary maintenance, operating limits, common-cause
failures, and the likelihood and consequences of failure or degradation
of the items and measures.
* * * * *
Configuration management (CM) means a management measure that
provides oversight and control of design information, safety
information, and records of modifications (both temporary and
permanent) that might impact the ability of items relied on for safety
to perform their functions when needed.
* * * * *
Critical mass of special nuclear material (SNM), as used in Subpart
H, means special nuclear material in a quantity exceeding 700 grams of
contained uranium-235; 520 grams of uranium-233; 450 grams of
plutonium; 1500 grams of contained uranium-235, if no uranium enriched
to more than 4 percent by weight of uranium-235 is present; 450 grams
of any combination thereof; or one-half such quantities if massive
moderators or reflectors made of graphite, heavy water, or beryllium
may be present.
* * * * *
Double contingency principle means that process designs should
incorporate sufficient factors of safety to require at least two
unlikely, independent, and concurrent changes in process conditions
before a criticality accident is possible.
* * * * *
Hazardous chemicals produced from licensed materials means
substances having licensed material as precursor compound(s) or
substances that physically or chemically interact with licensed
materials; and that are toxic, explosive, flammable, corrosive, or
reactive to the extent that they can endanger life or health if not
adequately controlled. These include substances commingled with
licensed material, and include substances such as hydrogen fluoride
that is produced by the reaction of uranium hexafluoride and water, but
do not include substances prior to process addition to licensed
material or after process separation from licensed material.
Integrated safety analysis (ISA) means a systematic analysis to
identify facility and external hazards and their potential for
initiating accident sequences, the potential accident sequences, their
likelihood and consequences, and the items relied on for safety. As
used here, integrated means joint consideration of, and protection
from, all relevant hazards, including radiological, nuclear
criticality, fire, and chemical. However, with respect to compliance
with the regulations of this part, the NRC requirement is limited to
consideration of the effects of all relevant hazards on radiological
safety, prevention of nuclear criticality accidents, or chemical
hazards directly associated with NRC licensed radioactive material. An
ISA can be performed process by process, but all processes must be
integrated, and process interactions considered.
Integrated safety analysis summary means a document or documents
submitted with the license application, license amendment application,
license renewal application, or pursuant to Sec. 70.62(c)(3)(ii) that
provides a synopsis of the results of the integrated safety analysis
and contains the information specified in Sec. 70.65(b). The ISA
Summary can be submitted as one document for the entire facility, or as
multiple documents that cover all portions and processes of the
facility.
Items relied on for safety mean structures, systems, equipment,
components, and activities of personnel that are relied on to prevent
potential accidents at a facility that could exceed the performance
requirements in Sec. 70.61 or to mitigate their potential consequences.
This does not limit the licensee from identifying additional
structures, systems, equipment, components, or activities of personnel
(i.e., beyond those in the minimum set necessary for compliance with
the performance requirements) as items relied on for safety.
* * * * *
Management measures mean the functions performed by the licensee,
generally on a continuing basis, that are applied to items relied on
for safety, to ensure the items are available and reliable to perform
their functions when needed. Management measures include configuration
management, maintenance, training and qualifications, procedures,
audits and assessments, incident investigations, records management,
and other quality assurance elements.
* * * * *
Unacceptable performance deficiencies mean deficiencies in the
items relied on for safety or the management measures that need to be
corrected to ensure an adequate level of protection as defined in 10
CFR 70.61(b), (c), or (d).
* * * * *
Worker, when used in Subpart H of this Part, means an individual
who receives an occupational dose as defined in 10 CFR 20.1003.
4. In Sec. 70.8 paragraph (b) is revised to read as follows:
Sec. 70.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Secs. 70.9, 70.17, 70.19, 70.20a, 70.20b, 70.21,
70.22, 70.24, 70.25, 70.32, 70.33, 70.34, 70.38, 70.39, 70.42, 70.50,
70.51, 70.52, 70.53, 70.57, 70.58, 70.59, 70.61, 70.62, 70.64, 70.65,
70.72, 70.73, 70.74, and Appendix A.
* * * * *
5. The undesignated center heading ``EXEMPTIONS'' is redesignated
as ``Subpart B--Exemptions.''
Secs. 70.13a and 70.14 [Redesignated]
6. Sections 70.13a and 70.14 are redesignated as Secs. 70.14 and
70.17, respectively.
7. The undesignated center heading ``GENERAL LICENSES'' is
redesignated as ``Subpart C--General Licenses.''
8. The undesignated center heading ``LICENSE APPLICATIONS'' is
[[Page 56226]]
redesignated as ``Subpart D--License Applications.''
9. The undesignated center heading ``LICENSES'' is redesignated as
``Subpart E--Licenses.''
10. The undesignated center heading ``ACQUISITION, USE AND TRANSFER
OF SPECIAL NUCLEAR MATERIAL, CREDITORS' RIGHTS,'' is redesignated as
``Subpart F--Acquisition, Use, and Transfer of Special Nuclear
Material, Creditors' Rights.''
