98-28066. Changes, Tests, and Experiments

[Federal Register Volume 63, Number 203 (Wednesday, October 21, 1998)]
[Proposed Rules]
[Pages 56098-56125]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-28066]

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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 50, 52 and 72

RIN 3150-AF94

Changes, Tests, and Experiments

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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SUMMARY: The Nuclear Regulatory Commission is proposing to amend its
regulations concerning the authority for licensees of production or
utilization facilities, such as nuclear reactors, and independent spent
fuel storage facilities, to make changes to the facility or procedures,
or to conduct tests or experiments, without prior NRC approval. The
proposed rule would clarify which changes, tests and experiments
conducted at a licensed facility require evaluation, and the criteria
that determine when NRC approval is needed before such changes to a
licensed facility can be implemented. The proposed rule would also add
definitions for terms that have been subject to differing
interpretations, reorganize the rule language for clarity, and revise
the criteria for when prior NRC approval is needed. The Commission is
also seeking comment on several specific issues as discussed below.

DATES: Submit comments by December 21, 1998. Comments received after
this date will be considered if it is practical to do so, but the
Commission is able to assure consideration only for comments received
on or before this date.

ADDRESSES: Send comments to: Secretary, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001. ATTN: Rulemakings and
Adjudications Staff.
Hand deliver comments to: 11555 Rockville Pike, Rockville,
Maryland, between 7:45 a.m. and 4:15 p.m. Federal workdays.

FOR FURTHER INFORMATION CONTACT: Eileen McKenna, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, telephone (301) 415-2189. (emm@nrc.gov) or Naiem Tanious,
Office of Nuclear Materials Safety and Safeguards, U.S. Nuclear
Regulatory Commission, Washington DC 20555-0001, telephone (301) 415-
6103 (nst@nrc.gov).

SUPPLEMENTARY INFORMATION:
I. Background
II. Proposed Rule Topics and Issues
A. Organization of the rule requirements
B. Change to the facility as described in the Safety Analysis
Report
C. Change to the procedures as described in the Safety Analysis
Report
D. Tests and experiments not described in the Safety Analysis
Report
E. Safety Analysis Report
F. Probability of occurrence or consequences of an accident or
malfunction of equipment important to safety previously evaluated in
the safety analysis report may be increased
G. More than a minimal increase in probability or consequences
H. Possibility of an accident of a different type from any
previously evaluated in the Safety Analysis Report may be created
I. Possibility of a malfunction of a different type from any
previously evaluated in the Safety Analysis Report may be created
J. Margin of safety as defined in the basis for any technical
specification is Reduced
K. Safety Evaluation
L. Reporting and record keeping requirements
M. Part 72 changes
III. Section by Section Analysis
IV. Commission Voting Record on SECY-98-171
V. Rule Language Proposed by the Nuclear Energy Institute
VI. Request for Public Comments
VII. Availability of Documents and Electronic Access
VIII. Finding of No Significant Environmental Impact
IX. Paperwork Reduction Act Statement
X. Regulatory Analysis
XI. Regulatory Flexibility Certification
XII. Backfit Analysis
XIII. Criminal Penalties
XIV. Compatibility Agreement State Regulations

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
10 CFR 72.48 for licensees of facilities for the independent storage of
spent nuclear fuel and high-level radioactive waste. This proposed
rulemaking affects the requirements for 10 CFR parts 50, 52 and 72; for
simplicity, the discussion will focus primarily on the language in 10
CFR 50.59). These regulations provide that licensees may make changes
to the facility or procedures as described in the safety analysis
report, 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 incorporated in the license or an unreviewed safety

[[Page 56099]]

question. Section 50.59(a)(2), as currently codified, states:

``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 record keeping and reporting requirements
associated with such changes, tests or experiments.
In order to understand the reasons for the provisions of the
current rule, and how the Commission proposes to revise it, it is
helpful to understand how this process fits within the overall
requirements undergirding licensing and oversight of nuclear reactors.

Overview of Licensing Process

The application for an operating license includes the final safety
analysis report (FSAR) which is to contain: a description of the
facility; the design bases and limits on operation; and the safety
analysis for the structures, systems, and components (SSC) and of the
facility as a whole. The safety analysis emphasizes performance
requirements, analytical bases and technical justifications, and
evaluations that show how safety functions will be accomplished. Design
bases include the specific functions that the SSC need to perform, the
parameters that need to be controlled to assure the function, and the
range of values for these parameters. As part of the FSAR, the
applicant is required to propose, for NRC approval, Technical
Specifications(TS) that will become part of the license.
The NRC issues a license after finding, among other things, that
the plant has been built according to its design and can be operated
within its design limits. The NRC prepares a safety evaluation report
that documents the basis for its findings, including its review of the
design information provided in the FSAR (and supporting documents) and
the applicable acceptance criteria (established either in regulations,
standards or guidance documents). In some cases, the NRC staff performs
independent analyses to confirm the adequacy of the facility design to
meet regulatory requirements. One example of this practice is the staff
calculation of radiological consequences (doses) for design basis
accidents.
The licensee is required to operate the facility in accordance with
NRC regulations and with requirements contained in the license. The
license describes the facility in general terms, and includes specific
conditions imposed on the facility and the licensee, as well as
incorporates the TS. Section 50.36 of the regulations defines for
inclusion in the TS, those limits and parameters of most immediate
significance for protection of public health and safety: safety limits,
limiting safety system settings, limiting conditions for operation,
surveillance requirements, and design features to which changes would
have a significant effect on safety, and administrative controls. The
TS are derived from the safety analysis, evaluations, and design bases
described in the FSAR. Any changes to the TS must receive NRC review
and approval before they are made.
Engineering evaluations demonstrate that the fundamental safety
principles of the plant design are met. Design basis events play a
central role in plant design. These are a combination of postulated
challenges and failure events against which plants are designed to
ensure adequate and safe plant response. Design basis events are
defined as conditions of normal operation, anticipated operational
occurrences and design basis accidents, external events and natural
phenomena for which the plant has been designed to ensure the integrity
of the pressure boundary, the capability to shutdown safely, and the
capability to prevent or mitigate the consequences of accidents. For
events with high frequency, NRC requires that consequences be low (such
as by preventing fuel damage). For more severe, but less probable
accidents, the allowable consequences are higher, but must still meet
the regulatory guidelines established in 10 CFR part 100. Adequacy of
the reactor design is evaluated by consideration of postulated design
basis events viewed as sufficiently credible that the facility should
be designed to prevent or mitigate their effects.
During the design process, plant response is evaluated using
assumptions that are intended to be conservative to account for
uncertainties in analysis or data. In the Final Safety Analysis Report
(FSAR), analyses are done conservatively to account for uncertainties
in the design, construction, and operation of nuclear power plants.
These conservatisms are introduced into FSAR analyses in numerous ways.
For example, some computer codes model systems and processes in a
simplified but bounding fashion. Analysis input assumptions are
typically worst case values (consistent with the design and operating
limits) of instrument drift or error, temperature, pressure, fluid
volume and enthalpy, flow rate, system response time, heat transfer
rate and heat capacity, reactivity coefficients, power history and
decay heat. An FSAR analysis also typically assumes the worst-case
single-active failure of equipment.
National standards and other regulatory policies, such as defense-
in-depth, constitute additional engineering considerations that
influence plant design and operation. Commensurate with expected
frequency and consequences of challenges to the system, defense-in-
depth could require: (1) Multiple means to accomplish safety functions
and prevent release of radioactive material (multiple barriers); (2)
reasonable balance among prevention of core damage, prevention of
containment failure and consequence mitigation; (3) system redundancy;
(4) independence; and (5) diversity.
Various margins exist in a facility design. These margins are based
on, for example, assumptions of initial conditions, conservatisms in
computer modeling and codes, allowance for instrument drift and system
response time, redundancy and independence of components in safety
trains, and plant response during operating transient and accident
conditions. Margin is provided by meeting codes and standards or
alternatives approved for use by NRC, including the safety analysis
acceptance criteria in the FSAR and in supporting analyses. Not all
margin that exists falls within the purview of ``reduction in margin of
safety \1\ as defined in the basis for any technical specification.''
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\1\ Margin of safety is not defined in the regulations, although
it is mentioned in Sec. 50.34(a) (``the margins of safety during
normal operations and transient conditions anticipated during the
life of the facility''); Sec. 50.92(c) (``No significant hazards
considerations if the proposed amendment would not involve a
significant reduction in a margin of safety'') as well as
Sec. 50.59.
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When a plant is licensed, the NRC states in its Safety Evaluation
Report (SER) why it found each FSAR analysis acceptable. An FSAR
analysis may be accepted because it was considered to be adequately
conservative and because the NRC's acceptance criteria for that
analysis are met. Frequently, the SER states specific conditions the
NRC relied upon for concluding that the analysis was conservative.
Examples of such conditions may be the use of an NRC-approved computer
code, correlation, or setpoint methodology, specific limitations on one
or more input assumptions, or penalties put into a calculation to
account for uncertainties. In addition to being stated in a plant-

[[Page 56100]]

specific SER, these conditions may be found in other safety evaluations
such as for an analysis method proposed by a topical report.
Changes to the basis for licensing occur over the life of the plant
through promulgation of new rules, plant-specific license amendments
and other analyses and reviews that may be conducted, such as in
response to NRC bulletins and generic letters. The NRC prepares a
safety evaluation for many of these issues based upon either licensee
requests for changes or licensee responses to NRC requests for
information. The licensee is required to periodically update the final
safety analysis report to reflect effects of these changes so that the
safety analysis report (as updated) remains a complete and accurate
description and analysis of the facility such that it can serve as the
reference document for evaluation of changes made under 10 CFR 50.59.

10 CFR 50.59 Evaluation Process

Section 50.59 was promulgated in 1962 to allow licensees to make
certain changes that affect systems, structures, components, 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 when 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 the level of safety documented
in the original licensing basis, such as in the safety analysis report.
The process is 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. Margins and equipment functionality,
reliability and availability also may be impacted by facility changes.
Therefore, the criteria for requiring NRC approval were directly
related to: (1) Preserving licensing assumptions concerning initiation
of design basis events by not allowing a different type of initiating
event or probability of occurrence larger than previously considered;
(2) preserving effectiveness (reliability) of the mitigation systems by
not allowing introduction of different equipment malfunctions and by
limiting increases in probability of malfunction, or reductions in the
margin of safety (which reflects the capability of the system); and (3)
preserving acceptability of consequences by limiting increases in
consequences of the postulated design basis events.

Implementation Guidance

In 1989, an industry guidance document, NSAC-125, ``Guidelines for
10 CFR 50.59 Safety Evaluations'' was published to assist licensees in
the conduct of the evaluations required under Sec. 50.59. The NRC
neither endorsed nor disapproved this document. While the staff
concluded that the evaluation process established in NSAC-125 was
generally sound, the staff was unable to endorse the document because
of some inconsistencies between the implementation guidance and the
language of Sec. 50.59.
On October 31, 1997, the Nuclear Energy Institute (NEI) submitted
for staff review a revised guidance document, NEI 96-07, ``Guidelines
for 10 CFR 50.59 Safety Evaluations.'' This document is an updated
version of NSAC-125 that NEI modified in response to some of the staff
positions, and other implementation issues arising from licensee use of
the NSAC-125 guidance. Along with the submittal of the guidance
document, NEI included an industry-wide initiative that would require
industry adoption and implementation of the revised guidance by June
1998. The NRC provided comments to NEI concerning this guidance in a
letter dated January 9, 1998. This letter noted that certain aspects of
this guidance were unacceptable for implementation of Sec. 50.59 as
presently written.
Staff efforts to develop guidance on implementation of Sec. 50.59
were prompted by a reassessment of the 10 CFR 50.59 evaluation process,
conducted in 1995, that examined existing guidance and practice, with
the goal of identifying how the process could be improved, or where
additional guidance was needed. The staff provided an action plan to
the Commission on April 15, 1996, outlining the actions the staff
proposed to complete with respect to guidance and oversight of
implementation of Sec. 50.59. The staff review identified a number of
areas in which the meaning of the rule language is not clear, or where
staff and industry interpretations (such as those in NSAC-125) are
different. In SECY-97-035, dated February 12, 1997, the staff forwarded
to the Commission proposed regulatory guidance on implementation of
Sec. 50.59. In this SECY, the staff presented positions on a number of
topic areas. These positions in some cases reaffirmed existing
regulatory practice or clarified staff expectations, and in other
areas, established positions where guidance did not previously exist.
In its proposed guidance, the staff compared its proposed regulatory
guidance to industry guidance contained in NSAC-125. In accordance with
a Commission Staff Requirements Memorandum dated April 25, 1997, the
staff guidance was published in the Federal Register as draft NUREG-
1606 (Proposed Regulatory Guidance Related to Implementation of 10 CFR
50.59), for public comment on May 7, 1997 (62 FR 24947).
In response to the Federal Register notice, many comments were
submitted that voiced strong opposition to a number of the positions
proposed by the staff. These comments were summarized in Attachment 1
to SECY-97-205, Integration and Evaluation of Results from Recent
Lessons-Learned Reviews, dated September 10, 1997. Since that time, the
NRC has conducted a more detailed review of the comments and concludes
that some issues can be resolved through guidance, while in other
areas, rulemaking is necessary to clarify the implementation issues. A
copy of this analysis of comments is available for review in the NRC
Public Document Room. As noted, the staff concluded that rulemaking was
necessary to resolve some of the issues associated with implementation
of the rule.

