[Federal Register Volume 64, Number 191 (Monday, October 4, 1999)]
[Rules and Regulations]
[Pages 53582-53617]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-25054]
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NUCLEAR REGULATORY COMMISSION
10 CFR Parts 50 and 72
RIN 3150-AF94
Changes, Tests, and Experiments
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations concerning the authority for licensees of production or
utilization facilities, such as nuclear reactors, and independent spent
fuel storage facilities, and for certificate holders for spent fuel
storage casks, to make changes to the facility or procedures, or to
conduct tests or experiments, without prior NRC approval. The final
rule clarifies the specific types of changes, tests, and experiments
conducted at a licensed facility or by a certificate holder that
require evaluation, and revises the criteria that licensees and
certificate holders must use to determine when NRC approval is needed
before such changes, tests, or experiments can be implemented. The
final rule also adds definitions for terms that have been subject to
differing interpretations, and reorganizes the rule language for
clarity. Additionally, the final rule grants in part and denies in
part, a petition for rulemaking (PRM-72-3) submitted by Ms. Fawn
Shillinglaw on December 9, 1995. This notice constitutes final NRC
action on this petition.
EFFECTIVE DATE: The amendments to sections 72.3, 72.9, 72.24, 72.56,
72.70, 72.80, 72.86, 72.244, 72.246, 72.248 of this rule are effective
February 1, 2000. Sections 50.59, 50.66, 50.71(e), and 50.90 become
effective 90 days after issuance of applicable regulatory guidance. The
NRC will publish a document in the Federal Register that announces the
issuance of the regulatory guidance and specifies that the final rule
becomes effective in 90 days. Section 72.212 and the amendments to
72.48 are effective April 5, 2001.
FOR FURTHER INFORMATION CONTACT: Eileen McKenna, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington,
[[Page 53583]]
DC 20555-0001, telephone (301) 415-2189; e-mail: [email protected].
SUPPLEMENTARY INFORMATION:
I. Background
II. Comments and resolution on proposed rule topics
A. Organization of the rule requirements
B. Change to the facility as described in the Safety Analysis
Report
B.1 Definition of change
B.2 Definition of facility
C. Change to the procedures as described in the safety analysis
report
D. Tests and experiments not described in the final safety
analysis report
E. Safety analysis report
F. Minimal increase principle
G. Section 50.59(c)(2) criteria on increases in probability or
consequences
H. Possibility of an accident of a different type from any
previously evaluated in the final safety analysis report (as
updated) is created
I. Possibility of a malfunction of a structure, system, or
component important to safety with a different result from any
previously evaluated in the final safety analysis report (as
updated) is created
J. Replacement criteria for ``margin of safety as defined in the
basis for any technical specification is reduced''
K. Safety evaluation
L. Reporting and recordkeeping requirements
M. No significant hazards consideration determinations
N. Part 52 changes
O.1 Part 72 changes
O.2 Petition for Rulemaking (PRM-72-3)
O.3 Part 71 (Transportation) Comments
P. Other topics discussed in the notice and comments not related
to preceding topic areas
Q Enforcement policy
R. Implementation
III. Section by section analysis
IV. Finding of no significant environmental impact
V. Paperwork Reduction Act statement
VI. Regulatory analysis
VII. Regulatory Flexibility Certification
VIII. Backfit analysis
IX. Small Business Regulatory Enforcement Fairness Act
X. National Technology Transfer and Advancement Act
XI. Criminal penalties
XII. Compatibility of Agreement State Regulations
List of Subjects
I. Background
The existing requirements governing the authority of production and
utilization facility licensees to make changes to their facilities and
procedures, or to conduct tests or experiments, without prior NRC
approval are contained in 10 CFR 50.59. Comparable provisions exist in
Sec. 72.48 for licensees of facilities for the independent storage of
spent nuclear fuel and high-level radioactive waste. These regulations
provide that licensees may make changes to the facility or procedures
as described in the safety analysis report (SAR), or conduct tests or
experiments not described in the safety analysis report, without prior
Commission approval, unless the proposed change, test, or experiment
involves a change to the Technical Specifications (TS) incorporated in
the license or an unreviewed safety question. Section 50.59(a)(2), as
codified, states the following:
A proposed change, test, or experiment shall be deemed to
involve an unreviewed safety question (i) if the probability of
occurrence or the consequences of an accident or malfunction of
equipment important to safety previously evaluated in the safety
analysis report may be increased; or (ii) if a possibility for an
accident or malfunction of a different type than any evaluated
previously in the safety analysis report may be created; or (iii) if
the margin of safety as defined in the basis for any technical
specification is reduced.
The rule also specifies recordkeeping and reporting requirements
associated with such changes, tests, or experiments.
Section 50.59 was promulgated in 1962 to allow licensees to make
certain changes that affect systems, structures, components (SSC), or
procedures described in the SAR without prior approval, provided
certain conditions were met. In 1968, the rule was revised to modify
some of the criteria for determining whether prior NRC approval was
required. The intent of the Sec. 50.59 process is to permit licensees
to make changes to the facility, provided the changes maintain
acceptable levels of safety as documented in the SAR. The process was
thus structured around the licensing approach of design basis events
(anticipated operational occurrences and accidents), safety-related
mitigation systems, and consequence calculations for the design basis
accidents.
On October 21, 1998 (63 FR 56098), the NRC published a proposed
rule to revise Secs. 50.59 and 72.48 to address a number of issues
concerning implementation of the current rule, and suitability of the
criteria used to determine when an unreviewed safety question exists.
Conforming changes were proposed in other portions of the regulations,
including Secs. 50.66, 50.71(e), and 50.90 for production and
utilization facilities licensed under part 50. Conforming changes were
also proposed in Sec. 72.212(b)(4).
The Commission proposed to make similar changes to appendices A and
B of part 52, the standard design certifications for the ABWR and CE
System 80+ designs respectively. These regulations contain a change
control process similar to that in Sec. 50.59. As noted in Section N,
``Part 52 changes'' below, the Commission has decided to defer
consideration of any changes to part 52 until a later date.
In addition, the Commission proposed to make parallel changes
applicable to independent spent fuel storage installations (ISFSIs)
licensed in accordance with part 72. As part of the proposed changes to
part 72, the Commission also proposed to extend the change control
authority granted to ISFSI or monitored retrievable storage (MRS)
license holders (in Sec. 72.48) to holders of NRC Certificates of
Compliance (CoC) for a spent fuel storage cask design.
II. Comments and Resolution on Proposed Rule Topics
The 60-day comment period for the proposed rule closed on December
21, 1998. Comments were received from 60 organizations or individuals.
Copies of the comments are available for public inspection and copying
for a fee at the Commission's Public Document Room, located at 2120 L
Street, NW., Washington DC. All comments were considered in formulating
the final rule. The comments were submitted by 35 utilities with power
reactor facilities; 2 representatives of nonpower reactor licensees; 3
law firms representing several utilities; 2 submittals from the Nuclear
Energy Institute (NEI); the U. S. Enrichment Corporation; a nuclear
industry group; 6 nuclear utility vendors, service companies or
consultants; 4 vendors or service companies for spent fuel storage
casks; and 6 individuals. Forty commenters endorsed (sometimes with
further comments) the NEI comments. NEI stated in its comment letter
that it generally supports the Commission's intent of the proposed rule
but had a number of comments or modifications for certain specific
provisions of the rule that it wished the Commission to consider in
preparing the final rule. Of those commenters who did not endorse the
NEI comments, most supported the concept of the proposed rule, and made
recommendations to enhance or modify certain elements of the rule. A
few commenters stated that the rule revision was unnecessary and
presented supporting arguments. These commenters felt that the
Commission should endorse NEI 96-07 ``Guidelines for 10 CFR 50.59
Safety Evaluations,'' as being sufficient to satisfy the existing rule
requirements. Many of the other comments related to the content of
regulatory guidance, suggesting that
[[Page 53584]]
examples be provided to amplify particular points.
In the following sections, the NRC presents a discussion and
resolution of the public comments, and the final rulemaking language in
a form that parallels the order of discussion of issues in the proposed
rulemaking. The organizational changes are discussed first, followed by
discussion of the revised provisions in the rule. Although the
discussion of many of the topics specifically focuses upon Sec. 50.59,
these matters are equally applicable to Sec. 72.48, except as noted.
Topics not related to particular rule sections are at the end of this
discussion.
A. Organization of the Rule Requirements
(1) Definitions
In the proposed rule, the Commission added a new paragraph (a) to
Sec. 50.59 that contains a number of definitions for terms used in the
rule. The Commission sought comment on the need for definitions as well
as on the specific definitions offered for the terminology. Most
commenters did not explicitly address whether they thought definitions
were needed. One commenter thought that adding definitions only added
confusion. Another stated that although the terms in the rule need to
be defined, having them in the rule means that any subsequent changes
in interpretation would require rulemaking. The Commission believes
that having the definitions in the rule adds clarity that improves
implementation of the rule, and, in some cases, are necessary for
completeness of requirements. Therefore the Commission has retained
several definitions in the final rule in Secs. 50.59(a) and 72.48(a).
The specific definitions are discussed in subsequent sections.
(2) Applicability
The Commission proposed to place all of the provisions concerning
applicability of the rule presently contained in several subsections
into Sec. 50.59(b), which is clearly labeled ``Applicability.'' The
rule applies to: production and utilization facilities (including power
and non-power reactors) that are authorized to operate, and reactors
(both power and non-power) that have permanently ceased operations. The
few commenters who addressed this topic were supportive of this
proposal. The final rule is unchanged from the proposed rule in this
regard (except that Sec. 72.48 now explicitly has a section with this
designation for consistency).
(3) Form of Prior Commission Approval
In the proposed rule, the Commission combined Secs. 50.59 (a) and
(c) and revised the regulation to state more clearly that a licensee
must apply for and obtain a license amendment, pursuant to Sec. 50.90,
before implementing changes, tests, or experiments that involve either
a change to the TS or that satisfy any of the criteria listed in new
section 50.59(c)(2). In addition, the Commission proposed relocating an
existing provision that refers to changes to the TS not associated with
a change, test, or experiment from Sec. 50.59 to Sec. 50.90. Parallel
changes to Sec. 72.48 and Sec. 72.56 were also proposed.
One aspect of the proposed rule that drew comment concerned the
requirement to obtain a license amendment before implementing a change
that involves a change to TS or meets Sec. 50.59(c)(2) criteria. In
particular, for those instances in which a licensee wishes to make a
modification to the facility, the use of which would require a TS
change (or meet one of the other criteria), the commenters believe that
it is acceptable for a licensee to install and test such a
modification, as long as such activities themselves do not place the
facility in a condition for which NRC review is needed, and as long as
the modification is not actually used until the amendment review has
been completed. These commenters believe that waiting for NRC approval
for use of such modifications before beginning any installation
activity is unduly restrictive. Typically this question arises for
plant modifications and installations or complex engineering changes
which may take months or years to complete.
In the Commission's view, the acceptability of such activities
depends upon the meaning of ``implementation'' and of which aspect of
the change requires NRC approval. If installing the modification, or
testing it after installation would violate a TS, NRC approval (of both
the modification and the revised TS) would be needed before the change
is implemented. In addition, the licensee would need to determine
whether the test itself meets the criteria in Sec. 50.59 so that prior
NRC approval of the test is not required. For changes that are not
inconsistent with existing TS, but for which the licensee plans to
submit an amendment to later revise TS to allow use of the modification
(as for instance a modification that may permit less restrictive TS
requirements), proceeding with the installation, before the approval is
received, is at the licensee's own risk with respect to whether the
Commission will approve use of the modification. If the NRC finds the
proposed TS or the modification unacceptable, the licensee would need
to appropriately revise the modification or may be unable to reap the
expected benefits. If the licensee establishes that installation and
testing of a modification do not require approval, but its use in
facility operations would, NRC approval would be needed before the
modification could be put into effect. With these clarifications, the
Commission accepts the comments on this aspect. The final rule text is
unchanged from that offered in the proposed rule.
(4) Criteria for Needing Commission Approval of Changes, Tests, and
Experiments and Unreviewed Safety Question (USQ) Designation
In the proposed rule, the Commission proposed to remove the
reference to the term ``unreviewed safety question'' and instead refer
to the need to obtain a license amendment. The Commission concluded
that this terminology has sometimes led to confusion about the purpose
of the evaluation required by Sec. 50.59. The purpose is to identify
possible changes that might affect the basis for licensing the facility
so that any changes that might pose a safety concern are reviewed by
NRC to confirm their safety before implementation. To avoid confusion
between a determination of safety and a determination of the need for
NRC approval, the Commission is removing the term ``unreviewed safety
question.'' In addition, the Commission proposed to list the criteria
(in the new Sec. 50.59(c)(2)) that, if met, would require prior
Commission approval for a proposed change, which would be in the form
of a license amendment. In the proposed rule, the compound statements
contained within the evaluation criteria of the current rule were
separated into several individual criteria. The deletion of the term
``unreviewed safety question'' also required a number of conforming
changes to other parts of the regulations.
Commenters generally supported these proposed changes. A few
commenters stated that the supplementary information should explain
that existing guidance referring to ``USQ'' (such as Generic Letter 91-
18, Revision 1), is still applicable. Further, commenters stated that a
simple process should be established by which licensee technical
specifications that use the term ``USQ'' could be revised.
The Commission agrees that the term USQ was used as a convenience
to describe those changes that met the rule criteria for prior NRC
review and
[[Page 53585]]
approval, and that any guidance referring to the same category of plant
changes is equally valid for describing plant changes that would
require prior NRC review and approval under the revised
Sec. 50.59(c)(2).
The Commission considered the merits of including specific language
in Sec. 50.59 that would address this point, but ultimately did not
include such language for a number of reasons. First, the NRC official
record copy would not be modified if licensees made changes on their
own (in accordance with the rule language). Second, the intent of the
specific provision would be to permit such changes; however, the fact
that the provision is contained in the rule may make it a requirement
to do so. This is clearly an unintended consequence and argues against
including such language. Finally, since there is no practical effect of
the wording as contained within the TS, there is no compelling reason
why licensees would need to promptly conform the wording of their TS.
For administrative convenience, the NRC requests that upon such
occasion as those sections of the TS require NRC approval for other
reasons or a licensee is requesting a license amendment in some other
area of the TS, the licensee should include any necessary changes to
the existing TS language to bring the plant-specific technical
specifications into conformance with the rule language. Such changes
could be made at any time if a general formulation of the requirement
is used, as for example, replacing ``USQ'' with ``requires NRC approval
pursuant to Sec. 50.59.'' Since these are viewed as editorial changes
only, effectiveness of the existing TS is not impacted. The
implementation period of the rule will give reasonable opportunity to
assure that the technical specifications are appropriately modified
without the need to file a separate amendment request.
(5) Changes in the Scope of the Rule
The Commission solicited public comment on the need to revise the
scope of the rule in the notice for the proposed rule. Specifically,
the Commission asked whether the scope of the rule should be linked to
the final safety analysis report (FSAR), as updated, or should the
focus of the rule be linked to another set of regulatory requirements.
Only a few commenters indicated interest in a redefinition of the
scope of the rule. These commenters suggested that any attempt to
redefine the scope of the rule should be considered as part of a longer
term revision that might be part of staff efforts to make the rule more
risk informed. Therefore, the NRC is not revising the scope of the rule
as part of the final rule. The NRC will reconsider the scope of the
rule as part of its ongoing initiatives to improve its regulations to
make them more risk informed.
B. Change to the Facility as Described in the Safety Analysis Report
In the proposed rule, the Commission created a new Sec. 50.59(a) to
contain definitions for terms such as ``change'' and ``facility as
described in the final safety analysis report (as updated).'' The
definitions in Sec. 50.59 of ``change'' and of ``facility as described
in the final safety analysis report (as updated)'' were written to more
explicitly establish that evaluation is required for changes to the
analyses and bases for the facility as well as for physical or hardware
changes to the facility. The proposed rule also explicitly stated that
additions were changes under the rule.
B.1 Definition of Change
In the proposed rule, the Commission concluded that a ``change'' is
a modification of an existing provision (e.g., structure, system, or
component design requirement, analysis method or parameter), an
addition or a removal (physical removals or non-reliance on a system to
meet a requirement) to the facility (or procedure) as described in the
FSAR.
Comment Summary: A number of comments related to the definition of
change. The major topic areas of the comments are summarized below. The
Commission's resolution of these matters follows.
(a) Screening: Most of the commenters were seeking revision of the
definition to allow screening of changes that would not affect design
functions. For instance, some commenters, while agreeing that additions
should be considered changes, also noted that additions, if not limited
by qualifiers such as ``inconsistent with FSAR or changing operation'',
could mean that even trivial additions to the facility or to a
procedure would require evaluations. A few commenters thought that
additions should instead be treated as ``tests or experiments,'' so
that evaluations would be needed only if the additions were
inconsistent with the FSAR or outside the design basis.
(b) Replacement components or maintenance: Other commenters sought
clarification as to whether particular activities, such as the
installation of ``equivalent'' components, or maintenance activities
are considered to be changes requiring evaluation against the criteria.
For instance, replacement equipment should only require review if the
replacement component has characteristics that are different from those
described in the FSAR. For maintenance, commenters stated that taking
SSC out of service for maintenance is adequately covered by maintenance
rule requirements or TS, and that a Sec. 50.59 evaluation should not be
required. Other commenters wanted clarification that requirements for
environmental qualification of electrical equipment were covered by
Sec. 50.49, such that equipment replacements that are qualified per
Sec. 50.49 are not ``reductions in margin of safety'' under Sec. 50.59.
(c) Interdependent changes: A number of comments concerned
``interdependent'' changes, that is, under what circumstances can more
than one change be considered together rather than individually. A few
commenters stated that the Commission should adopt a position with
respect to interdependent changes that multiple changes to the facility
or its procedures may be evaluated collectively if: (1) They are
interdependent as in the case where a modification to a system or
component necessitates additional changes to other systems or
procedures in order for the modified system to perform its function or
comply with its design or licensing basis; (2) they are performed
collectively to address a design or operational issue; or, (3) they are
otherwise planned as elements of a single project undertaken to
restore, maintain or improve plant performance or safety. Several
commenters also stated that examples would be helpful to illustrate how
closely related the changes needed to be in order to be viewed as
interdependent.
(d) Removal: One commenter stated that the term ``removal'' should
be clarified to include removal from service, physical removal,
retirement in place, discontinued availability, removal from the FSAR
text or tables, and removal from FSAR figures.
(e) De Facto Changes: One commenter stated that the NRC should
modify the definition or other rule language to explicitly state that
the requirements apply only to ``proposed'' changes and not to so-
called ``de facto'' changes.1 Another commenter thought the
rule language should explicitly codify the resolution process under
Generic Letter
[[Page 53586]]
(GL) 91-18, by including language in the rule such that the respective
requirements of Appendix B, criterion 16 and Sec. 50.59 do not
interfere.
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\1\ Under the NRC enforcement policy, Sec. 50.59 is sometimes
used to form the basis for a violation for circumstances under which
the as-built facility differs from the FSAR, in that the existing
condition is a ``change'' from the ``as-described FSAR condition'',
and no evaluation was performed supporting why the change could be
made without prior NRC approval. Such situations are referred to as
``de facto'' changes.
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(f) Changes made in response to NRC communications: Two commenters
asked if a proposed change that is the direct result of a response to
issues raised in generic communications requires evaluation under
Sec. 50.59 to determine the need for NRC approval, or if it is already
approved by the NRC. The Commission notes that this subject was also
raised by NEI during a meeting on guidance for minimal increases with
respect to changes being made to conform with changes to regulations.
Resolution: The Commission has modified the proposed rule language
for ``change'' to be responsive to the issues raised by these comments.
In particular, for comment (a), the Commission has incorporated into
the definition of ``change'' the phrase ``that affects design function,
method of performing or controlling a function, or an evaluation that
demonstrates that intended functions will be accomplished.'' The
Commission concluded that with this revision, other comments about
``additions'' and ``removals'' have been addressed (as for instance
comment (d)). The definition of change language will allow licensees to
eliminate the need to further assess specific changes against the
criteria in the rule because the nature of the change would never meet
the criteria of the rule and require prior NRC review before
implementation (known in the industry as a screening review). The
capability to perform such screening reviews for such minor changes
will reduce the burden of the review process.
With respect to comment (b) about whether specific types of
activities are ``changes'', the Commission agrees that clarification
would be useful and will work with affected stakeholders to address the
specific needs for regulatory guidance to successfully implement the
final rule. In particular, the Commission finds that guidance would be
useful on when ``replacement'' components must be treated as a change,
as for instance because the replacement component has characteristics
different from those described in the FSAR, compared to one that is
``equivalent'' and thus not a change. The Commission also agrees that
simply removing a component from service for maintenance does not
require a Sec. 50.59 evaluation, but notes that prolonged removal from
service appears indistinguishable in its effect from a change that
removes the component from the facility. Further, there may be
circumstances under which maintenance activities would place the
facility in a configuration not previously considered, or require
disabling of barriers or movement of heavy loads to accomplish. The
Commission further agrees that acceptability of environmental
qualification requirements would be determined with respect to
Sec. 50.49. However, use of different equipment would also require a
Sec. 50.59 review with respect to meeting the evaluation criteria as
now defined in the rule (as discussed elsewhere, the criterion on
``margin'' is being removed). The Commission notes that for certain
changes, such as a change that affects post-accident containment
conditions, although Sec. 50.49 may be the applicable regulation for
equipment qualification, other aspects (containment pressure) would
need to be evaluated under Sec. 50.59.
The Commission's previous comments on interdependent changes arises
from concern that if multiple changes were considered in a single
evaluation, certain aspects of the ``combined'' change could offset
other aspects and lead to a conclusion that the set of changes did not
require approval. Certain of the other changes being made to the final
rule alleviate much of the Commission's concern about this practice. In
particular, the Commission has described in section J how changes to
methods, input parameters, and facility changes should be evaluated in
determining whether the evaluation criteria are met. Although the
Commission agrees with many of the ideas offered by the commenters for
interdependent changes, the Commission further believes that providing
further discussion and examples in guidance on this point would be
useful.
The Commission did not modify the rule language to specifically
address comment (e) on ``de facto'' changes or GL 91-18 guidance,
believing that changes were not needed to allow the process under GL
91-18 to be implemented. The Commission did not revise the rule
language to specifically state that ``changes'' resulting from
corrective actions under Appendix B do not fall under the ``obtain
amendment prior to implementing'' requirement as suggested by the
commenter. The Commission acknowledges that in those instances of ``de
facto'' changes, it is not possible for the licensee to obtain NRC
approval prior to implementing a change that has already occurred. In
these cases, the ``proposed change'' that the licensee wishes to make
is to its FSAR such that it reflects the ``as-found'' condition of the
plant. The prior approval specified in Sec. 50.59 is the NRC's
agreement with the resolution of the nonconformance before the issue is
closed. For these instances, the Commission views ``implementing the
change'' as meaning closeout of the corrective action. Further, the
Commission does not plan to revise its enforcement policy concerning de
facto changes (see also section Q below for more discussion on
enforcement for Sec. 50.59).
With respect to item (f), the licensee has an obligation to comply
with the regulations (including any changes), and to respond
appropriately to any generic communication. The licensee must examine
the facility changes being made to determine how the facility will
function with the change and identify any potential impacts on safety.
A rule or generic communication may specify a requirement to be
satisfied, or the nature of a change to meet a particular intent, but
rarely is the specific issue presented at a level of detail necessary
for installation. For some facilities, or some configurations, the
``generic'' solution intended by the rule or generic communication may
not achieve the expected results, or there may be alternative ways that
would avoid other problems. These issues can be pursued in the
licensee's response to the generic communication or requirement.
The question about the need for NRC approval for the specific means
of implementation of an action prompted by NRC initiative (rule, order,
or generic communication) is less clear. As an example, NRC has issued
a rule requiring the licensee to cope with a station blackout. Suppose
that the means a licensee selects to meet the requirement is to cross-
connect a new non-safety-related diesel to safety-related buses. Before
implementing this modification, the licensee must evaluate the change
to determine whether the particular method of satisfying the rule has
created other circumstances that would warrant NRC review, such as if
the change would increase the likelihood of malfunction of the buses.
Given these considerations, the NRC concludes that changes made in
response to rules and generic communications must be evaluated in the
same way as other changes a licensee may wish to make, with the conduct
of Sec. 50.59 evaluations and submittal of license amendment requests
as needed. Where there are conflicts in requirements or schedules
resulting from these situations, the NRC has an obligation to take
timely and appropriate action on the licensee's submittals. To the
extent that the impacts of the generic communication or rule are within
the range of what the NRC had considered in its deliberations
[[Page 53587]]
on the rule or communication, the approval of the licensee's submittal
will be straightforward.
In summary, the Commission has included a definition of change as
meaning a modification or addition to, or removal from the facility or
procedures that affects a design function, method of performing or
controlling the function, or an evaluation that demonstrates that
intended functions will be accomplished. Other points raised by the
commenters, such as providing examples, will be handled in the
regulatory guidance to be developed.