11. The undesignated center heading ``SPECIAL NUCLEAR MATERIAL
CONTROL RECORDS, REPORTS AND INSPECTIONS'' is redesignated as ``Subpart
G--Special Nuclear Material Control Records, Reports, and
Inspections.''
12. In Sec. 70.50, paragraph (c) is revised and paragraph (d) is
added to read as follows.
Sec. 70.50 Reporting requirements.
* * * * *
(c) Preparation and submission of reports. Reports made by
licensees in response to the requirements of this section must be made
as follows:
(1) Licensees shall make reports required by paragraphs (a) and (b)
of this section, and by Sec. 70.74 and Appendix A of this part, if
applicable, by telephone to the NRC Operations Center.\1\ To the extent
that the information is available at the time of notification, the
information provided in these reports must include:
---------------------------------------------------------------------------
\1\ The commercial telephone number for the NRC Operations
Center is (301) 816-5100.
---------------------------------------------------------------------------
(i) Caller's name, position title, and call-back telephone number;
(ii) Date, time, and exact location of the event;
(iii) Description of the event, including:
(A) Radiological or chemical hazards involved, including isotopes,
quantities, and chemical and physical form of any material released;
(B) Actual or potential health and safety consequences to the
workers, the public, and the environment, including relevant chemical
and radiation data for actual personnel exposures to radiation or
radioactive materials or hazardous chemicals produced from licensed
materials (e.g., level of radiation exposure, concentration of
chemicals, and duration of exposure);
(C) The sequence of occurrences leading to the event, including
degradation or failure of structures, systems, equipment, components,
and activities of personnel relied on to prevent potential accidents or
mitigate their consequences; and
(D) Whether the remaining structures, systems, equipment,
components, and activities of personnel relied on to prevent potential
accidents or mitigate their consequences are available and reliable to
perform their function;
(iv) External conditions affecting the event;
(v) Additional actions taken by the licensee in response to the
event;
(vi) Status of the event (e.g., whether the event is on-going or
was terminated);
(vii) Current and planned site status, including any declared
emergency class;
(viii) Notifications, related to the event, that were made or are
planned to any local, State, or other Federal agencies;
(ix) Status of any press releases, related to the event, that were
made or are planned.
(2) Written report. Each licensee that makes a report required by
paragraph (a) or (b) of this section, or by Sec. 70.74 and Appendix A
of this part, if applicable, shall submit a written follow-up report
within 30 days of the initial report. Written reports prepared pursuant
to other regulations may be submitted to fulfill this requirement if
the report contains all the necessary information, and the appropriate
distribution is made. These written reports must be sent to the U.S.
Nuclear Regulatory Commission, Document Control Desk, Washington, DC
20555, with a copy to the appropriate NRC regional office listed in
Appendix D of 10 CFR Part 20. The reports must include the following:
(i) Complete applicable information required by Sec. 70.50(c)(1);
(ii) The probable cause of the event, including all factors that
contributed to the event and the manufacturer and model number (if
applicable) of any equipment that failed or malfunctioned;
(iii) Corrective actions taken or planned to prevent occurrence of
similar or identical events in the future and the results of any
evaluations or assessments; and
(iv) For licensees subject to Subpart H of this part, whether the
event was identified and evaluated in the Integrated Safety Analysis.
(d) The provisions of Sec. 70.50 do not apply to licensees subject
to Sec. 50.72. They do apply to those Part 50 licensees possessing
material licensed under Part 70 that are not subject to the
notification requirements in Sec. 50.72.
13. The undesignated center heading ``MODIFICATION AND REVOCATION
OF LICENSES'' is redesignated as ``Subpart I--Modification and
Revocation of Licenses.''
Secs. 70.61 and 70.62 [Redesignated and Amended]
14. Sections 70.61 and 70.62 are redesignated as Secs. 70.81 and
70.82, respectively.
15. The undesignated center heading ``ENFORCEMENT'' is redesignated
as ``Subpart J--Enforcement.''
Sec. 70.71 [Redesignated]
16. Section 70.71 is redesignated as Sec. 70.91.
17. Section 70.72 is redesignated as Sec. 70.92 and paragraph (b)
is revised to read as follows:
Sec. 70.92 Criminal Penalties
* * * * *
(b) The regulations in part 70 that are not issued under sections
161b, 161i, or 161o, for the purposes of section 223 are as follows:
Secs. 70.1, 70.2, 70.4, 70.5, 70.6, 70.8, 70.11, 70.12, 70.13, 70.14,
70.17, 70.18, 70.23, 70.31, 70.33, 70.34, 70.35, 70.37, 70.66, 70.73,
70.76, 70.81, 70.82, 70.63, 70.91, and 70.92.