II. Proposed Rule Topics and Issues

The NRC is proposing rulemaking on Sec. 50.59 (and Sec. 72.48) to
address a number of issues concerning implementation of the current
rule, and suitability of the criteria that determine when an unreviewed
safety question exists. The implementation issues primarily relate to
cases involving judgment as to whether a proposed change requires NRC
approval before it can be implemented. The differing interpretations of
the rule as it relates to an increase in probability of an accident, or
an increase in consequences have contributed to disputed inspection and
enforcement findings. Too stringent an interpretation of the meaning of
the requirements could result in diversion of licensee and staff
resources for review of inconsequential changes. Too high a threshold
for NRC review could lead to erosion of safety margins without NRC
review, particularly from the cumulative effect of more than one
change. In developing the proposed rule, the Commission has carefully
weighed these matters in trying to establish an appropriate threshold
for NRC review.
Conforming changes are proposed in other portions of the rules,
including Sec. 50.66, 50.71(e) for production and utilization
facilities licensed under part 50. Conforming changes are also

[[Page 56101]]

required in Sec. 72.212(b)(4) and Appendices A and B to part 52 (Design
Certification Rules for ABWR and System 80+ respectively).
In addition, the Commission is proposing to make parallel changes
applicable to facilities for independent spent fuel storage facilities
licensed in accordance with part 72. These changes are included in the
sections below (in some cases, the discussion of the issue focuses on
Sec. 50.59 for simplicity; except where noted, the discussion is also
applicable to the changes for Sec. 72.48). As part of the proposed
changes to part 72, the Commission is also proposing to extend the
change control process authority granted to ISFSI or MRS license
holders (in Sec. 72.48) to holders of NRC Certificates of Compliance
(CoC) for a spent fuel storage cask design.
In addition to changes to the requirements within Secs. 50.59 and
72.48, the Commission is also proposing to rearrange certain provisions
of these rules to provide a more logical structure. These changes do
not affect the substance of the requirements, but rather affect only
where they are located and how they are stated. These organizational
changes are discussed first, followed by discussion of each of the
issues where revisions to requirements are proposed by this rulemaking.
The proposed rule revisions are presented in the order that the issues
currently arise in the regulations.

A. Organization of the Rule Requirements

The organizational changes being proposed include the following:
(1) Applicability
In the existing rule, language concerning applicability to
different facilities is contained in three different paragraphs. These
facilities are: 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
Commission proposes to place all of these provisions in one paragraph
that is clearly labeled ``Applicability.'' ²
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\2\ Section 50.59(a) refers to holders of a license authorizing
operation of a production or utilization facility. Section 50.59(d)
explicitly refers to power reactor licensees who have submitted
certification of permanent cessation of operation required under
Sec. 50.82(a)(1)(i). As noted in Sec. 50.82(a)(iii), for power
reactors whose licenses were modified to allow possession but not
operation, before the effective date of this rule (that is of
Sec. 50.82), the certification of Sec. 50.82(a)(1)(i) shall be
deemed to have been submitted. Section 50.59(e) refers to non-power
reactors whose license no longer authorizes operation. The net
effect is that Sec. 50.59 applies to both power and nonpower
reactors, whether authorized to operate or no longer authorized to
operate (and to other production or utilization facilities).
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(2) Form of prior Commission approval
Existing Sec. 50.59(a) refers to the need for prior Commission
approval of changes, tests, and experiments under certain conditions,
but the method of receiving that approval is not discussed until
paragraph (c), which states that the licensee shall submit an
application for amendment under Sec. 50.90. The Commission proposes to
combine these two paragraphs and to revise the regulation to state more
clearly that a licensee must apply for and obtain a license amendment,
pursuant to Sec. 50.90, before implementing such changes, tests, or
experiments. This organizational change to the rule of combining
(existing) paragraphs (a) and (c) will also facilitate some of the
other proposed changes, such as the criteria for when approval is
needed.
(3) Criteria for needing Commission approval of changes, tests and
experiments and Unreviewed Safety Question (USQ) designation
The Commission proposes to remove the reference in the rule to the
term ``unreviewed safety question'' and instead to refer to the need to
obtain a license amendment. The Commission believes that the
terminology of ``USQ'' has sometimes led to confusion about the purpose
of the evaluation required by Sec. 50.59. Some licensees have concluded
that if they determined a change was safe, there could be no need for
NRC approval.
The Commission notes that the purpose of performing evaluations
against the criteria specified in Sec. 50.59 is to identify possible
changes that might affect the basis for licensing of the facility so
that any changes that might pose a safety concern are either reviewed
by the NRC or not implemented by the licensee. This evaluation process
will thus distinguish those changes which by their nature do not raise
safety concerns and therefore do not require prior NRC approval to
confirm their safety, from those that must be reviewed by the NRC to
independently confirm their safety before implementation. To avoid
confusion between a determination of safety and a determination of the
need for NRC approval, the Commission proposes to revise Sec. 50.59 to
delete use of the term ``unreviewed safety question'' and instead to
list the criteria (in new Sec. 50.59(c)(2)) that require prior
Commission approval, in the form of a license amendment. It is also
noted that many facility technical specifications refer to unreviewed
safety question determinations and such TS should ultimately be revised
in accordance with the final wording of Sec. 50.59. The deletion of
reference to USQ also requires a number of conforming changes to other
parts of the regulations, including Part 52 (Appendices A and B), in
which the term is presently used.
This proposed rule would revise the existing compound statements
contained with the evaluation criteria to state each specific criterion
individually. This will make the regulation more consistent with how it
is generally implemented by licensees. Changes to the criteria are
discussed in the sections below.
Finally, the Commission would simplify existing Sec. 50.59(c) by
removing the following statement: ``The holder of a license . . . who
desires (1) a change to its technical specifications . . . shall submit
an application for amendment of his license pursuant to Sec. 50.90.''
This statement refers to changes to the TS not associated with a
change, test or experiment. The Commission concludes that a more
suitable place for this provision is within Sec. 50.90, and therefore
as part of this rulemaking, proposes to modify Sec. 50.90 to state that
if a licensee wishes to amend its license (including the TS
incorporated into it), the licensee must file an application as
specified in Sec. 50.90. Revised Sec. 50.59(c)(i) would be revised to
state that if a proposed change, test, or experiment would involve a TS
change, the Sec. 50.90 process must be followed in order to change the
technical specification such that the proposed change, test or
experiment may be implemented.

B. Change to the Facility as Described in the Safety Analysis Report

Section 50.59 states that ``changes to the facility as described in
the safety analysis report'' must be evaluated to determine whether
prior approval is needed before implementation. As discussed in NUREG-
1606 and in the comment discussions, a common understanding between the
NRC and the industry on what constitutes a ``change to the facility as
described in the safety analysis report'' is necessary for effective
functioning of the review process. Guidance on preparation of
Sec. 50.59 evaluations provides the means for review of the effects of
changes, but these reviews are not conducted if the activity is not
considered to be a ``change . . .'' The Commission concludes that
modification of an existing provision (e.g., SSC, design requirement,
analysis method or

[[Page 56102]]

parameter), additions, and removals (physical removals or non-reliance
on a system to meet a requirement) are all changes to the facility as
described in the final safety analysis. The Commission believes that
additions to the facility which were not previously evaluated, could
adversely impact facility performance and the bases upon which the NRC
previously determined the acceptability of the design as described in
the SAR. Accordingly, the Commission concludes that additions should be
considered ``changes to the facility as described in the SAR'' in order
to assure that such changes are subject to evaluation using the
Sec. 50.59 criteria for determining whether prior NRC review and
approval are necessary.
Differences in interpretation have occurred about whether changes
that do not actually change the physical plant (the ``hardware'')
require a Sec. 50.59 evaluation. As an example, consider a change being
made to the basis (documented in the SAR) for demonstrating adequacy of
the facility without a physical change to the facility. Such changes
might include changes to evaluative methods, acceptance standards,
procurement specifications, or other information for SSC described in
the FSAR. The Commission believes that Sec. 50.59 does apply to the
requirements for design, construction and operation, and the safety
analyses for the facility that are documented in the FSAR. Section
50.34(b), ``Final safety analysis report,'' requires the FSAR to
contain a presentation of the design bases and the limits on its
operation, a description and analysis of the SSC of the facility, with
emphasis upon performance requirements, the bases, with technical
justifications therefore, upon which such requirements have been
established, and the evaluations required to show that safety functions
will be accomplished. The original licensing decision was based in part
upon the margins provided by performance requirements, analysis methods
and assumptions described in the SAR, and reviewed by the staff in the
SER. Therefore, the Commission concludes that changes to such
information (e.g., 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 SAR'' in order to assure that such
changes are subject to evaluation using the Sec. 50.59 criteria for
determining whether prior NRC review and approval are necessary.
If changes to methods and assumptions were not controlled, a
licensee might revise its analyses and then subsequently conclude that
a later facility change did not require NRC approval because the
results of the (new) analysis with this change were bounded by the
previous analysis. This proposed rulemaking would add definitions in
Sec. 50.59 of ``change'' and of ``facility as described in the final
safety analysis report(as updated)'' 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.
Accordingly, the Commission proposes to add the following as
definitions in section Sec. 50.59:
Change means a modification, addition, or removal.
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 (as updated), (ii)
design or performance requirements or methods of operation for such SSC
required to be included or described in the final safety analysis
report (as updated), and (iii) evaluations or methods of evaluation
required to be included in the FSAR (as updated) for such SSC that
demonstrate that their intended functions will be accomplished or that
their design bases can be met.
The Commission endorses the staff's previously stated position (in
draft NUREG-1606) about what constitutes a single change, as compared
to packaging of several changes with offsetting effects. Interdependent
changes (i.e., where a second change is caused by the first, with
respect to function or performance), can be treated as a single change,
whereas treating as one change the combination of changes (whether to
the facility directly or to the safety analysis) to offset one that
would otherwise require prior approval is not an appropriate
application of Sec. 50.59.

C. Change to the Procedures as Described in the Safety Analysis Report

The Commission proposes to provide 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 would
include the evaluations demonstrating that requirements are met, such
as assumed operator actions and response times.
The Commission also notes that Sec. 50.34(b) states that the final
SAR is to contain the managerial and administrative controls to be used
to meet Appendix B (Quality Assurance), and plans for coping with
emergencies, per Appendix E. Section 50.59 applies to changes to
procedures as described in the SAR. Quality assurance and emergency
planning program requirements are subject to the change control
provisions of Secs. 50.54(a) and 50.54(q) respectively. Based on this
set of rule provisions, it could be inferred that changes to quality
assurance or emergency plans would require both a Sec. 50.59 evaluation
and a Sec. 50.54 [either (a) or (q)] evaluation. The Sec. 50.54
³ regulations provide criteria and reporting requirements
specific to the plans and which were promulgated after Sec. 50.59. To
reduce duplication of effort, the Commission proposes that changes to
these programs be governed by Sec. 50.54 requirements, and that a
Sec. 50.59 evaluation would not be required unless other information
described in the FSAR is also being changed. The proposed rule would
add language to specifically exclude from the scope of Sec. 50.59
changes to procedures where other more specific requirements and
criteria have been established by regulation for controlling these
changes (e.g., for information required by Sec. 50.34(b)(6) (ii) and
(v)), through a provision in the Sec. 50.59(c)(1) of the proposed rule.
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\3\ Section 50.54(p) establishes change control requirements for
safeguards contingency plans. While these plans are part of the
application submitted pursuant to Sec. 50.34, they are not part of
the FSAR, and thus Sec. 50.59 would not apply to these plans.
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The proposed definition for ``procedures as described in the final
safety analysis report (as updated)'' is as follows:

Procedures as described in the final safety analysis report (as
updated) means information in the final safety analysis report (as
updated) regarding how systems, structures and components are
operated and controlled (including assumed operator actions and
response times), including assumed operator actions and response
times, and information on conduct of operations.

D. Tests and Experiments Not Described in the Safety Analysis Report

Section 50.59 also discusses the conduct of tests or experiments
not described in the safety analysis report. ``Test'' is, of course,
subject to many meanings including both routine verifications of
function, and also more unusual evolutions. In the former category,
there are many tests that are conducted that are not explicitly
described in the SAR. For example, a licensee conducts tests of
component and system performance that verify the

[[Page 56103]]

SSCs perform the functions as described or required. (Performance of
tests is typically controlled by procedure.) However, there also may be
tests of new materials or means of plant operation that may put the
plant in a situation that has not been previously evaluated and that
could affect the capability of SSC to perform their required functions.
The existing rule was designed to ensure that the latter type of tests
would be reviewed before they were conducted. Therefore, to assure that
there is clear definition with respect to the tests that are subject to
prior NRC review and approval before they are conducted, the Commission
proposes that a definition of ``tests and experiments not described in
the safety analysis report'' be provided in Sec. 50.59 as follows:

Tests or experiments not described in the final safety analysis
report (as updated) means any activity where the reactor or any of
its systems, structures, or components are used or controlled in a
manner which cannot be shown to be within (i) the controlling
parameters of their design bases as described in the final safety
analysis report (as updated) or (ii) consistent with the analyses in
the final safety analysis report (as updated).

E. Safety Analysis Report

In developing the proposed rule changes, the Commission noted the
varying references to the safety analysis report within related
sections of part 50. For example, in Sec. 50.59, the phrase used is
``safety analysis report,'' in Sec. 50.66, the reference is to the
``updated final safety analysis report;'' and Sec. 50.71(e) refers to
the updated FSAR. (Other sections and parts generally refer to the
final safety analysis report (e.g. part 55), but this is not
universally true (e.g. Sec. 50.54(a)). For purposes of Sec. 50.59,
``safety analysis report'' refers to the current revision of the FSAR,
so that the changes are evaluated against the most complete and
accurate description of the facility. When performing evaluations, a
licensee needs to consider changes already made for which the FSAR
update has not yet been submitted to the NRC. The Commission emphasizes
the need for as current a reference base as possible for Sec. 50.59
evaluations, in order that the evaluations appropriately consider other
changes already made that may have impacted the facility or procedures.
However, a licensee is not required to submit an update to its FSAR in
the form specified by Sec. 50.71(e) except at the required frequency.
To enhance consistency, the Commission is proposing to revise the rule
language in these sections 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. The
Commission did not use ``Updated FSAR'' for this purpose in order to
take into account two special circumstances: (1) Nonpower reactors, who
are not required to submit updates to the FSAR, although they still
need to consider other changes previously made when performing
Sec. 50.59 evaluations, and (2) a plant licensed to operate, during the
period between initial licensing and the first update. This revision is
reflected in the definitions in the earlier sections and in the
following sections. The definition also refers to ``Final Hazards
Summary Report,'' which is the applicable document for some early
plants whose application was submitted before the regulatory term
``safety analysis report'' was adopted.
The proposed definition is as follows:

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
modified as a result of changes made pursuant to Sec. 50.59 and
Sec. 50.90, and, as applicable, Sec. 50.71 (e) and (f).