B.2 Definition of Facility
In the proposed rule, the Commission concluded that changes to
information such as performance requirements, methods of operation, the
bases upon which the requirements have been established, and the
evaluations should be considered to constitute a change to the
``facility as described in the FSAR (as updated)''. The Commission
concludes that changes to methods and other requirements in the FSAR,
even if not physical changes to the facility, require evaluation under
Sec. 50.59. If changes to methods and performance requirements were not
so controlled, a licensee might revise its analyses or other
information, update its FSAR, and then subsequently conclude that a
later facility change does not require NRC approval because the revised
analysis or acceptance requirement can still be satisfied with the
facility change (that otherwise would have met the criteria as
requiring approval). Thus, the proposed definition specifically
itemized these points.
Comment Summary: A few commenters stated that it should be
clarified that changes, whether to analysis methods or to the physical
facility, are only subject to Sec. 50.59 requirements if they are
described in the FSAR. Other commenters stated that if the level of
discussion within the FSAR is unaffected by the change, there should be
no need for an evaluation.
NEI (as endorsed by other commenters) stated that ``methods of
operation'' should be removed from the definition of facility, as this
was better suited to the definition of ``procedures.''
Some commenters also were concerned that the phrase ``required to
be included in the FSAR'' used in the definition of facility was an
attempt to require licensees to look beyond the FSAR, or to undertake
actions to add information to its FSAR. These commenters thought such
matters were better handled as part of agency actions concerning
guidance for updating FSARs (see for instance, Draft Regulatory Guide
DG-1083 and NEI 98-03, ``Guidelines for Updating Final Safety Analysis
Reports'' ).
The Commission had included these words in the rule as an attempt
to limit what part of the FSAR needed to be considered for purposes of
Sec. 50.59 evaluations. If information was not required to be in the
FSAR, then as discussed under NEI 98-03, it could be removed from the
FSAR. On the other hand, a licensee may wish to retain such information
in its FSAR for purposes of completeness; then this part of the
definition would allow the licensee to screen out changes to the
information that does not meet the definition of facility as described.
In view of the confusion surrounding this phrase, and in light of other
proposed changes to these definitions, the Commission has deleted this
phrase from the final rule.
A commenter stated that such administrative changes as
organizational information, reporting relationships, and job titles
should be excluded from the scope of Sec. 50.59.
Resolution: The Commission considered these comments in selecting
the language that allows screening as to whether a change to the
facility affects the content of the FSAR. As previously noted in
implementation guidance, some SSC or subcomponents may not be
explicitly described in the FSAR, but they have the potential to affect
the function of an SSC that is described. The approach chosen by the
Commission for defining ``change'' as relating to those additions,
modifications, and removals that affect functions, methods of
performing or controlling functions and evaluation methods also
accomplishes an important purpose for these issues. Some changes a
licensee may wish to make to a component or procedure could affect the
functions or performance requirements of other SSC. Depending upon the
level of detail contained in the FSAR, the particular component being
changed may not be explicitly described. If a modification to that
(non-described) component could affect any SSC design function or
performance requirements that are described, that modification affects
the design function, and thus is a change as defined by Sec. 50.59(a)
and thus requires evaluation under Sec. 50.59. For example, the
bearings on a pump may not be specifically mentioned or described in
the FSAR. However, the pump function and performance requirement is
described. A change being made to the bearings would need to be
evaluated to determine if it affects the function or performance
requirements of the pump, and if so, whether the criteria in 50.59 (c)
are met.
Changes to the definition of ``facility'' were made in response to
the concerns noted above from the commenters, such as deletion of the
phrases ``required to be included * * *,'' and ``methods of
operation.'' The Commission has retained ``methods of evaluation'' as
being within the definition of ``facility,'' and as discussed under a
later section, added an evaluation criterion specifically designed to
provide a standard for evaluation of such changes.
The Commission believes that the definitions provided in the rule
for facility and procedures exclude the indicated administrative type
of changes from Sec. 50.59, and further notes that many of these
details would be part of a licensee's quality assurance plan that is
governed by the requirements of Sec. 50.54(a), and therefore excluded
from the purview of Sec. 50.59 by virtue of Sec. 50.59(c)(4).
The definition of facility includes performance requirements and
evaluations included in the FSAR which demonstrate that functions will
be accomplished. In part 54, ``Requirements for Renewal of Operating
Licenses for Nuclear Power Plants,'' Sec. 54.21(d) states that each
renewal application must contain an FSAR supplement that contains a
summary description of the programs and activities for managing the
effects of aging and the evaluation of time-limited aging analyses for
the period of extended operation. As discussed in the Statement of
Considerations for the final part 54, inclusion of the program
descriptions and analyses in the FSAR provides the appropriate
regulatory oversight such that subsequent changes are controlled by
Sec. 50.59. The Commission concludes that these summary descriptions
fall within the definition of ``facility'' as demonstrating that
functions will be accomplished in light of potential aging effects from
the period of extended operation. Therefore changes that affect this
information require evaluation under Sec. 50.59. The Commission further
finds that supplemental guidance or examples for implementation
specific to part 54 would be beneficial and NRC intends to consider
this as part of regulatory guidance.
C. Change to the Procedures as Described in the Safety Analysis Report
The Commission also proposed a definition of ``procedures as
described in the safety analysis report'' in order to have definitions
in the rule for all the major terms and criteria. This definition
includes the evaluations demonstrating
[[Page 53588]]
that requirements are met, such as assumed operator actions and
response times.
Commenters on the definition primarily expressed concern with the
phrase ``conduct of operations'' because licensees were concerned that
this language would inappropriately bring administrative procedures
within the scope of the rule. Other commenters suggested wording
changes to clarify the definition.
The Commission has decided to remove the phrase ``conduct of
operations'' from the definition. The Commission agrees that
administrative procedures are not intended to be within the scope of
the rule, and has made other minor wording changes to the final rule
for clarity.
Changes Governed by Other Regulatory Processes
In the proposed rule, the Commission proposed to exclude from the
scope of Sec. 50.59 review, specific types of changes to procedures
where other requirements and criteria have been established by
regulation for controlling these changes, through a proposed provision
in Sec. 50.59(c)(1).
Commenters supported this proposal, and suggested it be clarified
to also refer to plant changes in addition to procedure changes. As an
example, emergency response facilities are considered as part of the
emergency plans that are subject to Sec. 50.54(q). If also described in
the FSAR, there is a potential for confusion as to whether both a
Sec. 50.54(q) and Sec. 50.59 evaluation would be needed for a change to
an emergency response facility.
The Commission revised the rule language to make the requested
clarification. Further, this section was relocated to new
Sec. 50.59(c)(4) in the final rule. This language refers to situations,
such as Secs. 50.54(a) and 50.54(q), where the regulations explicitly
define how changes are to be reviewed, documented, and reported; and
thus, where a Sec. 50.59 evaluation would be duplicative. Another
example would be Sec. 50.46, which establishes criteria for reporting
and for action for changes involving methods for loss-of-coolant
analyses. A specific list of regulations was not included in the rule
so that if other such rule sections become available, Sec. 50.59 would
not need to be revised. The Sec. 50.59 obligation can only be replaced
in situations in which other rule requirements specify the governing
change process, in order to prevent duplication of reviews, not as a
means of avoiding change control requirements.
A few commenters stated that clarification should be included
concerning applicability of Sec. 50.59 for certain documents controlled
by a variety of processes (e.g., Core Operating Limit Reports contained
in TS; Technical Requirements Manual and other matters (e.g., offsite
dose calculation manual (ODCM)) that have been relocated from TS to
other controlled documents such as the FSAR; and vendor topical
reports, etc.).
The Commission notes that in NEI 98-03, which the NRC has proposed
to endorse through a regulatory guide, there is discussion about
incorporation by reference of other documents (such as ODCM, fire
protection plan, etc) into the FSAR. As discussed in Generic Letter 86-
10, ``Implementation of Fire Protection Requirements,'' licensees were
encouraged to consolidate their fire protection program documents and
incorporate them by reference into the FSAR. Then, by the terms of a
modified license condition, licensees could make changes to their fire
protection program. The vast majority of licensees have made this
change so that the program description is incorporated into the FSAR
and program changes can be made without NRC approval provided the
changes do not adversely affect the ability to achieve and maintain
safe shutdown in the event of a fire (or require an exemption). The
Commission sees no need to provide additional clarification as the
processes for control of most of these documents are already defined.
D. Tests and Experiments Not Described in the Safety Analysis Report
The Commission proposed a definition for ``tests and experiments
not described in the final safety analysis report (as updated)'' to be
included in Sec. 50.59. The intent of the requirement is that tests
that put the facility in a situation that has not previously been
evaluated or that could affect the capability of SSC to perform their
intended functions should be evaluated before they are conducted. Thus,
the definition focused upon the facility being outside its design basis
values or inconsistent with the safety analyses in the FSAR.
A few comments were made on this topic, with some indicating that a
definition was not needed, and with some noting that certain terms were
unclear or stating that the term ``activity'' should be used instead of
condition, to avoid confusion between planned tests and identification
of degraded or nonconforming conditions. (Note: because of
administrative error, the proposed rule text used the term
``condition,'' although in the proposed rule supplementary information,
the term used was ``activity.'')
The Commission agrees with the commenters and has used ``activity''
in the final rule. Further, the Commission believes that the phrase
``reactor, or any of its structures, systems or components'' is
sufficiently clear to reflect the intent that the determination as to
whether the activity is a test not described in the FSAR, is not
affected by whether it is limited to only one component, or involves a
wider set, up to and including the entire facility. Therefore, the
final rule has been revised to contain a definition of ``test or
experiment not described in the final safety analysis report (as
updated)'' which has minor changes from the definition offered in the
proposed rule.
E. Safety Analysis Report
The Commission proposed to revise the rule language to add a
definition of the ``final safety analysis report (as updated)'' and to
clarify in the evaluation criteria that evaluations need to account for
changes made through other processes that have not yet been included in
an update to the FSAR. Thus, each of the evaluation criteria contained
a phrase referring to evaluations and analyses performed since the last
FSAR update was submitted. The rule referred to FSAR (as updated),
rather than to updated FSAR to account for both non-power reactors who
are not required to submit updates to their FSARs, and to any reactors
between the time of initial licensing and the first required update.
The definition also refers to Final Hazards Summary Report, because a
few facilities were licensed before the rules were revised to require
submittal of FSARs.
Commenters generally supported the idea that the FSAR changes since
the last update submittal needed to be considered in the Sec. 50.59
evaluations, but sought clarification on a few details. Further,
commenters thought the rule language could be simplified by defining in
one place that ``FSAR (as updated)'' includes such information, rather
than including in each evaluation criterion the phrase ``or in
evaluations performed pursuant to this section and safety analyses
performed pursuant to Sec. 50.90 after the last final safety analysis
report was updated pursuant to Sec. 50.71 of this part.''
The Commission has modified the rule text in response to these
comments by adding a new paragraph (c)(3) to explicitly state that the
``FSAR (as updated)'' for purposes of implementing this paragraph, also
includes the FSAR update pages resulting from analyses
[[Page 53589]]
and evaluations performed since the last update was submitted.
Accordingly, the statements of the individual evaluation criterion have
been simplified.
Two commenters were concerned that the requirement to consider
other evaluations since the last update submittal would require a
review of all past evaluations to find the most conservative result as
the baseline for these evaluations.
The Commission does not believe that the rule requires such action.
The Commission's intent in stating that for purposes of implementation
of Sec. 50.59, the FSAR (as updated) is considered to include FSAR
changes resulting from evaluations of changes made since the FSAR
update is to ensure that decisions about particular changes are made
with the most complete and accurate information. If other changes did
not impact upon the accuracy of the FSAR, they would not need to be
examined. If as a result of other changes, the licensee will need to
revise the FSAR at the next update because the present information is
no longer accurate following that change, that information may be
relevant to evaluation of a future change that involves that part of
the FSAR. Indeed, for nonpower reactors, this process has already been
necessary because these facilities are not required to submit updates
to their safety analysis report. Nevertheless, they must ensure that
proposed changes are judged with respect to the existing facility, not
the facility as originally described in the FSAR at time of licensing.
This requirement does not make these evaluations part of the updated
FSAR pursuant to Sec. 50.71(e); that rule requires that the FSAR be
updated to reflect the effects of the changes and evaluations, not that
the evaluations themselves become part of the updated FSAR. Rather, the
intent of the requirement is that the changes that were the subject of
these evaluations be considered in the process of determining what the
``facility as described'' now is such that the reference for subsequent
evaluations is complete and accurate.
One commenter stated that it should be made clear that the FSAR (as
updated) includes the TS and bases because these documents sometimes
contain information, such as applicable operating modes, not in the
FSAR that is relevant to the evaluation process. A few other commenters
thought the definition for ``FSAR'' should include other documents such
as staff safety evaluations, selected commitments and other licensing
documents.
The Commission does not agree that these documents fall within the
required scope of the rule, or that they are part of the FSAR. However,
as noted in existing guidance, licensees are free to refer to other
documents to assist in understanding the implications of the change,
but the rule language does not require such reviews.
F. Minimal Increase Principle
Strict interpretation of the existing rule language related to the
probability of an accident or a malfunction has lead to significant
burden to the industry with no clear safety benefits. Therefore, in the
proposed rule, the Commission relaxed the standard for which prior NRC
review would be required by revising existing paragraph
Sec. 50.59(a)(2)(i) of the rule. The specific proposal was to replace
the phrase ``may be increased'' with ``would result in more than a
minimal increase.'' As previously discussed, the present
Sec. 50.59(a)(2)(i) is being expanded into four separate criteria, two
for occurrence of accidents and malfunctions and two for consequences.
The information that can be revised under Sec. 50.59 is limited to
that which does not require review under any other sections of the
regulations; thus, it is information is of less direct importance to
public health and safety. In consideration of the conservatisms in NRC
design and analysis requirements and acceptance criteria, ``minimal''
variations in probability of occurrence or consequences of accidents
and malfunctions should not affect the basis for the previous licensing
decision. During the plant licensing process, accident probabilities
were assessed in relative frequencies (such as likely to occur more
than once, likely to occur once during the life of the plant, or
limiting fault that is not likely to occur during the life of the
plant). System train and equipment failures were generally postulated
to gauge the robustness of the design, without estimating their
likelihood of occurrence. In this light, minimal increases in
probability would not significantly change the licensing basis of the
facility and could not impact the conclusions reached about
acceptability of the facility design.
Further, the limits for radiological consequences established in
the regulations and in the Standard Review Plan are conservatively
chosen, so that minimal increases also would not impact the safety
determination if demonstrated by a suitably conservative analysis. The
Commission therefore concluded that the proposed criteria would provide
reasonable assurance that those changes that would affect the NRC's
basis for licensing would be identified as requiring NRC approval
before implementation. The proposed revisions to the Sec. 50.59
criteria would provide some degree of flexibility for licensees to make
changes with smaller impacts without the need to obtain a license
amendment.
On the other hand, the Commission intends to limit the amount of
increase in probability or consequences of accidents such that it
remains substantially less than a ``significant increase'' as referred
to in Sec. 50.92. In accordance with Sec. 50.92, a license amendment
involving a significant increase in the probability or consequences of
an accident previously evaluated would be categorized as a
``significant hazards considerations'' and any hearing must be
completed prior to issuance of the amendment.
Although the final rule allows minimal increases, licensees still
must meet applicable regulatory limits and other acceptance criteria to
which they are committed (such as are contained in Regulatory Guides
and nationally recognized industry consensus standards, e.g., the ASME
B&PV Code and IEEE Standards). Further, departures from the design,
fabrication, construction, testing, and performance requirements as
outlined in the General Design Criteria (appendix A to part 50) are not
compatible with a ``no more than minimal increase'' standard. Because
the ``no more than minimal'' standard allows for there to be some
increase compared to the current requirement, which would have required
any increase to be submitted for prior staff review, NRC needs to
establish a point beyond which one would conclude that the increase is
not minimal. Application of the ``minimal increase'' concept to the
specific criteria in the revised final rule is discussed in the next
sections.
G. Section 50.59 (c)(2) Criteria on Increases in Probability or
Consequences
For each of the four evaluation criteria replacing existing
Sec. 50.59(a)(i), the Commission presented language in the proposed
rule reflecting the ``minimal increase'' principle. Resolution of each
of these criteria is discussed below, including consideration of the
public comments.
For each criterion proposed, the Commission had presented guidance
on how the rule could be met, including values as to when the
Commission would conclude that each revised criterion is not met.
Comments received on this guidance are discussed below. The Commission
also notes that regulatory guidance will be provided that is derived
from this discussion.
[[Page 53590]]
As the rule provides a qualitative standard of ``no more than
minimal,'' quantitative calculations are not required except for those
instances in which a licensee decides to offer quantitative arguments
as part of its evaluation. This is expected to occur for some instances
involving increases in consequences, where licensees may perform
calculations of the predicted dose from postulated accidents.
(i) More Than a Minimal Increase in the Frequency of Occurrence of an
Accident Previously Evaluated
For criterion (i), the final rule requires prior NRC approval if
the change results in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the FSAR (as
updated). Several commenters agreed with the premise that ``minimal''
increases in probability of accidents should not require prior NRC
approval. No specific comments were received on the rule language
itself. Issues about guidance are discussed below.
The only change made by the Commission in the final rule language
from the proposed rule is the substitution of ``frequency'' for
``probability.'' This was done to provide a better representation of
the attribute of concern, that is, occurrence over some period of time,
and to emphasize that what is of interest is whether the proposed
change has the effect of making the accident occur more often.
Guidance for Frequency of Accidents
In the proposed rule, the Commission offered guidance concerning
``minimal'' with respect to increases in probability (now frequency).
Several comments were received on certain of these statements, as noted
below.
First, the Commission had noted that the current guidance in NEI
96-07 stating: ``Where a change in probability is so small or the
uncertainties in determining whether a change in probability has
occurred are such that it cannot be reasonably concluded that the
probability has actually changed (i.e. there is no clear trend towards
increasing the probability), the change need not be considered an
increase in probability'' satisfies the proposed NRC standard for
increases in frequency of an accident. Commenters agreed with the
characterization that this guidance would satisfy the rule, but also
noted that the rule language provides more flexibility than is
presently afforded by the NEI guidance.
Second, the Commission had stated that in order to be considered as
a minimal increase, the resulting frequency of occurrence (considering
the change, test, or experiment) must still satisfy the event frequency
classification provided in the licensee's FSAR (as updated). Typically,
these would be anticipated operational occurrence (expected once a
year) or design basis accidents (not expected during life of plant, but
sufficiently credible to require mitigation). The use of frequency
classifications will not apply for all facilities subject to
Secs. 50.59 or 72.48, but is included here because it was a
consideration in the licensing of most operating power plants. Some
commenters sought clarification as to whether increases that remain
within the frequency classification would satisfy the ``no more than
minimal increase'' criterion. Changes that result in a change in
classification do not meet the standard; however, remaining within the
classification is not sufficient to conclude that no more than a
minimal increase has occurred because qualitative judgments are not as
rigorous as quantitative assessments and the accident categories and
their uncertainties may be large. The Commission agrees that the effect
of the change on the frequency of the accident must be discernible and
attributable to the change in order to exceed the ``more than minimal''
increase standard, as compared to uncertainty about the existing
frequency value and how it might be quantified.
Some commenters stated that the ``minimal increase in probability''
standard was too vague and sought more explicit criteria. Others
requested quantitative standards for determining minimal increases in
probability, and in particular, guidance for using risk insights or
probabilistic risk analysis to determine when a more than minimal
increase in probability has occurred. For instance, commenters thought
that the values for changes in core damage frequency or large early
release frequency in Regulatory Guide (RG) 1.174, ``An Approach for
Using Probabilistic Risk Assessment in Risk-Informed Decisions on
Plant-Specific Changes to the Licensing Basis,'' might be used.
However, this RG was developed for the purpose of guiding changes to
the licensing basis where the staff was reviewing and approving the
change, not for changes made under Sec. 50.59. The Commission concludes
that if use is to be made of PRA in Sec. 50.59, more fundamental
changes to the rule would be necessary to provide a coherent set of
requirements, in that Sec. 50.59 deals with design basis events, and RG
1.174 deals with risk including that from severe accidents beyond the
design basis. In addition, RG 1.174 is specifically dealing with
operating power reactors. Applicability to other facilities would need
to be examined. The Commission acknowledges that it may be possible to
develop more guidance that could be used in a quantitative sense to
judge minimal increases. As part of development of the guidance, the
NRC will consider using the values developed as part of the revised
oversight process (SECY-99-07), so that if the resultant likelihood of
occurrence remains well within the acceptable ranges given for
initiating events, that the increase is ``minimal.''
(ii) Minimal Increase in Likelihood of Malfunction of Structures,
Systems or Components
In the proposed rule, Sec. 50.59(c)(2)(ii) would require NRC
approval for a change that would result in ``more than a minimal
increase in the probability of malfunction of equipment important to
safety previously evaluated in the FSAR (as updated).'' Similar changes
were proposed in Sec. 72.48(c)(2)(ii), except for use of the term
``structures, systems, and components'' (SSCs) rather than equipment.
These differences in wording reflected differences between existing
language in Secs. 50.59 and 72.48. Commenters supported the idea that
``minimal'' increases should not require approval. Commenters also
suggested that the terminology in Secs. 50.59 and 72.48 should be made
more consistent between the two sections.
In the final rule, the Commission has revised the criterion in
Sec. 50.59 by referring to SSC rather than to equipment. The Commission
concludes that the term ``SSC'' is commonly used in both parts 50 and
72 and is well understood, and that ``equipment'' was an older term
that does not have a unique meaning requiring its use. For the final
rule, the Commission has also substituted the term ``likelihood'' for
``probability.'' This change was made to acknowledge that while the
criterion refers to ``minimal'' increases, the Commission is not
implying that quantitative assessments are expected. The Commission
concludes that the word ``likelihood'' is more generally understood to
represent qualitative judgments.
Guidance for Likelihood of Occurrence of Malfunction
In the proposed rule, the Commission discussed the following
positions as guidance for implementing the criterion of a ``more than
minimal'' increase in probability (now likelihood) of a malfunction of
equipment (now SSC).
First, the Commission noted that the existing guidance in NEI 96-07
states:
[[Page 53591]]
``Where a change in probability is so small or the uncertainties in
determining whether a change in probability has occurred are such that
it cannot be reasonably concluded that the probability has actually
changed (i.e. there is no clear trend towards increasing the
probability), the change need not be considered an increase in
probability.'' Continued use of this guidance for a determination of
whether criterion (i) has been met is satisfactory. Commenters agreed
with this guidance, but also believe that this does not represent the
outer bound of what would be acceptable to meet the rule. The
Commission agrees with this comment.
Second, the Commission concluded that the likelihood of malfunction
of SSC important to safety previously evaluated in the FSAR (as
updated) would not be more than minimally increased if ``design bases''
assumptions and requirements are still satisfied (i.e., the seismic or
wind loadings, qualification specifications, etc). Thus, for instance,
a change that would cause piping stresses to exceed their code
allowable values would be more than a minimal increase in likelihood of
malfunction. Commenters stated that if design basis requirements are
met, there is no increase in probability. The Commission agrees with
the essence of this comment, but was attempting to help licensees
comply with the rule language by offering ways of demonstrating that
the criterion is satisfied. Changes that would invalidate specific
commitments made for redundancy, diversity, separation, and other such
design characteristics, would be considered as ``more than a minimal
increase in likelihood of malfunction,'' and thus would require prior
NRC approval.
In the proposed rule, the Commission stated that for purposes of
determining whether this criterion has been satisfied, the probability
of malfunction would be no more than minimally increased if a new
failure mode as likely as existing modes is introduced. Some commenters
indicated that the presence of new failure modes should not be a
determinant as to whether probability of malfunction has increased;
rather, it is whether the effects of the failure modes have previously
been considered that would determine the need for NRC review consistent
with Sec. 50.59(c)(2)(vi). The Commission finds that the question of
likelihood is not addressed if new failure modes are only examined with
respect to criterion (vi), since that criterion looks only at whether
the effects of the failure are bounded, not how likely it is to occur.
However, since likelihood can be increased regardless of whether new
failure modes are involved, the Commission has deleted this statement
as proposed guidance for assessing increases in likelihood.
Additions of components to a system (cabling, manual valves,
protective features) would not generally be viewed as more than a
minimal increase in likelihood of malfunction, provided that applicable
design and quality standards are followed. For example, adding
protective devices to breakers, or installing an additional drain line
(with appropriate isolation capability) would not be increases in
likelihood of malfunction. However, there could be situations where
such additions would impact upon how a system performs its functions
that might not satisfy the Sec. 50.59 criteria (for example, a cross-
connect between trains that is not suitably isolated).
Substitution of one type of component for another (as for instance,
an air-operated valve for a motor-operated valve), would also be viewed
as no more than a minimal increase in likelihood of malfunction,
provided requirements for redundant motive force, quality, and other
requirements are met (and of course that any new failure modes are
already bounded by the analysis).