18. In part 70, a new subpart H (Secs. 70.60-70.76) is added to
read as follows:
Subpart H--Additional Requirements for Certain Licensees Authorized
To Possess a Critical Mass of Special Nuclear Material
Sec.
70.60 Applicability.
70.61 Performance requirements.
70.62 Safety program and integrated safety analysis.
70.64 Requirements for new facilities or new processes at existing
facilities.
70.65 Additional content of applications.
70.66 Additional requirements for approval of license application.
70.72 Facility changes and change process.
70.73 Renewal of licenses.
70.74 Additional reporting requirements.
70.76 Backfitting
Sec. 70.60 Applicability.
The regulations in Sec. 70.61 through Sec. 70.76 apply, in addition
to other applicable Commission regulations, to each applicant or
licensee that is or plans to be authorized to possess greater than a
critical mass of special nuclear material, and engaged in enriched
uranium processing, fabrication of uranium fuel or fuel assemblies,
uranium enrichment, enriched uranium hexafluoride conversion, plutonium
processing, fabrication of mixed-oxide fuel or fuel assemblies, scrap
recovery of special nuclear material, or any other activity that the
Commission determines could significantly affect public health and
safety. The regulations in Sec. 70.61 through Sec. 70.76 do not apply
to decommissioning activities performed
[[Page 56227]]
pursuant to other applicable Commission regulations including
Sec. 70.25 and Sec. 70.38 of this part. Also, the regulations in
Sec. 70.61 through Sec. 70.76 do not apply to activities that are
certified by the Commission pursuant to part 76 of this chapter or
licensed by the Commission pursuant to other parts of this chapter.
Unless specifically addressed in Sec. 70.61 through Sec. 70.76,
implementation by current licensees of the Subpart H requirements shall
be completed no later than the time of the ISA Summary submittal
required in Sec. 70.62(c)(3)(ii).
Sec. 70.61 Performance requirements.
(a) Each applicant or licensee shall evaluate, in the integrated
safety analysis performed in accordance with Sec. 70.62, its compliance
with the performance requirements in paragraphs (b), (c), and (d) of
this section.
(b) The risk of each credible high-consequence event must be
limited. Engineered controls, administrative controls, or both, shall
be applied to the extent needed to reduce the likelihood of occurrence
of the event so that, upon implementation of such controls, the event
is highly unlikely or its consequences are less severe than those in
paragrahs (b)(1)-(4) of this section. High consequence events are those
internally or externally initiated events that result in:
(1) An acute worker dose of 1 Sv (100 rem) or greater total
effective dose equivalent;
(2) An acute dose of 0.25 Sv (25 rem) or greater total effective
dose equivalent to any individual located outside the controlled area
identified pursuant to paragraph (f) of this section;
(3) An intake of 30 mg or greater of uranium in soluble form by any
individual located outside the controlled area identified pursuant to
paragraph (f) of this section; or
(4) An acute chemical exposure to an individual from licensed
material or hazardous chemicals produced from licensed material that:
(i) Could endanger the life of a worker, or
(ii) Could lead to irreversible or other serious, long-lasting
health effects to any individual located outside the controlled area
identified pursuant to paragraph (f) of this section. If an applicant
possesses or plans to possess quantities of material capable of such
chemical exposures, then the applicant shall propose appropriate
quantitative standards for these health effects, as part of the
information submitted pursuant to Sec. 70.65 of this subpart.
(c) The risk of each credible intermediate-consequence event must
be limited. Engineered controls, administrative controls, or both shall
be applied to the extent needed so that, upon implementation of such
controls, the event is unlikely or its consequences are less than those
in paragraphs (c)(1)-(4) of this section. Intermediate consequence
events are those internally or externally initiated events that are not
high consequence events, that result in:
(1) An acute worker dose of 0.25 Sv (25 rem) or greater total
effective dose equivalent;
(2) An acute dose of 0.05 Sv (5 rem) or greater total effective
dose equivalent to any individual located outside the controlled area
identified pursuant to paragraph (f) of this section;
(3) A 24-hour averaged release of radioactive material outside the
restricted area in concentrations exceeding 5000 times the values in
Table 2 of Appendix B to Part 20; or
(4) An acute chemical exposure to an individual from licensed
material or hazardous chemicals produced from licensed material that:
(i) Could lead to irreversible or other serious, long-lasting
health effects to a worker, or
(ii) Could cause mild transient health effects to any individual
located outside the controlled area as specified in paragraph (f) of
this section. If an applicant possesses or plans to possess quantities
of material capable of such chemical exposures, then the applicant
shall propose appropriate quantitative standards for these health
effects, as part of the information submitted pursuant to Sec. 70.65 of
this subpart.
(d) In addition to complying with paragraphs (b) and (c) of this
section, the risk of nuclear criticality accidents must be limited by
assuring that under normal and credible abnormal conditions, all
nuclear processes are subcritical, including use of an approved margin
of subcriticality for safety. Preventive controls and measures must be
the primary means of protection against nuclear criticality accidents.