F. Probability of Occurrence or Consequences of an Accident or
Malfunction of Equipment Important to Safety Previously Evaluated in
the Safety Analysis Report may be Increased

The current language of the rule states that an unreviewed safety
question exists when the probability of occurrence or consequences of
an accident or malfunction of equipment important to safety previously
evaluated may be increased [emphasis added]. Many of the concerns with
current implementation relate to the appropriate interpretation of the
words ``probability of occurrence . . . or consequences . . . may be
increased.'' In the draft NUREG-1606, the NRC staff stated that the
plain reading of the words would mean that uncertainty about whether
there has been an increase must lead to the conclusion that the
criterion is met. As a result of trying to deal with the question of
uncertainty, licensees were placed in the position of having to prove
there could not be an increase, even when there was no reason to
believe that the proposed change, test or experiment would have that
effect. A similar problem was experienced in considering whether the
possibility of an accident or malfunction of a different type may be
created.
Many of the commenters on the staff's proposed positions viewed
this as overly restrictive and stated that it would result in many
changes requiring prior NRC approval that are below the level of
significance warranting such review. The position espoused in the
revised industry guidance document (NEI 96-07) is that an increase in
probability or consequences must be discernable in order for approval
to be needed. The Commission concludes that the plain reading of the
existing rule language is not consistent with this interpretation.
Although the current rule language would not permit discernable
increases in probability or consequences, the Commission has concluded
that at minimum, this would be a reasonable standard for requiring
prior approval of changes, tests or experiment for increases in
probability of occurrence of an accident or malfunction. The existing
rule language dates from early in the development of reactor
regulation, where with the knowledge base at the time, the then-AEC
found it appropriate to set a very low threshold for changes. Over the
last thirty years, the Commission has garnered experience with
implementation of Sec. 50.59 and insights from probabilistic risk
assessments, both of which indicate that this threshold can be adjusted
without adversely impacting safety. Further, the analytical
capabilities to calculate probabilities have greatly advanced, such
that the effect of even minor changes on probabilities can be
evaluated. Therefore, the Commission proposes to revise existing
paragraph Sec. 50.59(a)(2)(i) of the rule by replacing ``may be
increased'' with ``would result in more than a minimal increase,'' in
order to provide that there must be a clearly discernable change to
require approval, the ``minimal increase'' concept is described in the
next section. As noted above, the (a)(2) paragraph would be broken into
four statements and renumbered as (c)(2)(i) through (iv).

G. More than a Minimal Increase in Probability or Consequences

The Commission notes that Sec. 50.59 permits changes that do not
otherwise require approval (such as would be the case if the provisions
being changed are in TS or license, quality assurance or emergency
plans, or inservice inspection and testing programs). Because the
information being revised is of less immediate importance to public
health and safety, and in consideration of the conservatisms in NRC
design and analysis requirements, acceptance criteria, and the
precision with which safety analyses are performed, ``minimal''
variations in probability of occurrence or consequences of accidents
and malfunctions should not affect the

[[Page 56104]]

basis for the licensing decision. This conclusion is based upon the
qualitative consideration of probability during plant licensing;
accident probabilities were assessed in relative frequencies; equipment
failures were generally postulated to gauge the robustness of the
design, without estimating their likelihood of occurrence. Therefore,
minimal increases in probability could not even have been identifiable,
and could not impact the conclusions reached about acceptability of the
facility design. Radiological consequences for accidents are calculated
and reported at a level of precision such that minimal increases also
would not impact the safety determination. The Commission therefore
concludes 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 revised criteria would also 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 involves a ``significant hazards
considerations;'' any hearing for an amendment constituting a
``significant hazards consideration'' must be completed prior to the
grant of the amendment.) The standard in the proposed rule is
qualitative (probability or consequences no more than minimally
increased). The intent of this proposed rule is to allow changes that
are small enough that they would not affect the facility's licensing
basis, or adversely affect safety performance. While the proposed rule
would allow minimal increases, licensee still must meet applicable
regulatory limits and other acceptance criteria to which they are
committed (such as contained in Regulatory Guides, etc.) Because the
``more than minimal'' standard allows for there to be a discernable
increase, NRC needs to establish a point beyond which one would
conclude that the increase is not minimal. The following guidance is
offered, including values as to when the Commission would conclude that
the revised criteria are not met. Quantitative calculations are not
required except for those instances in which a licensee offers other
than qualitative arguments as part of its evaluation.

Probability of Occurrence of an Accident

The current guidance in NEI 96-07 states: ``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.'' The Commission
believes this satisfies the proposed NRC standard.
In order to be considered as a minimal increase, the resulting
probability (considering the change, test or experiment) must still
satisfy the event frequency classification provided in the licensee's
FSAR (as updated), e.g., for an anticipated operational occurrence
(expected once a year) or for a design basis accident (not expected
during life of plant, but sufficiently credible to require mitigation).

Probability of Equipment Malfunction

The Commission believes that the probability of malfunction is more
than minimally increased if a new failure mode as likely as existing
modes is introduced. The determination should be made either at the
component level, or consistent with the failure modes and effects
analyses, taking into account single failure assumptions, and the level
of the change being made.
Guidance in NEI 96-07 states: ``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.'' The Commission believes this
satisfies this criterion.
The probability of malfunction of equipment important to safety
previously evaluated in the FSAR (as updated) is no more than minimally
increased if ``design bases'' assumptions and requirements are still
satisfied (i.e., the seismic or wind loadings, qualification
specifications, procurement requirements). As part of this guidance,
note that NRC concludes that licensees can treat changes in external
hazard design requirements as potentially affecting equipment
malfunction probability rather than as ``accident probability.''

Consequences of Accident or Malfunction

Guidance in NEI 96-07 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.'' The NRC believes this
satisfies the revised NRC standard.
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, NRC concludes that such a change
satisfies the criteria of no more than a minimal increase in
consequences.
As a quantitative measure, the Commission is considering some
options. One would be to establish that a 0.5 rem increase in
calculated dose as a result of the change be used to assess whether a
minimal increase has occurred. This range of change would generally be
in the decimal place for accident analyses where doses are reported in
rem. The facility must still satisfy applicable acceptance values
(e.g., the SRP) or regulatory requirements (e.g., part 100) for the
particular accident. If a licensee would need to change its design
basis assumptions or analytical methods, or both, to demonstrate that
the change in consequences is less than 0.5 rem, then the NRC does not
view the change as minimal and would expect the licensee to submit a
license amendment for such a change.
In addition, the Commission is considering a graduated approach,
consistent with the concept of ``minimal'' being small enough so as not
to impact the basis for acceptability. When the facility is far from
the limit, a larger increase can be accommodated without concern about
impact on the basis for acceptability. The values proposed take into
account such factors as differences between licensee calculated values
and staff estimation of existing performance, potential for a single
change with a large increase, or for several ``minimal'' increases to
approach the regulatory limits. The specific proposal offered for
comment is:

[[Page 56105]]

Example using 300 rem thyroid dose as the limit.

----------------------------------------------------------------------------------------------------------------
Existing calculated dose ``Minimal'' change Pre-change After the change
----------------------------------------------------------------------------------------------------------------
<50% of="" limit.......................="">10% increase..... 140 rem............... 170 rem.
80% of limit............. 5% increase...... 205 rem............... 220 rem.
more than 80%....................... 1% increase (NTE 245 rem............... 248 rem.
limit).
----------------------------------------------------------------------------------------------------------------

A third option under consideration, similar to option 2, would
limit the fraction of remaining margin that can be consumed by a
particular change. By defining ``minimal'' as being 10% of the
remaining margin between current conditions and acceptance guidelines,
the amount of change would decrease as the limit is approached, and the
limit could not be exceeded.

Cumulative Effect

The Commission is concerned about the cumulative effect of minimal
increases. Since some increases are allowed, the Commission believes
that the proposed process 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 then becomes how the NRC can best oversee the process
such that several ``minimal'' changes do not result in unacceptable
results. The Commission has decided to require licensees to report
effects of changes in a different manner to facilitate evaluation of
cumulative effect, as discussed in a later section on reporting
requirements, in which the Commission proposes to require that the SAR
update in accordance with Sec. 50.71(e) discuss the effects of the
changes upon calculated doses and other information.

H. Possibility of an Accident of a Different Type from any Previously
Evaluated in the Safety Analysis Report may be Created

As noted in Section F above, the uncertainty connected with
demonstrating that no accident or malfunction may have been created is
a major source of confusion and difficulty in implementing the existing
rule; and is unnecessary for purposes of identifying when NRC review of
a change is needed. Accordingly, the Commission proposes that the
language in existing Sec. 50.59(a)(2)(ii) be revised as discussed below
in this section and the following one. As noted earlier, the Commission
is proposing to separate the requirements into distinct criteria for
clarity. This criterion would now read ``if a possibility for an
accident of a different type from any previously evaluated in the final
safety analysis report (as updated) is created.'' Under the proposed
rule, a license amendment would be needed only if the licensee
reasonably concluded that the possibility of an accident of a different
type is created. This contrasts with the current rule, which would
require a license amendment if the licensee is uncertain or unable to
reasonably conclude that a new accident of a different type is not
created. The Commission concludes that this proposed rule change will
still identify those proposed changes, tests, or experiments that the
NRC should review, without also including other changes of lesser
significance that may be viewed as meeting the existing criteria.

Need for Definition of Accident

In determining whether a proposed change requires prior NRC
approval under Sec. 50.59, the rule refers to whether ``accidents''
previously evaluated in the SAR are impacted, or whether an accident of
a different type may be created (see also Sec. 50.92 criteria for ``no
significant hazards consideration)''. Those accidents evaluated in the
SAR, that is, those events that a plant must show that it can
withstand, are derived from a number of regulatory requirements, and
the safety analyses are included in the FSAR.
The regulations and NRC guidance documents, refer to ``a design
basis accident'' (Sec. 50.36), to design basis events (Sec. 50.49), to
loss-of-coolant accidents (Appendix A), to anticipated operational
occurrences (Appendix A) and to accidents that could result in release
of significant quantities of radioactive fission products (part 100).
The PSAR, and by extension the FSAR, pursuant to Sec. 50.34, is to
contain ``analysis and evaluation of the design and performance of SSC
of the facility with the objective of assessing the risk to public
health and safety resulting from operation of the facility and
including determination of (i) the margins of safety during normal
operations and transient conditions anticipated during the life of the
facility and (ii) the adequacy of SSC provided for the prevention of
accidents and the mitigation of the consequences of accidents.'' RG
1.70 states that the FSAR is to include postulated anticipated
operational occurrences; postulated off-design transients that induce
fuel failures above those expected for normal operational experience,
and design basis accidents. The Standard Review Plan for Chapter 15,
refers to anticipated operational occurrences and to postulated
accidents, and also to ``transients and accidents'' (the SRP notes that
other events, such as response to external phenomena, are covered in
other chapters).
Design basis accident(s) has been used in regulatory practice both
singularly and generally. The regulations also include the concept of a
design basis accident (DBA), for purposes of evaluating siting, which
is an assumed fission product release, based upon a major accident that
would result in potential hazards not exceeded by those from any
accident considered credible. Such accidents have generally been
assumed to result in substantial meltdown of the core with subsequent
release of appreciable quantities of fission products. The set of
``accidents'' that a plant must postulate for purposes of FSAR design
and safety analyses, including LOCA, other pipe ruptures, rod ejection,
etc., are often referred to as ``design basis accidents''.
The terms of accidents and transients are often used in regulatory
documents (as for example in Chapter 15 of the Standard Review Plan),
where transients are viewed as the more likely, low consequence events
and accidents as more serious. In the context of probabilistic risk
assessment, transients are typically viewed as initiating events, and
accidents as the sequences that result from various combinations of
plant and safety system response.
However, the meaning of the term ``accident'' as it is used more
generally in Part 50, is somewhat obscured by the

[[Page 56106]]

use of the term ``design basis event.'' In Sec. 50.49, design basis
event is defined as:

normal operations including anticipated operational occurrences,
design basis accidents, external events, natural phenomena
(earthquakes, tornados, hurricanes, floods, tsunami and seiches),
for which the plant must be designed to ensure safety-related
functions.

In view of the range of language presently used to describe the
types of events evaluated as part of the licensing basis, the
Commission is contemplating the need to clarify its intent as to the
extent of events that are within the purview of the criteria in
Sec. 50.59 and in Sec. 72.48). For purposes of stimulating discussion,
the Commission offers two proposals. One would be to set forth a
definition for the term ``accident'' as follows:

an initiating event or combination of events and/or conditions that
could occur from equipment failure, human error, natural or manmade
hazards which challenges the integrity of one or more fission
product barriers (fuel, reactor coolant system, release of
radionuclides (confinement/containment)), required to be analyzed
and/or accounted for by the Commission and addressed in the
licensee's safety analysis report.

Such a definition would make it clear that the Commission's intent
in referring to ``accidents'' in Sec. 50.59 (and in Sec. 72.48) is to
refer to the design basis accidents that are addressed in the SAR. The
second approach is to add the phrase ``design basis accident'' into the
existing criteria. This could be done for each of the three criteria
that refer to ``accident'' or just for the one on accident of a
different type. Since the criteria on probability and consequences also
contain language about ``previously evaluated in the SAR,'' there may
be less need for a reference to ``design basis accident'' in these
criteria. The proposed rule language includes use of the phrase
``design basis accident'' in the one criterion, for purposes of
obtaining public comment.