(iii) and (iv) Minimal Increases in Consequences of Accident or
Malfunction
In the proposed rule, the Commission revised the existing criterion
concerning increases in consequences from a standard of ``may be
increased'' to ``more than minimally increased,'' and separated the two
statements on consequences within Sec. 50.59(a)(2)(i) into separate
criteria. Only a few comments were received concerning the rule
language itself. One commenter stated that the two criteria on
consequences should not be separate, since consequences would only
result from accidents, and having another criterion might force
evaluators either to duplicate their documentation, or struggle to
explain why consequences were not increased for malfunctions. The
Commission concludes that having separate criteria provides greater
clarity and is consistent with common practice. Further, the criteria
cover different types of changes, that is, some that arise from
malfunctions (such as failure of a waste tank or filter systems), and
others that might arise from changes in source term or timing of
mitigation systems, that are more pertinent to ``accidents.'' Licensees
may combine their responses to questions and reference other sections
when preparing evaluations.
Commenters requested two areas of clarification. First, they asked
if consequences refers only to radiological consequences (dose), and
second whether consequences refers only to those associated with
accidents and not from normal operations or anticipated operational
occurrences. The rule reference to consequences is intended to relate
directly to radiological consequences, and not to other outcomes that
are covered by the remaining criteria. Secondly, the Commission notes
that 10 CFR part 20 establishes requirements for protection against
radiation during normal operations. For anticipated occupational
occurrences, NRC requirements are such that there should not be any
radiological consequences. However, the Commission also wishes to
clarify that ``consequences of accidents'' includes not only offsite
exposure, but also dose to operators in the control room (in accordance
with General Design Criterion 19 of appendix A to 10 CFR part 50) or
other onsite personnel, resulting from accidents and malfunctions
previously evaluated in the FSAR.
The language in the rule for criterion (iii) was unchanged from the
proposed rule; for criterion (iv), the term ``systems, structures, or
components'' was substituted for ``equipment'' as it was for criterion
(ii), for the reasons already discussed.
Guidance for Minimal Increase in Consequences
In the proposed rule, the Commission had discussed several
positions that might be helpful in developing guidance that would
successfully implement the revised rule. First, the Commission agreed
with the guidance in NEI 96-07 which states: ``Where a change in
consequences is so small or the uncertainties in determining whether a
change in consequences has occurred are such that it cannot be
reasonably concluded that the consequences have actually changed (i.e.,
there is no clear trend towards increasing the consequences), the
change need not be considered an increase in consequences.'' No
specific comments were received on this point.
Second, if a licensee has performed an analysis with certain
bounding assumptions, and the change would increase a specific
parameter from its present value to a different value that is still
bounded by the value assumed in the analysis, the NRC concludes that
such a change satisfies the criterion of ``no more than a minimal
increase in consequences.'' In fact, as noted by some of the comments,
this is no
[[Page 53592]]
increase in consequences, because the bounding analysis is what
determines the value from which a change is being judged.
Third, if a licensee would need to change its design basis
assumptions or analytical methods, or both, to demonstrate that the
change in consequences satisfies this guidance, then the NRC does not
view the change as minimal and would expect the licensee to submit a
license amendment for such a change. This position is consistent with
the logic presented as the basis for implementing new criterion
Sec. 50.59(c)(2)(viii), which will be discussed in greater detail
below. Some commenters thought that adopting methodologies that have
been approved by NRC in certain contexts (such as use of International
Conference on Radiation Protection (ICRP) dose conversion factors, or
credit for suppression pool scrubbing) should be allowable under
Sec. 50.59. New criterion (viii), discussed in section J below,
specifies under what conditions changes to evaluation methods can be
changed without prior NRC approval.
In the proposed rule, the Commission proposed a graduated approach,
consistent with the concept of ``minimal'' being small enough so as not
to impact the basis for the acceptability of the previous licensing
decision. The Commission proposed that when the facility is far from
the limit, a larger increase could be accommodated without concern
about impact on the basis for acceptability. The Commission did not
believe that allowing increases up to the regulatory values without
approval was consistent with a ``minimal'' increase standard, and was
not consistent with the purpose of the rule, that is, to allow the NRC
the opportunity to confirm the adequacy of the licensee's review of the
change before it is implemented.
The proposed rule offered three different ways to define what would
constitute a minimal increase in consequences. Most commenters favored
the third method (10% of the difference between the calculated value
and the regulatory guidelines) over the other two. Other commenters
thought the limits themselves should be the point at which NRC review
would be needed, or offered other suggestions, such as allowing 20
percent of the difference. Comments were also received about the use of
Standard Review Plan guideline values 2 as they are not in
the regulations and that for some plants, the existing analysis may
exceed the guideline such that no changes would be allowed. Some
commenters also expressed concern about the criterion for those
situations where a previous change may have resulted in a decrease in
consequences, and a subsequent change that increased consequences would
exceed the 10 percent difference, but would not have done so if the
first change had not occurred.
---------------------------------------------------------------------------
\2\ In the Standard Review Plan, NUREG-0800, the NRC established
acceptance criteria for certain events that are considered of
greater likelihood than the limiting accidents as a small fraction
of the part 100 guidelines. Thus, for instance, for a steam
generator tube rupture, the SRP guideline is that the dose be 10
percent of the part 100 value. For the postulated accident with an
assumed preaccident iodine spike in the reactor coolant at the time
the tube rupture occurs, the full part 100 value is the acceptance
criterion.
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During the comment period, some commenters were concerned that as
the rule is currently planned to be implemented, they would have no
flexibility under the rule if their calculated consequence values were
already in excess of the current SRP guidelines. In general, the
Commission agrees that for cases where a licensee is licensed with
calculated consequences in excess of the established SRP guidelines,
only limited flexibility under this provision of the revised rule would
exist for changes that increased the calculated radiological
consequences of accidents. In this regard, the Commission does view
differences of about 0.1 rem as being within the error or uncertainty
of design basis-type radiological consequences analysis such that NRC
review of such changes is not needed.
The Commission has taken these comments into account in revising
the ``minimal'' increases in consequences aspects of the final rule.
The Commission will conclude that the requirements of the rule are met
if the calculated doses from a change at a facility would be less than
10 percent of the remaining margin between current calculated dose
values and acceptance values in the regulations 3 (e.g., GDC
19 or part 100) for the particular accident. Under this approach, the
threshold for what constitutes a minimal change varies as a licensee
approaches the regulatory limit. The amount of change allowed would
decrease as the limit is approached, and the limit could not be
exceeded without prior NRC review. Specifically, it is no more than a
minimal increase in consequences if the increase is less than or equal
to the more limiting of either 10 percent of the difference between the
existing calculated value and the regulatory guideline value (10 CFR
part 100 or GDC 19 as applicable), or has reached the SRP guideline
value for the particular design basis event.
---------------------------------------------------------------------------
\3\ GDC 19 requires adequate radiation protection to permit
access and occupancy of the control room under accident conditions
without personnel receiving radiation exposure in excess of 5 rem
whole body or its equivalent to any part of the body, for the
duration of the accident. Part 100 establishes requirements for
exclusion area and low population zones around the reactor so that
an individual located at any point on its boundary immediately
following onset of the postulated fission product release would not
receive a total radiation dose to the whole body in excess of 25 rem
or a total radiation dose of 300 rem to the thyroid for iodine
exposure. For future applications, as noted in subpart B to 10 CFR
part 100, the radiological consequences are to meet the criteria
stated in Sec. 50.34(a)(1), which sets a dose of 25 rem total
effective dose equivalent (TEDE).
---------------------------------------------------------------------------
Examples
The Commission has selected several examples to illustrate the
implementation of this criterion. In each example, the Commission
assumes that the calculated consequences do not include changes in
methodology. As discussed later, changes in methodology used to
calculate radiological consequences would fail new criterion (viii) of
the revised rule and require prior NRC review regardless of how small
the increase would be in the calculated radiological consequences.
Example 1 involves a case in which a licensee has a calculated fuel
handling accident (FHA) dose of 50 rem to the thyroid at the exclusion
area boundary. Because of some change in the facility, the calculated
FHA dose increases to 70 rem. Under the revised final rule, ten percent
of the difference between the calculated value and the regulatory
limits is 25 rem (10% of 250). The SRP acceptance guideline is 75 rem.
Since the calculated increase is less than 25 rem and the total is less
than the SRP acceptance guidelines, then the revised Sec. 50.59
consequence criterion would not trigger the need for a prior NRC review
and a licensee may make the change to the facility.
Example 2 involves a case in which the calculated consequences for
a steam generator tube rupture accident are 25 rem at the exclusion
area boundary. Because of a change in the plant, the calculated
consequences increase to 29 rem. The implementation of the revised rule
language would permit these changes to occur because the new calculated
doses do not exceed the established SRP acceptance criteria nor does
the incremental change in consequences (4 rem) exceed 10 percent of the
difference between the previous calculated value and the regulatory
limit of 300 rem. Ten percent of the difference between the acceptance
criteria (300 rem) and the calculated value (25) is 27.5 (10% of 275)
rem;
[[Page 53593]]
since 4 is less than 27.5, this change satisfies the criterion.
Example 3 involves a case in which the calculated consequences of a
fuel handling accident are 25 rem to the thyroid at the exclusion area
boundary. Because of a proposed change in the facility, the calculated
consequences increase to 65 rem. For this case, the revised calculated
consequences are still less than the SRP acceptance guidelines of 75
rem; however, the incremental increase in consequences (40 rem) exceeds
the 10 percent of the difference to the regulatory limit of 300 rem
(which would be 27.5 rem). For this example, the change results in more
than a minimal increase in consequences and thus requires NRC approval
pursuant to Sec. 50.59(c)(2)(iii).
If Example 3 had been an event for which no SRP value was
specifically established, so that the part 100 guideline was the only
applicable standard, the rationale would be that an increase up to 52.5
(25+27.5) rem would meet the ``minimal increase'' criterion.
Example 4 involves a case where the calculated dose to the control
room operators following a loss of coolant accident is 4 rem whole
body. A change is made to the control room ventilation system such that
the calculated dose increases to 4.5 rem. The regulations dictate that
the control room doses are to be controlled to less than 5 rem by
General Design Criterion 19. Although the new calculated doses are less
than the regulatory limits for the operators, the incremental increase
in dose (0.5 rem) exceeds the value of 10 percent of the difference
between the previously calculated value and the regulatory value (10%
of 1 rem = 0.1 rem). This change would require prior NRC review before
the licensee could implement the change.
As an example of the ``calculational error'' concept, suppose the
existing approved analysis for a fuel handling accident at a plant
predicts an offsite dose to the thyroid of 77 rem. The SRP acceptance
guideline for this event is 75 rem. The change that a licensee wishes
to make would predict an increase in the calculated dose from 77 to
77.1 rem. In this case, the proposed change could be made under
Sec. 50.59 because the calculated value, even though greater than the
SRP value, is satisfied within the level of uncertainty specified
above. However, for this example, the Commission notes that increases
in consequences that would increase the calculated consequences to 77.2
rem would require prior NRC review before the specific change could be
implemented.
H. Possibility of an Accident of a Different Type From Any Previously
Evaluated in the Final Safety Analysis Report (as Updated) Is Created
The Commission had proposed that the language in existing
Sec. 50.59(a)(2)(ii), renumbered to Sec. 50.59(c)(2)(v) in the proposed
rule, be revised to read ``(would) create the possibility for a design
basis accident of a different type from any previously evaluated in the
final safety analysis report (as updated).'' This change had two
parts--the first, changing from may be created to ``would create'' and
the second being the insertion of the phrase ``design basis.'' The
purpose of the first change was to provide some flexibility to
licensees. Thus, rather than having to prove that an accident had not
been created, under this rule language, a licensee would need to
request a license amendment only if it could be reasonably concluded
that the possibility of an accident of a different type is created by
the change, test, or experiment. The intent of the second change was to
indicate that in referring to ``accidents'' in Secs. 50.59 and 72.48,
the Commission had in mind creation of accidents of the likelihood and
significance of those that, had the possibility already existed, would
have been a design basis accident in the FSAR. Thus, ``accidents'' that
would require multiple independent failures or other circumstances in
order to ``be created'' would not fall within this criterion.
For an accident to be of a different type, a few commenters thought
that the accident must result in a new or greater release path than
originally considered, result in a new fission product barrier failure
mode, or create a new sequence of events that results in significant
cladding failure, ``such that the accident would have been included if
the FSAR were being written today.'' The Commission agrees that these
are useful considerations for determining whether a change results in
an accident of a different type.
One commenter noted that for certain older facilities, the term
``design basis accident'' was only applied to a very small set of
events. Other commenters thought that accidents must be ``credible'' to
be ``created.'' Another commenter was concerned that a slightly
different initiator leading to the same design basis accident might be
viewed as an accident of a different type.
One commenter stated that ``accident of a different type'' should
be changed to ``accident with a different result,'' for consistency
with the criterion on malfunction. However, the Commission also notes
the similarity with the criterion in Sec. 50.92 (for no significant
hazards consideration determination). Allowing changes that result in
an accident of a different type (even if the result has previously been
analyzed) appears inconsistent with the criterion in Sec. 50.92.
The Commission has concluded that use of the modifier ``design
basis'' with respect to accidents of a different type in the rule
language may be confusing because, by the terms of the rule, accidents
of a different type are distinct from those (design basis) accidents
evaluated in the FSAR. Therefore, in the final rule, the Commission
removed the phrase ``design basis.'' The Commission agrees that the
accident must be credible in the sense noted above, of having been
created within the range of assumptions previously considered (e.g.,
random single failure, loss of offsite power, no reliance on non-
safety-grade equipment, etc.), and that a new initiator of the same
accident is not a ``different type'' (but may affect the frequency of
that accident under Sec. 50.59(c)(2)(i)).
Therefore, the final rule uses the same language as is currently
contained in the existing rule, concerning accidents of a different
type, except for changing the phrase ``possibility * * * may be
created'' to ``would create the possibility.''
Need for Definition of Accident
In addition, the Commission had requested comment as to the need
for a definition of accident, and offered a specific definition for
comment. The term ``accident'' also appears in other evaluation
criteria, specifically, Secs. 50.59(c)(2)(i) and 50.59(c)(2)(iii), in
the context of accidents previously evaluated in the FSAR.
Several comments were received on the proposed definition of
accident. Most commenters felt that a definition in the rule was not
necessary, and most also disagreed with the specific definition offered
in some respect. Commenters generally agreed that accidents include
design basis accidents (typically analyzed in Chapters 6 and 15 of the
FSAR), anticipated occupational occurrences, external events that the
plant is required to withstand and other special events that are
analyzed to demonstrate safety. Included within the set of accidents
are those scenarios for which requirements have been established for
the facility either to withstand or cope with the event. Notable
examples include pressurized thermal shock events (Sec. 50.61),
anticipated transient without scram (Sec. 50.62) and station blackout
(Sec. 50.63).
[[Page 53594]]
Commenters also noted that external events, such as earthquakes, high
winds, floods, and missiles can be treated as causes of malfunctions of
SSC, rather than accidents. Some suggested that examples or a list of
accidents could be presented in the implementation guidance.
The Commission concludes that a definition of accident is not
necessary in the final rule and that examples of accidents are best
discussed in rule implementation guidance.
I. Possibility of a Malfunction of Structures, System, or Components
Important to Safety With a Different Result From Any Previously
Evaluated in the Final Safety Analysis Report (as Updated) is Created
In the proposed rule, the Commission modified the remaining part of
existing Sec. 50.59(a)(2)(ii), concerning malfunctions of a different
type by creating a new criterion (vi), that would require approval if a
change, test, or experiment would ``create a possibility for a
malfunction of equipment important to safety with a different result
than any evaluated previously in the final safety analysis report (as
updated).''
Comments were supportive of the change from ``different type'' to
``different result,'' and of the change from ``may be'' to ``is''
created. Some commenters objected to the insertion of the phrase
``important to safety'' and suggested other phrases, such as ``safety-
related'' or ``FSAR-described.'' Others suggested that the terminology
in Secs. 50.59 and 72.48 should be made consistent (the former refers
to equipment; the latter to systems, structures or components).
In the final rule, The Commission has revised the existing
criterion to read ``create a possibility for a malfunction of an SSC
important to safety with a different result from any previously
evaluated in the final safety analysis report (as updated).'' The
Commission concludes that the term ``SSC'' is commonly used in both
parts 50 and 72 and is well-understood, and that equipment was an older
term that does not have a unique meaning requiring its use. The
modifier ``important to safety'' was considered as always being part of
the criterion in practice, and that its omission from the rule was
viewed as editorial and not substantive. Other terms might have the
effect of limiting or broadening the scope of SSC to be considered. The
Commission notes that since the overall scope of Sec. 50.59 is the
facility as described in the FSAR, there is no need to use that phrase
in characterizing which SSC need be considered with respect to
malfunctions.
Guidance for Malfunction With a Different Result
The proposed rule discussion further stated that this determination
should be made either at the component level, or consistent with the
failure modes and effects analyses (FMEA), taking into account single
failure assumptions, and the level of the change being made. Several
commenters stated that this guidance should be revised to refer only to
the failure modes and effects analysis in the FSAR, and not to specify
the component level. The Commission agrees that this criterion should
be considered with respect to the FMEA, but also notes that certain
changes may require a new FMEA, which would then need to be evaluated
as to whether the effects of the malfunctions are bounding.
J. Replacement Criteria for ``Margin of Safety as Defined in the Basis
for Any Technical Specification is Reduced''
The phrases ``margin of safety'' and ``as defined in the basis for
any technical specification'' in the third criterion in existing
Sec. 50.59(a)(2) have been the subject of differing interpretations for
a number of years because Sec. 50.59 does not define what constitutes a
margin of safety or a basis for any technical specification in the
context of Secs. 50.59 and 72.48.
The Commission continues to believe that changes representing a
potentially significant decrease in certain margins should require NRC
review and approval prior to their implementation. Margins within the
plant design and in the established licensing basis exist on many
levels. There are margins from the assumptions of initial conditions,
conservatisms such as computer modeling and codes to account for
uncertainties, allowances for instrument drift and system response
time, redundancy and independence of components. Margins are built into
the facility to account for routine plant fluctuations and transients
and response to accident conditions. Margins also exist in the
established regulatory acceptance criteria to be met for response to
various accidents and transients. The acceptance criteria are
established at a value that accounts for uncertainty about physical
properties and other variability. As a result, substantial margins are
provided by the regulatory envelope within which a plant has
demonstrated its ability to respond to a spectrum of design basis
accidents. In sum, not every margin is important to assuring safety
such that changes in that margin must be reviewed and approved by the
NRC prior to their implementation. However, the Commission recognizes
that precisely delineating the margins for which changes would require
prior NRC review and approval is a difficult task. A change criterion
which does not directly refer to margins, but which nonetheless
indirectly assures that important design and licensing basis margins
are not changed without prior NRC review and approval, is an acceptable
alternative that would meet the Commission's goal of assuring
regulatory review of potentially significant changes to certain
margins. Such an approach avoids having to describe in the rule the
margins of regulatory interest, and the nature of the change in margin
for which prior NRC review and approval would be required.
In the proposed rule, the Commission solicited public comment on
several options. The Commission also requested the public to provide
alternative means for control of margin.
Option 1 in Proposed Rule
The first option in the proposed rule was to control inputs to
analyses and the methods and criteria that establish TS. Under this
option, the Commission would conclude that the analyses and information
in the FSAR establish the basis for the margins of safety for the TS.
Thus, the Commission's proposal would have added a definition for
``reduction in margin of safety associated with any technical
specification'' and conformed the criterion for needing a license
amendment in new Sec. 50.59(c)(2). Although this option would maintain
the safety analyses that underlie the TS, this approach also would have
the effect of giving all input values and assumptions within the FSAR
the weight of TS (even though they are not included in the TS), which
is inconsistent with the philosophy in Sec. 50.36. In many instances,
changes to inputs can be accommodated by other available margins so
that the licensing envelope is preserved. Several comments expressed
strong concern that this option would be too restrictive, for the
reasons noted above. The Commission agrees with these concerns and
concludes that the approach is not consistent with the intent of the
original rule. In this light, this option of requiring prior NRC
approval for any change to input parameters associated with TS was
rejected as an approach for the final rule.
[[Page 53595]]
Option 2 in Proposed Rule
The proposed rule contained a second option that was a proposal to
delete the ``margin of safety'' criterion completely. Instead, the
Commission would rely upon the other criteria in Sec. 50.59, as well as
the regulatory requirement that all changes to TS be reviewed and
approved by the NRC, to assure that there are no significant adverse
changes to margins in design and operation. If this option were
adopted, the Commission would argue that there is no need for prior
review of changes that do not satisfy any of the other evaluation
criteria in view of ``risk-informed'' insights and greater
understanding of the margins that exist through meeting the body of
regulatory requirements. The Commission also sought comment on whether
any of the other evaluation criteria should be revised if this approach
were adopted.
A significant number of comments were received in support of the
proposal to delete margin of safety as an evaluation criterion. In
support of their position, commenters noted that TS and the other six
evaluation criteria, in conjunction with other regulatory requirements
for design, testing, and operation, make the margin question moot. The
Commission did not adopt this proposal because of the variability in
existing TS, and uncertainties about how licensees might gauge the
other evaluation criteria for specific changes.
Option 3 in Proposed Rule
In the Federal Register notice, the NRC also offered a set of
options that focused on control of margins associated with results of
analyses. Instead of focusing on the inputs to safety analyses, these
options would focus on the results of the safety analyses in order to
determine whether changes to operational characteristics or other
information described in the FSAR (as updated) would reduce the level
of protection reflected by the results of safety analyses.
In developing which results would be governed by this evaluation
criterion, the Commission considered what aspects of the facility
safety are controlled by other requirements and thus what other
information might a ``margin'' criterion be intended to capture. As
part of the licensing review for a facility, the NRC established a
level of required performance (which will be referred to in this
discussion as acceptance criteria) for certain physical parameters,
such as those that define the integrity of the fission product barriers
(e.g., fuel cladding, reactor coolant system boundary, and
containment). Satisfying these acceptance criteria produces a margin of
safety to loss of barrier integrity. The safety analyses presented in
the FSAR (as updated) demonstrate that the response of the barriers to
the postulated accidents, transients, and malfunctions meets the
acceptance criteria. Thus, in constructing the options for comment, the
Commission suggested a more explicit linkage between when ``margin of
safety'' needed to be preserved to the response of the fission product
barriers relied upon to provide protection from uncontrolled release of
radioactivity.
In the range of options, the Commission also suggested that certain
mitigation system capability, as, for instance engineered safety
feature performance parameters (flow rates, efficiencies, etc.) also
might be considered with respect to margin, and asked for comment
whether there were other parameters that should be explicitly accounted
for in any criterion on ``margin of safety.''
As part of these options, the Commission also offered different
approaches to how much flexibility should be allowed, as for instance,
minimal reductions, or use of limits as the point at which reductions
in margin would be determined. Also, as discussed later, the Commission
asked in the proposed rule whether changes to evaluation methods should
also be controlled.
Comment Summary for Option 3: The Commission received a large
number of comments on the various suboptions under Option 3 concerning
results of analyses. With respect to the identification of those
parameters to control, many of the commenters who supported a
``margin'' concept based upon limits for results, believed that the
parameters should be limited to those that directly affect fission
product barriers and for which there are clearly defined limits. One
commenter thought that a criterion on margin is not needed for a
reactor that was being decommissioned. Commenters also thought that
mitigation system performance was best controlled by other criteria,
such as those concerning malfunction of SSC, or consequences of
accidents. It was also noted that important characteristics of
mitigation systems are governed by TS. With respect to parameters that
might be used under part 72, commenters stated that these should be
those with the potential to increase the likelihood or the amount of
offsite release, specifically, such things as fuel and cladding
temperature, cask temperature and internal pressure, and cask stresses.
For the question as to when NRC approval is needed, comments can be
grouped into two main themes: those that are supporting the position
currently included in NEI 96-07 related to acceptance limits as being
the point of departure for reduction in margin, and those supporting a
new proposal from NEI. No commenters supported either a ``no reduction
in results'' or a ``minimal'' standard, or any type of graduated
approach such as that discussed earlier for consequences. As part of
its comments on the proposed rule, the NEI proposed to replace the
existing margin of safety criterion with one that states that a change
requires prior NRC approval if it would result in a design basis limit
directly related to integrity of the fuel cladding, the reactor coolant
system boundary, or the containment boundary being exceeded or altered.
Their proposal is similar in several respects to the guidance offered
in NEI 96-07, with respect to using ``limits'' as the point at which a
reduction in margin occurs, and in focusing on parameters for fission
product barriers as being the instances where there is margin to
protect. The difference is the concept of ``design basis limits'' as
represented in the FSAR instead of acceptance limits that might be
found in other documents. Further, NEI suggested that as part of the
rule changes to adopt this criterion, the NRC should also delete the
third criterion in Sec. 50.92, which states that a determination of
``no significant hazards consideration'' cannot be made for amendments
that would involve a significant reduction in a margin of safety.
Resolution
In SECY-99-054, dated February 22, 1999, the staff presented an
alternate proposal for the margin of safety criterion. The staff
proposal employed a concept that used the design basis capability for a
SSC as the determinant for when prior staff review would be required.