(e) Each engineered or administrative control or control system
necessary to comply with paragraphs (b), (c), or (d) of this section
shall be designated as an item relied on for safety. The safety
program, established and maintained pursuant to Sec. 70.62 of this
subpart, shall ensure that each item relied on for safety will be
available and reliable to perform its intended function when needed and
in the context of the performance requirements of this section.
(f) Each licensee must establish a controlled area, as defined in
Sec. 20.1003. In addition, the licensee must retain the authority to
exclude or remove personnel and property from the area. For the purpose
of complying with the performance requirements of this section,
individuals who are not workers, as defined in Sec. 70.4, may be
permitted to perform ongoing activities (e.g., at a facility not
related to the licensed activities) in the controlled area, if the
licensee:
(1) Demonstrates and documents, in the integrated safety analysis,
that the risk for those individuals at the location of their activities
does not exceed the performance requirements of paragraphs (b)(2),
(b)(3), (b)(4)(ii), (c)(2), and (c)(4)(ii) of this section; or
(2) Provides training that satisfies 10 CFR 19.12(a)(1)-(5) to
these individuals and ensures that they are aware of the risks
associated with accidents involving the licensed activities as
determined by the integrated safety analysis, and conspicuously posts
and maintains notices stating where the information in 10 CFR 19.11(a)
may be examined by these individuals. Under these conditions, the
performance requirements for workers specified in paragraphs (b) and
(c) of this section may be applied to these individuals.
Sec. 70.62 Safety program and integrated safety analysis.
(a) Safety program. (1) Each licensee or applicant shall establish
and maintain a safety program that demonstrates compliance with the
performance requirements of Sec. 70.61. The safety program may be
graded such that management measures applied are graded commensurate
with the reduction of the risk attributable to that item. Three
elements of this safety program; namely, process safety information,
integrated safety analysis, and management measures, are described in
paragraphs (b) through (d) of this section.
(2) Each licensee or applicant shall establish and maintain records
that demonstrate compliance with the requirements of paragraphs (b)
through (d) of this section.
(3) Each licensee or applicant shall maintain records of failures
readily retrievable and available for NRC inspection, documenting each
discovery that an item relied on for safety or management measure has
failed to perform its function upon demand or has degraded such that
the performance requirements of Sec. 70.61 are not satisfied. These
records must identify the item relied on for safety or management
measure that has failed and the safety function affected, the date of
discovery, date (or estimated date) of the failure, duration (or
estimated duration) of the
[[Page 56228]]
time that the item was unable to perform its function, any other
affected items relied on for safety or management measures and their
safety function, affected processes, cause of the failure, whether the
failure was in the context of the performance requirements or upon
demand or both, and any corrective or compensatory action that was
taken. A failure must be recorded at the time of discovery and the
record of that failure updated promptly upon the conclusion of each
failure investigation of an item relied on for safety or management
measure.
(b) Process safety information. Each licensee or applicant shall
maintain process safety information to enable the performance and
maintenance of an integrated safety analysis. This process safety
information must include information pertaining to the hazards of the
materials used or produced in the process, information pertaining to
the technology of the process, and information pertaining to the
equipment in the process.
(c) Integrated safety analysis. (1) Each licensee or applicant
shall conduct and maintain an integrated safety analysis, that is of
appropriate detail for the complexity of the process, that identifies:
(i) Radiological hazards related to possessing or processing
licensed material at its facility;
(ii) Chemical hazards of licensed material and hazardous chemicals
produced from licensed material;
(iii) Facility hazards that could affect the safety of licensed
materials and thus present an increased radiological risk;
(iv) Potential accident sequences caused by process deviations or
other events internal to the facility and credible external events,
including natural phenomena;
(v) The consequence and the likelihood of occurrence of each
potential accident sequence identified pursuant to paragraph (c)(1)(iv)
of this section, and the methods used to determine the consequences and
likelihoods; and
(vi) Each item relied on for safety identified pursuant to
Sec. 70.61(e) of this subpart, the characteristics of its preventive,
mitigative, or other safety function, and the assumptions and
conditions under which the item is relied upon to support compliance
with the performance requirements of Sec. 70.61.
(2) Integrated safety analysis team qualifications. To assure the
adequacy of the integrated safety analysis, the analysis must be
performed by a team with expertise in engineering and process
operations. The team shall include at least one person who has
experience and knowledge specific to each process being evaluated, and
persons who have experience in nuclear criticality safety, radiation
safety, fire safety, and chemical process safety. One member of the
team must be knowledgeable in the specific integrated safety analysis
methodology being used.
(3) Requirements for existing licensees. Individuals holding an NRC
license on September 18, 2000 shall, with regard to existing licensed
activities:
(i) By April 18, 2001, submit for NRC approval, a plan that
describes the integrated safety analysis approach that will be used,
the processes that will be analyzed, and the schedule for completing
the analysis of each process.