I. Possibility of a Malfunction of a Different Type from any Previously
Evaluated in the Safety Analysis Report may be Created

In a similar fashion, the Commission proposes to modify the
remaining part of existing Sec. 50.59(a)(2)(ii), concerning
malfunctions of a different type by creating a new criterion that would
read ``if 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) is created.'' This criterion
involves three revisions to the existing rule. The first change is the
use of the phrase ``is created'' which would require a determination
that the possibility has been created, rather than uncertainty as to
exclusion.
The second change is to insert the words ``of equipment important
to safety.'' The existing rule does not provide this characterization
within paragraph (ii), but it is included in paragraph (i). It has
generally been inferred that the statement in paragraph (ii) is an
abbreviated version of that in paragraph (i). A review of the history
of the 1968 rulemaking adopting revisions to Sec. 50.59 did not
disclose any discussion suggesting that the Commission intended to
distinguish between the (a)(2)(i) and the (a)(2)(ii) criteria with
respect to the scope of equipment covered. Therefore, the Commission
concludes that the rule was intended to apply to the same scope of
equipment in each cases, and therefore, proposes to include the words
in this criterion to eliminate any doubt.
The final change is being proposed in response to the comments on
the staff-proposed guidance (NUREG-1606) on the interpretation of
malfunction (of equipment important to safety) of a different type. The
commenters believe that the cause of the malfunction should be a
consideration in determining whether the probability of the malfunction
may have increased, and that a malfunction of a different type would
only be created if the effects of the malfunction are not already
bounded by the FSAR analysis. The recent industry guidance states that
if a component were subject to failure from a new failure mode but the
failure of the component is already considered in the safety analysis,
then there would not be a failure of a different type. The Commission
does not agree that the industry interpretation is consistent with the
rule as written, which refers to creation or possibility of a
malfunction of a different type, not of a different result. However,
the Commission recognizes that in its reviews, equipment malfunctions
are generally postulated as potential single failures to evaluate plant
performance; thus, the focus of the NRC review was on the result,
rather than the cause/type of malfunction. Unless the equipment would
fail in a way not already evaluated in the safety analysis, there is no
need for NRC review of the change that led to the new type of
malfunction. Therefore, as the third change in Sec. 50.59(a)(2)(ii),
the Commission is proposing to change the phrase ``of a different
type'' to ``with a different result''. Therefore, this criterion would
read: ``if a possibility for a malfunction of equipment important to
safety with a different result . . . is created.''
In implementing this position, attention must be given to whether
the malfunction is evaluated at the component level or the overall
system level. While the evaluation should take into account the level
that was previously evaluated in terms of malfunctions and resulting
event initiators or mitigation impacts, it also needs to consider the
nature of the change. Thus for instance, if failures were previously
postulated on a train level because the trains were independent, a
change that introduces a cross-tie might need to be evaluated to see
whether new outcomes have been introduced. The staff has provided
guidance on this issue in Generic Letter (GL) 95-02, concerning
replacement of analog systems with digital instrumentation. The GL
states that in considering whether new types of failures are created,
this must be done at the level of equipment being replaced--not at the
overall system level. Further, it is not sufficient for a licensee to
state that since failure of a system or train was postulated in the
SAR, any other equipment failure is bounded by this assumption, unless
there is some assurance that the mode of failure can be detected and
that there are no consequential effects (electrical interference,
materials interactions, etc), such that it can be reasonably concluded
that the SAR analysis was truly bounding and applicable. Otherwise, the
Commission would conclude that there was increase in probability of
malfunction or that a malfunction with a different result has been
created.

J. Margin of Safety as Defined in the Basis for any Technical
Specification is Reduced

Two criteria in the current regulations (Sec. 50.59) specifically
focus upon accidents and equipment malfunction (creation, consequences
and likelihood) as the measures for determining when a change requires
prior NRC approval. However, the phrases ``margin of safety'' and ``as
defined in the basis for any technical specification'' in the third
criterion have been the subject of differing interpretations because
the rule 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. In addition, some have questioned the need for the third
criterion on ``margin of safety.''
The Commission has under consideration a number of proposals on
margin. In the proposed rule text specifically being offered for
comment, one option has been inserted so that commenters can examine
the

[[Page 56107]]

relationship of this aspect of the proposed rule to other changes being
offered. This should not be viewed as meaning that this option is
preferred by the Commission. The range of options under consideration
is discussed in more detail below.
Questions of margin are commonly judged in terms of the degree of
confidence that the response of the facility, or of particular SSC, to
postulated challenges is acceptable. Various margins exist in a
facility design. These margins are based on, for example, assumptions
of initial conditions, conservatisms in computer modeling and codes,
allowance for instrument drift and system response time, redundancy and
independence of components in safety trains, and plant response during
operating transient and accident conditions. Margin to conditions that
might be detrimental to safety is also determined by establishing
acceptance criteria to be met for response to various accidents and
transients. Acceptance criteria are established at a value that
accounts for uncertainty about physical properties and other
variability and thus provides margin to unacceptable plant conditions.
Margins are built into the facility to account for routine plant
fluctuations and transients. Margins are also built into the plant to
establish the regulatory envelope within which a plant has demonstrated
its ability to respond to a spectrum of design basis accidents. It is
in this category termed the ``regulatory envelope,'' that the NRC
believes that regulatory oversight of changes in margin may be needed
from the standpoint of Sec. 50.59. Thus the Commission notes that not
all margins fall within the purview in which changes to the margin
require prior NRC approval. As part of this rulemaking, the Commission
wants to clarify which margins fall within the regulatory envelope and
how possible reductions in margin resulting from facility or procedure
changes, or from conduct of tests and experiments should be evaluated.
In defining in the rule a standard for NRC review and approval of
changes to margins in the regulatory envelope, the Commission may want
to preserve the NRC's ability to review changes when there is a
potentially significant reduction in a margin of safety,\4\ but clearly
would not want to unduly affect licensee operations. Therefore, for
this proposed rulemaking, the Commission is offering the public the
opportunity to comment on a range of options for treating margin.
Commenters are requested to present opinions about the merits, or
concerns about the specific proposals, or both, and also to offer any
other suggestions for wording.
---------------------------------------------------------------------------

\4\ In accordance with 10 CFR 50.92(c)(3), license amendments
involving a significant reduction in a margin of safety do not meet
the criteria for a ``no significant hazards consideration''
determination; thus, changes involving a significant reduction in a
margin of safety are not to be performed under 10 CFR 50.59.
---------------------------------------------------------------------------

Option 1: Control Inputs to Analyses and Methods that Establish TS

The Commission believes it is reasonable to interpret the specific
reference to ``basis for any technical specification'' in the 1968
rulemaking that added the ``margin of safety'' criterion as preserving
the margins in the analyses that established the TS requirements. For
instance, the minimum plant performance conditions and configurations
stated in the TS are the limiting conditions for operation, limiting
safety system settings, and safety limits. Margins of safety exist
within the safety analyses as a result of the specific input
assumptions, methods, or other limits that were used. These parameters
and methods were proposed by the licensee and reviewed by NRC to
account for uncertainties, instrumentation response, and ranges of
possible operating conditions. Because Sec. 50.59 requires prior NRC
approval for a change to the TS, a change that could invalidate the
basis upon which the TS values were established should also receive
prior approval. In accordance with this interpretation, changes that
invalidate these specific conditions described in the FSAR for analyses
that established the TS requirement (such as a limiting condition of
operation, or a limiting safety system setting) would reduce the margin
of safety associated with the 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 would propose to add a
definition for ``reduction in margin of safety associated with any
technical specification'' and to conform the criterion for needing a
license amendment in new Sec. 50.59(c)(2). The existing terminology of
``basis for any TS'' would be replaced by ``associated with any TS.''
The following definition would be added:

Reduction in margin of safety associated with any technical
specification means that the input assumptions, analytical methods,
acceptance conditions, criteria and limits of the safety analyses,
presented in the final safety analysis report (as updated), that
established any technical specification requirement, are altered in
a nonconservative manner.

Although this option would maintain the safety analyses that
underlie the TS, this approach would also have the effect of giving
input values and assumptions the weight of TS, which is inconsistent
with the philosophy in Sec. 50.36 of establishing TS only on those
values of most immediate safety importance. In many instances, changes
to inputs can be accommodated by other available margins so that the
licensing envelope is preserved.

Option 2: Delete ``margin of safety'' as a Criterion.

Under this option, the Commission would delete any criterion
focusing upon margins. 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. 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 seeks comment on whether any of
the other evaluation criteria should be revised were this approach to
be adopted.

Option 3: Control margins associated with results of analyses

Instead of focusing on the inputs to safety analyses, another
interpretation would be to examine the results of the safety analyses,
and to determine whether changes to operational characteristics or
other information described in the FSAR (as updated) would reduce the
level of protection afforded by the TS (i.e., by the limiting safety
system settings and limiting conditions of operation), as reflected in
the results of safety analyses.
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
(fuel cladding, reactor coolant system boundary and containment).
Satisfying these acceptance criteria (or regulatory limits) 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. For

[[Page 56108]]

certain of these parameters, TS safety limits have been established;
these safety limits are limits upon important process variables that
are found necessary to reasonably protect the integrity of physical
barriers that guard against the uncontrolled release of radioactivity.
However, for other parameters, a licensee must determine the
licensing basis of the parameter in question by reviewing the plant-
specific safety analyses. The acceptance criterion is that value
approved by the NRC for a particular parameter or process variable
(e.g., ASME Code stress limits, a departure from nucleate boiling ratio
limit or maximum critical power ratio limit or containment design
pressure). These acceptance criteria may be stated in the FSAR, may be
in NRC regulations, or may be presented in the NRC Standard Review
Plan. (Note: This approach may require some licensees to revise their
FSAR to accurately describe the regulatory values for the set of
critical parameters. For example, licensees would need to identify the
expected operating or design values and then specify the minimum
performance capabilities for the related parameters, which cannot be
modified with NRC review).
In constructing the requirements for controlling margin through
consideration of results of analyses, there are three aspects to take
into account: (a) Which results/parameters are to be controlled through
the Sec. 50.59 process, (b) the degree of change to be allowed without
review, and (c) how the changes should be evaluated in demonstrating
that the criterion is satisfied.
In the sections below, these three aspects are separately discussed
in order to amplify upon the issues under consideration. However, any
rule language option would need to include some provision for each of
the three aspects.
(a) Which parameters should be controlled?
The margins of safety that would be controlled by the 10 CFR 50.59
process can be characterized in different ways.

OPTION 3(A)(1)--Safety and Regulatory Limits

The margin between regulatory limits and the failure of physical
barriers is protected in the regulations (and also in the portion of
the Technical Specifications (TSs) called ``safety limits''). The
margin, as reflected in approved safety and accident analyses, between
the protection afforded by the TSs (e.g., the limiting safety system
settings and limiting conditions of operations) and the associated
regulatory limits is a possible interpretation as to ``the margin of
safety as defined in the basis for any TS'', which would be subject to
the 10 CFR 50.59 evaluation process. Thus, one proposal under
consideration would be to define ``margin of safety'' as follows:

With this option, before changing operational characteristics
described in the UFSAR (not directly controlled by TS), a safety
evaluation must be performed to determine, among other things, if the
change results in a reduction in the level of protection afforded by
the TS (margin of safety as defined in any TS). Such a reduction would
typically occur only if the operational characteristic had been used as
a bounding condition in the analysis upon which the selection of TS was
based, or in analysis where the acceptability of selected TS values was
demonstrated. Licensees could make desired changes to operational
characteristics without prior NRC approval, provided that the change
does not result in accident analysis results that are nearer the
regulatory, or safety, limits than the corresponding results that the
NRC used in evaluating the acceptability of the TS during licensing of
the facility.

OPTION 3(A)(2)--Fission product barriers--definition

The NRC notes that Sec. 50.36 (requirements for Technical
Specifications) has criteria for when TS are to be provided that
specifically are tied to design basis accident or transient analysis
that either assumes the failure of or presents a challenge to the
integrity of a fission product barrier. Thus, the margin as defined in
the basis for any TS can be reasonably viewed as that margin associated
with preserving integrity of these barriers. Therefore, the NRC is also
considering a more explicit linkage to the response of the three
fission product barriers generally relied upon to provide protection
from uncontrolled release of radioactive materials from a reactor
facility. Under such a proposal, the text of the rule would explicitly
state that it is the response of fission product barriers (fuel,
reactor coolant system, and containment) to accidents, transients, and
malfunctions that is being controlled.
The following could be given as a definition of margin of safety
and of fission product barrier response. Regulatory guidance would
explicitly list the parameters (for PWRs and BWRs) that are to be
controlled.

The margin of safety for any fission product barrier response is
the difference between the calculated value and its associated
acceptance criteria. Fission product barrier response means those
parameters that must be satisfied in the event of postulated design
basis events to demonstrate integrity of the fuel, reactor coolant
system and containment system barriers.

The following parameters would be included: Fuel and cladding
performance (peak cladding temperature, or energy deposition, DNBR or
MCPR, oxidation), RCS performance (pressure, flows, stress), and
containment performance (peak pressure, containment leakage).

OPTION 3(A)(3)--Specified Parameters

A variant on the previous option would be to actually list the
parameters of interest directly in the criterion for prior review, as
for instance, the criterion could read:

(vii) Result in a change to the FSAR (as updated) calculated
value of RCS peak pressure, containment peak pressure, or fuel
performance (DNBR/MCPR, others), etc.

This variant has the advantage of being more precise, but the rule
language would need to be crafted to account for various reactor types.

OPTION 3(A)(4)--Include Mitigation Capability

The Commission is interested in preserving the integrity of both
prevention and mitigation capabilities available in the plant, and is
therefore considering an option that would include both features within
the ``margin'' criterion if the margin criterion is maintained. If this
approach were adopted, the definition or the list of parameters would
be supplemented with the performance parameters for the

[[Page 56109]]

accident mitigation capability of the plant, as for instance, ECCS
performance (pressures, flows, actuation values), engineered safety
feature performance (flows, pressures, spray effectiveness, system
efficiencies).
Finally, in conjunction with any of these approaches, the
Commission is also considering whether there are other parameters
important to preservation of barriers that should be explicitly
defined. For instance, for fuel stored in spent fuel pools, or for the
reactor during periods of shutdown or refueling, there may be other
analysis results (water level, pool temperature) in lieu of reactor
coolant system pressure. Therefore, the Commission seeks input as to
whether there are other parameters of interest beyond those previously
offered that should be included within the ``margin of safety''
criterion if that criterion is maintained, and how should the rule
language be revised to specify what those parameters might be.
(b) Determination of reduction in margin requiring review
Once the parameters of interest are determined, it is also
necessary to define when a reduction in margin warranting NRC review
and approval has occurred. The Commission is evaluating options ranging
from any ``nonconservative change in calculated values,'' to a
``minimal change'' standard, and ultimately an option that would allow
increases up to ``specified limits (acceptance criteria)'' for those
parameters that may be established in the regulations or NRC guidance
(such approaches to the limits might be controlled in a graduated
fashion as was discussed in the section of this notice relating to
``minimal increases''). An option for the degree of reduction would be
paired with an option (such as one of those listed in (a) above) to
provide the text of the rule.

OPTION 3(B)(1)--No Reduction

One approach would be require that the safety analysis, considering
the effect of the change, must show that the accident analysis results
are not nearer to any safety or regulatory limit, thus, a ``no
reduction in margin'' standard. Possible rule text:

Changes, or the net effect of multiple changes, which result in
a reduction in the margin of safety require prior NRC approval.
Changes, or the net effect of multiple changes, which do not cause a
reduction in the margin of safety do not require prior NRC approval.