As presented in the final safety analysis report, there is a design
basis (functions and controlling values of parameters) that determines
the minimum performance requirements for SSCs. The controlling value
for a parameter is the point at which confidence in the capability of
the structure, system or component to perform its intended safety
functions begins to decrease. For many parameters, requirements have
been established in TS; for others, which are not directly controlled
or measured, while certain TS requirements may have been imposed to
keep values within required ranges, inclusion of a criterion
[[Page 53596]]
that verifies that facility changes have not adversely impacted design
basis capability provides assurance of completeness beyond the
requirements for approval of TS changes.
The staff was supportive of the NEI concept of using the design
basis as the determinant of when prior NRC approval was needed. The
staff proposal was a modification of the suggested NEI approach that
would focus on the effectiveness of systems to protect barriers. The
staff thought that the rule language as offered by NEI could be viewed
too narrowly, and might not ensure that changes affecting performance
of mitigation and support systems were appropriately evaluated with
respect to their roles in protecting integrity of the barriers.
Therefore, the staff's proposal was more explicit about the design
basis capabilities of the SSC being used to determine whether approval
of a change was needed. The principal difficulty with this proposal was
uniquely identifying the design basis capabilities for all SSCs that
would need to be satisfied in order to implement the concept.
Since the time that SECY-99-054 was submitted to the Commission,
the NRC has gained a greater understanding of the NEI proposal and how
it would be implemented, and, in particular, how it would be used to
assess changes to mitigation systems and support systems. Although the
NRC agreed that the process described in the NEI comment letter of
December 21, 1998, would be sufficient to ensure that changes to other
systems are appropriately examined with respect to impact upon the
barriers, it was not apparent that the specific rule language suggested
would require licensees to implement such a systematic approach to
examination of design basis limits.
Therefore, the approach contained in the final rule is a
combination of the NEI proposal contained in its comment letter and the
staff proposal contained in SECY-99-054. In the final rule, the
Commission is eliminating the existing criterion on reduction of margin
of safety. In its place, the Commission is adding a new criterion (vii)
that requires prior NRC review of changes that result in a design basis
limit related to the integrity of the fission product barriers being
exceeded or altered.
The final rule also contains a new criterion (viii) related to the
use and control of evaluation methods (see below). These two criteria
together in place of a criterion on margin of safety explicitly cover
those margins that the Commission believes are important to address in
this evaluation process--the first being the margin that exists in the
limits that are to be met, and the second being the margin that exists
from the conservatisms included in the methods used to demonstrate that
requirements are met. Each of these criteria are discussed below.
The Commission concludes that the new criteria (vii) and (viii)
together will maintain safety because they will preserve the design
basis capabilities that protect the integrity of important fission
product barriers, and thus those features that protect against release
of radioactive material. The rule will also control the analyses and
assessment process through control of the methods and will assure that
the required response of the barriers as previously established by NRC
review will be maintained.
The Commission does not plan to make any changes to the criterion
in Sec. 50.92(c)(3), which provides that license amendments involving a
significant reduction in a margin of safety do not meet the criteria
for a ``no significant hazards consideration'' determination as
discussed in section M below.
Final Rule Language
New Criterion (vii)
New criterion (vii) would require a prior NRC review of any change
that would ``result in a design basis limit for a fission product
barrier as described in the FSAR (as updated) being exceeded or
altered.'' For purposes of implementation of this criterion, the
Commission defines design basis limit for a fission product barrier as
the controlling numerical value for a parameter established during the
licensing review as presented in the final safety analysis report for
any parameter(s) used to determine the integrity of a barrier.
Typically, the controlling value for the parameter is set at a point
far enough away from failure that there is confidence in the integrity
of the barrier. As a partial substitute for the previous ``reduction in
margin'' criterion in the former Sec. 50.59(a)(2)(iii), a change which
does not exceed or alter a design basis limit for a fission product
barrier does not involve any reduction in the margin of safety.
The Commission did not retain the suggested wording from commenters
for criterion (vii) which might suggest that the evaluation can be
limited to those changes that are directly related to fuel cladding,
reactor coolant system boundary, and containment boundary. The
Commission believes that a broader initial assessment of parameters is
necessary than that which might be suggested by the term ``directly
related.'' All changes that might affect the design basis limits,
including changes to parameters within mitigation and support systems,
must be evaluated for their effects upon the design basis limits for
the barriers. Further, the Commission used the term ``fission product
barrier,'' rather than listing the specific barriers for operating
power reactors as used by NEI, so that the rule language would be
appropriate for all Part 50 facilities (including non-power reactors,
and reactors undergoing decommissioning). The more general terminology
is also appropriate for the part 72 facilities.
New criterion (vii) narrows the focus for when prior NRC approval
is required to those changes which result in the specific limits that
relate directly to the performance of fission product barriers being
exceeded or altered. For power reactors, these barriers are generally
limited to the fuel cladding, the reactor coolant system pressure
boundary and containment. For a reactor undergoing decommissioning,
where the fuel is stored in the spent fuel pool, the barrier would be
the fuel cladding. For non-power reactors, the fission product barriers
would include, as applicable to the specific reactor, the fuel
cladding, the reactor tank, and the reactor room, building,
confinement, or containment.
The proposed criterion (vii) is equally applicable to independent
spent fuel storage facilities or spent fuel storage cask designs in
part 72. The particular parameters or barriers would be specified in
terms of the barriers against release of radioactivity afforded by fuel
storage facilities. For instance, these would include calculated fuel
temperature or cladding oxidation, and stresses (or pressures) on the
cask structure.
Although the list of fission product barriers includes containment
and other features that prevent the release of radiation, the design
basis limits for these barriers are for parameters such as pressure.
The determination of resultant radiological consequences from leakage
through or breech of these barriers is the subject of criteria (iii)
and (iv), rather than criterion (vii).
Further, design basis limits for certain fission product barriers
may not be applicable to particular facilities or conditions of the
facility (such as permanently shutdown facilities). The determination
as to the need for evaluation of particular barrier parameters or
limits depends upon the safety analyses and information presented in
the FSAR (as updated).
The Commission notes that the new criterion (vii) does not
incorporate the use of a minimal change concept. The
[[Page 53597]]
modification of the criterion to reflect design basis limits as a point
for evaluating when prior NRC review is necessary would not permit
small changes beyond the limits without review.
With respect to changes relating to the design basis capability of
SSCs to perform their functions in those circumstances in which the
change does not cause any design basis limits to be exceeded or
altered, the other evaluation criteria in Sec. 50.59 (as well as other
requirements such as TS or ASME code requirements) provide the
standards for prior NRC approval of such changes.
The rule language that provides that a design basis limit may not
be altered provides important and needed assurance. Changes that
involve alteration of the design basis limit for a fission product
barrier involve such a fundamental alteration of the facility design
that a change, even in the conservative direction, should receive prior
NRC review.
Guidance for Implementation
To satisfy new criterion (vii), licensees must determine the
parameters that would be affected by the proposed change. The affected
parameters are not limited to the specific parameters in the system in
which the change is being made or to parameters that are only directly
linked to the actual fission product barrier. Rather, the design
parameters must include an assessment of all affected parameters,
including design parameters of mitigation and support systems. Once the
parameters are identified, the licensee must establish whether the
parameters have values established in the FSAR, whether the parameters
are controlling parameters that are reference bounds for the design,
and whether the parameter has the potential to affect the performance
of the fission product barrier. If the specific parameter values are
already subject to controls established by the TS or other rules or
regulation, those requirements shall be followed.
After a licensee assesses the information discussed above, it would
need to identify the specific design basis limits that could be
affected for each of the identified parameters. After the licensee
completes its assessment of the change against each design basis limit,
if no design basis limit is altered or exceeded, criterion (vii) is
satisfied, and a licensee may make the change without prior NRC review.
Examples
The NRC has selected several examples to illustrate how the new
criterion (vii) would be implemented. In these examples, it is assumed
that NRC approval is not required because of other reasons, such as
need for a TS change, section 50.55a requirements etc.
Example 1: A plant FSAR states that the function of the auxiliary
feedwater system (AFW) is to provide feedwater flow to the steam
generators following postulated accidents (e.g., main steam line break,
feed line break, small break loss-of-coolant accident), or when a
reactor trip occurs coincident with a loss-of-offsite power. The FSAR
states that 700 gallons per minute (gpm) will be delivered to the steam
generators. The licensee's accident analyses used 700 gpm to assess the
acceptability of the plant to respond to the accidents and concluded
that no safety limits were challenged if 500 gpm were supplied. As a
result of recent testing of the AFW system, the licensee determines
that the pumps can no longer deliver 700 gpm. The licensee determines
that the AFW pumps can deliver only 500 gpm at the required pressure
and temperature. The licensee performs the necessary safety analyses
and confirms that 500 gpm is sufficient to meet all necessary functions
and that no safety limits would be challenged as a result of the flow
reduction. The licensee decides to leave the pumps in the plant as is
rather than replace the pumps to restore the originally stated
capability. The licensee revises the FSAR to state that the AFW system
will deliver 500 gpm during postulated accidents or for transients
involving a loss-of-offsite power.
Under the new criterion (vii), the licensee would have to assess
the impact of the reduced flow rate on the design limits of the fission
product barriers. The licensee would have to identify the system
parameters that would vary as a result of the changes in AFW system
performance, identify the specific design limits that have the
potential to affect the fission product barrier performance, and
complete the analyses to determine whether the specific design limits
for the fission product barriers would be challenged. In this example,
it is assumed that the licensee did not change the method of evaluation
for the safety analyses. If the licensee had used a different
methodology from that used initially in establishing that the limits
were met, then, the licensee may have to submit the revised analyses
under criterion (viii) of the revised rule.
For this example, the licensee would have to complete the
evaluations required by Sec. 50.59 but would not have to submit a
license amendment request to lower the expected flow rate of the AFW
system, from that stated in the FSAR, to the lower as-found value, nor
would a licensee have to request an amendment to remove the old pumps
and replace the pumps with new pumps that provide the lower capacity
assumed in this example. The basis for this conclusion is that the
licensee analyses determined that the design limits of the fission
product barriers would not be challenged and, therefore, that the
fundamental basis for the staff's initial safety conclusion is
maintained.
Example 2: A facility FSAR states that some of the functions of the
component cooling water system are to provide cooling water flow to the
reactor coolant pump seals and to the shell side of the residual heat
removal system (RHR) heat exchangers. The FSAR states that the CCW
system provides 400 gallons per minute, 100 gpm for the seals and 300
gpm for the RHR heat exchanger. The licensee has recently obtained a
new reactor coolant pump seal which requires an additional 25 gpm of
cooling flow. The licensee plans to revise the flow distribution such
that 125 gpm is directed to the seals, and 275 gpm to the RHR heat
exchangers. The licensee performs analyses to determine that with the
reduced CCW flow to the RHR heat exchangers, the RHR system can still
perform its required functions with required limits, as for example,
removing sufficient decay heat to cool down within required time
frames, keeping post-accident temperatures within required limits, etc.
The licensee would satisfy criterion (vii) and be able to make this
change under Sec. 50.59.
Example 3: A licensee discovers an error in the primary system
pressure boundary piping fatigue calculation performed to demonstrate
compliance with the ASME Code requirements. A corrected calculation
shows that the fatigue criterion would be exceeded (for the postulated
FSAR events). A change to the licensing basis to accept revised fatigue
criteria would require review under criterion (vii) because the design
basis limit for one of the fission product barriers (reactor coolant
system piping) would be exceeded or altered. (This change would also
not satisfy criterion (i), ``minimal increase in frequency of
occurrence of an accident'' because of potential failure of piping due
to fatigue cracking, leading to loss of piping system integrity.)
[[Page 53598]]
New Criterion (viii)--Control of Evaluation Methods
In the proposed rule notice as part of the options presented on
margin of safety, the Commission had discussed the issue of controlling
methods (also, as noted, the proposed rule had explicitly stated that
changes to methods were changes to the facility, and as such, required
Sec. 50.59 evaluations). Specifically, the Commission sought comment on
whether the rule should include a statement that ``all analyses and
evaluations for assessing the impact of plant changes must be performed
using methodology and analytical techniques which are either reviewed
and approved by the NRC or which are shown to meet applicable review
guidance and standards for such analyses.''
Five commenters stated that methods should not be controlled by
Sec. 50.59 because the limits (e.g., acceptance limits) are
conservative. These commenters thought that licensees should be allowed
to use methods that are accepted by the NRC Standard Review Plan or
other processes, without the need for prior NRC approval. A few
commenters agreed that methods should either be reviewed and approved
by NRC (or meet applicable standards); produce results that are
consistent with the licensing basis methods; or that changes to methods
should be reviewed as separate changes under Sec. 50.59.
The Commission concludes that control of methods is essential in
assuring a consistent application of the change review process,
especially in light of the flexibility being provided by changes to the
other evaluation criteria, such as having criterion (vii) that uses
design basis limits being exceeded as the point at which NRC review is
required instead of the ``margin of safety'' criterion. Although the
Commission agreed that changes to methods should be reviewed as
separate changes, the other evaluation criteria do not provide a
standard that could be used to determine when changes to methods should
be reviewed by NRC. While the NEI proposal would have controlled the
methodologies through regulatory guidance, the Commission did not judge
that process to provide sufficient rigor to assure uniform
implementation of the requirement. A statement that the analysis should
meet applicable standards was considered, but was ultimately rejected
as being too vague. Therefore, the Commission has added criterion
(viii) to be specifically used for changes to methods of evaluation.
Final Rule Language
New criterion (viii) will require prior NRC review of any change in
a methodology or evaluation method that ``results in a departure from a
method of evaluation described in the FSAR (as updated) used in
establishing the design bases or in the safety analyses.''
Definitions and Guidance
For the purposes of this rule, a departure from a method of
evaluation described in the FSAR (as updated) used in establishing the
design bases or in the safety analyses means (1) changing any of the
elements of the method described in the FSAR (as updated) unless the
results of the analysis are conservative or essentially the same; or
(2) changing from a method described in the FSAR to another method
unless that method has been approved by NRC for the intended
application. Results from a changed method are conservative relative to
results from the previous method, if closer to the limits or values
that must be satisfied to meet the design bases.
Results are ``essentially the same'' if they are within the margin
of error needed for the type of analysis being performed, even if
tending in the non-conservative direction. Results are essentially the
same if the variation in results because of the change to the method is
explainable as routine analysis sensitivities, and the differences in
the results are not a factor in determining whether any limits or
criteria are satisfied. The determination can be made through
benchmarking (new vs. old method), or may be apparent from the nature
of the changes between the methods. When benchmarking a method to
determine how it compares to the previous one, the analyses that are
done must be for the same set of plant conditions, otherwise, the
results may not be comparable. Approval for intended application
includes assuring that the approved method was approved for the type of
analysis being conducted, generically approved for the type of facility
using it, and that all terms and conditions for use of the method are
satisfied.
The rule words were chosen to allow licensees only a small degree
of flexibility in methods where the results are tending in the non-
conservative direction, without burdening either the licensee or the
NRC with the need to review very small changes that are not important
with respect to the demonstrations of performance that the analyses are
providing. The intent is to limit the need for review to those changes
to methods that could impact upon the acceptability of performance were
the results to be at the limiting values.
By limiting the methods to those described in the FSAR, and to
those used for design bases and safety analyses, the Commission
concludes that the burden of requiring review is justified in view of
the relaxations in the other evaluation criteria. Unless the methods
are used in FSAR safety analyses, as demonstrating that the facility
performance continues to meet requirements, or to verify conformance
with the design bases, they would not meet the rule requirements for
approval. Thus, for example, if a licensee chose to perform sensitivity
studies, or to examine alternative approaches for a change being
contemplated, or included other analyses in the FSAR for reference
purposes, these methods would not be subject to the rule. It is at the
point in time that the revised method becomes the means used for
purposes of satisfying FSAR safety analysis or design bases
requirements that the approval (if the noted conditions are not met)
would become necessary.
The Commission has included a definition of ``departure'' in the
definitions section of the rule such that the intended meaning for
purposes of Sec. 50.59 is clearly understood.
Design bases as used in criterion (viii) is that information
meeting the definition contained in 10 CFR 50.2, and in particular,
those controlling values that are restraints derived from generally
accepted practices for achieving functional goals, or requirements
derived from analysis of the effects of a postulated accident for which
a SSC must meet its functional goals. Safety analyses are those
evaluations that demonstrate that acceptance criteria for the
facility's capability to withstand or to respond to postulated events
are met.
Thus, this criterion applies to those methods of evaluation used
for demonstrating that design basis limits for fission product barriers
are met, for other analyses such as radiological consequences that are
part of the safety analyses, and for analyses that demonstrate that
functional goals for SSC are met. These would include those analyses
that show that SSC will function under limiting conditions such as
natural phenomena, environmental conditions, dynamic effects, and so
forth. However, as noted in the rule language, only those methods that
are used in establishing the design bases or in the safety analyses
fall within the criterion. In addition, the Commission notes that
changes to time-limited aging
[[Page 53599]]
analyses and evaluations of aging management programs required by
Secs. 54.21(d) and 54.37(b), require evaluation with respect to
criterion (viii) to the extent that evaluation methods for these
analyses are described in the FSAR supplement.
To assure consistent implementation of criterion (viii), the
Commission believes that it is important to clearly distinguish between
methods of evaluation and input parameters to the methods. Methods of
evaluation means the calculational framework for evaluating behavior or
response of the reactor or any SSC. This includes the following (to the
extent that they are described or applicable for a particular method):
--Data correlations
--Means of data reduction
--Physical constants or coefficients
--Mathematical models
--Specific assumptions in a computer program
--Specified factors to account for uncertainty in measurements or data
--Statistical treatment of results
--Dose conversion factors and assumed source term(s)
Input parameters are defined as those values derived directly from
the physical characteristics of structures, systems or components, or
processes in the plant. These would include such things as: Flow rates,
temperatures, pressures, dimensions or measurements (e.g., volume,
weight, size), or system response times. Changes to input parameters
(that are described in the FSAR) are to be evaluated as facility
changes, and criterion (viii) would not be applicable. Additional
guidance will be provided in the implementation guidance to describe
the specific elements of the evaluation methods or methodology that
would require review and to clearly define specific types of input
parameters. The NRC intends to work closely with stakeholders to revise
the existing guidance related to implementation of Sec. 50.59 to
reflect these definitions.
The rule requirements for evaluation methods would allow for use of
generic topical reports as not being a ``departure,'' provided that the
topical report is applicable to the facility, and is used within the
terms and conditions specified in the approved topical report.
The Commission believes that with the guidance concerning
``evaluation methods'' and the definition of departure, licensees have
the capability to perform analyses as needed without being unduly
burdened by the need for NRC review, while still preserving those
inherent conservatisms in the methods that provide the confidence that
safety is maintained when the parameters are calculated to be at their
design basis limits and that SSC capability continues to meet design
basis requirements.
Examples
Example 1: The FSAR states that a damping value of 0.5 percent is
used in the seismic analysis of safety-related piping. The licensee
wishes to change this value to 2 percent to reanalyze the seismic loads
for the piping. Using a higher damping value to represent the response
of the piping to the acceleration from the postulated earthquake in the
analysis would result in lower calculated stresses because the
increased damping reduces the loads. Since this analysis was used in
establishing the seismic design bases for the piping, and since this is
a change to an element of the method that is not conservative and is
not essentially the same, the NRC concludes that this change would
require approval under criterion (viii). On the other hand, had NRC
approved an alternate method of seismic analysis that allowed 2 percent
damping provided certain other assumptions were made, and the licensee
used the complete set of assumptions to perform its analysis, then the
use of the 2 percent damping under these circumstances would not be a
departure, under the second part of the definition.
Example 2: The licensee wishes to use an inelastic analysis
procedure, not previously used in its seismic analyses as described in
the FSAR, to demonstrate that the structural acceptance criteria are
met for cable trays. NRC concludes that this would be a departure from
the methods of evaluation and that it would not be essentially the same
because the revised analysis would predict greater capacity than would
the previous analysis. Therefore, this change would require NRC
approval.
Example 3: The licensee wishes to change a non-LOCA FSAR Chapter 15
transient methodology. The methodology is being changed to a different
vendor's NRC approved method. The new vendor's method has been approved
generically for the particular reactor type (e.g., 2 loop PWR) and for
the particular transient being analyzed. The analysis is being
performed in accordance with all the applicable limitations and
restrictions. The licensee can make this change without prior NRC
approval because using a generically approved method for the purpose it
was approved, while meeting all the limitations and restrictions, is
not a ``departure.'' Subsequent plant changes can then be evaluated
using this new method and the other seven criteria in Sec. 50.59.
Example 4: The licensee wishes to change an analysis described in
the FSAR which states that adequate net positive suction head (NPSH) is
verified by analysis without crediting containment overpressure. The
new analysis will assume that five pounds of overpressure is credited
in calculation of available NPSH. The revised analysis predicts more
(five additional pounds of) available NPSH for the pumps, a result
further from the limit (the required NPSH) for an analysis that
establishes part of the design bases for the pumps as being capable of
performing their required function under the range of expected
conditions. This change can not be made without prior NRC approval
because a change in an element of a method described in the FSAR, used
to establish the design basis, that is not conservative, or essentially
the same, is a ``departure.''
Example 5: The licensee wishes to change an evaluation method
described or incorporated by reference in the FSAR Chapter 15 transient
analysis. In an attempt to remove some of the conservatism associated
with the analysis, the change the licensee is contemplating is removal
from the analysis of consideration of certain instrument uncertainties
for a few parameters, by assuming nominal values instead. By not
accounting for the greater range of the parameter (including the
uncertainties), the analysis predicts response further from the limit
to be satisfied. The treatment of uncertainties was an element of the
method described in the FSAR, and, therefore, this change can not be
made without prior NRC approval because a change in an element of a
method described in the FSAR, used in the safety analysis, that is not
essentially the same is a ``departure.''
On the other hand, if an instrument in the plant were replaced with
a different one, the assumed uncertainty in the analysis for that
instrument could be used in the analysis without prior NRC review,
using the other seven Sec. 50.59 criteria rather than criterion (viii),
because this is an input change rather than a model change. How the
uncertainties are treated in the analysis is part of the method. The
range of values of the uncertainties associated with particular
instruments is a characteristic of the facility and is thus an input
parameter.
K. Safety Evaluation
The Commission proposed to delete the word ``safety'' in referring
to the
[[Page 53600]]
required evaluation for determining whether the change, test, or
experiment requires a license amendment. A similar change was proposed
for Sec. 50.71(e), which presently refers to safety evaluations either
in support of license amendments or of conclusions that changes did not
involve USQs.
The Commission also proposed to change ``safety evaluation in
support of license amendments'' to ``safety analysis in support of
license amendments.'' The second part of the existing phrase would be
revised to refer to the ``evaluation that changes did not require a
license amendment in accordance with Sec. 50.59(c)(2) of this part.''
Conforming changes in Part 72 to revise the language to refer to
``evaluation'' were also proposed.
Commenters were generally supportive of these proposed changes. A
few noted that as with the term ``USQ,'' a simple process should be
adopted for revision of TS that use the term safety evaluation (this
issue is discussed under Section A(4)). Other clarifying wording
changes were included as a result of the comments, as for instance,
referring to ``approved'' license amendments rather than to
``requested'' license amendments to make clear that the updates, as
well as subsequent Sec. 50.59 evaluations, should be based upon what
has been approved (and implemented), not on what a licensee may have
proposed for approval, but that has not been approved.
The final rule includes these changes offered in the proposed rule
for Sec. 50.71(e); in addition, the term ``approved'' was used in
reference to license amendments. The final rule language for
Sec. 50.71(e) is presented in Section L, which also discusses other
aspects of the requirements for FSAR updating.
L. Reporting and Recordkeeping Requirements
Records
Requirements for records for evaluations performed under
Sec. 50.59, and for submittal of a summary report are being moved to
paragraph (d) as part of this rulemaking. In the final rule, the
Commission has simplified the rule text concerning records. Although
the text is simpler, there is no change in which records are being
required. That is, the Commission views the phrase ``made pursuant to
paragraph (c)'' as referring to those changes, tests, and experiments
that require evaluation against the criteria (for example, because they
involve the facility as described in the FSAR), but not to those other
activities or changes that are determined to not fall within these
required evaluations (as for instance, being screened out). As noted in
Section K above, the rule now refers to ``evaluations'' not to ``safety
evaluations.''
In addition, the Commission had proposed a change to the record
retention requirements in existing paragraph Sec. 50.59(b)(3)
(renumbered by this rulemaking to (d)(3)). The change would add to the
requirement that the records of changes to the facility be maintained
until the termination of the license, the following statement ``or
until the termination of a license issued pursuant to 10 CFR part 54,
whichever is later.'' Commenters were supportive of this proposal, and
the final rule section is unchanged from the proposed rule in this
regard.
Summary Report
Simplified text was also included in Sec. 50.59(d)(2), concerning
submittal of the summary report. The existing text required submittal
annually, or along with the FSAR update (which could be up to 24 months
between submittals), or at such other frequencies as specified in the
license. The Commission sees no need for such variability in submittal
dates, and believes that a 24 month interval is acceptable for
submittal of the summary report. Licensees may submit reports more
often if they wish. If a licensee has a shorter time specified in its
license, that licensee may request that the requirement be removed so
that the rule frequency would be applicable. The 24 month frequency is
also included in the part 72 sections, as requested by several
commenters.