(ii) By October 18, 2004, or in accordance with the approved plan
submitted under Sec. 70.62(c)(3)(i), complete an integrated safety
analysis, correct all unacceptable performance deficiencies, and
submit, for NRC approval, an integrated safety analysis summary,
including a description of the management measures, in accordance with
Sec. 70.65. The Commission may approve a request for an alternative
schedule for completing the correction of unacceptable performance
deficiencies if the Commission determines that the alternative is
warranted by consideration of the following:
(A) Adequate compensatory measures have been established;
(B) Whether it is technically feasible to complete the correction
of the unacceptable performance deficiency within the allotted 4-year
period;
(C) Other site-specific factors which the Commission may consider
appropriate on a case-by-case basis and that are beyond the control of
the licensee.
(iii) Pending the correction of unacceptable performance
deficiencies identified during the conduct of the integrated safety
analysis, the licensee shall implement appropriate compensatory
measures to ensure adequate protection.
(d) Management measures. Each applicant or licensee shall establish
management measures to ensure compliance with the performance
requirements of Sec. 70.61. The measures applied to a particular
engineered or administrative control or control system may be graded
commensurate with the reduction of the risk attributable to that
control or control system. The management measures shall ensure that
engineered and administrative controls and control systems that are
identified as items relied on for safety pursuant to Sec. 70.61(e) of
this subpart are designed, implemented, and maintained, as necessary,
to ensure they are available and reliable to perform their function
when needed, to comply with the performance requirements of Sec. 70.61
of this subpart.
Sec. 70.64 Requirements for new facilities or new processes at
existing facilities.
(a) Baseline design criteria. Each prospective applicant or
licensee shall address the following baseline design criteria in the
design of new facilities. Each existing licensee shall address the
following baseline design criteria in the design of new processes at
existing facilities that require a license amendment under Sec. 70.72.
The baseline design criteria must be applied to the design of new
facilities and new processes, but do not require retrofits to existing
facilities or existing processes (e.g., those housing or adjacent to
the new process); however, all facilities and processes must comply
with the performance requirements in Sec. 70.61. Licensees shall
maintain the application of these criteria unless the analysis
performed pursuant to Sec. 70.62(c) demonstrates that a given item is
not relied on for safety or does not require adherence to the specified
criteria.
(1) Quality standards and records. The design must be developed and
implemented in accordance with management measures, to provide adequate
assurance that items relied on for safety will be available and
reliable to perform their function when needed. Appropriate records of
these items must be maintained by or under the control of the licensee
throughout the life of the facility.
(2) Natural phenomena hazards. The design must provide for adequate
protection against natural phenomena with consideration of the most
severe documented historical events for the site.
(3) Fire protection. The design must provide for adequate
protection against fires and explosions.
(4) Environmental and dynamic effects. The design must provide for
adequate protection from environmental conditions and dynamic effects
associated with normal operations, maintenance, testing, and postulated
accidents that could lead to loss of safety functions.
(5) Chemical protection. The design must provide for adequate
protection against chemical risks produced from licensed material,
facility conditions which affect the safety of licensed material, and
hazardous chemicals produced from licensed material.
[[Page 56229]]
(6) Emergency capability. The design must provide for emergency
capability to maintain control of:
(i) Licensed material and hazardous chemicals produced from
licensed material;
(ii) Evacuation of on-site personnel; and
(iii) Onsite emergency facilities and services that facilitate the
use of available offsite services.
(7) Utility services. The design must provide for continued
operation of essential utility services.
(8) Inspection, testing, and maintenance. The design of items
relied on for safety must provide for adequate inspection, testing, and
maintenance, to ensure their availability and reliability to perform
their function when needed.
(9) Criticality control. The design must provide for criticality
control including adherence to the double contingency principle.
(10) Instrumentation and controls. The design must provide for
inclusion of instrumentation and control systems to monitor and control
the behavior of items relied on for safety.
(b) Facility and system design and facility layout must be based on
defense-in-depth practices. \1\ The design must incorporate, to the
extent practicable:
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\1\ As used in Sec. 70.64, Requirements for new facilities or
new processes at existing facilities, defense-in-depth practices
means a design philosophy, applied from the outset and through
completion of the design, that is based on providing successive
levels of protection such that health and safety will not be wholly
dependent upon any single element of the design, construction,
maintenance, or operation of the facility. The net effect of
incorporating defense-in-depth practices is a conservatively
designed facility and system that will exhibit greater tolerance to
failures and external challenges. The risk insights obtained through
performance of the integrated safety analysis can be then used to
supplement the final design by focusing attention on the prevention
and mitigation of the higher-risk potential accidents.
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(1) Preference for the selection of engineered controls over
administrative controls to increase overall system reliability; and
(2) Features that enhance safety by reducing challenges to items
relied on for safety.
Sec. 70.65 Additional content of applications.