OPTION 3(B)(2)--Minimal Amount--Definition of Margin Reduction

As discussed in other sections of this notice, the Commission
concludes that the revised rule should allow licensees some flexibility
in making changes, through development of a ``minimal increase''
standard. In considering margins, the Commission is thus weighing how
such a concept could be applied. One option would be that NRC approval
would be required for a change, test, or experiment if the output
values (calculated in the SAR) are altered by more than a minimal
amount. The ``margin'' criterion would be modified to state that a
change in calculated result of ``more than a minimal amount'' would
require prior review and approval. Either in the rule itself, or in
guidance, the Commission would define ``minimal amount'', modeled upon
the options offered for minimal increases in consequences (see section
II.G. of this notice). For example, there could be a fixed amount
(percent change) in margin, as long as regulatory limits are still met.
If guidance itemizes the parameters, such guidance could also customize
how ``minimal'' should be judged for each particular parameter
(allowing greater amounts for certain parameters depending on precision
of calculations, sensitivity of results and other considerations).
For instance, the definition of ``margin of safety reduction * *
*'' might be stated as follows:

Reduction in margin of safety means that as a result of a
change, the [MARGIN] is altered in a nonconservative manner by more
than a minimal amount.

OPTION 3(B)(3)--Minimal Determined With Respect to Acceptance Criteria
(Available Margin)

It is also possible to achieve this result by removing the language
referring to margin of safety (and to TS), and defining ``minimal'' in
the rule itself in terms of the results or analyses for barrier
response, with respect to meeting the acceptance criteria for those
barriers. For example, rule language could read as follows:
License amendment needed if as a result of a change, test or
experiment:

(vii) there is more than a 10% reduction in the difference
between the calculated value and the acceptance criteria for fission
product barrier response to accidents evaluated in the SAR.

If such an approach is followed, the Commission would propose to
include a definition of acceptance criteria, such as follows:

Acceptance criteria are those values, established by NRC
regulation or review guidance, to which the licensee is committed
through its FSAR (as updated), as the basis for acceptability of
response to the postulated accident, transient or malfunction.

(c) Evaluation of effect of the change upon analysis results.
The Commission also notes that the results of safety analyses are
subject to variance depending upon the assumptions, analysis methods or
analytical techniques used. In many instances, these factors were
reviewed by the NRC during its licensing deliberations, and their use
may have formed part of the basis for the conclusion that acceptable
safety margins were demonstrated. Therefore, the Commission wishes to
ensure that proposed changes by a licensee would not invalidate these
conclusions by requiring a demonstration that the evaluation techniques
and analyses are suitable.
To accomplish this, the Commission is considering having as part of
whichever definition of ``margin of safety reduction'' is selected the
following statement [Option 3(c)]:

All analyses and evaluations for assessing the impacts of
proposed 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.

The alternative to this proposed language would be to rely upon a
licensee's design control processes under their quality assurance
requirements and program, to provide the assurance that any evaluative
work has been conducted with methods and techniques commensurate with
the safety significance of the analyses being performed.

Impacts for Part 72 Changes

Certain of the options discussed above may need to be modified for
application to independent spent fuel storage facilities or spent fuel
storage cask designs in Part 72. While the overall philosophy would be
the same, the particular outputs or barriers that would be specified
for reductions in margin would have to be defined in terms of the
barriers against release of radioactivity afforded by fuel storage
facilities. For instance, these might include calculated fuel
temperature or cladding oxidation, and stresses (or pressures) on the
cask structure. Comment is also requested on the appropriate parameters
for facilities licensed under Part 72.

K. Safety Evaluation

Section 50.59(b)(1) requires licensees to maintain records that
must include a written safety evaluation that provides

[[Page 56110]]

the bases for the determination that the change, test, or experiment
does not involve an unreviewed safety question. Section 50.59(b)(2)
requires submittal of a report containing a brief description of any
changes, tests, or experiment, including a summary of the safety
evaluation of each. In the interest of emphasizing the regulatory
purpose of the evaluation required under Sec. 50.59, which led the
Commission to propose deletion of the term ``unreviewed safety
question,'' the Commission proposes to delete the word ``safety'' in
referring to the required evaluation for determining whether the
change, test, or experiment requires a license amendment. For purposes
of the summary report of tests and experiments submitted to NRC, the
staff would propose that the rule specify that a summary of the
evaluation be provided (rather than a summary of the safety
evaluation).
A similar change is 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
proposes to change ``safety evaluation in support of license
amendments'' to ``safety analysis in support of license amendments,''
to reduce confusion between the information prepared by the licensee
for the amendment (safety analysis) and the NRC review (safety
evaluation). The second part of this 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.'' (In this case, it is a
licensee evaluation against the regulatory criteria in Sec. 50.59 that
is being referred to). In addition, other minor wording changes are
proposed such as with respect to terminology on ``final safety analysis
report'' and ``effects of'' (see reporting requirements discussion
below). Conforming changes in the appendices to part 52 and in part 72
to revise language to refer to ``evaluation'' are also proposed.

L. Reporting and Recordkeeping Requirements

In view of the ``minimal increase'' criteria in Sec. 50.59, the
Commission concludes that the reporting requirements for the SAR update
should be enhanced to enable the NRC to better understand the potential
cumulative impact of changes that might have been made since the last
update. Therefore, the Commission proposes to supplement the reporting
requirements 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. Accordingly, the Commission proposes to revise Sec. 50.71(e) to
read as follows:

``(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. The submittal must
describe the effects \1\ of: (1) All changes made in the facility or
procedures as described in the FSAR; (2) all safety analyses and
evaluations performed by the licensee either in support of requested
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; (3) all analyses of new safety issues performed by or
on behalf of the licensee at Commission request; and (4) the net
effect of all changes made since the last update on the safety
analyses, including probabilities, consequences, calculated values,
system or component performance, that are in the FSAR (as updated).
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.
---------------------------------------------------------------------------

Finally, the Commission is proposing a change to the record
retention requirements in existing Sec. 50.59 (b)(3) (renumbered by
this rulemaking to (c)(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 statement ``or until the termination of
a license issued pursuant to 10 CFR part 54, whichever is later.'' This
change would make more clear the requirement that records must be
maintained through the life of the facility so that they will remain
available until such time as they are no longer needed (that is, when
the license is terminated, not just at the end of the initial licensing
term).

M. Part 72 Changes

In part 72 the Commission is proposing to make conforming changes
to Sec. 72.48 with those made to Sec. 50.59 and to expand the scope of
Sec. 72.48 so that holders of a Certificate of Compliance (CoC) are
also subject to it. In addition to the proposed changes to Sec. 72.48,
the Commission proposes to make changes in other sections of part 72.
When subpart L--Approval of Spent Fuel Storage Casks, was originally
added to part 72, no provisions were included to address potential
amendments of CoCs. However, regulations in this area are necessary to
provide requirements for certificate holders in instances where a
proposed change does not meet the tests of Sec. 72.48, and an amendment
to the CoC is necessary. Therefore Secs. 72.244 and 72.246 would be
added to subpart L, to provide regulations on applying for, and
approving, amendments to CoCs. Section 72.248 would also be added to
provide regulations for the certificate holder submitting an updated
final safety analysis report, which would document the changes it made
to procedures or structures, systems, and components under the
provisions of Sec. 72.48. The Commission notes that a general licensee
is not precluded from loading spent fuel into an approved spent fuel
storage cask during the 90-day period allowed for the certificate
holder to submit a final safety analysis report. This approach is the
same as that required for part 72 license holders to update their final
safety analysis report under Sec. 72.70. The Commission also notes,
that for dual-purpose spent fuel casks (i.e., casks which have been
issued CoCs for transportation and storage under parts 71 and 72,
respectively), no regulation equivalent to Sec. 72.48 exists in part
71. Consequently, a certificate holder could make changes to the design
of a spent fuel storage cask under the authority of Sec. 72.48 (i.e.,
without prior NRC approval); however, if the change also affected the
transportation aspects of the cask's design and involved a modification
to the part 71 certificate, then NRC approval and amendment of the
transportation CoC would be required before the cask could be used to
transport spent fuel to another site. Additionally, a transportation
cask CoC has a term of 5 years, compared to the 20-year term for a
storage CoC. Consequently, the Commission envisions that most of this
type of change would be captured during the periodic renewal of a
transportation CoC and this delay would not have a significant adverse
impact on a licensee's ability to transport spent fuel in a dual
purpose cask.
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. The existing
requirement in Sec. 72.24(a)--Contents of application: Technical
Information,

[[Page 56111]]

would be revised to reference shared common utilities and services in
the applicant's assessment of potential interactions between the ISFSI
and another facility. The Commission would remove the existing
requirement in Sec. 72.3 for the applicant to evaluate the impact of
sharing common utilities and services on the ``other facility.'' The
Commission believes that evaluation of the impact on the ``other
facility'' should not be part of the licensing process for an ISFSI.
Rather, such evaluation should be part of the license amendment process
for that ``other facility'' and should be performed under the
regulations used to license that ``other facility.''
Changes to Sec. 72.56 would be conforming changes to those made to
Sec. 50.90. Changes to Sec. 72.70 are also conforming changes to those
made to Sec. 50.71(e); additionally, requirements would be added to
Sec. 72.70 on standards for submitting revised Final Safety Analysis
Report (FSAR) pages. The Commission notes that the proposed Sec. 72.70
would retain the requirement that the site-specific licensee submit a
final safety analysis report at least 90 days prior to the planned
receipt of spent fuel or high-level waste. The Commission has not
received any requests for exemption from this regulation and believes
that this regulation does not impose an undue burden or schedule impact
on licensees. The proposed rule also modifies the requirements for
filing of updates (through reference to Sec. 72.4) to be consistent
with other changes being made to part 72. Changes to Sec. 72.216 for a
general licensee are similar to the changes made to Sec. 72.70 for a
site-specific licensee and are also conforming changes to those made to
Sec. 50.71(e). The Commission also envisions that a general licensee
who wishes to adopt a change to the design of a spent fuel storage cask
it possesses--which 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. The
changes to Secs. 72.9 and 72.86 are conforming changes due to the
addition of new Secs. 72.244, 72.246, and 72.248.
Changes to part 72 Record keeping requirements would include the
clarification that records required by Sec. 72.48 shall also include
determinations that significant increases in occupational exposure or
unreviewed environmental impacts did not exist, such that a license
amendment would have been required. (The existing language linked the
written evaluation only to the ``unreviewed safety question''
determination, and thus did not explicitly require Record keeping for
the determinations of whether the change would cause a significant
increase in occupational exposure or a significant unreviewed
environmental impact). Certificate holders would also be required to
keep records of such changes as would be allowed under Sec. 72.48.
Requirements in Sec. 72.70 would be established for reporting
changes to procedures. The Commission notes that Sec. 72.70 presently
requires that the update include ⁵ a description and
analysis of changes in the structures, systems, and components with
emphasis upon performance requirements; the bases, with technical
justification therefor, upon which such requirements are based; and
evaluations showing that safety functions will be accomplished. It also
requires an analysis of the significance of any changes to codes,
standards, regulations, or regulatory guides which the licensee has
committed to meeting the requirements of which are applicable to the
design, construction, or operation of the facility. 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 reporting requirements for
general licensees would be added as Sec. 72.216(d), similar to existing
reporting requirements for site-specific licensees in Sec. 72.70 and
proposed requirements for certificate holders in Sec. 72.248. In both
of these sections, the Commission is adding a requirement that the
entity making a change to the cask, either the general licensee or the
certificate holder, provide a copy of the submittal to the other party
for their information.
---------------------------------------------------------------------------

\5\ The similarity in the language between Secs. 72.24 and
50.34(a) and between Secs. 72.70 and 50.34(b)(2) is noteworthy.
---------------------------------------------------------------------------

III. Section By Section Analysis

10 CFR Part 50

10 CFR 50.59

As discussed in more detail above, Sec. 50.59 would be restructured
and revised to have the following components.
Paragraph (a)--This is a new paragraph that provides definitions of
terms such as ``change'', ``facility as described * * *,'' in order to
specify more clearly which changes, tests and experiments require
further evaluation and how reductions in margin of safety are to be
determined. The references to ``safety analysis report'' are being
revised to ``final safety analysis report (as updated)'' to state that
the evaluations are to be performed that take into account other
changes made that have affected the final safety analysis report since
its original submittal.
Paragraph (b)--Relocation of existing applicability provisions.
Paragraph (c)(1)--Relocation of existing provisions establishing
which changes, tests, or experiments require evaluation, using the
defined terms. The terminology of ``unreviewed safety question'' has
been replaced by referring to the need to obtain a license amendment.
This paragraph also clarifies that the licensee must submit its request
for license amendment, and obtain the amendment prior to implementing
those changes, tests or experiments that involve TS or otherwise meet
the criteria for prior NRC approval as specified in (new) paragraph
(c)(2).
Paragraph (c)(2)--Reformatting of the evaluation requirements into
seven distinct statements of the criteria and revision of the criteria
for when prior NRC approval of a change, test or experiment is
required. Specifically, language of ``more than a minimal increase''
was inserted in the criteria concerning increases in probability and
consequences, and revisions to the rule requirements were made
concerning creation of accidents of a different type and malfunctions
of equipment with a different result. Clarification is also being
provided that the margins of safety are those associated with TS
requirements established by the FSAR analyses, and are not confined to
the BASES section of the TS. 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 (d)(1)--Renumbered paragraph with record keeping
requirements. Also includes change from ``safety evaluation'' to
``evaluation.''
Paragraph (d)(2)--Renumbered paragraph with reporting requirements.
Paragraph (d)(3)--Renumbered and revised paragraph on retention of
records, to cover the term of any renewed license.

10 CFR 50.66

The proposed changes for Sec. 50.66 are to conform existing
language referring to unreviewed safety questions, and references to
updated final safety analysis report, to the language

[[Page 56112]]

proposed in revised Sec. 50.59 for consistency.

10 CFR 50.71(e)

The proposed changes to this section are to conform language with
respect to unreviewed safety question, safety evaluation, and reference
to final safety analysis report (as updated), with the proposed
language in Sec. 50.59, and to clarify reporting requirements relating
to ``effects of'' changes such that cumulative effects of minimal
increases in probability and consequences are included in the update to
the FSAR.

10 CFR 50.90

A portion of existing Sec. 50.59(c) would be relocated into this
section. This change would place the requirements for changes to
technical specifications in the rule section on amendments to licenses.

10 CFR Part 52

Appendix A and Appendix B to 10 CFR Part 52

The proposed changes to these sections are to conform references to
unreviewed safety question, safety evaluation and the evaluation
criteria concerning when prior NRC approval is needed, to the language
in the proposed revision to Sec. 50.59.

10 CFR Part 72

10 CFR 72.3

The definition for independent spent fuel storage installation
would be 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 also proposed to be
revised to include this evaluation).