Updates to the Final Safety Analysis Report
In the proposed rule, the Commission proposed to supplement the
reporting requirements in Sec. 50.71(e) on ``effects'' of changes to
require that in the FSAR update submittal (with the replacement pages),
the licensee shall include a description of each change affecting that
part of the SAR that provides sufficient information to document the
effect of the change upon the probability or consequences of accidents
or malfunctions, or reductions in margin associated with that part of
the SAR.
The reason for this proposal was that the Commission was concerned
about the potential cumulative effect of minimal increases. Since some
increases are allowed in probability and consequences, the Commission
thought that these rule changes would place greater importance on: (1)
Complete and accurate SAR updating; (2) the licensee's evaluation
process taking into account other changes made since last update; (3)
the licensee's screening process examining plant changes to determine
whether they are indeed changes requiring evaluation; and (4) reporting
requirements so that staff can assess the ongoing nature of cumulative
impact.
The issue discussed in the proposed rule was how the NRC could best
oversee the process such that several ``minimal'' changes do not result
in unacceptable results. In the proposed rule, the Commission proposed
requiring licensees to report effects of changes in the FSAR update
submittal in accordance with Sec. 50.71(e) in a different manner to
facilitate evaluation of cumulative effect.
A large number of commenters stated that this proposal was
burdensome and unnecessary in view of the minimal standards. Further,
commenters thought that this provision would require them to perform
additional evaluations of the cumulative effects, or to numerically
gauge the result of increases to probability that were judged on a
qualitative basis. Others stated that when analyses were performed,
such as for consequences or performance of SSC against limits, the
existing update requirements would specify that the effects of these
analyses be included in the update. The Commission agrees that the
burden associated with the proposed rule change is not warranted in
view of the specific criteria adopted and the existing update
requirements. Therefore, the final rule does not contain such language.
Other wording changes for Sec. 50.71(e) were discussed under
section K. Therefore, the following language is in the final rule for
this section:
(e) Each person licensed to operate a nuclear power reactor
pursuant to the provisions of Sec. 50.21 or Sec. 50.22 of this part
shall update periodically, as provided in paragraphs (e)(3) and (4)
of this section, the final safety analysis report (FSAR) originally
submitted as part of the application for the operating license, to
assure that the information included in the FSAR (as updated)
contains the latest information developed. This submittal shall
contain all the changes necessary to reflect information and
analyses submitted to the Commission by the licensee or prepared by
the licensee pursuant to Commission requirement since the last
submittal of the original FSAR, or as appropriate the last update to
the FSAR under this section. The submittal shall include the effects
\1\ of: all changes made in the facility or procedures as described
in the FSAR; all safety analyses and evaluations performed by the
licensee either in support of approved license amendments, or in
[[Page 53601]]
support of conclusions that changes did not require a license
amendment in accordance with Sec. 50.59(c)(2) of this part; and all
analyses of new safety issues performed by or on behalf of the
licensee at Commission request. The updated information shall be
appropriately located within the update to the FSAR.
---------------------------------------------------------------------------
\1\ Effects of changes includes appropriate revisions of
descriptions in the FSAR such that the FSAR (as updated) is complete
and accurate.
---------------------------------------------------------------------------
M. No Significant Hazards Consideration Determinations
Under Sec. 189.a(2)(A), the Commission may issue and make
immediately effective an amendment to an operating license if the
Commission has made a determination that the amendment involves a ``no
significant hazards consideration'' (NSHC), despite the pendancy of a
request for a hearing or the completion of such a hearing. The
Commission's criteria for determining whether an amendment involves a
NSHC, as set forth in Sec. 50.92(c), are similar to the current USQ
criteria in Sec. 50.59:
(c) The Commission may make a final determination * * * that a
proposed amendment to an operating license * * * involves no
significant hazards consideration, if operation of the facility in
accordance with the proposed amendment would not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
(2) Create the possibility of a new or different kind of
accident from any accident previously considered; or
(3) Involve a significant reduction in a margin of safety.
The Commission has evaluated whether the NSHC criteria in
Sec. 50.92(c) must be modified if the existing criteria in Sec. 50.59
are altered, deleted or supplanted. The AEA does not define NSHC, nor
does any provision of the AEA conceptually link the NSHC concept to any
particular standard or concept. A review of the legislative history of
the ``Sholly amendment'' which modified Section 189.a did not disclose
any reference to Sec. 50.59 or a discussion which links the NSHC
concept and the Sec. 50.59 criteria. H.R. Conf. Rep. No. 97-884, 97th
Cong., 2d Sess. (1982), Sen. Rep. No. 97-113, 97th Cong., 2d Sess.
(1981), H. Rep. No. 97-22, Part 2, 97th Cong., 2d. Sess. (1981).
The Commission has also evaluated whether changes to the NSHC
criteria to conform more closely to the revised Sec. 50.59 would
facilitate implementation of the revisions to Sec. 50.59, even if
changes to the NSHC criteria are not required by the AEA. There are
three areas where the current NSHC criteria diverge from the revised
Sec. 50.59 criteria: (i) The current NSHC criteria do not include the
``malfunction of components'' criterion in the revised Sec. 50.59; (ii)
the NSHC criteria retains a ``significant reduction in margin of
safety'' criterion, which is no longer part of the revised Sec. 50.59;
and (iii) the NSHC criteria do not include the revised Sec. 50.59
criteria (vii) and (viii) concerning changes to fission barrier design
basis limits, and changes to and departures from evaluation methods.
Although there may be some conceptual tidiness in utilizing the same
evaluation factors for changes under Sec. 50.59 and NSHC determinations
under Sec. 50.92, nothing in the AEA or the legislative history
requires that the criteria be identical. Furthermore, the Commission
notes that Sec. 50.59 and NSHC address issues which are fundamentally
different in purpose. Section 50.59 is focused upon the NRC's
regulatory needs with respect to its review and approval of licensee-
initiated changes, tests and experiments. By contrast, the NSHC
determination is directed at determining what license amendments will
require the Congressionally-mandated 30-day notice in the Federal
Register and completion of any hearing granted pursuant to the
Congressionally-mandated opportunity for hearing in Section 189.a. In
the Commission's view, the existing NSHC criteria have been
demonstrated through years of application to provide a workable
standard for determining the potential safety significance of a
proposed amendment for the purposes of determining whether issuance of
a license amendment must await notice in the Federal Register and
completion of any requested hearing. On balance, the Commission
believes that no changes to the existing NSHC criteria are necessary in
order to implement the revised change criteria in the revised
Sec. 50.59.
Recognizing the difference between the two sections, the Commission
notes that if a change does not require a license amendment by virtue
of the new Sec. 50.59(c)(2)((vii) and (viii) criteria, then the change
cannot be regarded as involving a ``significant reduction in a margin
of safety'' under Sec. 50.92(c)(3). If a change does require a license
amendment by virtue of either Sec. 50.59(c)(2)((vii) or (viii), the NRC
would be required to determine whether the design basis limit for a
fission product barrier being exceeded or altered, or the departure
from the method of evaluation used in establishing the design bases or
safety analyses, constitutes a significant reduction in a margin of
safety. With respect to new Sec. 50.59(c)(2)(ii) and (iv), the
Commission regards these criteria as a substitute for and refinement of
the ``malfunction of equipment'' aspect of the existing
Sec. 50.59(a)(2)(ii) criterion, for which there is no parallel
provision in Sec. 50.92(c)(2). Therefore, the NSHC evaluation for
license amendments necessitated by the new Sec. 50.59(c)(2)(ii) and
(iv) criteria will be largely the same as the current process for
evaluating license amendments necessitated by the ``malfunction of
equipment'' provision in the existing Sec. 50.59(a)(2)(ii).
N. Part 52 Changes
In the proposed rule, the Commission had proposed to revise
appendices A and B to part 52 to conform with the proposed changes to
Sec. 50.59 concerning the evaluation criteria for when prior NRC
approval is required for changes to certain Tier 2 information in
plant-specific design control documents.
Two commenters believe that the changes to part 52 needed to be
expanded to either include certain provisions or definitions, or to
refer to Sec. 50.59 to incorporate them. The Commission has decided to
defer consideration of the changes in the proposed rule for part 52.
The Commission anticipates other rule changes for Part 52 arising from
an ongoing lessons-learned review. Further, the proposed design
certification rule for the AP600 design being issued for public comment
will emulate the two design certification rules in appendices A and B.
Accordingly, the Commission will consider these proposed changes in an
integrated manner later.
O.1. Part 72 Changes
This section first discusses the changes offered in the proposed
rule on part 72, then discusses the comments received and the
resolution and final rule language. The comments and rule language are
discussed under subheadings relating to the specific requirements, such
as for evaluation of changes, FSAR updating, and other conforming
changes. A discussion of petition for rulemaking (PRM 72-3), submitted
by Ms. Fawn Shillinglaw, and how it relates to the changes to part 72
is contained in section O.2.
Changes Presented in the Proposed Rule
For part 72, in the proposed rule, the Commission proposed changes
to Sec. 72.48 conforming with those made to Sec. 50.59 and proposed to
expand the scope of Sec. 72.48 so that holders of a Certificate of
Compliance (CoC) approving a spent fuel storage cask design also would
be subject to the requirements of this section. The Commission
envisioned that a general licensee who wants to adopt a change to the
design of a spent fuel storage cask
[[Page 53602]]
it possesses--which change was previously made to the generic design by
the certificate holder under the provisions of Sec. 72.48--would be
required to perform a separate evaluation under the provisions of
Sec. 72.48 to determine the suitability of the change for itself.
Certificate holders would be required to keep records of such
changes as are allowed under Sec. 72.48. New reporting requirements for
certificate holders would be added in Secs. 72.244 and 72.248, similar
to existing requirements imposed on licensees in Secs. 72.56 and 72.70,
respectively.
In addition to these changes to Sec. 72.48, the Commission proposed
making changes in other sections of part 72 as follows:
In Sec. 72.3 the definition for independent spent fuel storage
installation (ISFSI) would be revised to remove the tests for
evaluation of the acceptability of sharing common utilities and
services between the ISFSI and other facilities; and the existing
requirement in Sec. 72.24(a) revised to reference shared common
utilities and services in the applicant's assessment of potential
interactions between the ISFSI and another facility. Proposed changes
to Sec. 72.56 would be conforming changes to those made to Sec. 50.90.
Changes to Secs. 72.9 and 72.86 are conforming changes due to the
proposed addition of new Secs. 72.244, 72.246, and 72.248. The change
to Sec. 72.212(b)(4) would be a conforming change necessitated directly
by the change to Sec. 50.59, as this section in part 72 refers to
Sec. 50.59 with respect to evaluations for the reactor facility at
which site the ISFSI is located.
In the proposed rule, Sec. 72.70 was proposed for revision to
conform to Sec. 50.71(e). Requirements would be added on standards for
submitting revised Final Safety Analysis Report (FSAR) pages.
Requirements would also be established for reporting changes to
procedures. New reporting requirements for certificate holders would be
added in Secs. 72.244 and 72.248, similar to existing requirements
imposed on licensees in Secs. 72.56 and 72.70, respectively.
New Secs. 72.244 and 72.246 would be added to subpart L, to provide
regulations on applying for, and approving, amendments to CoCs. A new
Sec. 72.248 would also be added to provide regulations for the
certificate holder on submitting and updating the FSAR, which would
document the changes it made to procedures or SSC under the provisions
of Sec. 72.48. The new Sec. 72.248(c) would also require, in part, that
updates to the FSAR use revision numbers, change bars, and a list of
current pages.
Resolution of Comments Received: Of the 60 comment letters, 10
raised issues related to part 72. The following is a summary of those
comments and the Commission's responses:
1. Overall Changes to Part 72
All ten of the commenters were generally supportive of the changes
to part 72 and the expansion of scope of Sec. 72.48 to include part 72
certificate holders. Nevertheless, the commenters indicated that the
regulations in part 72 were more restrictive than similar regulations
in part 50. The commenters pointed to certain part 72 requirements
(i.e., release limits, Sec. 72.48 evaluation criteria on occupational
exposure and environmental impact, and update frequency and content for
Sec. 72.48 evaluations and FSAR changes) that do not exist in part 50
or that are more stringent than similar part 50 regulations. Overall,
the commenters believe the risk from spent fuel storage casks and
facilities is much less than from reactors. The commenters generally
recommended that Secs. 72.48 and 72.70 should be more consistent with
Secs. 50.59 and 50.71(e).
The Commission agrees that where possible the language used in the
respective sections in parts 50 and 72 should be similar. Therefore,
except where unique requirements exist (e.g., because Sec. 72.48
involves both licensees and certificate holders, as well as facilities
and spent fuel storage cask designs, and Sec. 50.59 only involves
licensees and facilities), the final rule has used consistent language
in both parts 50 and 72. The NRC also notes that the comments on
revising the release limits for part 72 are clearly beyond the scope of
the proposed rule and no further response is made.
2. Sec. 72.48 (Changes, Tests, and Experiments)
The ten commenters suggested that the tests in Sec. 72.48 should be
same as are used in Sec. 50.59; in particular, five commenters said
that the significant increase in occupational exposure and significant
unreviewed environmental impact tests were unnecessary and therefore
should be removed. One commenter indicated the unreviewed environmental
impact test should be retained, but only for specific licensees.
The Commission agrees that the occupational exposure test is
unnecessary because licensees are currently required by Sec. 20.1101(b)
to take actions to maintain occupational exposure as low as is
reasonably achievable. The Commission also agrees that the significant
unreviewed environmental impact test is unnecessary. As stated in the
Finding of No Significant Environmental Impact for this rule, the
changes being made in Sec. 72.48 will allow only minimal increases in
probability or consequences of accidents (still satisfying regulatory
limits) without prior NRC review. Further, changes which result in more
than minimal increases in radiological consequences will continue to
require prior NRC approval, including NRC consideration of potential
impact on the environment. Therefore, consistent with Sec. 50.59, there
is no need for this criterion to be included with respect to
consideration of a change under Sec. 72.48 and it has been deleted from
the final rule.
One commenter suggested that the scope of Sec. 72.48 should be
limited to only ``important to safety'' structures, systems, and
components (SSCs), not all SSCs described in the FSAR. One commenter
suggested the Sec. 50.59 term ``equipment important to safety'' should
be used rather than ``SSC important to safety.'' One commenter
suggested the term ``evaluations'' should be removed from the
definition of the facility in proposed paragraph Sec. 72.48(a)(3)(iii).
The Commission disagrees with these comments. The term SSCs
provides a better description than equipment and is consistent with
other regulations in both parts 50 and 72 (as noted earlier, the
Commission is revising Sec. 50.59 to refer to SSC instead of to
equipment). The scope of these Sec. 72.48 evaluations should include
all SSCs described in the FSAR, not just those that are important to
safety. The current regulations in Sec. 72.48 require a scope that
includes all structures, systems, and components described in the FSAR
not just those ``important to safety.'' The Commission continues to
believe that this approach is necessary to insure that changes to SSCs
considered ``not important to safety'' do not have a negative impact on
SSCs considered important to safety due to interactions and interfaces,
and do not cause any adverse impact on public health and safety. The
term ``evaluations and methods of evaluation'' is necessary for the
reasons previously discussed for Sec. 50.59 changes, and is retained in
final Sec. 72.48(a)(2)(iii).
One commenter stated that the term FSAR should not be used because
Part 72 is a one step licensing process and using the term implies a
second review step is required by staff. The same commenter added that
the discussion of the FSAR (in the rule) could also imply that the
Sec. 72.48 process is not required to address changes until the
licensee has an FSAR. (The commenter thought the
[[Page 53603]]
proposed rule language suggested that Sec. 72.48 would not apply until
after the FSAR was submitted). Two commenters identified concerns with
the current requirement for a specific licensee to update its SAR every
6 months and its role as a hold point (requiring staff review) and the
requirement to update the SAR 90 days prior to loading fuel. Two other
commenters suggested that the order of Secs. 72.48 (a)(2) and (a)(3)
should be reversed and that the term ``required to be included'' should
be deleted from proposed paragraph (a)(3)(iii).
The Commission has revised Secs. 72.48, 72.70 and 72.248 in
response to these comments. These changes have clarified the use of the
term FSAR to avoid the interpretation that multiple staff reviews of
this document will be required. The FSAR being submitted 90 days after
license issuance precludes both a hold point and an additional staff
review. Further the Commission agrees that providing a periodic FSAR
update every 6 months and a final one 90 days prior to fuel load was an
unnecessary burden, which does not exist in Sec. 50.71(e), and these
requirements have been eliminated. The Commission agrees that language
was needed to indicate that the facility or design can be changed using
the new process in Sec. 72.48 after a license is issued and prior to
issuing the FSAR and that has been reflected in the final rule.
Sections 72.48 a(2) and a(3) have been reversed in order and the phrase
``required to be included'' has been deleted for clarity and for
consistency with Sec. 50.59.
Several commenters suggested that a different approach be taken on
the margin of safety; that the terms ``minimal'', ``more than minimal''
or ``significant'' required further clarification and should be
consistent with Sec. 50.59; suggested reports of Sec. 72.48 changes,
tests, and experiments be submitted every 24 months: and that an
implementation schedule be provided for the final rule.
The NRC agrees that Secs. 50.59 and 72.48 should be as consistent
as possible. Therefore Sec. 72.48 has used the language adopted in
response to comments on Sec. 50.59 (see comments on Sec. 50.59 on the
use of minimal and margin of safety terminology). The NRC agrees that a
24 month reporting frequency is appropriate. The NRC has also provided
direction in implementing the final rules.
One commenter suggested that licensees and certificate holders
should inform each other of changes implemented under Sec. 72.48 that
affect a particular cask design, through the summary reports rather
than through the FSAR update, as was stated in the proposed rule. One
commenter also suggested that guidance on the timeliness of the review
to be performed upon receipt of such changes be provided.
The NRC agrees with both comments and has added Sec. 72.48
(d)(6)(i)--(iii) on providing copies of Sec. 72.48 evaluations to other
interested persons who use the particular cask design within 60-days of
implementing the change (the proposed language in Secs. 72.216 and
72.248 on this point has been deleted). Guidance on the timeliness of
the reviews will be provided by the NRC along with other guidance
information for Secs. 50.59 and 72.48.
General licensees who have evaluated a proposed change under
Sec. 72.48 and concluded that a CoC amendment is required, must request
that the certificate holder submit the application for amendment under
Sec. 72.244. Clarifying language was included in Sec. 72.48 on this
point.
As a result of other changes made earlier in Sec. 72.48, the
section on recordkeeping was reformatted to include subsection
numbering. As part of this revision, the text in paragraphs (d)(3)(i)
and (d)(3)(ii) was clarified to acknowledge those situations where the
facility is no longer being used, but for which the license has not yet
been terminated.
3. Secs. 72.70, 72.216, and 72.248 (FSAR Updating)
Several commenters suggested that the language in Secs. 72.70,
72.216, and 72.248 on updating the FSAR conform to the language in
Sec. 50.71(e). Specific changes requested included requiring a 24-month
reporting period, adding a 6-month cutoff for reporting changes,
clarifying requirements for the initial submittal of the FSAR, and how
no changes to the FSAR are to be reported by stating that there are no
changes. One commenter felt that requiring a general licensee to
maintain its own FSAR (i.e., potentially separate and distinct from the
certificate holder) was unnecessary and would cause confusion. One
commenter felt that the process for revising the FSAR for a general
licensee was confusing.
The NRC agrees that providing a 24-month FSAR update and adding the
6-month cutoff for bringing the FSAR up to date for changes made are
consistent with Sec. 50.71(e), are appropriate, and are a reduction in
unnecessary regulatory burden. Lastly, the NRC believes that providing
a written confirmation when no changes to the FSAR have been made
provides a clear and timely record of the status of the FSAR to both
the staff and the public and agrees with this comment. The NRC also
agrees that having a general licensee keep a separate FSAR from that of
a certificate holder is redundant and believes that requiring a
separate FSAR is not necessary for the staff to maintain its regulatory
oversight over general licensees. Accordingly, proposed paragraph (d)
to Sec. 72.216 has been withdrawn. In withdrawing this section, the NRC
wishes to clarify that the certificate holder is not expected to
incorporate Sec. 72.48 changes made by general licensees into its FSAR;
rather the certificate holder is responsible for updating the FSAR for
any changes it has made under the provisions of Sec. 72.48.
Furthermore, the NRC expects certificate holders to maintain the FSAR
current for any version of its cask design, which is being used to
store spent fuel.
Two commenters suggested that the proposed rule language in
Secs. 72.70, and 72.248 that the FSAR update include a ``description
and analysis of changes in procedures or in [SSC]'', was more
burdensome than the existing language in Sec. 50.71(e) that the update
is to ``contain all the changes necessary to reflect information and
analyses submitted. * * *''
The NRC agrees that this language could be read as requiring a
separate discussion of the effects of changes beyond the SAR updates
themselves, which was not the intent of the proposed rule. The language
in Secs. 72.70 and 72.248 has been revised to be as consistent with
Sec. 50.71(e) as possible and, in particular, refers to ``include the
effects of'' changes, analyses and evaluations, but not stating that
the update needs to describe each change.
In the current rule, a licensee must submit to the NRC its FSAR 90
days prior to the receipt of fuel or high level waste and this action
serves as a formal notification to the regulator that fuel (or high
level waste) is planned to be loaded. A number of comments viewed this
requirement as overly restrictive because many changes related to cask
loading included in a FSAR will not be identified or analyzed until
preoperational testing is performed and, thus, the 90 day FSAR update
requirement could be interpreted as another holdpoint before loading.
The NRC agrees that the requirement that a FSAR be submitted at least
90 days prior to fuel load was not intended to serve as a holdpoint and
in the final rule, this has been changed to require a specific licensee
to submit a FSAR 90 days after receiving a license. To maintain the
notification aspect of the current regulation, a new requirement
[[Page 53604]]
was added to Sec. 72.80(g) to notify the NRC of the licensee's
readiness to begin operation at least 90 days prior to the first
loading of spent fuel or high-level radioactive waste. Specific
licensees will update their FSAR every two years. Because the FSAR will
be submitted before construction and preoperational testing of the
ISFSI would be completed, a requirement was retained in Sec. 72.70 to
provide a final analysis and evaluation of the design and performance
of SSCs taking into account information since the submittal of the
application (i.e., information developed during final design,
construction, and preoperational testing), in the next periodic update
to the FSAR. This information is not required by the final
Sec. 50.71(e); however, it is necessary to require these actions to
complete the description of the ISFSI, because of the single-step
licensing process in part 72.
New reporting requirements for certificate holders will be added in
Secs. 72.244 and 72.248, similar to existing requirements imposed on
licensees in Secs. 72.56 and 72.70, respectively.
4. Secs. 72.3, 72.9, 72.24, 72.56, 72.86, and 72.212 (Miscellaneous
Sections of Part 72)
No specific comments were received on Secs. 72.3, 72.9, 72.24 and
72.86, and the final rule language is unchanged from the proposed rule
language for these sections.
Two commenters believed that Sec. 72.56 was not clear on whether
this regulation applied to specific licensees, general licensees, or
both.
The NRC agrees and has revised this section to indicate it applies
to specific licensees only.
One commenter suggested that Sec. 72.56 be revised to allow
licensees to apply for emergency or exigency processing of license
amendment requests, similar to that allowed under certain conditions
for Part 50 licensees under Sec. 50.91(a)(5) and (6).
The NRC disagrees. The NRC currently has the authority under
Sec. 72.46(b)(2) to immediately issue an amendment to a part 72 license
upon a finding that no genuine issue exists that could adversely affect
public health and safety. Consequently, the NRC's authority to
immediately issue an amendment to a part 72 license obviates the need
for a separate emergency or exigency amendment process.
One commenter recommended that any changes to the written
evaluations performed by a general licensee in accordance with
Sec. 72.212(b), in determining whether a spent fuel storage cask design
can be used at a particular part 50 reactor site, should be
accomplished using the requirements of Sec. 72.48.
The NRC agrees and has revised Sec. 72.212(b)(2)(ii) to require the
general licensee evaluate any changes to the written evaluations
required by Sec. 72.212 using the requirements of Sec. 72.48(c).
O.2 Petition for Rulemaking (PRM-72-3)
The NRC received a petition for rulemaking submitted by Ms. Fawn
Shillinglaw in the form of two letters addressed to Chairman Jackson
dated December 9 and December 29, 1995. The Office of General Counsel
determined on March 5, 1996, that the issues presented in these letters
would be treated as a petition for rulemaking. The petition requested
that the NRC amend its regulations in 10 CFR part 72, ``Licensing
Requirements for the Independent Storage of Spent Fuel and High-Level
Radioactive Waste.'' The petition was docketed as PRM-72-3 on March 14,
1996. Ms. Shillinglaw supplemented her petition with additional
information in a letter dated April 15, 1996. The NRC published in the
Federal Register on May 14, 1996, a notice of receipt of this petition
and stated the issues contained in the petition (61 FR 24249).