(a) In addition to the contents required by Sec. 70.22, each
application must include a description of the applicant's safety
program established under Sec. 70.62.
(b) The integrated safety analysis summary must be submitted with
the license or renewal application (and amendment application as
necessary), but shall not be incorporated in the license. However,
changes to the integrated safety analysis summary shall meet the
conditions of Sec. 70.72. The integrated safety analysis summary must
contain:
(1) A general description of the site with emphasis on those
factors that could affect safety (i.e., meteorology, seismology);
(2) A general description of the facility with emphasis on those
areas that could affect safety, including an identification of the
controlled area boundaries;
(3) A description of each process (defined as a single reasonably
simple integrated unit operation within an overall production line)
analyzed in the integrated safety analysis in sufficient detail to
understand the theory of operation; and, for each process, the hazards
that were identified in the integrated safety analysis pursuant to
Sec. 70.62(c)(1)(i)-(iii) and a general description of the types of
accident sequences;
(4) Information that demonstrates the licensee's compliance with
the performance requirements of Sec. 70.61, including a description of
the management measures; the requirements for criticality monitoring
and alarms in Sec. 70.24; and, if applicable, the requirements of
Sec. 70.64;
(5) A description of the team, qualifications, and the methods used
to perform the integrated safety analysis;
(6) A list briefly describing each item relied on for safety which
is identified pursuant to Sec. 70.61(e) in sufficient detail to
understand their functions in relation to the performance requirements
of Sec. 70.61;
(7) A description of the proposed quantitative standards used to
assess the consequences to an individual from acute chemical exposure
to licensed material or chemicals produced from licensed materials
which are on-site, or expected to be on-site as described in
Sec. 70.61(b)(4) and (c)(4);
(8) A descriptive list that identifies all items relied on for
safety that are the sole item preventing or mitigating an accident
sequence that exceeds the performance requirements of Sec. 70.61; and
(9) A description of the definitions of unlikely, highly unlikely,
and credible as used in the evaluations in the integrated safety
analysis.
Sec. 70.66 Additional requirements for approval of license
application.
(a) An application for a license from an applicant subject to
subpart H will be approved if the Commission determines that the
applicant has complied with the requirements of Secs. 70.21, 70.22,
70.23, and 70.60 through 70.65.
(b) Submittals by existing licensees in accordance with
Sec. 70.62(c)(3)(i) will be approved if the Commission determines that:
(1) The integrated safety analysis approach is in accordance with
the requirements of Secs. 70.61, 70.62(c)(1), and 70.62(c)(2); and
(2) The schedule is in compliance with Sec. 70.62(c)(3)(ii).
(c) Submittals by existing licensees in accordance with
Sec. 70.62(c)(3)(ii) will be approved if the Commission determines
that:
(1) The requirements of Sec. 70.65(b) are satisfied; and
(2) The performance requirements in Sec. 70.61 (b), (c) and (d) are
satisfied, based on the information in the ISA Summary, together with
other information submitted to NRC or available to NRC at the
licensee's site.
Sec. 70.72 Facility changes and change process.
(a) The licensee shall establish a configuration management system
to evaluate, implement, and track each change to the site, structures,
processes, systems, equipment, components, computer programs, and
activities of personnel. This system must be documented in written
procedures and must assure that the following are addressed prior to
implementing any change:
(1) The technical basis for the change;
(2) Impact of the change on safety and health or control of
licensed material;
(3) Modifications to existing operating procedures including any
necessary training or retraining before operation;
(4) Authorization requirements for the change;
(5) For temporary changes, the approved duration (e.g., expiration
date) of the change; and
(6) The impacts or modifications to the integrated safety analysis,
integrated safety analysis summary, or other safety program
information, developed in accordance with Sec. 70.62.
(b) Any change to site, structures, processes, systems, equipment,
components, computer programs, and activities of personnel must be
evaluated by the licensee as specified in paragraph (a) of this
section, before the change is implemented. The evaluation of the change
must determine, before the change is implemented, if an amendment to
the license is required to be submitted in accordance with Sec. 70.34.
(c) The licensee may make changes to the site, structures,
processes, systems, equipment, components, computer programs, and
activities of personnel, without prior Commission approval, if the
change:
[[Page 56230]]
(1) Does not:
(i) Create new types of accident sequences that, unless mitigated
or prevented, would exceed the performance requirements of Sec. 70.61
and that have not previously been described in the integrated safety
analysis summary; or
(ii) Use new processes, technologies, or control systems for which
the licensee has no prior experience;
(2) Does not remove, without at least an equivalent replacement of
the safety function, an item relied on for safety that is listed in the
integrated safety analysis summary;
(3) Does not alter any item relied on for safety, listed in the
integrated safety analysis summary, that is the sole item preventing or
mitigating an accident sequence that exceeds the performance
requirements of Sec. 70.61; and
(4) Is not otherwise prohibited by this section, license condition,
or order.