10 CFR 72.9

Paragraph (b) would be revised as a conforming change to include in
the list of information collection requirements the new reporting
requirements in Secs. 72.244 and 72.248 for reports of changes made by
CoC holders and for updates to the safety analysis reports by CoC
holders.

10 CFR 72.24

This section would be 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

New definitions have been added for terms such as ``change'' and
``facility as described in the Final Safety Analysis Report (as
updated).'' The specific criteria in existing paragraph (a)(2) have
been revised to separate out the various statements, to insert the
language of ``more than a minimal increase,'' and to modify the
criterion from ``malfunction of a different type'' to ``malfunction of
a different result.'' The text for Record keeping requirements was
revised to refer to the need for license or certificate of compliance
(CoC) amendments, rather than involving an unreviewed safety question.
As part of this revision, the Commission is also clarifying that the
records shall also provide a basis for why a proposed change, test, or
experiment did not require a license or CoC amendment with respect to
significant increases in occupational exposure or significant
unreviewed environmental impacts. Additionally, the term ``Final Safety
Analysis Report (FSAR) (as updated)'' has been used to provide greater
clarity and consistency with Sec. 50.59 and other sections of Part 72.
The filing requirements for the summary reports are modified to be
consistent with Sec. 72.4 (Communications).

10 CFR 72.56

Existing Sec. 72.48 (c)(2) is being relocated into this section.
This is a parallel change to that proposed for Sec. 50.59 and
Sec. 50.90, wherein the Commission would place the requirements for
changes to license conditions in the rule section on amendments to
licenses.

10 CFR 72.70

Paragraphs (a) and (b) would be revised to use the terms ``Final
Safety Analysis Report,'' ``FSAR,'' and ``as updated.'' Paragraph
(b)(2) would be revised to add changes to procedures to the annual
updates of the FSAR. New paragraph (c) would be added to provide
requirements on submitting revisions to the FSAR.

10 CFR 72.86

Paragraph (b) currently includes those sections under which
criminal sanctions are not issued. This paragraph would be revised by
adding 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)(4)

The change to this section is to conform the reference to 10 CFR
50.59 provisions, specifically to change from the terminology of
unreviewed safety question to referring to need for 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

New paragraph (d) provides 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 cask that are used by the
general licensee. The general licensee is also required to provide a
copy of its submittal to the certificate holder. This section is
similar to the requirements in Secs. 72.70 and 72.248 for submission of
annual updates to the FSAR associated with a site-specific Part 72
licensee or a certificate holder, respectively.

10 CFR 72.244

This new section provides requirements for a certificate holder 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 provides 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 provides requirements for submittal of annual
updates to a FSAR associated with the design of a spent fuel storage
cask which has been issued a CoC. This new section also provides that
the changes to procedures and structures, systems, and components
associated with the spent fuel storage cask and which are made pursuant
to Sec. 72.48 would be included in the annual update. The proposed
revisions would also require that the certificate holder provide a copy
of the FSAR submittal to each general licensee using that cask. This
section is similar to the requirements in Sec. 72.70 for submission of
annual updates to the FSAR associated with a site-specific part 72
license and new section 72.216 for general licensees to provide updates
to the FSAR.

[[Page 56113]]

IV. Commission Voting Record on SECY-98-171

The staff forwarded to the Commission a proposed rulemaking package
on Sec. 50.59 and related regulations in SECY-98-171, dated July 10,
1998. This document was placed in the Public Document Room on July 29,
1998. Subsequently, the Commission voted to approve issuance of a
proposed rule for public comments with several additions and changes
that are reflected in this notice. The Commission also directed that
the record of their decision on SECY-98-171 be included as part of this
notice to clearly inform stakeholders on preliminary positions taken by
the Commission. The text of the resultant staff requirements memorandum
and of the individual Commissioner vote sheets, is presented below.

Commission SRM on SECY-98-171, Dated September 25, 1998

The Commission has approved publication, for a 60 day public
comment period, the proposed rulemaking that would revise 10 CFR 50.59
and related provisions in parts 50, 52 and 72 concerning the processes
controlling licensee changes, tests and experiments for production and
utilization facilities and for facilities for independent storage of
spent nuclear fuel and high-level radioactive waste. The Voting Record,
which includes the Commissioner votes and this Staff Requirements
Memorandum, should be published in the Federal Register notice to
clearly inform stakeholders on preliminary positions taken by the
Commission (Enclosed).
The Commission also approves the staff's recommendations for
handling violations of 10 CFR 50.59 and 72.48, including staff plans
for exercise of enforcement discretion, while rulemaking is underway.
The Commission requested that the staff specifically solicit public
comment in the Federal Register notice on:
1. A wide array of options for the margin of safety criterion
(50.59(c)(2)(vii) in the proposed rule) and its definition including:
(a) Deleting the criterion and definition, (b) a new definition as
described in Chairman Jackson's vote, and (c) an option which would
decouple the last criterion from technical specifications and focus
instead on a new criterion relating to performance of fission product
barriers (e.g., reactor coolant system pressure, containment pressure,
etc), with minimal changes being allowed up to specified limits,
perhaps utilizing a graduated approach similar to the approaches
proposed for other criteria.
2. Options for defining ``minimal'' as it pertains to ``probability
of occurrence of an accident'' or ``probability of equipment
malfunction.''
3. The definitions of ``facility,'' ``procedures,'' and ``tests or
experiments,'' including elimination of the definitions.
4. A clear definition of ``accident.''
(This action scheduled for completion October 9, 1998).
The Commission requests the staff to complete the revised 50.59
rule on an expedited schedule.
(This action scheduled for completion February 19, 1999).
All Commissioners approved in part and disapproved in part the
proposed rulemaking on 10 CFR parts 50, 52 and 72 requirements
concerning changes, tests and experiments and staff recommendations on
changes to other regulations and enforcement policy, and provided
additional comments. In their vote sheets, all Commissioners approved
the staff's recommendations to approve publication of the proposed rule
for public comment, and use of the enforcement discretion guidance in
its assessment of severity levels for violations while the rulemaking
is underway, and provided some additional comments. In particular, all
Commissioners disapproved the staff's proposed margin of safety
criterion (Sec. 50.59(c)(2)(vii) in the proposed rule) and its
definition and each Commissioner provided an option for evaluation
during the comment period. The Commissioners also specifically
requested comments on a number of other issues. Because of the need to
finalize this rule as expeditiously as possible and because SECY-98-171
has already been publicly available since July 29, 1998, the Commission
agreed to a 60 day comment period, and that the staff complete the
revised Sec. 50.59 rule by February 19, 1999. Subsequently, the
comments of the Commission were incorporated into the guidance to staff
as reflected in the SRM issued on September 25, 1998.

Chairman Jackson's Comments on SECY-98-171

I approve, in part, and disapprove, in part, the staffs proposal
for rulemaking. I approve the staff's proceeding with issuance of the
proposed rule language for public comment in order to support the
expedited finalization of a revision to these processes. I disapprove
of the specific language proposed by the staff for
Sec. 50.59(c)(2)(vii), ``reductions in the margin of safety.''
I agree with the recent letter from ACRS on this rulemaking, in
that: (1) 10 CFR 50.59 can accommodate risk-informed decisionmaking.
(2) the positions, as presented, on margin of safety may add regulatory
burden without a commensurate safety benefit.
I disagree with ACRS in that I believe:
(1) The rulemaking should go out for public comment to foster
comment on this high priority issue, and
(2) The regulatory guidance can be worked in parallel with the
rulemaking.
I note that a further reason for issuing this package for public
comment at this time is that the paper calls for the proper use of
enforcement discretion as this rulemaking progresses, thereby providing
further stability in the implementation of this rule in the industry.
Further, I propose that the SRM on this SECY, and the voting
record, be placed in the FR notice to clearly inform stakeholders on
preliminary positions taken by the Commission.

Giving Definition to Minimal

Attached to the recent ACRS letter was ``A Proposal for the
Development of a Risk-Informed Framework for 10 CFR 50.59 and Related
Matters.'' The proposal forwarded by the ACRS parallels an existing
risk-informed approach described in Regulatory Guide 1.174. Regulatory
Guide 1.174 describes a method for determining the level of review,
based on severe accident implications, for proposed licensing actions.
The proposal forwarded by the ACRS describes methodology for creating
frequency-consequence curves for Class 1-8 accidents. The proposal
states that existing processes could be extended to provide appropriate
context for whether the results of a change are ``minimal.'' The
proposal also notes that aspects of this type of approach are in use in
the international regulatory community. The approach utilized in the
proposal forwarded by the ACRS is consistent with the Commission
guidance in the Staff Requirements Memorandum of March 24, 1998 on
SECY-97-205.
Without commenting on the specifics of the proposal forwarded by
the ACRS, I am convinced that changes to nuclear plants can be
evaluated in a risk-informed context. Any such approach would benefit
from paralleling existing methodology. Careful consideration would be
required to ensure that the ``consequence'' and ``frequency'' standards
are appropriate for a Sec. 50.59 type application. For instance,
``consequences'' could be evaluated at one of the following levels:
Fractional releases, off-site or on-site doses, or

[[Page 56114]]

challenges to fission product release barriers. ``Frequency'' could be
evaluated for Class 1-8 accidents or for design basis accidents using
existing guidelines for risk-informed regulation. The level at which
consequences and frequency of events were tracked would also impact the
type of parallel, deterministic (e.g., protection of redundancy,
defense in depth, etc.), considerations against which changes would
have to be evaluated. For instance, evaluating consequences at the
level of the loss of a single barrier, or occurrences of accident
sequence initiators, might allow elimination of parallel,
deterministic, considerations such as ``margin.''
It is of some concern to me that the whole staff has pursued risk-
informed approaches to issues like the review of TSs, the use of Graded
Quality Assurance, and programs like Inservice Inspection and Inservice
Testing, the staff appears to be more reluctant to allow risk-informed
approaches if the result is the relinquishment of review and approval
authority. Because prior NRC review and approval impacts the cost and
schedule of licensed activities, we must ensure that we require such
prior review and approval only when justified or required by mandate.
We should not limit the application of risk-informed regulation as a
means to ensure continued NRC reviews and approvals of licensed
activities. This message is complimentary to my oft repeated message to
industry that the use of risk information is ``double-edged,'' that is
that relief and additional regulatory scrutiny may both result from its
use.

Margin of safety

The staff proposes to provide a specific definition of ``Reduction
in margin of safety associated with any technical specification,'' and
to revise the current provisions of 10 CFR 50.59(a)(2)(iii) to
explicitly refer to this definition. While I commend the staff on its
efforts to provide clear, definitive, requirements in this proposed
rulemaking, I am concerned that the proposed rule is not consistent
with policy direction established by the Commission in the SRM dated
March 24, 1998. I concur that it is important that the staff has the
independence to (and, I believe, has the responsibility to) inform the
Commission when there are concerns with Commission guidance (as it did
in COMSECY 98-013). However, I believe that when the staff proposes to
take action that is inconsistent with Commission direction, it is
obliged to provide a clear and complete rationale for the proposed
departure. I do not feel that the staff has met that obligation for the
``margin of safety'' aspect of this proposed rule. However, this said,
I do not disagree with the staff's conclusion that we should be careful
to understand, and maintain, a consistent regulatory basis on ``margin
of safety.'' We must proceed in a manner that does not call into
question the existing deterministic basis for ``reasonable assurance''
of public safety embodied in plants Technical Specifications (TSs).
My previous discussions with the staff have indicated that it is
extremely difficult (and probably not legally defensible) to allow
decreases in the ``margin of safety'' when the upper and lower limits
between which ``margin'' may exist are not defined in relation to the
regulatory requirements for safe operation. Based upon these
discussions, I can only assume that the staff is hesitant to allow
direct reductions in margin within the ``basis'' for TSs because some
such changes could create a de-facto change in the TSs themselves. The
staff may also be concerned by the lack of consistency in the ``margin
of safety in the basis for TSs'' associated with the different
generations of existing licenses (e.g., older customized TSs compared
to improved standardized TSs), and associated with the different
methods utilized in the technical review and approval of the TS (e.g.,
some TSs might be based on maintaining margin between accident analysis
results and acceptance limits, while other TSs might be based on margin
which was built into analytical techniques and methodologies used in
the accident and safety analysis, with no ``margin'' between the
results and the acceptance limits, etc.).
The staff's proposed method of requiring prior agency approval to
changes of input assumptions, analytical methods, etc., for those
parameters which affected the selection of TSs, results in the newly
controlled parameters being treated essentially the same way as values
in the TSs. It also appears that implementation of the staffs proposed
control over a broad range of parameters used in the safety analysis
would effectively prevent any change to the facility that would result
in a ``minimal change in consequence,'' a condition allowed elsewhere
in the proposed rule. In other words, it is not clear what type of
changes would successfully pass the 10 CFR 50.59 test for allowed
``minimal increases in consequences,'' without failing the test for
``no reductions in the margin of safety.'' I do not believe that the
potential safety significance of all the parameters to be covered under
the proposed definition of a reduction in the margin of safety always
justify the requirement of prior NRC approval.
The staff should continue to work to establish a technically sound
method for allowing licensees to make plant changes where there is only
``minimal'' impact on safety. If fundamental conflicts exist with
allowing reductions in some ``margins of safety,'' especially those on
which the validity of TSs are based, then staff should provide a clear
explanation of this, and should address how other changes to the
structure of the regulation, which do not create fundamental conflicts,
can be made in a manner which achieves the Commission's objective of
removing unnecessary burdens from licensees.
Attachment ``A'' to this vote describes one alternate method for
addressing the issue of ``margin of safety.'' This alternative would
maintain existing margins of safety (associated with TSs), while
providing greater flexibility to licensees in implementing changes to
their facilities. This alternative is based on methodology similar to
that described in NEI 96-07. This methodology requires evaluating the
effect of proposed tests and changes on the accident analysis results
(rather than inputs, as proposed by the staff), in cases where TSs are
based on accident analysis considerations. Prior NRC approval of
changes, tests, and experiments would be limited to those cases where
there was a net effect on the accident analysis results. The
alternative also recognizes the significance of the analytical
techniques used in the safety or accident analysis, and would require
some form of prior approval for analytical methods used to support
changes when the change did not have prior NRC approval. This approach
could provide staff reasonable assurance that the assumptions made by
the license reviews are not invalidated. The staff should evaluate this
option, along with other comments in this area, during the comment
period.
In considering the technical and regulatory underpinning of this
clause of Sec. 50.59, I have become concerned that we are evaluating
incremental changes to a provision which is not well suited to such
changes. I am concerned that the result may be the addition of yet
another layer of regulatory process rather than the elimination of any
unnecessary layers. For this reason, the staff should be receptive to
internal or public comments on feasible alternatives which eliminate
the discussion of ``the margin of safety in the basis of TSs,'' while
maintaining the integrity of the plant's licensing basis. I envision
that it may be possible to eliminate the rule

[[Page 56115]]

language criteria on ``margin of safety'' if evaluations of
``frequency'' and ``consequences'' are performed at a level of
significance which bounds allowable ``minimal'' reductions in margin.