Specifically, the petitioner requested that the NRC amend those
regulations which govern independent storage of spent nuclear fuel in
dry storage casks to require that: (1) The safety analysis report (SAR)
for a dry storage cask design fully conforms with the associated NRC
safety evaluation report (SER) and Certificate of Compliance (CoC)
before NRC certification (i.e., approval) of the dry storage cask
design; (2) the revision date and number of an SAR be specified
whenever that report is referenced in documents; (3) the NRC clarify
the process for modification of an SAR after a cask has been certified;
and (4) the NRC make available to the public, the licensees' unloading
procedures. In her supplemental letter, the petitioner recommended that
to eliminate confusion, the term ``CSAR'' (i.e., cask safety analysis
report) be used when referring to the SAR for any dry storage cask
design which has been approved by the NRC and issued a CoC.
The Commission received ten comment letters on PRM-72-3. The
commenters included five members of the public, three public interest
groups, and the Nuclear Energy Institute (NEI). Copies of the public
comments on PRM-72-3 are available for review in the NRC Public
Document Room, 2120 L Street, NW (Lower Level), Washington, DC 20003-
1527. No comments were received objecting to the petition. Eight of the
commenters were supportive of all, or some, of the four issues raised
in PRM-72-3. One commenter (NEI), neither supported nor opposed the
petition and recommended that any rulemaking action based on the
petition be delayed until the NRC addressed issues in 10 CFR part 50
relating to the use of the ``FSAR'' as a licensing basis document and
the application of Sec. 50.59 in 10 CFR part 50. One commenter objected
to NEI's recommendation to delay rulemaking on PRM-72-3.
The Commission has determined that PRM-72-3 issues (1), (2), and
(3) should be granted, in part; and issue (4) should be denied. This
notice constitutes the Commission's final action on this petition. The
basis for the Commission's actions on each issue and responses to
public comments received on the petition are described below.
Issue (1): Part 72 should be amended to require that the safety
analysis report (SAR) for a spent fuel dry storage cask design fully
conforms with the associated NRC safety evaluation report (SER) and
certificate of compliance (CoC) before NRC certification (i.e.,
approval) of the cask design.
Five comment letters were received supporting Issue (1) of PRM-72-
3.
Resolution of Issue (1): In this final rule the Commission has
granted, in part, the petitioner's request on this issue. This rule
adds new Sec. 72.248 to part 72 and this section addresses this issue
by requiring a certificate holder to submit a final safety analysis
report (FSAR) after issuance of the CoC. This rule also describes the
process for periodic updates of the FSAR. Section 72.248, paragraphs
(a)(1) and (a)(2) state, in part:
Each certificate holder shall submit an original FSAR to the
Commission * * * within 90 days after the spent fuel storage cask
design has been approved pursuant to Sec. 72.238. This original FSAR
shall be based on the safety analysis report submitted with the
application and reflect any changes and applicant commitments
developed during the cask design review process. The original FSAR
shall be updated to reflect any changes to requirements contained in
the issued Certificate of Compliance (CoC). * * *
The Commission agrees with the petitioner that the FSAR should be
fully conformed (i.e., consistent) with the operating limits contained
in the CoC, because the FSAR contains the design information the staff
used to make its safety finding and to approve the dry storage cask
design for use. The Commission disagrees with the petitioner's request
that the FSAR be conformed to the NRC SER for the dry storage cask
design, and that the FSAR be submitted to the NRC before approval
[[Page 53605]]
of the cask design (i.e., issuance of the CoC). The NRC SER contains
staff conclusions on the adequacy of the cask design, not applicant
commitments to the NRC on the cask design. Therefore, the Commission
believes it is not necessary to conform the FSAR to the issued NRC SER
before the CoC can be issued. The NRC SER is available in the NRC
Public Document Room for public review.
The Commission disagrees with the petitioner's request that
issuance of the CoC (i.e., placement of the CoC in the list at
Sec. 72.214 which enables a general licensee to use the cask design) be
delayed until after the certificate holder has submitted an FSAR to the
NRC (i.e., updated the topical safety analysis report, submitted with
its application for approval of a dry storage cask design, to ensure
that the SAR is consistent (fully conforms) with the approved CoC).
This final rule codifies as a regulation the NRC's current approach
which, administratively, requires a certificate holder to update its
SAR after issuance of the CoC to ensure it is consistent with the
issued CoC. For administrative purposes, the Commission prefers that
the original FSAR be submitted to the NRC, within 90 days after the CoC
is issued, so that the certificate holder can include [conform] in the
FSAR any conditions from the issued CoC. The FSAR does not need to be
conformed to the CoC, before the CoC is issued, because this action
does not provide any new information the NRC would need to make a
determination that the cask design meets the requirements of part 72,
subpart L, and is acceptable for use.
The Commission also disagrees with the petitioner's supplemental
information to use the term ``cask safety analysis report (CSAR)'' when
referring to the SAR submitted after the NRC approves a cask design.
Instead, the Commission is using the term ``final safety analysis
report (FSAR)'' to identify the SAR submitted after the NRC approves a
cask design. The use of the term ``FSAR'' is the accepted practice by
industry and will not cause confusion. Further, this approach will
ensure consistency between parts 50 and 72, because the term ``FSAR''
is used by Secs. 50.59, 50.71(e), 72.48, and 72.70 in this final rule.
Issue (2): Part 72 should be amended to require that the revision
date and number of an SAR be specified whenever that report is
referenced in documents.
Five comment letters were received supporting Issue (2) of PRM-72-
3.
Resolution of Issue (2): In this final rule the Commission has
granted, in part, the petitioner's request on this issue. This rule
adds new Sec. 72.248 to part 72 which requires that revision numbers,
change bars, and a list of current pages be included in any revisions
to the FSAR. Section 72.248, subparagraphs (c)(2) and (c)(3) state:
The update [of the FSAR] shall include a list that identifies
the current pages of the FSAR following page replacement. Each
replacement page shall include both a change indicator for the area
changed, e.g., a bold line vertically drawn in the margin adjacent
to the portion actually changed, and a page change identification
(date of change or change number or both).
These features will clearly identify what has been changed, as well
as the date of the change, in any revision to a FSAR. While Sec. 72.248
will provide a process for requiring revisions to the FSAR be clearly
indicated, the Commission has denied the portion of the petitioner's
request to amend part 72 to require a FSAR revision number and date be
specified when the FSAR is referenced in other documents (e.g., an
application for a part 72 license or CoC). Instead, the NRC will revise
guidance documents for part 72 activities (e.g., regulatory guides and
standard review plans) to require specification of the FSAR revision
date and number whenever a FSAR is referenced in another document. The
Commission believes addressing this portion of the petitioner's request
in guidance documents rather than in a regulation is more appropriate
and meets the intent of the request.
Issue (3): The NRC must clarify the process for modification of a
safety analysis report after a cask [design] has been certified (i.e.,
approved by the NRC).
Five comment letters were received supporting Issue (3) of PRM-72-3
including a comment from the petitioner clarifying that she believed
that ``any changes to the SAR (FSAR) should be done by the amendment
process of rulemaking.'' Four commenters also recommended that any
changes made to the SAR (including a generic SAR), the cask design, or
the CoC should require rulemaking and public comment or a public
hearing. One commenter also suggested that the regulations be amended
to include more detail on who can make changes to dry storage cask
designs and whether vendors (i.e., certificate holders) can make these
changes.
Resolution of Issue (3): The Commission is revising Sec. 72.48 to
allow a certificate holder to make certain types of changes to a cask
design, or procedures, or to conduct tests and experiments, not
described in the FSAR (as updated) without requiring prior NRC approval
if the criteria in Sec. 72.48(c) are met. If these criteria are not
met, a certificate holder must obtain a CoC amendment pursuant to
Sec. 72.244. Following such changes (either resulting from the
Sec. 72.48 process or the CoC amendment process), the certificate
holder must update the FSAR as required by Sec. 72.248. Section 72.248,
paragraphs (b), (b)(2), and (b)(3) state, in part:
The (FSAR) update shall include the effects of: All safety
analyses and evaluations performed by the certificate holder either
in support of approved CoC amendments, or in support of conclusions
that the changes did not require a CoC amendment in accordance with
Sec. 72.48. All analysis of new safety issues performed by or on
behalf of the certificate holder at Commission request. The
information shall be appropriately located with the updated FSAR.
The Commission is seeking to reduce any unnecessary regulatory
burden placed on its licensees and certificate holders without
compromising safety. The dry storage cask design review process and the
analysis acceptance criteria are defined in the NRC's standard review
plans. This final rule allows licensees and certificate holders to make
changes to the cask design, without obtaining prior NRC approval, for
changes which do not significantly impact the ability of the cask to
perform its intended functions. The impact of these changes are then
incorporated into an updated FSAR, which is submitted to the NRC.
Requiring that all changes to a cask design or changes to a FSAR be
reviewed and approved by the NRC through the rulemaking amendment
process, including either a public comment period or a public hearing,
defeats these efforts with no discernable increase in safety. Further,
while rulemaking is currently utilized to amend a CoC, the Commission
is presently re-examining the appropriateness of this procedure.
Therefore, the Commission has granted petitioner's request to clarify
the process for modification of an FSAR after the NRC has approved the
cask design and issued the CoC, but has rejected the request to require
all changes to a cask design, or the FSAR, be made via a rulemaking
amendment process.
Issue (4): The NRC should make cask unloading procedures publicly
available.
Five comment letters were received supporting Issue (4) of PRM-72-
3. One commenter also requested that the NRC review, approve, and have
tested unloading procedures prior to their being implemented. One
commenter suggested suspending all cask loading
[[Page 53606]]
activities until the NRC reviews procedures [for loading and unloading]
and appropriate tests are completed.
Resolution of Issue (4): The NRC does not approve or test a
licensee's loading or unloading procedures, rather the licensee is
responsible for development, verification, and validation of the
loading and unloading procedures. The NRC inspects the licensee's
procedures (i.e., reviews the procedures and observes the licensee
implementing them) to determine whether the procedures will provide
reasonable assurance that public health and safety will be adequately
protected.
The Commission does not agree that cask unloading procedures should
be required to be public documents. First, in order to make these
procedures publicly available, either the NRC must possess the
procedures, or the licensee must place the procedures in the public
domain. The Commission's position is that only those documents
necessary to demonstrate that a dry storage cask is designed to meet
the requirements of part 72, subpart L, need to be submitted to the NRC
on the docket (i.e., to allow the NRC to determine that the cask design
is acceptable for use). Cask loading and unloading procedures are
implementing documents required by the CoC which are developed and
implemented by the licensee.
Although the NRC does not possess the procedures, they are subject
to inspection by NRC staff. However, even during inspection activities,
NRC generally does not take possession of the procedures. Therefore,
the unloading procedures remain the property of the licensees and are
not available to the public. The NRC's inspection program for part 72
licensees requires the inspection of loading and unloading activities,
including a review of applicable procedures, before a licensee begins
cask loading. NRC inspection personnel perform these activities at the
licensee's site and observe the licensee's preoperational testing and
dry run activities to assess the adequacy of these procedures and the
readiness of the licensee to begin loading spent fuel. The results of
these inspections are documented in reports which are placed in the NRC
Public Document Room and are available for public review.
Furthermore, requiring part 72 licensees to submit their
implementing procedures to the NRC (i.e., operating procedures such as
loading and unloading procedures, maintenance procedures, surveillance
procedures, radiation protection procedures, security procedures,
emergency procedures, and administrative procedures), as well as any
revisions to these procedures, would impose a huge paperwork burden on
both the licensee and on NRC staff without a corresponding safety
benefit. Therefore, Issue (4) is denied.
Additional Public Comments on the Petition
In addition to the specific comments that were received on the
petition that are discussed above, a number of comments were received
on related and unrelated subjects.
Comment: Five comments were received on the VSC-24 cask design
being used at the Palisades and Point Beach plants and incidents
related to the VSC-24 cask design.
Response: The Commission considers these comments beyond the scope
of this petition and this rulemaking.
Comment: Two comments were received suggesting that when a change
to an approved dry storage cask design is requested, that the existing
CoC be suspended until the changes are approved by the NRC.
Response: The Commission considers these comments would impose an
unreasonable burden on part 72 licensees. Suspending a CoC solely on
the basis of receiving a change and not on the basis of a compelling
safety need, would imply that any casks manufactured under the CoC,
which are in use by part 72 licensees, should be taken out of service
(i.e., unloaded) upon receipt of any request to revise the cask design.
Requiring that a cask be unloaded in these circumstances would impose
an unreviewed backfit on the part 72 licensees using that cask design
and would also result in unnecessary occupational exposure to licensee
workers.
Comment: One comment was received recommending that any rulemaking
action based on PRM-72-3 be delayed until the NRC addressed issues in
10 CFR part 50 relating to the use of the ``FSAR'' as a licensing basis
document and the application of Sec. 50.59 in 10 CFR part 50. Another
commenter disagreed with this recommendation to delay rulemaking on
PRM-72-3.
Response: The Commission believes that issuance of this final rule
resolves this comment.
Comment: One commenter requested that the NRC prohibit general
licensees from using Sec. 72.48 and only permit cask design changes via
rulemaking. One commenter recommended that any identification of an
unreviewed safety question submitted to the NRC should require that NRC
conduct a hearing on the issue. One commenter suggested that the NRC
approve each Sec. 72.48 safety evaluation and place each evaluation in
the public document room. One commenter suggested that the NRC ``vacate
the generic ruling procedure'' subpart L and require that public
hearings be held prior to NRC cask certification. One commenter
suggested a moratorium on additional dry cask storage cask designs.
Response: Petitioner's concerns related to cask certification
issues; in particular, the process for modifying a SAR for a dry cask
storage design before and after issuance of the CoC. These comments
raise broad policy issues that go well beyond the scope of this
petition and rulemaking.
O.3 Part 71 (Transportation) Comments
Several commenters stated that a change control process similar to
Sec. 72.48 should be established in part 71 for transportation. These
commenters noted that for dual-purpose casks, used for both
transportation and storage, the lack of a process in part 71 would
limit the usefulness of the authority provided under Sec. 72.48.
Although the Commission agrees that this comment has merit, adding this
authority to part 71 is beyond the scope of the proposed rule. In
response to these comments, the Commission will consider adding
``Sec. 71.48-type'' change authority as part of a currently planned
rulemaking for part 71 intended to update requirements for
compatibility with the most recent International Atomic Energy Agency
transportation standards.
P. Other Topics Discussed in the Notice and Comments Not Related to
Preceding Topic Areas
The Federal Register notice containing the proposed rule also
solicited comments on particular topics that were discussed in the
preceding sections. In addition, comments were received on a number of
aspects not directly related to the rule language itself, such as
guidance, enforcement policy, the regulatory (and backfit) analysis, or
on other issues.
Guidance
Many comments were received on the subject of guidance. Many
suggested that NEI and NRC work together to develop guidance, and that
the guidance be endorsed before the revised rule becomes effective.
Commenters also requested examples of such matters as interdependent
changes, minimal increases, and screening of changes (as discussed in
Sections B and G).
The NRC agrees that guidance is important, and notes that NEI has
stated its willingness to revise existing guidance to conform with the
final rule such that NRC could endorse it. The
[[Page 53607]]
NRC will work with interested stakeholders to agree upon guidance that
includes consideration of these issues. Further, NRC is delaying the
required implementation of the rule for several months to allow time
for guidance to be revised.
Fuel Burnup Limits
One commenter stated that NRC should clarify the acceptance limits
of Sec. 51.55 concerning burnup assumptions for the transportation of
spent fuel for BWRs, as well as clarifying if this is subject to
Sec. 50.59 evaluations.
The Commission notes that a proposed rule (Sec. 51.52, not
Sec. 51.55 as cited by the commenter) was recently published on
February 26, 1999 (64 FR 9884), concerning environmental implications
of higher burnup fuel for transportation of spent fuel. Transportation
of fuel is not covered by Sec. 50.59 (as noted elsewhere in this
notice, the Commission is considering revisions to part 71 that would
add a change control process similar to Sec. 50.59 that could be used
for changes to transportation requirements under part 71). If the
commenter was asking whether higher burnup fuel can be used without NRC
approval, it is unlikely that such a change would satisfy the criteria
of Sec. 50.59, either because TS changes would be involved, other
requirements (e.g., Sec. 50.46) would not be met, or the burnup being
considered would be outside the range of what was approved in the
topical reports for the fuel.
Alternative Criteria
Two commenters proposed the use of alternate criteria for reactors
that are being decommissioned. One commenter suggested that a
``margin'' criterion is not necessary, but that a criterion on
environmental impact might be appropriate.
The Commission notes that the new criteria in the final rule that
replace the ``margin'' criterion are appropriate for a reactor being
decommissioned. Further, Sec. 50.82(a)(6) specifies that licensees
shall not perform any decommissioning activities that result in
significant environmental impact not previously reviewed. Section
50.82(a)(4) requires that the post-shutdown decommissioning activities
report include a discussion that provides the reasons for concluding
that the environmental impacts associated with site-specific
decommissioning activities will be bounded by appropriate, previously
issued environmental impact statements. For these reasons, the
Commission concludes that a criterion on environmental impact is not
needed.
The second commenter stated that the scope of Sec. 50.59 should be
limited to systems related to spent fuel pool cooling or radiological
waste.
The Commission notes that the staff involved in requirements for
decommissioning are developing guidance on the scope of information
required to be in an updated FSAR for a reactor undergoing
decommissioning. This effort is examining what information should be
retained in an FSAR for these facilities. The Commission believes that
defining the scope of information required to be in the FSAR for a
reactor undergoing decommissioning would be the best way to address the
apparent concern raised in this comment, rather than by modifying
Sec. 50.59 as recommended by the commenter.
Regulatory Analysis
Some comments were received on the regulatory analysis, primarily
that NRC underestimated the impacts on NRC and licensees of the number
of license amendments that would result, or the burden on part 72
licensees. These comments would appear to reflect a view that the
proposed rule would require more amendments than are currently
required, perhaps because of differences between the proposed rule
language and existing practice of some licensees using NEI 96-07, or
depending upon which formulation of ``margin of safety'' was ultimately
adopted. The Commission has prepared a final regulatory analysis that
reflects the final rule language and consideration of the public
comments. The Commission does not agree that the final rule language
will result in more amendments than presently arise under the existing
rule.
Need for Further Notice and Comment
Two commenters stated that the Commission should ensure that the
final rule is within the bounds of the proposed rule notice, or should
provide opportunity for public comment on substantive changes. The
Commission has examined the final rule for consistency with the
proposed rule and concludes that the final rule is within the bounds of
the proposed rule, taking due consideration of the public comments that
sought clarification and revisions in some respects, as well as greater
consistency between the Part 50 and Part 72 requirements.
Different Process for non-TS Issues
Several commenters believe that the license amendment process is
not well suited to the type of changes that require review under
Sec. 50.59(c)(2), but that do not involve changes to the TS or the
license directly. They believe that the Commission should establish a
different review process for such changes, such as letter approval.
The Commission notes that at one time (until 1974), Sec. 50.59 did
contain two approval processes, one for license amendments, and the
other for ``authorizations.'' The rule was revised in 1974 to delete
the ``authorization'' process and to handle all the required approvals
as license amendments. The Commission notes that the present rulemaking
provides some relaxation in the evaluation criteria. Therefore, the NRC
has responded to concerns about having to process a license amendment
for ``minimal'' changes. The current process provides opportunity for
public participation in the process under the provisions of Sec. 50.90
for changes that exceed the criteria, and for public knowledge, through
the summary reports, of those matters that did not require prior
approval. Therefore, the Commission does not plan to establish a
different process.
Other Definitions
Some commenters felt that NRC should provide better definitions of
certain terms that appear in Sec. 50.59 (and elsewhere), specifically,
for ``design bases'' and for ``important to safety.''
The Commission notes that Sec. 50.2 does define design bases, but
also notes that efforts are underway within the agency to enhance
understanding of what constitutes design basis information, through
possible development of criteria and examples. Concerning ``important
to safety,'' the Commission does not believe that a definition is
critical to implementation of the rule, since the set of SSCs viewed as
important to safety was arrived at during the license review and are
described in the FSAR. Thus, lack of an established definition is not
an impediment to implementation of the rule (the Commission notes that
for part 72, a definition is provided for SSC important to safety).
Applicability to Part 76
In its development of the proposed rule, as discussed in SECY-98-
171, the staff recommended exclusion of part 76 (``Certification of
Gaseous Diffusion Plants'') from those regulations for which rule
changes were being proposed. The basis for this recommendation was a
lack of design detail currently available in the safety analysis
reports for these plants. One commenter argued that the flexibility
provided by the revised evaluation criteria should also be included in
Sec. 76.68 (this section contains
[[Page 53608]]
requirements very similar to existing Secs. 50.59 and 72.48). This
commenter stated that the process by which changes are evaluated should
not vary based on the detail of the description being changed.
The Commission notes that the gaseous diffusion plants (GDP) have
significantly less design basis information than is currently available
for reactor facilities. The lack of design detail and lack of
understanding of the design basis has been documented in the Compliance
Plans for the GDPs, in NRC inspection reports, and is evident in the
GDP SARs. The Commission concludes that successful implementation of a
change control process is dependent upon the level of knowledge about
the design basis of the plant equipment or operation being changed. At
the present time, the Commission does not believe that additional
flexibility is appropriate for part 76 facilities.
Q. Enforcement Policy
Some commenters raised issues about how enforcement decisions would
be made during the transition period, and following implementation,
particularly with respect to evaluations performed in the past.
The Commission recognizes that it will take time to revise existing
industry guidance and to revise procedures, and conduct training on the
new rule provisions before the rule can be fully implemented. There
will still be the possibility of finding previous plant changes
performed prior to the implementation of the new rule that would be
potential violations of the previous rule. The Commission has concluded
that enforcement of potential violations of Secs. 50.59 and 72.48 for
past evaluations will be handled as described below, and also in
accordance with the NRC Enforcement Policy, NUREG-1600, Revision 1.
Following publication of the revised rule, for situations that
violate the ``old'' requirements, but that would not be violations had
the evaluation been performed under the revised rule, the NRC will
exercise enforcement discretion pursuant to VII.B.6 of the Enforcement
Policy and not issue citations against the ``old'' rule. The staff will
document in inspection reports that the issue was identified, but that
no enforcement action is being taken because the revised rule
requirements are met. However, for those situations identified prior to
the effective date of the revised rule that involve a violation of the
existing rule requirements but that would not be violations under the
revised rule, licensees still need to take the required corrective
action within a reasonable time frame commensurate with safety
significance to avoid the potential for a willful violation of NRC
requirements.
The NRC plans to maintain an enforcement panel made up of NRR (and
NMSS as applicable), OE, and OGC representatives for some months after
publication to maintain consistency. Additional enforcement policy
changes that may be applicable to violations of Secs. 50.59 or 72.48
are under consideration. The Commission intends to revise NUREG-1600,
Rev. 1, ``General Statement of Policy and Procedures for NRC
Enforcement Actions,'' consistent with this enforcement approach prior
to the effective date of the rule.
R. Implementation
The Commission recognizes the role that regulatory guidance will
play in effective implementation of the revisions to the rule. Existing
guidance (e.g., NEI 96-07 and NRC inspection guidance) needs to be
revised to conform with the rule changes. To allow time for the
guidance to be revised, and for licensees to implement the revised rule
provisions using the revised guidance, the Commission has established
that the rule changes to part 50 will become effective 90 days after
promulgation of the final regulatory guidance.
For part 72 facilities, current schedules for guidance would result
in availability at a time later than that anticipated for the guidance
for part 50. Accordingly, the effective date for these sections is
longer, set at 18 months from publication of the rule in the Federal
Register. For those sections in part 72 for which no guidance is
needed, as for instance, Secs. 72.244 and 72.246, the effective date is
120 days from publication.
III. Section by Section Analysis
10 CFR Part 50
10 CFR 50.59
As discussed in more detail above, Sec. 50.59 is being restructured
and revised to have the following components:
Paragraph (a): This is a new paragraph that contains definitions of
terms used in the rule. The terms establish requirements for when
evaluations are to be conducted to determine if the proposed changes,
tests, or experiments meet the criteria to require prior NRC approval.
Accordingly, definitions are given for ``change,'' ``facility as
described in the final safety analysis report (as updated) * * *,''
``procedures as described * * *,'' ``tests and experiments not
described * * *'' etc. The specific definitions were discussed in the
preceding sections.
Paragraph (b): Relocation into one paragraph of existing
applicability provisions. Section 50.59 applies to facilities licensed
under part 50, including power reactors and non-power reactors, whether
operating or being decommissioned.
Paragraph (c)(1): Relocation and clarification of existing
provisions establishing which changes, tests, or experiments require
evaluation and process for receiving approval when necessary. The
provisions now use the terms defined in paragraph (a), and refer to the
``final safety analysis report (as updated),'' rather than to ``safety
analysis report.'' The terminology of ``unreviewed safety question''
has been replaced by referring to the need to obtain a license
amendment.
Paragraph (c)(2): Reformatting of the (existing) evaluation
requirements into seven distinct statements of the criteria, addition
of an eighth criterion, and revision of the existing criteria for when
prior NRC approval of a change, test, or experiment is required.
Specifically, language of ``more than a minimal increase in frequency
(or likelihood),'' and of ``more than a minimal increase in
consequences'' was inserted in the criteria concerning accidents and
malfunctions, and rule requirements were revised from ``may be
created'' to ``would create'' concerning creation of accidents of a
different type and malfunctions of structures, systems, and components
important to safety with a different result (instead of existing
language of malfunction of equipment of a different type). In addition,
the existing criterion on ``margin of safety'' was replaced by a
criterion focusing upon design basis limits for fission product
barriers being exceeded or altered, and a new criterion was added to
control evaluation methods. These revisions clarify the criteria for
when prior approval is needed and allow some flexibility for licensees
to make changes that would not affect the NRC basis for licensing of
the facility.
Paragraph (c)(3): This is a new paragraph containing the
requirement that evaluations and analyses performed since the last FSAR
update was submitted need to be considered in performing evaluations of
changes to the facility or procedures, or for conduct of tests and
experiments. This paragraph is consistent with the terminology of
``final safety analysis report (as updated).''
Paragraph (c)(4): This is a new paragraph that states that
Sec. 50.59 requirements do not apply to changes to
[[Page 53609]]
the facility or procedures when other regulations establish more
specific criteria for such changes. Thus, this paragraph clarifies that
duplicative reviews in accordance with Sec. 50.59 are not necessary for
information that is described in the FSAR, but for which other
regulations provide standards for change control.
Paragraph (d)(1): Renumbered paragraph with (existing)
recordkeeping requirements. The text was simplified concerning which
records are needed, and conforming changes were made for the change in
terminology from ``safety evaluation'' to ``evaluation.''
Paragraph (d)(2): Renumbered paragraph with (existing) reporting
requirements. The text was simplified to state that summary reports
must be submitted at least once every 24 months, instead of the
existing statement that refers to submitting the summary report along
with the FSAR update submittal or annually. This revision will allow
all facilities to submit the report on a 24 month frequency.
Paragraph (d)(3): Renumbered paragraph on retention of records. The
text was revised to cover retention of records required by Sec. 50.59
until the term of any renewed license has expired.
10 CFR 50.66
This section specifies requirements for thermal annealing of a
reactor pressure vessel. The changes to Sec. 50.66 are to conform
existing language referring to unreviewed safety questions, and to
updated final safety analysis report, to the language in revised
Sec. 50.59.
10 CFR 50.71(e)
This section discusses requirements for periodic updating of the
final safety analysis report, to reflect the effects of changes made
either under Sec. 50.59, or through license amendments, or effects of
new analyses. The changes to this section are to conform language with
respect to unreviewed safety question, safety evaluation, and reference
to the final safety analysis report (as updated), with the language in
revised Sec. 50.59, as well as other minor wording changes as noted
above (e.g., ``approved'' license amendments).
10 CFR 50.90
A portion of existing Sec. 50.59(c) is being relocated into this
section. This change places the requirements for changes to technical
specifications themselves (not a result of a change, test or experiment
as defined in Sec. 50.59), into the rule section on amendments to
licenses rather than retaining the requirement in the section on
changes to the facility.
10 CFR Part 72
Most of the revisions in part 72 mirror those made to Sec. 50.59.
As for part 50, other changes are needed with respect to updating of
safety analysis reports, and in other sections for consistent
terminology.
10 CFR 72.3
The definition of ``independent spent fuel storage installation''
is being revised to remove the tests for evaluation of the
acceptability of sharing common utilities and services between the
ISFSI and other facilities. (Section 72.24 is being revised to include
this evaluation.)
10 CFR 72.9
Paragraph (b) is being revised as a conforming change to include in
the list of information collection requirements the new requirements in
Secs. 72.244 and 72.248 for amendments and for updates to the safety
analysis reports by CoC holders.
10 CFR 72.24
This section is being revised to reference shared common utilities
and services in the applicant's assessment of potential interactions
between the ISFSI and another facility (previously covered by
Sec. 72.3).
10 CFR 72.48
This section is being totally reformatted and revised, as discussed
above for Sec. 50.59. Specifically, it contains the following:
Paragraph (a): This paragraph now specifies definitions for terms
such as ``change'' and ``facility as described in the Final Safety
Analysis Report (as updated).'' Additionally, the term ``Final Safety
Analysis Report (FSAR) (as updated)'' has been defined to provide
greater clarity and consistency with Sec. 50.59 and other sections of
part 72.
Paragraph (b): This paragraph specifies that this section is
applicable to general and specific licensees for an ISFSI or MRS, and
to spent fuel storage cask certificate holders.
Paragraph (c): Paragraph (c)(1) establishes the conditions a
licensee or certificate holder must meet in order to (1) make changes
to the facility or spent fuel storage cask design as described in the
FSAR, or (2) make changes to the procedures as described in the FSAR,
or (3) conduct tests or experiments not described in the FSAR, without
prior NRC approval. Those conditions are that: (1) A change to the
technical specifications is not required; (2) a change in the terms,
conditions or specifications incorporated in the CoC is not required;
and (3) the change, test, or experiment does not meet any of the
criteria in paragraph (c)(2).
Paragraph (c)(2) lists the specific criteria which, if met, permit
a licensee or certificate holder to make the changes, or conduct the
tests or experiments, described in paragraph (c)(1) without NRC
approval. These new criteria revise existing criteria and conform with
the criteria adopted in Sec. 50.59(c)(2). Two existing criteria
involving a significant increase in occupational exposure or a
significant environmental impact have been deleted. Paragraph (c)(3)
states that changes made but not yet reflected in the FSAR update also
need to be considered in making the determination under paragraph
(c)(2). Paragraph (c)(4) states that Sec. 72.48 does not apply to
changes to the facility or procedures when the regulations establish
other change control processes for such changes.
Paragraph (d): This paragraph contains the recordkeeping
requirements and reporting requirements. In the final rule, subsection
numbers were included for clarity. For records, the rule is revised to
refer to the records of determinations of the need for license or
certificate of compliance (CoC) amendments, rather than to records
involving unreviewed safety question determinations. The time frame for
submitting summary reports in (renumbered) paragraph (d)(2) was revised
from 12 months to 24 months. The filing requirements for the summary
reports are modified to be consistent with Sec. 72.4 (Communications).
Paragraphs (d)(3), (d)(4) and (d)(5) contain record retention
requirements. The retention requirements for changes to procedures and
conduct of tests and experiments were revised to be 5 years (instead of
until termination). These time frames are more consistent with those in
Sec. 50.59, and also reflect that while facility changes need to be
maintained until termination, other records are of less importance
after a period of time such as 5 years. Paragraph (d)(3)(i) and
(d)(3)(ii) are renumbered and clarified with respect to when records no
longer need to be maintained.
New paragraph (d)(6) requires licensees who make changes under
Sec. 72.48 to provide copies of the records of such changes to the
certificate holder for the cask, and for the certificate holders who
make changes to provide
[[Page 53610]]
records to the general and specific licensees using that cask, within
60 days of implementing the changes.
10 CFR 72.56
Existing Sec. 72.48(c)(2) is being relocated into this section.
This is a parallel change to that for Secs. 50.59 and 50.90. The
Commission is placing the requirements for changes to license
conditions in the rule section on amendments to licenses instead of in
the section on changes to the facility.
10 CFR 72.70
This section contains requirements for updating of safety analysis
reports by licensees. Section 72.70 was reformatted and revised to
conform more closely with the update requirements in Sec. 50.71(e), as
well as those in (new) Sec. 72.248. The update frequency is being
revised from 12 months to 24 months. Paragraphs (a) and (b) are being
revised to use the terms ``Final Safety Analysis Report,'' ``FSAR,''
and ``as updated.'' Paragraph (a) is also being revised to indicate the
original FSAR for a specific licensee will be submitted within 90 days
of issuance of the license. Final analyses associated with completion
of construction or preoperational testing will be provided in the next
periodic update of the FSAR. The requirement for a licensee to submit a
FSAR 90 days before planned receipt of spent fuel has been removed, in
lieu of a notification under Sec. 72.80(g) by the licensee 90 days
before ISFSI operation commences. The section is also being revised to
add the requirement that changes to procedures be reflected in the
periodic updates of the FSAR. New paragraph (c) is being added to
provide requirements on submitting revisions to the FSAR for specific
licensees, including provisions for replacement pages, a cut off date
for changes, time frame to file, and provisions for updating if no
changes were made.
10 CFR 72.80
New paragraph (g) is being added to this section to require a
specific licensee to notify the NRC at least 90 days in advance of its
readiness to commence ISFSI (or MRS) operations This requirement
replaces a requirement in present Sec. 72.70(a) that an FSAR be
submitted to the Commission at least 90 days prior to the planned
receipt of spent fuel or high-level waste. This requirement thus
ensures that the NRC is informed in advance of licensee plans to use
the facility so that appropriate oversight activities can be conducted.
10 CFR 72.86
Paragraph (b) currently includes those sections under which
criminal sanctions are not issued. This paragraph is being revised to
add Secs. 72.244 and 72.246 as a conforming change to reflect that
certificate holders who fail to comply with these new sections would
not be subject to the criminal penalty provisions of section 223 of the
Atomic Energy Act (AEA). New Sec. 72.248 has not been included in
paragraph (b) to reflect that certificate holders who fail to comply
with this new section would be subject to the criminal penalty
provisions of section 223 of the AEA.
10 CFR 72.212(b)(2)
Paragraph (b)(2)(i) retains the current rule language but has been
renumbered and reordered for clarity as a result of the addition of
paragraph (b)(2)(ii). Paragraph (b)(2)(ii) was added to require that
the general licensee evaluate any changes to the written evaluations
required by Sec. 72.212 using the requirements of Sec. 72.48(c).
10 CFR 72.212(b)(4)
The change to this section is to conform the reference to
Sec. 50.59 provisions, specifically to change from the terminology of
unreviewed safety question to referring to the need for a license
amendment for the facility (that is, the reactor facility at whose site
the independent spent fuel storage installation is located).
10 CFR 72.216
In the proposed rule, a new paragraph (d) would have been added to
present requirements for a general licensee to submit annual updates to
a final safety analysis report (FSAR) for the cask or casks approved
for spent fuel storage that are used by the general licensee. In the
final rule, this section was withdrawn because the Commission concluded
that it was not necessary for general licensees to submit updates to
the safety analysis report for the approved cask design that they are
using for storage.
10 CFR 72.244
This new section presents requirements for how a certificate holder
is to submit an application to amend the certificate of compliance
(CoC). This section is similar to the requirements in Sec. 72.56 for
licensees to apply for an amendment to their license.
10 CFR 72.246
This new section presents requirements for approval of an amendment
to a CoC. This section is similar to the requirements in Sec. 72.58 for
approval of an amendment to a license.
10 CFR 72.248
This new section presents requirements for submittal of periodic
updates to an FSAR associated with the design of a spent fuel storage
cask which has been issued a CoC. This new section also states that the
changes to procedures and SSC associated with the spent fuel storage
cask and which are made pursuant to Sec. 72.48 would be included in the
update. This section is similar to the requirements in Sec. 72.70 for
submission of updates to the FSAR associated with a part 72 license and
to the requirements in Sec. 50.71(e) for power reactor FSAR updates.
IV. Finding of No Significant Environmental Impact
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, as adopted, will not have
a significant impact on the environment. The rule changes are of two
types: those that relate to the processes for evaluating and approving
changes to licensed facilities and those that involve the degree of
potential change in safety for which changes can proceed without NRC
review. The process changes will make it more likely that planned
changes are properly reviewed and approved by NRC when necessary. With
respect to the criteria changes, only minimal increases in frequencies
of postulated design basis accidents will be allowed without prior NRC
review. All changes to the Technical Specifications, which are the
operating limits and other parameters of most immediate concern for
public health and safety, will continue to require prior NRC review and
approval. Changes to the facility that would involve an accident of a
different type from any already analyzed require prior approval.
Further, changes that result in more than minimal increases in
radiological consequences will continue to require prior NRC approval,
including NRC consideration as to whether there is a potential impact
on the environment. Therefore, the Commission concludes that there will
be no significant impact on the environment from this rule. This
discussion constitutes the environmental assessment and finding of no
significant impact for this rulemaking.
V. Paperwork Reduction Act Statement
This rule amends information collection requirements that are
subject
[[Page 53611]]
to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). The
proposed rule was submitted to the Office of Management and Budget for
review and approval of the information collection requirements.
Existing requirements were approved by the Office of Management and
Budget approval numbers 3150-0011 and 3150-0132.
The rule changes affect information collection requirements through
the existing reporting requirements in Sec. 50.59 for a summary report
of changes, tests and experiments, performed under the authority of
Sec. 50.59 as well as recordkeeping requirements. Similar requirements
exist in Sec. 72.48 for licensees under part 72. In addition, revisions
are being made to the requirements in Sec. 72.70 and (new) 72.248 for
submittal of updates to the safety analysis reports. Further, the final
rule establishes recordkeeping and reporting requirements for CoC
holders who make changes to an approved storage cask design in
accordance with Sec. 72.48.
The public reporting burden for this information collection request
was estimated in the proposed rule to average 3100 hours per response,
including the time for reviewing instructions, searching existing data
sources, gathering and maintaining the data needed, and completing and
reviewing the information collection. The Commission had estimated that
there would be only a slight increase in burden associated with these
proposed changes over the existing burden. For the final rule, certain
of the provisions that might have resulted in an increase in burden
have been removed; therefore, the Commission now concludes that the
final rule would result in an overall reduction in reporting and
recordkeeping burden, other than for the estimated effort required for
a one-time revision to procedures and training. Therefore, the present
estimate of the public reporting burden for this information collection
request under the final rule is 2900 hours per response.
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.
VI. Regulatory Analysis
The Commission has prepared a regulatory analysis for this
rulemaking. The analysis sets forth the objectives of the rulemaking,
the alternatives considered, and examines the values and impacts of the
alternatives considered by the Commission. The alternatives considered
in this analysis include no action, issuance of guidance only, or
rulemaking. The analysis is available for inspection in the NRC Public
Document Room, 2120 L Street NW., (Lower Level), Washington, D.C.
VII. Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act of 1980, (5
U.S.C. 605(b)), the Commission certifies that this rule will not, have
a significant economic impact on a substantial number of small
entities. This rule affects only the licensing, operation and
decommissioning of nuclear power plants, nonpower reactors, and
independent spent fuel storage facilities (including cask certificate
holders). The companies that own these facilities do not fall within
the scope of the definition of ``small entities'' set forth in the
Regulatory Flexibility Act or the Small Business Size Standards set out
in regulations issued by the Small Business Administration at 13 CFR
part 121.
VIII. Backfit Analysis
The Commission has evaluated these rule changes under the
backfitting requirements in Secs. 50.109 and 72.62. The Commission does
not regard the changes to be backfits as defined in Secs. 50.109(a)(1)
and 72.62(a), as applicable. Accordingly, a backfit analysis applicable
to these changes has not been prepared. However, the Commission has
prepared a regulatory analysis which sets forth the objectives of the
rulemaking changes, the alternatives that were considered, and the
expected benefits and costs associated with the rulemaking changes. The
Commission regards this analysis as providing for a disciplined
approach for evaluating the impacts of the proposed changes, which
satisfies the underlying purposes of the backfitting requirements in
Secs. 50.109 and 72.62.
Changes to Section 50.59
Section 50.59 defines the circumstances under which holders of
nuclear power plant operating licenses may make changes to and conduct
tests or experiments at their facilities without prior NRC review and
approval. In this rulemaking, new definitions are added to Sec. 50.59
(e.g., the definitions for ``change,'' and ``facility as described in
the final safety analysis report (as updated)''), and the structure and
language of the rule were modified (e.g., the addition of a new
applicability section, and the removal of the term, ``unreviewed safety
question''). These changes constitute clarifications of the existing
rule, and codification of existing NRC practice and interpretations of
terminology which are undefined by the current rule. Clarifications and
codification of existing NRC interpretation and practice do not
constitute a generic backfit (although the application of the revised
rule may constitute a plant-specific backfit). The new criteria in
Sec. 50.59(c)(2)(i), (ii), (iii), (iv), (v) and (vi) are being added
primarily 4 for the purpose of providing additional
flexibility to licensees to make changes and conduct tests without
having to obtain prior NRC review and approval. Each of these changes
constitute permissive relaxations 5 from the superseded
Sec. 50.59(a)(2)(i) and (ii) criteria. Permissive relaxations are not
considered to be backfits, inasmuch as a licensee will continue to be
in compliance with the final rule even if it uses its existing
procedures and the superseded criteria for implementing Sec. 50.59. The
new criteria in Sec. 50.59(c)(2)(vii) and (viii) together constitute
replacements for the superseded Sec. 50.59(a)(2)(iii) criterion on
``margin of safety.'' As noted in Section J, these two criteria
together, in place of a criterion on margin of safety, explicitly cover
those margins that the Commission believes are important to address in
this evaluation process--the first being the margin that exists in the
limits that are to be met, and the second being the margin that exists
from the conservatisms included in the methods used to demonstrate that
requirements are met. The replacement criteria were thus developed to
accomplish two complementary goals: (1) Defining with more precision
the important safety margins which should be the focus of a Sec. 50.59
determination, rather than the problematic term, ``margin of safety as
defined in the basis for any technical specification;'' and (2)
assuring that the relaxations embodied in the Sec. 50.59(c)(2)(i),
(ii), (iii), (iv), (v) and (vi) criteria will not result in changes
approaching the adequate protection threshold without prior NRC review
and approval. As such, the new criteria (vii) and (viii) are
fundamentally part of the overall regulatory scheme in the revisions to
Sec. 50.59 which relax and clarify the thresholds for licensee-
initiated changes and tests requiring
[[Page 53612]]
prior NRC review and approval before their implementation. In sum, the
Commission has determined that the changes to Sec. 50.59 constitute
clarifications and codifications of existing practices, or constitute
permissive relaxations from the existing Sec. 50.59 criteria, and
therefore do not constitute backfits as defined in Sec. 50.109(a)(1).
---------------------------------------------------------------------------
\4\ In some cases, these changes coincide with other changes
intended to clarify and codify existing practice, and to make the
rule easier to understand (e.g., separating the ``frequency of
occurrence'' of an accident from the ``consequences'' of an accident
as a criterion for NRC review and approval.
\5\ ``Permissive'' relaxations are relaxations which licensees
may voluntarily choose (but are not compelled) to comply.
---------------------------------------------------------------------------
Changes to Part 72
Section 72.48 defines the circumstances under which a holder of a
ISFSI license may make changes and conduct tests and experiments,
analogous to the criteria in Sec. 50.59. The change to Sec. 72.48 will
conform the criteria for ISFSI and storage cask changes to that in
Sec. 50.59. Therefore, as with the changes to Sec. 50.59, the changes
to Sec. 72.48 constitute a permissive relaxation as compared with the
existing criteria in Sec. 72.48. Furthermore, there will be consistency
in regulatory approach in changes to nuclear power plants and ISFSIs.
Such consistency is appropriate since most ISFSIs are licensed to
nuclear power plant licensees; there are resource efficiencies for such
licensees using the same criteria for evaluating changes, tests and
experiments. The change criteria in Sec. 72.48 are also extended by the
final rule to holders of CoCs., which contributes to regulatory
stability and predictability since known standards will be utilized in
determining whether a change to a CoC may be made without prior NRC
review and approval. The existing backfitting provision in Sec. 72.62
only apply to licensees and not to CoC holders. However, even if the
backfitting provisions in Sec. 72.62 applied to CoC holders, the
changes in Sec. 72.48 would not be regarded as backfits since the
extension of Sec. 72.48 to CoC holders represents a permissive
relaxation. For similar reasons, the changes in part 72 applicable to
CoC holders, which are necessary to support the extension of the change
criteria in Sec. 72.48 to CoC holders, are not considered to be
backfits under Sec. 72.62.
The Commission is deferring consideration of conforming changes to
the design certifications in part 52, appendices A and B, which are the
design certifications for the ABWR and System 80+ designs. The
Commission will conduct a broader rulemaking to amend part 52, whose
purpose will be to correct typographic errors, clarify language, and
reflect lessons learned as a result of the ABWR, System 80+, and AP600
design certification rulemakings. If conforming changes to appendices A
and B are made, in a future rulemaking, the Commission regards this
rulemaking amending Sec. 50.59 as satisfying the Commission's
obligations under the backfit rule for any conforming changes made to
part 52, inasmuch as the backfitting issues associated with the
adoption of the new criteria are being addressed in this rulemaking.
IX. 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.
X. National Technology Transfer and Advancement Act
The National Technology Transfer and Advancement Act of 1995, Pub.
L. 104-113, requires that Federal agencies use technical standards
developed by or adopted by voluntary consensus standards bodies unless
the use of such a standard is inconsistent with applicable law or
otherwise impractical. There are no consensus standards that apply to
the change control process requirements established in this rulemaking.
Thus the provisions of the Act do not apply to this rulemaking.
XI. Criminal Penalties
For the purposes of section 223 of the Atomic Energy Act (AEA), the
Commission is issuing this rule to amend 10 CFR part 50:50.59, : 50.66,
and :50.71; and 10 CFR part 72:72.48, : 72.70, :72.212, and :72.248,
under one or more of sections 161b, 161i, or 161o of the AEA. Willful
violations of the rule would be subject to criminal enforcement.
XII. Compatibility of Agreement State Regulations
Under the ``Policy Statement on Adequacy and Compatibility of
Agreement State Programs'' approved by the Commission on June 30, 1997,
and published in the Federal Register (62 FR 46517, September 3, 1997),
this rule is classified as compatibility Category ``NRC.''
Compatibility is not required for Category ``NRC'' regulations. The NRC
program elements in this category are those that relate directly to
areas of regulation reserved to the NRC by the AEA or the provisions of
Title 10 of the Code of Federal Regulations, and although an Agreement
State may not adopt program elements reserved to NRC, it may wish to
inform its licensees of certain requirements via a mechanism that is
consistent with the particular State's administrative procedure laws,
but that does not confer regulatory authority on the State.
List of Subjects
10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
reactors, Radiation protection, Reactor siting criteria, Reporting and
record keeping requirements.
10 CFR Part 72
Criminal penalties, Manpower training programs, Nuclear materials,
Occupational safety and health, Reporting and recordkeeping
requirements, Security measures, Spent fuel.
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 10 CFR parts 50 and 72.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
1. The authority citation for part 50 continues to read as follows:
Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,
83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246, (42 U.S.C. 5841, 5842, 5846).
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951, as amended by Pub. L. 102-486, sec. 2902, 106 Stat. 3123, (42
U.S.C. 5851). Sections 50.10 also issued under secs. 101, 185, 68
Stat. 936, 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L.
91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd),
and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42
U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued
under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a,
50.55a, and Appendix Q also issued under sec. 102, Pub. L. 91-190,
83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued
under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58,
50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42
U.S.C. 2239). Sections 50.78 also issued under sec. 122, 68 Stat.
939 (42 U.S.C. 2152). Sections 50.80, 50.81 also issued under sec.
184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also
issued under sec. 187, 66 Stat. 955 (42 U.S.C. 2237).
2. Section 50.59 is revised to read as follows:
[[Page 53613]]
Sec. 50.59 Changes, tests, and experiments.
(a) Definitions for the purposes of this section:
(1) Change means a modification or addition to, or removal from,
the facility or procedures that affects a design function, method of
performing or controlling the function, or an evaluation that
demonstrates that intended functions will be accomplished.
(2) Departure from a method of evaluation described in the FSAR (as
updated) used in establishing the design bases or in the safety
analyses means:
(i) Changing any of the elements of the method described in the
FSAR (as updated) unless the results of the analysis are conservative
or essentially the same; or
(ii) Changing from a method described in the FSAR to another method
unless that method has been approved by NRC for the intended
application.
(3) Facility as described in the final safety analysis report (as
updated) means:
(i) The structures, systems, and components (SSC) that are
described in the final safety analysis report (FSAR) (as updated),
(ii) The design and performance requirements for such SSCs
described in the FSAR (as updated), and
(iii) The evaluations or methods of evaluation included in the FSAR
(as updated) for such SSCs which demonstrate that their intended
function(s) will be accomplished.
(4) Final Safety Analysis Report (as updated) means the Final
Safety Analysis Report (or Final Hazards Summary Report) submitted in
accordance with Sec. 50.34, as amended and supplemented, and as updated
per the requirements of Sec. 50.71(e) or Sec. 50.71(f), as applicable.
(5) Procedures as described in the final safety analysis report (as
updated) means those procedures that contain information described in
the FSAR (as updated) such as how structures, systems, and components
are operated and controlled (including assumed operator actions and
response times).
(6) Tests or experiments not described in the final safety analysis
report (as updated) means any activity where any structure, system, or
component is utilized or controlled in a manner which is either:
(i) Outside the reference bounds of the design bases as described
in the final safety analysis report (as updated) or
(ii) Inconsistent with the analyses or descriptions in the final
safety analysis report (as updated).
(b) Applicability. This section applies to each holder of a license
authorizing operation of a production or utilization facility,
including the holder of a license authorizing operation of a nuclear
power reactor that has submitted the certification of permanent
cessation of operations required under Sec. 50.82(a)(1) or a reactor
licensee whose license has been amended to allow possession but not
operation of the facility.
(c)(1) A licensee may make changes in the facility as described in
the final safety analysis report (as updated), make changes in the
procedures as described in the final safety analysis report (as
updated), and conduct tests or experiments not described in the final
safety analysis report (as updated) without obtaining a license
amendment pursuant to Sec. 50.90 only if:
(i) A change to the technical specifications incorporated in the
license is not required, and
(ii) The change, test, or experiment does not meet any of the
criteria in paragraph (c)(2) of this section.
(2) A licensee shall obtain a license amendment pursuant to
Sec. 50.90 prior to implementing a proposed change, test, or experiment
if the change, test, or experiment would:
(i) Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the final safety
analysis report (as updated);
(ii) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component (SSC)
important to safety previously evaluated in the final safety analysis
report (as updated);
(iii) Result in more than a minimal increase in the consequences of
an accident previously evaluated in the final safety analysis report
(as updated);
(iv) Result in more than a minimal increase in the consequences of
a malfunction of an SSC important to safety previously evaluated in the
final safety analysis report (as updated);
(v) Create a possibility for an accident of a different type than
any previously evaluated in the final safety analysis report (as
updated);
(vi) Create a possibility for a malfunction of an SSC important to
safety with a different result than any previously evaluated in the
final safety analysis report (as updated);
(vii) Result in a design basis limit for a fission product barrier
as described in the FSAR (as updated) being exceeded or altered; or
(viii) Result in a departure from a method of evaluation described
in the FSAR (as updated) used in establishing the design bases or in
the safety analyses.
(3) In implementing this paragraph, the FSAR (as updated) is
considered to include FSAR changes resulting from evaluations performed
pursuant to this section and analyses performed pursuant to Sec. 50.90
since submittal of the last update of the final safety analysis report
pursuant to Sec. 50.71 of this part.
(4) The provisions in this section do not apply to changes to the
facility or procedures when the applicable regulations establish more
specific criteria for accomplishing such changes.
(d)(1) The licensee shall maintain records of changes in the
facility, of changes in procedures, and of tests and experiments made
pursuant to paragraph (c) of this section. These records must include a
written evaluation which provides the bases for the determination that
the change, test, or experiment does not require a license amendment
pursuant to paragraph (c)(2) of this section.
(2) The licensee shall submit, as specified in Sec. 50.4, a report
containing a brief description of any changes, tests, and experiments,
including a summary of the evaluation of each. A report must be
submitted at intervals not to exceed 24 months.
(3) The records of changes in the facility must be maintained until
the termination of a license issued pursuant to this part or the
termination of a license issued pursuant to 10 CFR part 54, whichever
is later. Records of changes in procedures and records of tests and
experiments must be maintained for a period of 5 years.
3. In Sec. 50.66, paragraph (b), introductory text, paragraphs
(b)(4), (c)(2), and (c)(3)(iii) are revised to read as follows:
Sec. 50.66 Requirements for thermal annealing of the reactor pressure
vessel.
* * * * *
(b) Thermal Annealing Report. The Thermal Annealing Report must
include: a Thermal Annealing Operating Plan; a Requalification
Inspection and Test Program; a Fracture Toughness Recovery and
Reembrittlement Trend Assurance Program; and an Identification of
Changes Requiring a License Amendment.
(1) * * *
(4) Identification of Changes Requiring a License Amendment. Any
changes to the facility as described in the final safety analysis
report (as updated) which requires a license amendment pursuant to
Sec. 50.59(c)(2) of this part, and any changes to the Technical
Specifications, which are necessary to either conduct the thermal
annealing or to operate the nuclear
[[Page 53614]]
power reactor following the annealing must be identified. The section
shall demonstrate that the Commission's requirements continue to be
complied with, and that there is reasonable assurance of adequate
protection to the public health and safety following the changes.
(c) * * *
(2) If the thermal annealing was completed but the annealing was
not performed in accordance with the Thermal Annealing Operating Plan
and the Requalification Inspection and Test Program, the licensee shall
submit a summary of lack of compliance with the Thermal Annealing
Operating Plan and the Requalification Inspection and Test Program and
a justification for subsequent operation to the Director, Office of
Nuclear Reactor Regulation. Any changes to the facility as described in
the final safety analysis report (as updated) which are attributable to
the noncompliances and which require a license amendment pursuant to
Sec. 50.59(c)(2) and any changes to the Technical Specifications shall
also be identified.
(i) If no changes requiring a license amendment pursuant to
Sec. 50.59(c)(2) or changes to Technical Specifications are identified,
the licensee may restart its reactor after the requirements of
paragraph (f)(2) of this section have been met.
(ii) If any changes requiring a license amendment pursuant to
Sec. 50.59(c)(2) or changes to the Technical Specifications are
identified, the licensee may not restart its reactor until approval is
obtained from the Director, Office of Nuclear Reactor Regulation and
the requirements of paragraph (f)(2) of this section have been met.
(3) * * *
(iii) If the partial annealing was not performed in accordance with
the Thermal Annealing Operating Plan and the Requalification Inspection
and Test Program, the licensee shall submit a summary of lack of
compliance with the Thermal Annealing Operating Plan and the
Requalification Inspection and Test Program and a justification for
subsequent operation to the Director, Office of Nuclear Reactor
Regulation. Any changes to the facility as described in the final
safety analysis report (as updated) which are attributable to the
noncompliances and which require a license amendment pursuant to
Sec. 50.59(c)(2) and any changes to the technical specifications which
are required as a result of the noncompliances, shall also be
identified.
(A) If no changes requiring a license amendment pursuant to
Sec. 50.59(c)(2) or changes to Technical Specifications are identified,
the licensee may restart its reactor after the requirements of
paragraph (f)(2) of this section have been met.
(B) If any changes requiring a license amendment pursuant to
Sec. 50.59(c)(2) or changes to Technical Specifications are identified,
the licensee may not restart its reactor until approval is obtained
from the Director, Office of Nuclear Reactor Regulation and the
requirements of paragraph (f)(2) of this section have been met.
* * * * *
4. In Sec. 50.71, paragraph (e), introductory text is revised to
read as follows:
Sec. 50.71 Maintenance of records, making of reports.
* * * * *
(e) Each person licensed to operate a nuclear power reactor
pursuant to the provisions of Sec. 50.21 or Sec. 50.22 of this part
shall update periodically, as provided in paragraphs (e) (3) and (4) of
this section, the final safety analysis report (FSAR) originally
submitted as part of the application for the operating license, to
assure that the information included in the report contains the latest
information developed. This submittal shall contain all the changes
necessary to reflect information and analyses submitted to the
Commission by the licensee or prepared by the licensee pursuant to
Commission requirement since the submittal of the original FSAR, or as
appropriate the last update to the FSAR under this section. The
submittal shall include the effects \1\ of: All changes made in the
facility or procedures as described in the FSAR; all safety analyses
and evaluations performed by the licensee either in support of approved
license amendments, or in support of conclusions that changes did not
require a license amendment in accordance with Sec. 50.59(c)(2) of this
part; and all analyses of new safety issues performed by or on behalf
of the licensee at Commission request. The updated information shall be
appropriately located within the update to the FSAR.
---------------------------------------------------------------------------
\1\ Effects of changes includes appropriate revisions of
descriptions in the FSAR such that the FSAR (as updated) is complete
and accurate.
---------------------------------------------------------------------------
(1) * * *
* * * * *
5. Section 50.90 is revised to read as follows:
Sec. 50.90 Application for amendment of license or construction
permit.
Whenever a holder of a license or construction permit desires to
amend the license (including the Technical Specifications incorporated
into the license) or permit, application for an amendment must be filed
with the Commission, as specified in Sec. 50.4, fully describing the
changes desired, and following as far as applicable, the form
prescribed for original applications.
PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE
6. The authority citation for part 72 continues to read as follows:
Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183,
184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953,
954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C.
2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233,
2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat.
688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846);
Pub. L. 95-601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851); sec. 102,
Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133,
135, 137, 141, Pub. L. 97-425, 96 Stat. 2229, 2230, 2232, 2241, sec.
148, Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152,
10153, 10155, 10157, 10161, 10168).
Section 72.44(g) also issued under secs. 142(b) and 148 (c),
(d), Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C.
10162(b), 10168(c), (d)). Section 72.46 also issued under sec. 189,
68 Stat. 955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat.
2230 (42 U.S.C. 10154). Section 72.96(d) also issued under sec.
145(g), Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)).
Subpart J also issued under secs. 2(2), 2(15), 2(19), 117(a),
141(h), Pub. L. 97-425, 96 Stat. 2202, 2203, 2204, 2222, 2224 (42
U.S.C. 10101, 10137(a), 10161(h)). Subparts K and L are also issued
under sec. 133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96
Stat. 2252 (42 U.S.C. 10198).
7. Section 72.3 is amended by revising the definition for
independent spent fuel storage installation or ISFSI to read as
follows:
Sec. 72.3 Definitions.
* * * * *
Independent spent fuel storage installation or ISFSI means a
complex designed and constructed for the interim storage of spent
nuclear fuel and other radioactive materials associated with spent fuel
storage. An ISFSI which is located on the site of another facility
licensed under this part or a facility licensed under part 50 of this
chapter and which shares common utilities and services with such a
facility or is physically connected with such other
[[Page 53615]]
facility may still be considered independent.
* * * * *
8. In Sec. 72.9, paragraph (b) is revised to read as follows:
Sec. 72.9 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Secs. 72.7, 72.11, 72.16, 72.19, 72.22 through
72.34, 72.42, 72.44, 72.48 through 72.56, 72.62, 72.70 through 72.82,
72.90, 72.92, 72.94, 72.98, 72.100, 72.102, 72.104, 72.108, 72.120,
72.126, 72.140 through 72.176, 72.180 through 72.186, 72.192, 72.206,
72.212, 72.216, 72.218, 72.230, 72.232, 72.234, 72.236, 72.240, 72.244,
and 72.248.
9. In Sec. 72.24, paragraph (a) is revised as follows:
Sec. 72.24 Contents of application: Technical information.
* * * * *
(a) A description and safety assessment of the site on which the
ISFSI or MRS is to be located, with appropriate attention to the design
bases for external events. Such assessment must contain an analysis and
evaluation of the major structures, systems, and components of the
ISFSI or MRS that bear on the suitability of the site when the ISFSI or
MRS is operated at its design capacity. If the proposed ISFSI or MRS is
to be located on the site of a nuclear power plant or other licensed
facility, the potential interactions between the ISFSI or MRS and such
other facility--including shared common utilities and services--must be
evaluated.
* * * * *
10. Section 72.48 is revised to read as follows:
Sec. 72.48 Changes, tests, and experiments.
(a) Definitions for the purposes of this section:
(1) Change means a modification or addition to, or removal from,
the facility or spent fuel storage cask design or procedures that
affects a design function, method of performing or controlling the
function, or an evaluation that demonstrates that intended functions
will be accomplished.
(2) Departure from a method of evaluation described in the FSAR (as
updated) used in establishing the design bases or in the safety
analyses means:
(i) Changing any of the elements of the method described in the
FSAR (as updated) unless the results of the analysis are conservative
or essentially the same; or
(ii) Changing from a method described in the FSAR to another method
unless that method has been approved by NRC for the intended
application.
(3) Facility means either an independent spent fuel storage
installation (ISFSI) or a Monitored Retrievable Storage facility( MRS).
(4) The facility or spent fuel storage cask design as described in
the Final Safety Analysis Report (FSAR) (as updated) means:
(i) The structures, systems, and components (SSC) that are
described in the FSAR (as updated),
(ii) The design and performance requirements for such SSCs
described in the FSAR (as updated), and
(iii) The evaluations or methods of evaluation included in the FSAR
(as updated) for such SSCs which demonstrate that their intended
function(s) will be accomplished.
(5) Final Safety Analysis Report (as updated) means:
(i) For specific licensees, the Safety Analysis Report for a
facility submitted and updated in accordance with Sec. 72.70;
(ii) For general licensees, the Safety Analysis Report for a spent
fuel storage cask design, as amended and supplemented; and
(iii) For certificate holders, the Safety Analysis Report for a
spent fuel storage cask design submitted and updated in accordance with
Sec. 72.248.
(6) Procedures as described in the Final Safety Analysis Report (as
updated) means those procedures that contain information described in
the FSAR (as updated) such as how SSCs are operated and controlled
(including assumed operator actions and response times).
(7) Tests or experiments not described in the Final Safety Analysis
Report (as updated) means any activity where any SSC is utilized or
controlled in a manner which is either:
(i) Outside the reference bounds of the design bases as described
in the FSAR (as updated) or
(ii) Inconsistent with the analyses or descriptions in the FSAR (as
updated).
(b) This section applies to:
(1) Each holder of a general or specific license issued under this
part, and
(2) Each holder of a Certificate of Compliance (CoC) issued under
this part.
(c)(1) A licensee or certificate holder may make changes in the
facility or spent fuel storage cask design as described in the FSAR (as
updated), make changes in the procedures as described in the FSAR (as
updated), and conduct tests or experiments not described in the FSAR
(as updated), without obtaining either:
(i) A license amendment pursuant to Sec. 72.56 (for specific
licensees) or
(ii) A CoC amendment submitted by the certificate holder pursuant
to Sec. 72.244 (for general licensees and certificate holders) if:
(A) A change to the technical specifications incorporated in the
specific license is not required; or
(B) A change in the terms, conditions, or specifications
incorporated in the CoC is not required; and
(C) The change, test, or experiment does not meet any of the
criteria in paragraph (c)(2) of this section.
(2) A specific licensee shall obtain a license amendment pursuant
to Sec. 72.56, a certificate holder shall obtain a CoC amendment
pursuant to Sec. 72.244, and a general licensee shall request that the
certificate holder obtain a CoC amendment pursuant to Sec. 72.244,
prior to implementing a proposed change, test, or experiment if the
change, test, or experiment would:
(i) Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the FSAR (as
updated);
(ii) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a system, structure, or component (SSC)
important to safety previously evaluated in the FSAR (as updated);
(iii) Result in more than a minimal increase in the consequences of
an accident previously evaluated in the FSAR;
(iv) Result in more than a minimal increase in the consequences of
a malfunction of an SSC important to safety previously evaluated in the
FSAR (as updated);
(v) Create a possibility for an accident of a different type than
any previously evaluated in the FSAR (as updated);
(vi) Create a possibility for a malfunction of an SSC important to
safety with a different result than any previously evaluated in the
FSAR (as updated);
(vii) Result in a design basis limit for a fission product barrier
being exceeded or altered as described in the FSAR (as updated); or
(viii) Result in a departure from a method of evaluation described
in the FSAR (as updated) used in establishing the design bases or in
the safety analyses.
(3) In implementing this paragraph, the FSAR (as updated) is
considered to include FSAR changes resulting from evaluations performed
pursuant to this section and analyses performed pursuant to Sec. 72.56
or Sec. 72.244 since the
[[Page 53616]]
last update of the FSAR pursuant to Sec. 72.70, or Sec. 72.248 of this
part.
(4) The provisions in this section do not apply to changes to the
facility or procedures when the applicable regulations establish more
specific criteria for accomplishing such changes.
(d)(1) The licensee and certificate holder shall maintain records
of changes in the facility or spent fuel storage cask design, of
changes in procedures, and of tests and experiments made pursuant to
paragraph (c) of this section. These records must include a written
evaluation which provides the bases for the determination that the
change, test, or experiment does not require a license or CoC amendment
pursuant to paragraph (c)(2) of this section.
(2) The licensee and certificate holder shall submit, as specified
in Sec. 72.4, a report containing a brief description of any changes,
tests, and experiments, including a summary of the evaluation of each.
A report shall be submitted at intervals not to exceed 24 months.
(3) The records of changes in the facility or spent fuel storage
cask design shall be maintained until:
(i) Spent fuel is no longer stored in the facility or the spent
fuel storage cask design is no longer being used, or
(ii) The Commission terminates the license or CoC issued pursuant
to this part.
(4) The records of changes in procedures and of tests and
experiments shall be maintained for a period of 5 years.
(5) The holder of a spent fuel storage cask design CoC, who
permanently ceases operation, shall provide the records of changes to
the new certificate holder or to the Commission, as appropriate, in
accordance with Sec. 72.234(d)(3).
(6)(i) A general licensee shall provide a copy of the record for
any changes to a spent fuel storage cask design to the applicable
certificate holder within 60 days of implementing the change.
(ii) A specific licensee using a spent fuel storage cask design,
approved pursuant to subpart L of this part, shall provide a copy of
the record for any changes to a spent fuel storage cask design to the
applicable certificate holder within 60 days of implementing the
change.
(iii) A certificate holder shall provide a copy of the record for
any changes to a spent fuel storage cask design to any general or
specific licensee using the cask design within 60 days of implementing
the change.
11. Section 72.56 is revised to read as follows:
Sec. 72.56 Application for amendment of license.
Whenever a holder of a specific license desires to amend the
license (including a change to the license conditions), an application
for an amendment shall be filed with the Commission fully describing
the changes desired and the reasons for such changes, and following as
far as applicable the form prescribed for original applications.
12. Section 72.70 is revised to read as follows:
Sec. 72.70 Safety analysis report updating.
(a) Each specific licensee for an ISFSI or MRS shall update
periodically, as provided in paragraphs (b) and (c) of this section,
the final safety analysis report (FSAR) to assure that the information
included in the report contains the latest information developed.
(1) Each licensee shall submit an original FSAR to the Commission,
in accordance with Sec. 72.4, within 90 days after issuance of the
license.
(2) The original FSAR shall be based on the safety analysis report
submitted with the application and reflect any changes and applicant
commitments developed during the license approval and/or hearing
process.
(b) Each update shall contain all the changes necessary to reflect
information and analyses submitted to the Commission by the licensee or
prepared by the licensee pursuant to Commission requirement since the
submission of the original FSAR or, as appropriate, the last update to
the FSAR under this section. The update shall include the effects \1\
of:
---------------------------------------------------------------------------
\1\ Effects of changes includes appropriate revisions of
descriptions in the FSAR such that the FSAR (as updated) is complete
and accurate.
---------------------------------------------------------------------------
(1) All changes made in the ISFSI or MRS or procedures as described
in the FSAR;
(2) All safety analyses and evaluations performed by the licensee
either in support of approved license amendments, or in support of
conclusions that changes did not require a license amendment in
accordance with Sec. 72.48;
(3) All final analyses and evaluations of the design and
performance of structures, systems, and components that are important
to safety taking into account any pertinent information developed
during final design, construction, and preoperational testing; and
(4) All analyses of new safety issues performed by or on behalf of
the licensee at Commission request. The information shall be
appropriately located within the updated FSAR.
(c)(1) The update of the FSAR shall be filed in accordance with
Sec. 72.4, on a replacement-page basis;
(2) The update shall include a list that identifies the current
pages of the FSAR following page replacement;
(3) Each replacement page shall include both a change indicator for
the area changed, e.g., a bold line vertically drawn in the margin
adjacent to the portion actually changed, and a page change
identification (date of change or change number or both);
(4) The update shall include:
(i) A certification by a duly authorized officer of the licensee
that either the information accurately presents changes made since the
previous submittal, or that no such changes were made; and
(ii) An identification of changes made under the provisions of
Sec. 72.48, but not previously submitted to the Commission;
(5) The update shall reflect all changes implemented up to a
maximum of 6 months prior to the date of filing; and
(6) Updates shall be filed every 24 months from the date of
issuance of the license.
(d) The updated FSAR shall be retained by the licensee until the
Commission terminates the license.
13. In Sec. 72.80, paragraph (g) is added to read as follows:
Sec. 72.80 Other records and reports.
* * * * *
(g) Each specific licensee shall notify the Commission, in
accordance with Sec. 72.4, of its readiness to begin operation at least
90 days prior to the first storage of spent fuel or high-level waste in
an ISFSI or MRS.
14. In Sec. 72.86, paragraph (b) is revised to read as follows:
Sec. 72.86 Criminal penalties.
* * * * *
(b) The regulations in this part 72 that are not issued under
sections 161b, 161i, or 161o for the purposes of section 223 are as
follows: Secs. 72.1, 72.2, 72.3, 72.4, 72.5, 72.7, 72.8, 72.9, 72.16,
72.18, 72.20, 72.22, 72.24, 72.26, 72.28, 72.32, 72.34, 72.40, 72.46,
72.56, 72.58, 72.60, 72.62, 72.84, 72.86, 72.90, 72.96, 72.108, 72.120,
72.122, 72.124, 72.126, 72.128, 72.130, 72.182, 72.194, 72.200, 72.202,
72.204, 72.206, 72.210, 72.214, 72.220, 72.230, 72.238, 72.240, 72.244,
and 72.246.
15. In Sec. 72.212, paragraphs (b)(2) and (b)(4) are revised to
read as follows:
Sec. 72.212 Conditions of general license issued under Sec. 72.210.
* * * * *
[[Page 53617]]
(b) * * *
(2)(i) Perform written evaluations, prior to use, that establish
that:
(A) conditions set forth in the Certificate of Compliance have been
met;
(B) cask storage pads and areas have been designed to adequately
support the static load of the stored casks; and
(C) the requirements of Sec. 72.104 have been met. A copy of this
record shall be retained until spent fuel is no longer stored under the
general license issued under Sec. 72.210.
(ii) The licensee shall evaluate any changes to the written
evaluations required by this paragraph using the requirements of
Sec. 72.48(c). A copy of this record shall be retained until spent fuel
is no longer stored under the general license issued under Sec. 72.210.
* * * * *
(4) Prior to use of this general license, determine whether
activities related to storage of spent fuel under this general license
involve a change in the facility Technical Specifications or require a
license amendment for the facility pursuant to Sec. 50.59(c)(2) of this
chapter. Results of this determination must be documented in the
evaluation made in paragraph (b)(2) of this section.
16. Section 72.244 is added to read as follows:
Sec. 72.244 Application for amendment of a certificate of compliance.
Whenever a certificate holder desires to amend the CoC (including a
change to the terms, conditions or specifications of the CoC), an
application for an amendment shall be filed with the Commission fully
describing the changes desired and the reasons for such changes, and
following as far as applicable the form prescribed for original
applications.
17. Section 72.246 is added to read as follows:
Sec. 72.246 Issuance of amendment to a certificate of compliance.
In determining whether an amendment to a CoC will be issued to the
applicant, the Commission will be guided by the considerations that
govern the issuance of an initial CoC.
18. Section 72.248 is added to read as follows:
Sec. 72.248 Safety analysis report updating.
(a) Each certificate holder for a spent fuel storage cask design
shall update periodically, as provided in paragraph (b) of this
section, the final safety analysis report (FSAR) to assure that the
information included in the report contains the latest information
developed.
(1) Each certificate holder shall submit an original FSAR to the
Commission, in accordance with Sec. 72.4, within 90 days after the
spent fuel storage cask design has been approved pursuant to
Sec. 72.238.
(2) The original FSAR shall be based on the safety analysis report
submitted with the application and reflect any changes and applicant
commitments developed during the cask design review process. The
original FSAR shall be updated to reflect any changes to requirements
contained in the issued Certificate of Compliance (CoC).
(b) Each update shall contain all the changes necessary to reflect
information and analyses submitted to the Commission by the certificate
holder or prepared by the certificate holder pursuant to Commission
requirement since the submission of the original FSAR or, as
appropriate, the last update to the FSAR under this section. The update
shall include the effects \1\ of:
---------------------------------------------------------------------------
\1\ Effects of changes includes appropriate revisions of
descriptions in the FSAR such that the FSAR (as updated) is complete
and accurate.
---------------------------------------------------------------------------
(1) All changes made in the spent fuel storage cask design or
procedures as described in the FSAR;
(2) All safety analyses and evaluations performed by the
certificate holder either in support of approved CoC amendments, or in
support of conclusions that changes did not require a CoC amendment in
accordance with Sec. 72.48; and
(3) All analyses of new safety issues performed by or on behalf of
the certificate holder at Commission request. The information shall be
appropriately located within the updated FSAR.
(c)(1) The update of the FSAR shall be filed in accordance with
Sec. 72.4, on a replacement-page basis;
(2) The update shall include a list that identifies the current
pages of the FSAR following page replacement;
(3) Each replacement page shall include both a change indicator for
the area changed, e.g., a bold line vertically drawn in the margin
adjacent to the portion actually changed, and a page change
identification (date of change or change number or both);
(4) The update shall include:
(i) A certification by a duly authorized officer of the certificate
holder that either the information accurately presents changes made
since the previous submittal, or that no such changes were made; and
(ii) An identification of changes made by the certificate holder
under the provisions of Sec. 72.48, but not previously submitted to the
Commission;
(5) The update shall reflect all changes implemented up to a
maximum of 6 months prior to the date of filing;
(6) Updates shall be filed every 24 months from the date of
issuance of the CoC; and
(7) The certificate holder shall provide a copy of the updated FSAR
to each general and specific licensee using its cask design.
(d) The updated FSAR shall be retained by the certificate holder
until the Commission terminates the certificate.
(e) A certificate holder who permanently ceases operation, shall
provide the updated FSAR to the new certificate holder or to the
Commission, as appropriate, in accordance with Sec. 72.234(d)(3).
Dated at Rockville, Maryland, this 20th day of September, 1999.
For the Nuclear Regulatory Commission.
Annette Vietti-Cook,
Secretary of the Commission.
[FR Doc. 99-25054 Filed 10-1-99; 8:45 am]
BILLING CODE 7590-01-P