(d)(1) For changes that require pre-approval under Sec. 70.72, the
licensee shall submit an amendment request to the NRC in accordance
with Sec. 70.34 and Sec. 70.65 of this chapter.
(2) For changes that do not require pre-approval under Sec. 70.72,
the licensee shall submit to NRC annually, within 30 days after the end
of the calendar year during which the changes occurred, a brief summary
of all changes to the records required by Sec. 70.62(a)(2) of this
subpart.
(3) For all changes that affect the integrated safety analysis
summary, the licensee shall submit to NRC annually, within 30 days
after the end of the calendar year during which the changes occurred,
revised integrated safety analysis summary pages.
(e) If a change covered by Sec. 70.72 is made, the affected on-site
documentation must be updated promptly.
(f) The licensee shall maintain records of changes to its facility
carried out under this section. These records must include a written
evaluation that provides the bases for the determination that the
changes do not require prior Commission approval under paragraph (c) or
(d) of this section. These records must be maintained until termination
of the license.
Sec. 70.73 Renewal of licenses.
Applications for renewal of a license must be filed in accordance
with Secs. 2.109, 70.21, 70.22, 70.33, 70.38, and 70.65 of this
chapter. Information contained in previous applications, statements, or
reports filed with the Commission under the license may be incorporated
by reference, provided that these references are clear and specific.
Sec. 70.74 Additional reporting requirements.
(a) Reports to NRC Operations Center. (1) Each licensee shall
report to the NRC Operations Center the events described in Appendix A
to Part 70.
(2) Reports must be made by a knowledgeable licensee representative
and by any method that will ensure compliance with the required time
period for reporting.
(3) The information provided must include a description of the
event and other related information as described in Sec. 70.50(c)(1).
(4) Follow-up information to the reports must be provided until all
information required to be reported in Sec. 70.50(c)(1) of this subpart
is complete.
(5) Each licensee shall provide reasonable assurance that reliable
communication with the NRC Operations Center is available during each
event.
(b) Written reports. Each licensee that makes a report required by
paragraph (a)(1) of this section shall submit a written follow-up
report within 30 days of the initial report. The written report must
contain the information as described in Sec. 70.50(c)(2).
Sec. 70.76 Backfitting.
(a) For each licensee, this provision shall apply to Subpart H
requirements as soon as the NRC approves that licensee's ISA Summary
pursuant to Sec. 70.66. For requirements other than Subpart H, this
provision applies regardless of the status of the approval of a
licensee's ISA Summary.
(1) Backfitting is defined as the modification of, or addition to,
systems, structures, or components of a facility; or to the procedures
or organization required to operate a facility; any of which may result
from a new or amended provision in the Commission rules or the
imposition of a regulatory staff position interpreting the Commission
rules that is either new or different from a previous NRC staff
position.
(2) Except as provided in paragraph (a)(4) of this section, the
Commission shall require a systematic and documented analysis pursuant
to paragraph (b) of this section for backfits which it seeks to impose.
(3) Except as provided in paragraph (a)(4) of this section, the
Commission shall require the backfitting of a facility only when it
determines, based on the analysis described in paragraph (b) of this
section, that there is a substantial increase in the overall protection
of the public health and safety or the common defense and security to
be derived from the backfit and that the direct and indirect costs of
implementation for that facility are justified in view of this
increased protection.
(4) The provisions of paragraphs (a)(2) and (a)(3) of this section
are inapplicable and, therefore, backfit analysis is not required and
the standards in paragraph (a)(3) of this section do not apply where
the Commission finds and declares, with appropriately documented
evaluation for its finding, any of the following:
(i) That a modification is necessary to bring a facility into
compliance with Subpart H of this part;
(ii) That a modification is necessary to bring a facility into
compliance with a license or the rules or orders of the Commission, or
into conformance with written commitments by the licensee;
(iii) That regulatory action is necessary to ensure that the
facility provides adequate protection to the health and safety of the
public and is in accord with the common defense and security; or
(iv) That the regulatory action involves defining or redefining
what level of protection to the public health and safety or common
defense and security should be regarded as adequate.
(5) The Commission shall always require the backfitting of a
facility if it determines that the regulatory action is necessary to
ensure that the facility provides adequate protection to the health and
safety of the public and is in accord with the common defense and
security.
(6) The documented evaluation required by paragraph (a)(4) of this
section must include a statement of the objectives of and reasons for
the modification and the basis for invoking the exception. If immediate
effective regulatory action is required, then the documented evaluation
may follow, rather than precede, the regulatory action.
(7) If there are two or more ways to achieve compliance with a
license or the rules or orders of the Commission, or with written
license commitments, or there are two or more ways to reach an adequate
level of protection, then ordinarily the licensee is free to choose the
way that best suits its purposes. However, should it be necessary or
appropriate for the Commission to prescribe a specific way to comply
with its requirements or to achieve adequate protection, then cost may
be a factor in selecting the way, provided that the objective of
compliance or adequate protection is met.