Accident of a Different Type

In determining the effect of any proposed change to Sec. 50.59, it
will be necessary to more clearly understand what an ``accident of a
different type'' is. The staff should provide a more definitive
definition of an accident than was included in COMSECY-98-013. The
information provided by the staff should address, as a minimum, the
following:
(1) What is an ``accident'' under this section, and is it
consistent with other existing regulations (e.g., Sec. 50.92,
Sec. 50.34, Appendix A of part 50, etc.)?
(2) Is an ``accident of a different type'' better described as an
``initiating event (e.g., loss of feedwater, loss of offsite power, new
common mode failure mechanism, etc.) of a different Type?''
(3) What are the bounds which limit those ``accidents'' which are
the subject of this Section (e.g., only those initiating events which,
when evaluated using approved analytical techniques, result in
transients with the potential to challenge fission product barriers,
etc.)?

Procedures

I commend staff on inserting a definition for the term ``Procedures
as described in the final safety analysis report (as updated).''
However, I am concerned that the definition provided may cloud the
distinction between: (1) Those procedures which must be screened, or
evaluated, under Sec. 50.59, and (2) the criteria which necessitates a
full safety evaluation. I believe that staff seeks to indicate that all
procedures which are described as being required in the FSAR are
subject to a Sec. 50.59 screening. The screening would identify the
need for a full safety evaluation only if a proposed procedure change
created a change to the ``information in the FSAR regarding how
structures, systems, and components are operated and controlled. . .
.'' Staff should solicit comment on this definition and clarify the
proposed definition, as required, in the final rule.

Making the Rule Risk Informed

I note with interest that members of the ACRS believe that there
are substantial barriers in the existing deterministic framework of 10
CFR part 50 to the concept of allowing ``minimal'' changes in accident
probabilities or consequences. In my previous vote on SECY-97-205,
``Integration and Evaluation of Results from Recent Lessons-Learned
Reviews,'' I approved the staff's proposal to develop the framework for
risk-informed regulatory processes. In particular, I called for the
staff to develop a series of milestones by which the Commission could
``chart its course in its move to more risk-informed regulatory
processes.'' Additionally, I promoted the idea of promulgating a new
regulation in 10 CFR part 50, that would make clear how the Commission
uses risk information in its decision-making. In proceeding with the
``short-term'' changes to 10 CFR 50.59 (and related regulations;
``short-term'' actions from SECY-97-205), and in responding to the
ACRS, the staff should re-evaluate whether the Agency should initiate
action to provide for a risk-informed framework that would allow for
the efficiencies to be gained through use of risk-informed,
performance-based revisions to our regulatory processes.

Attachment ``A'' to Chairman Jackson's vote sheet on SECY-98-171

``Straw Man'' on Margin of Safety

Regarding margin:
The margin between regulatory limits and the failure of
physical barriers is protected in the regulations (and also in the
portion of the Technical Specifications (TSs) called ``safety
limits'').
The margin, as reflected in approved safety and
accident analyses, between the protection afforded by the TSs (e.g.,
the limiting safety system settings and limiting conditions of
operations) and the associated regulatory limits is ``the margin of
safety as defined in the basis for any TS.''
The margin between normal plant or system operation and
the ``bounding'' assumptions used in accident analysis is below the
threshold of safety significance that requires NRC prior approval
for changes.
The results of safety and accident analyses are subject
to significant variance, depending on the analytical techniques and
methods used in the analysis. Where a licensee wishes to make a
change in their facility without prior NRC approval, the effects of
the change must be evaluated using analytical techniques and methods
which are NRC approved for the application, or which are reviewed
and vetted (but not subject to specific NRC approval) in a NRC
approved manner.

Direct changes to technical specifications require prior NRC
approval. Before changing other operational characteristics described
in the UFSAR, a safety evaluation must be performed to determine, among
other things, if the change results in a reduction in the level of
protection afforded by the TS (margin of safety as defined in any TS).
Such a reduction would typically occur only if the operational
characteristic had been used as a bounding condition in the analysis
upon which the selection of TS was based, or in analysis where the
acceptability of selected TS values was demonstrated. Licensees can
make desired changes to operational characteristics without prior NRC
approval, provided that the change does not result in accident analysis
results that are nearer the regulatory, or safety, limits than the
corresponding results that the NRC used in evaluating the acceptability
of the TS during licensing of the facility.
This regulatory position could be codified by adding the following
footnote to Section 50.59(a)(2)(iii):

The ``margin of safety as defined in any technical
specification'' (margin of safety) is the amount (quantitative or
qualitative) of margin between the operation of the facility as
described in the technical specifications and the exceedance of
safety limits listed in the technical specifications or other
regulatory limits. In relation to accident analysis, the margin of
safety is typically the difference between calculated parameters
(e.g., peak fuel clad temperature, maximum RCS pressure, etc.) and
the associated regulatory or safety limit. The margin of safety is a
product of specific values and limits contained in the technical
specifications (which cannot be changed without NRC approval) and
other values, such as assumed accident or transient initial
conditions or assumed safety system response times, which are not
specifically contained in the technical specifications. Any change
to the values not specifically contained in technical specifications
must be evaluated for impact on the margin between the calculated
result of an accident or transient and the safety or regulatory
limit. Changes, or the net effect of multiple changes, which result
in a reduction in the margin of safety require prior NRC approval.
Changes, or the net effect of multiple changes, which do not cause a
reduction in margin of safety do not require prior NRC approval. All
evaluatory work in assessing the impact of proposed changes must be
performed using methodology and analytical techniques which are
either reviewed and approved by the NRC or which are reviewed and
vetted in a manner approved by the NRC.

Commissioner Diaz's Comments on SECY-98-171

I consider this rulemaking effort to be our short term fix for the
50.59 rule, not the longer term risk-informed rule enhancement
discussed in SECY-97-205.
I approve the publication of this rulemaking package for a 90-day
public comment period, contingent upon the additions described in the
last paragraph of my comments. I propose that the package also include
the Commissioners' votes for public consideration. The purpose of
issuing the rulemaking package is to expedite rulemaking by opening the
process for

[[Page 56116]]

public comments during the Commission's continuing deliberation on this
matter. It should be made very clear to all stakeholders that
publication of the package is an invitation to participate in improving
the rulemaking. In fact, I do not agree with several of the proposed
positions in this paper, as delineated in my specific comments below.
I agree with the staff's recommendation to remove the reference to
``unreviewed safety question'' from Sec. 50.59 and to make conforming
changes in parts 50, 52, and 72. I also agree with staff's proposal to
allow a minimal increase in the probability of occurrence or
consequence of an accident or malfunction previously evaluated, and to
not allow the creation of an accident of a different type or
malfunction of equipment important to safety with a different result
than any previously evaluated.
I agree with the ACRS comments in their June 16, 1998, letter
regarding the definition of ``reduction in margin of safety.''
Notwithstanding the staff's suggestion of a possible Commission
interpretation, the language ``altered in a nonconservative manner''
can still be interpreted as a de facto ``zero increase'' standard for
the 50.59 criterion on margin of safety. I believe the risk-informed
Sec. 50.59 approach suggested in the ACRS letter deserves serious
consideration as part of longer term improvements and should be
considered in the staff's response, due in February 1999, to the SRM
for SECY-97-205.
The current language in Sec. 50.59(a)(2)(iii) (``margin of safety
as defined in the basis for any technical specification'') is, in fact,
defined and bounded by the technical specifications. Therefore, as long
as the licensee proposed change, test, or experiment under Sec. 50.59
is not in violation of the technical specification requirements, the
requisite margin of safety is maintained, and it is possible to
eliminate ``reduction of margin of safety'' from the rule as a
condition requiring prior staff approval. This change will eliminate
the existing ambiguity in the use of Sec. 50.59 for changes with
minimal safety significance. This alternative should also be published
for public comment; it is consistent with the safety envelope provided
by the technical specifications and is a straightforward improvement
that will match with the eventual conversion to a risk-informed rule.
I support the staff's recommended changes in the reporting and
record keeping requirements relating to Sec. 50.59. The enforcement
policy and its corresponding implementation guidance should be changed
in accordance with the revised Sec. 50.59 rule. I recommend that,
during the rulemaking period, the enforcement policy be revised to
grant discretion (i.e., suspend issuance of Level IV violations) under
Section VII.B.6 for those Sec. 50.59 violations of little or no safety
significance.
I do not agree with the recommended definitions of ``facility'',
``procedures'', ``reduction in margin of safety'', and ``tests or
experiments.'' These definitions appear to increase prescriptiveness at
the input of the licensees' change process instead of the output, and
therefore, are more broad-based than the definitions to date. I believe
that these definitions will create more burden for the NRC and
licensees, are not consistent with the original intent of the
Sec. 50.59 rule, i.e., to evaluate whether the licensee proposed
changes will result in inadequate protection of public health and
safety, and therefore, are not necessary.
On the other hand, the ``accident'' in the proposed revisions to
Sec. 50.59 should be defined. The ``accident of a different type than
any previously evaluated'' as described in the proposed
Sec. 50.59(c)(2)(v) should be of the same safety significance as the
``accident'' in the proposed Sec. 50.59(c)(2)(I) and (c)(2)(iii). The
staff should determine if the anticipated operational transients and
the postulated design basis accidents described in the FSAR form a
sufficient basis for the Sec. 50.59 evaluation.
The staff should continue its interactions with NEI in resolving
the differences between the NRC's position on Sec. 50.59 implementation
guidance and that contained in NEI 96-07. The regulatory guide for
Sec. 50.59 that endorses a revised NEI 96-07, with exceptions and
clarifications, as appropriate, should be developed concurrently with
the rulemaking process.
In summary, the staff should proceed with publishing the existing
rulemaking package, and concurrently solicit public comment on the
following alternatives: (1) eliminate ``reduction of margin of safety''
as a condition requiring prior staff approval, (2) eliminate the
broadened definitions of ``facility'', ``procedures'', ``reduction in
margin of safety'', and ``tests or experiments,'' and (3) clearly
define ``accident'' in the proposed revisions to Sec. 50.59. I urge the
staff to complete the revised Sec. 50.59 rule and the associated
regulatory guide by the end of March, 1999.

Commissioner McGaffigan's Comments on SECY-98-171

I approve publishing this rulemaking package for a ninety-day
public comment period. However, like my colleagues, I do not agree with
the staff proposal regarding ``reduction in the margin of safety
associated with any technical specification.''
As the Chairman points out, the definition of ``reduction in margin
of safety * * *'' would extend the requirements for prior agency
approval to underlying aspects (e.g., input assumptions) of parameters
that affected the selection of technical specifications, and result in
the newly controlled parameters being treated essentially the same way
as values in the technical specifications. This is the wrong way to go.
It is clear from my colleagues' and my vote that the margin of
safety criterion (Sec. 50.59(c)(2)(vii) in the proposed rule) and the
definition will need to be fixed in the final rule. My concern at this
point is that the staff discuss a wide enough array of options in the
Federal Register notice to ensure that the proposed rule will not have
to be renoticed before being finalized. Commissioner Diaz has proposed
to simply delete the criterion and definition as not needed. The
Chairman has proposed essentially a new definition. Another option
would decouple the last criterion from technical specifications and
focus instead on a new criterion relating to performance of fission
product barriers (e.g., RCS pressure, containment pressure. etc), with
minimal changes being allowed up to specified limits, perhaps utilizing
a graduated approach similar to the approaches proposed for other
criteria. Comment should be solicited on this option as well.
I believe that the staff has done a good job in proposing options
for defining ``minimal'' for consequences of an accident or
malfunction. On probability, however, the staff has essentially only
said that NEI 96-07 satisfies the proposed NRC standard for a
``minimal'' increase. That is a good step forward, and will bring
regulatory stability. I believe that in choosing the word ``minimal''
the Commission intended to grant greater flexibility than the NEI 96-07
``so small'' or negligible standard. The staff should continue to try
to give better definition to ``minimal'' as it pertains to
``probability of occurrence of an accident'' or ``probability of
equipment malfunction'' and solicit comment on this.
Finally, I endorse the use of enforcement discretion under Section

[[Page 56117]]

VII of the Enforcement Policy as the rulemaking proceeds for those
Sec. 50.59 violations of little or no safety/risk significance. The
staff should treat (vice ``consider treating'' as proposed by staff) as
minor violations cases where the violation of existing rule
requirements would not constitute a violation under the rule were it
revised as proposed. I do not object to documenting such minor
violations in inspection reports because the rule is still in a
proposed revision stage.

V. Rule Language Proposed by The Nuclear Energy Institute

In a letter dated November 14, 1997, the Nuclear Energy Institute
provided to the NRC suggested language for revising 10 CFR 50.59 that
they believed would enable the NRC to endorse NEI 96-07. This language
is included here in this Statement of Considerations so that interested
parties can offer comment on whether this language should be adopted by
the NRC. The supporting information for NEI's proposal is contained in
the referenced letter which is available for review in the Public
Document Room.
Specifically, NEI proposed that [existing] section 50.59(a)(2) be
revised to read:

(a)(2) A proposed change, test, or experiment shall be deemed to
involve an unreviewed safety question: (i) If there is more than a
negligible increase in the probability of occurrence of an accident
or malfunction of equipment important to safety previously evaluated
in the safety analysis report; or (ii) if the consequences of an
accident or malfunction important to safety previously evaluated in
the safety analysis report exceeds the established acceptance limit;
or (iii) if a possibility for an accident of a different type or
malfunction with a different result from any evaluated previously in
the safety analysis report may be created; or (iv) if the margin of
safety provided by any technical specification is reduced.

In this rulemaking, the Commission is proposing to adopt certain
aspects of the changes offered by NEI (e.g., on malfunction with a
different result). The Commission is seeking comment as to whether
other aspects of this proposal should be adopted. The Commission also
offers the following observations about this proposal for consideration
as part of the comment process:

A. Negligible Increase in Probability of Occurrence

NEI proposes that the rule be revised to state that a change would
be an USQ ``if there is more than a negligible increase in the
probability of occurrence of an accident or malfunction of equipment
important to safety previously evaluated in the safety analysis
report.'' As discussed above, the Commission is proposing a ``more than
minimally increased'' criterion, which is considered comparable in
overall intent to what was proposed by NEI.

B. Increase in Consequences of an Accident or Malfunction

NEI proposes that the rule be revised such that a change would be a
USQ if the consequences of an accident or malfunction previously
evaluated exceed the established acceptance limit. As NEI discusses
further in its letter, the established acceptance limit would be the
value that was previously reviewed and approved by the NRC generally as
documented in the staff's safety evaluation report (SER).⁶
---------------------------------------------------------------------------

\6\ Attempting to use values from the staff's SER as acceptance
limits would be difficult since SERs were not written for the
purpose of establishing such limits. In a literal sense, neither the
SAR nor the SER set an ``acceptance limit.'' Rather, the SAR
documents an applicant's/licensee's analytically derived conclusion
that a given event has a certain consequence which is within the
regulatory bounds set by NRC regulations. The SER is intended only
to confirm or modify that conclusion. The SAR value as modified
through the staff's review and approval then becomes the baseline
for future analyses.
---------------------------------------------------------------------------

The current industry guidance, NEI 96-07, would permit, in some
instances, increases in consequences up to the regulatory thresholds
(such as Part 100), without review. As discussed in (draft) NUREG-1606,
the staff typically performs independent evaluations of radiological
consequences of accidents, rather than an in-depth review of the
licensee's calculations, during licensing of the plant. As a result,
the degree of conservatism in the licensee calculations differs from
that used in the staff's assessments. As noted above, the Commission is
proposing to revise the rule to allow ``minimal'' increases in
consequences without prior approval, provided that the regulatory
limits are still met. The Commission has some concerns about allowing
licensee changes without review, which when evaluated with licensee
assumptions and methods, result in doses at or very close to the
regulatory guidelines (e.g., part 100). This is because such changes,
if reviewed with staff assumptions (or starting from the staff's
previous estimation of the accident dose), might result in the
regulatory guidelines not being met. Rather than allowing one change to
result in an increase in consequences up to the guidelines, the
Commission concludes that minimal increases, along with NRC oversight
of cumulative effects, is the appropriate standard for review.

C. Malfunction with a Different Result

As discussed above, the Commission is proposing to adopt this
particular proposed change to the rule.

D. Margin of Safety Provided by Any Technical Specification

NEI proposes to replace the existing language of ``as defined in
the basis for any technical specifications,'' with ``as provided by any
technical specification'' with respect to reductions in the margin of
safety. The proposed change is intended to clarify that the margin of
safety is not necessarily limited to information in the BASES section
of the technical specification. NEI 96-07 guidance notes that the SAR,
staff SERs and other licensing basis documents should be reviewed to
determine if a proposed change would result in a reduction in margin of
safety. NEI intended to use this rule language in conjunction with
guidance that the margin of safety is the range of values between the
acceptance limit reviewed by the NRC (e.g., ASME code stress limits,
containment design pressure, etc.) and the failure point. The
Commission is seeking comment on a range of options relating to margin
of safety, including the option proposed by NEI.

VI. Request for Comment

The Commission requests comments on the proposed rule, as discussed
in Section II above. In addition, the Commission is seeking comment on
a number of specific issues related to this rulemaking. All commenters
are encouraged to provide specific comments on the following issue
areas:
1. The Commission is seeking input on a number of options relating
to the criterion of margin of safety reduction, and its definition.
Some possible alternatives are presented in Section II.J as being
representative of the range of approaches under consideration, but the
Commission is open to other proposals that commenters may wish to put
forth as representing the best means to provide a clear understanding
of which margins should fall within the regulatory envelope of
requiring approval if they would be reduced as a result of a change,
test or experiment, if the margin of safety criterion were to be
retained.
2. The Commission is interested in options for defining what
constitutes a ``minimal'' increase in the probability of occurrence of
an accident previously evaluated in the FSAR or in the probability of
equipment malfunction (refer to Section II.G). This might include
suggested examples of changes

[[Page 56118]]

that commenters believe represent only a ``minimal increase'' in
probability.
3. The Commission is interested in comments upon the proposed
definitions for such terms as ``facility as described in the FSAR,''
``procedures as described in the FSAR,'' and ``tests or experiments''
(refer to Sections II.B, C, and D). The Commission is soliciting views
on whether (1) definitions are necessary, (2) the proposed definitions
are desirable, even if not necessary, and (3) whether the suggested
definitions are clear and focused upon the appropriate changes that
should be evaluated. In this light, the Commission is also interested
in comments on a broader view of the scope of changes that should be
evaluated; for instance, should the scope be linked to the SAR, or
should the focus of changes to the facility be linked to another set of
regulatory information?
4. As part of the present rulemaking, the Commission is seeking
comment on the need for a clear definition of accident as it is used in
Sec. 50.59 to reflect the Commission's intent that the ``accidents''
referred to are those dealt with in the safety analysis report (see
Section II.H of this notice for discussion of issues related to
definition of accident).
5. In addition to the NRC proposals in Sections II and III, the
Commission is also interested in receiving comments on the proposals
and language suggested by NEI (Section V).

VII. Availability of Documents and Electronic Access

Certain documents related to this rulemaking, including comments
received and the regulatory analysis, may be examined at the NRC Public
Document Room, 2120 L Street NW. (Lower Level), Washington, DC NRC
documents also may be viewed and downloaded electronically via the
interactive rulemaking website established by NRC for this rulemaking.
You may also provide comments via the NRC's interactive rulemaking
web site through the NRC home page (http://www.nrc.gov). This site
provides the availability to upload comments as files (any format), if
your web browser supports that function. For information about the
interactive rulemaking site, contact Ms. Carol Gallagher, (301) 415-
5905; e-mail [email protected]

VIII. 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, if adopted, will not have
a significant impact on the environment. The proposed 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 being proposed 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 probability or consequences of accidents (still satisfying
regulatory limits) would 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, or reductions in defined margins of safety
require prior approval. 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, the Commission concludes that there will be
no significant impact on the environment from this proposed rule. This
discussion constitutes the environmental assessment and finding of no
significant impact for this proposed rule.

IX. Paperwork Reduction Act Statement

This proposed rule amends information collection requirements that
are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). This rule has been 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 proposed rule changes would 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 and in Sec. 50.71(e) for submittal of
updates to the FSAR, as well as record keeping requirements. To the
extent that the definitions provided in the proposed revisions would
require evaluations that are not presently being performed, there may
be an increase in record keeping and reporting. The Commission
estimates that this is a small increment over the existing burden. On
the other hand, some changes might be screened out as not needing
evaluation on the basis of these definitions, and thus there would
overall be at most a small increase in the record keeping required.
In addition, the requirements under Sec. 72.48 are also being
revised to explicitly require records of determinations concerning
occupational dose and environmental impact (the existing rules required
the evaluations but did not explicitly specify record retention
requirements for these evaluations). The Commission does not believe
this that this change will significantly impact record keeping burden
because records of evaluations of changes are already required (as to
whether they involve a USQ), and the evaluation itself is already
required by the rule. The part 72 burden associated with the
definitions of when evaluations are required should be significantly
less than for Sec. 50.59 since the number of licensees is smaller and
the expected number of changes is also smaller. Further, there is a
recordkeeping requirement established for CoC holders who make changes
to an approved storage cask design in accordance with Sec. 72.48.
With respect to reporting requirements, the Commission is proposing
to modify the FSAR update requirement to state that the updates must
include specific information on the effects of changes made. This was
not explicitly stated in the current rule, although it could be
inferred that this was what the update rule intended, as follows. In
the Statement of Considerations for Sec. 50.71(e),(45 FR 30615), the
NRC commented on the relationship between changes made under Sec. 50.59
and FSAR updating, stating: ``The Sec. 50.59(b) reporting may not be
detailed sufficiently to be considered adequate to fulfill the FSAR
updating requirement. The degree of detail required for updating the
FSAR will be generally greater than a `brief description' and a
`summary of the safety evaluation'.'' Thus, the Commission clearly
expected the update submittal to include sufficient information to
appropriately reflect the changes that were made. The burden associated
with explicitly documenting in the update the effects of the changes on
event probabilities and consequences is therefore small.
The public reporting burden for this information collection request
is estimated to average 3100 hours per response, including the time for
reviewing instructions, searching

[[Page 56119]]

existing data sources, gathering and maintaining the data needed, and
completing and reviewing the information collection. The Commission
estimates that there is only a slight increase in burden associated
with these proposed changes over the existing burden. The U.S. Nuclear
Regulatory Commission is seeking public comment on the potential impact
of the collection of information contained in the proposed rule and on
the following issues:
1. Is the proposed collection of information necessary for the
proper performance of the functions of the NRC, including whether the
information will have practical utility?
2. Is the estimate of the burden correct?
3. Is there a way to enhance the quality, utility, and clarity of
the information to be collected?
4. How can the burden of the collection of information be
minimized, including the use of automated collection techniques?
Send comments on any aspect of this proposed collection of
information, including suggestions for reducing the burden, to the
Information and Records Management Branch (T-6 F33), U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, or by Internet
electronic mail at [email protected]; and to the Desk Officer, Office of
Information and Regulatory Affairs, NEOB-10202, (3150-0017, -0020, -
0011, -0009, and -01320), Office of Management and Budget, Washington,
DC 20503.
Comments to OMB on the collections of information or on the above
issues should be submitted by November 20, 1998. Comments received
after this date will be considered if it is practical to do so, but
assurance of consideration cannot be given to comments received after
this date.

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless it displays a currently
valid OMB control number.

X. Regulatory Analysis

The Commission has prepared a draft regulatory analysis on this
proposed regulation. The analysis examines the values and impacts of
the alternatives considered by the Commission and includes the backfit
analysis required by Sec. 50.109 (and Sec. 72.62). The alternatives
considered in this analysis include no action, issuance of guidance
only, or rulemaking. The draft analysis is available for inspection in
the NRC Public Document Room, 2120 L Street NW. (Lower Level),
Washington, DC and is available through the NRC interactive rulemaking
website. Single copies of the analysis may be obtained from Eileen
McKenna, [email protected] (301) 415-2189, Mail stop O-11-F-1, U.S. Nuclear
Regulatory Commission, Washington DC 20555.
The Commission requests public comment on the draft analysis.
Comments on the draft analysis may be submitted to the NRC as indicated
under the ADDRESSES heading.

XI. 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, if
promulgated, have a significant economic impact on a substantial number
of small entities. This proposed rule affects only the licensing and
operation and decommissioning of nuclear power plants, nonpower
reactors, and independent spent fuel storage facilities. 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.

XII. Backfit Analysis

As required by Sec. 50.109 and Sec. 72.62, the Commission has
completed a backfit analysis for the proposed rule, which is included
within the regulatory analysis. The Commission has determined, based on
this analysis, that in most respects, the proposed rule does not impose
new requirements, but provides more flexibility or clarification of
existing requirements. In other respects, such as the definitions of
change to the facility and ``reduction of margin of safety* * *'', some
licensees may view the revised rule as imposing new requirements.
Therefore, the Commission has prepared an analysis considering the
factors in Sec. 50.109(c), which is included in the Regulatory
Analysis.

XIII. Criminal Penalties

For the purposes of Section 223 of the Atomic Energy Act (AEA), the
Commission is issuing the proposed 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.

XIV. 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 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 52

Administrative practice and procedure, Antitrust, Backfitting,
Combined license, Early site permit, Emergency planning, Fees,
Inspection, Limited work authorization, Nuclear power plants and
reactors, Probabilistic risk assessment, Prototype, Reactor siting
criteria, Redress of site, Reporting and record keeping requirements,
Standard design, Standard design certification.

10 CFR Part 72

Manpower training programs, Nuclear materials, Occupational safety
and health, Reporting and record keeping 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. 553, the NRC is proposing to
adopt the following amendments to 10 CFR parts 50, 52 and 72.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES

1. The authority citation for part 50 continues to read as follows:

[[Page 56120]]

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. 1244, 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 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101,
185, 68 Stat. 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, and
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). Section 50.37
also issued under E.O. 12829, 3 CFR 1993 Comp., P. 570; E.O. 12958,
Sections 50.58, 50.91, and 50.92 also issued under Pub. L. 97-415,
96 Stat. 2073 (42 U.S.C. 2239). Section 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, 68 Stat. 955 (42 U.S.C 2237).

2. Section 50.59 is revised to read as follows:

Sec. 50.59 Changes, tests and experiments.

(a) Definitions for the purposes of this section:
(1) Change means a modification, addition, or removal.
(2) Facility as described in the final safety analysis report (as
updated) means:
(i) The systems, structures, and components that are described in
the final safety analysis report(as updated),
(ii) The design, performance requirements and methods of operation
for such systems, structures and components required to be included or
described in the final safety analysis report (as updated), and
(iii) The evaluations or methods of evaluation required to be
included in the FSAR (as updated) for such SSC and which demonstrate
that their intended function(s) will be accomplished.
(3) 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
modified as a result of changes made pursuant to Sec. 50.59 and
Sec. 50.90, and, as applicable, Sec. 50.71 (e) and (f).
(4) Procedures as described in the final safety analysis report (as
updated) means information in the final safety analysis report (as
updated) regarding how structures, systems, and components are operated
and controlled (including assumed operator actions and response times)
and information describing the conduct of operations.
(5) Reduction in margin of safety associated with any technical
specification means that the input assumptions, analytical methods,
acceptance conditions, criteria and limits of the safety analyses,
presented in the final safety analysis report (as updated), that
established any technical specification requirement, are altered in a
nonconservative manner.
(6) Tests or experiments not described in the final safety analysis
report (as updated) means any condition where the reactor or any of its
systems, structures or components are utilized or controlled in a
manner which is either:
(i) Outside the controlling parameters of the design bases as
described in the final safety analysis report (as updated) or
(ii) Inconsistent with the analyses in the final safety analysis
report (as updated).
(b) Applicability. The provisions of this section apply 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
permanently modified 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. The provisions in this
section do not apply to changes in procedures when the applicable
regulations establish more specific criteria for accomplishing such
changes.
(2) A licensee shall obtain an amendment to the license pursuant to
Sec. 50.90 prior to implementing a change, test or experiment if it
would:
(i) