[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

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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

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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

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(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.
---------------------------------------------------------------------------

    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