(b) In reaching the determination required by paragraph (a)(3) of
this section, the Commission will consider how the backfit should be
scheduled in light of other ongoing regulatory
[[Page 56231]]
activities at the facility and, in addition, will consider information
available concerning any of the following factors as may be appropriate
and any other information relevant and material to the proposed
backfit:
(1) Statement of the specific objectives that the proposed backfit
is designed to achieve;
(2) General description of the activity that would be required by
the licensee in order to complete the backfit;
(3) Potential change in the risk to the public from the accidental
release of radioactive material and hazardous chemicals produced from
licensed material;
(4) Potential impact on radiological exposure or exposure to
hazardous chemicals produced from licensed material of facility
employees;
(5) Installation and continuing costs associated with the backfit,
including the cost of facility downtime;
(6) The potential safety impact of changes in facility or
operational complexity, including the relationship to proposed and
existing regulatory requirements;
(7) The estimated resource burden on the NRC associated with the
proposed backfit and the availability of such resources;
(8) The potential impact of differences in facility type, design,
or age on the relevancy and practicality of the proposed backfit; and
(9) Whether the proposed backfit is interim or final and, if
interim, the justification for imposing the proposed backfit on an
interim basis.
(c) No license will be withheld during the pendency of backfit
analyses required by the Commission's rules.
(d) The Executive Director for Operations shall be responsible for
implementation of this section, and all analyses required by this
section shall be approved by the Executive Director for Operations or
his or her designee.
19. Appendix A to part 70 is added to read as follows:
Appendix A to Part 70--Reportable Safety Events
Licensees must comply with reporting requirements in this
appendix, except for (a)(1), (a)(2), and (b)(4), after they have
submitted an ISA Summary in accordance with Sec. 70.62(c)(3)(ii).
Licensees must comply with (a)(1), (a)(2), and (b)(4) after October
18, 2000. As required by 10 CFR 70.74, licensees subject to the
requirements in subpart H of part 70, shall report:
(a) One hour reports. Events to be reported to the NRC
Operations Center within 1 hour of discovery, supplemented with the
information in 10 CFR 70.50(c)(1) as it becomes available, followed
by a written report within 30 days:
(1) An inadvertent nuclear criticality.
(2) An acute intake by an individual of 30 mg or greater of
uranium in a soluble form.
(3) An acute chemical exposure to an individual from licensed
material or hazardous chemicals produced from licensed material that
exceeds the quantitative standards established to satisfy the
requirements in Sec. 70.61(b)(4).
(4) An event or condition such that no items relied on for
safety, as documented in the Integrated Safety Analysis summary,
remain available and reliable, in an accident sequence evaluated in
the Integrated Safety Analysis, to perform their function:
(i) In the context of the performance requirements in
Sec. 70.61(b) and Sec. 70.61(c), or
(ii) Prevent a nuclear criticality accident (i.e., loss of all
controls in a particular sequence).
(5) Loss of controls such that only one item relied on for
safety, as documented in the Integrated Safety Analysis summary,
remains available and reliable to prevent a nuclear criticality
accident, and has been in this state for greater than eight hours.
(b) Twenty-four hour reports. Events to be reported to the NRC
Operations Center within 24 hours of discovery, supplemented with
the information in 10 CFR 70.50(c)(1) as it becomes available,
followed by a written report within 30 days:
(1) Any event or condition that results in the facility being in
a state that was not analyzed, was improperly analyzed, or is
different from that analyzed in the Integrated Safety Analysis, and
which results in failure to meet the performance requirements of
Sec. 70.61.
(2) Loss or degradation of items relied on for safety that
results in failure to meet the performance requirement of
Sec. 70.61.
(3) An acute chemical exposure to an individual from licensed
material or hazardous chemicals produced from licensed materials
that exceeds the quantitative standards that satisfy the
requirements of Sec. 70.61(c)(4).
(4) Any natural phenomenon or other external event, including
fires internal and external to the facility, that has affected or
may have affected the intended safety function or availability or
reliability of one or more items relied on for safety.
(5) An occurrence of an event or process deviation that was
considered in the Integrated Safety Analysis and:
(i) Was dismissed due to its likelihood; or
(ii) Was categorized as unlikely and whose associated
unmitigated consequences would have exceeded those in Sec. 70.61(b)
had the item(s) relied on for safety not performed their safety
function(s).
(c) Concurrent Reports. Any event or situation, related to the
health and safety of the public or onsite personnel, or protection
of the environment, for which a news release is planned or
notification to other government agencies has been or will be made,
shall be reported to the NRC Operations Center concurrent to the
news release or other notification.
Dated at Rockville, Maryland, this 6th day of September, 2000.
For the Nuclear Regulatory Commission.
Annette Vietti-Cook,
Secretary of the Commission.
[FR Doc. 00-23354 Filed 9-15-00; 8:45 am]
BILLING CODE 7590-01-P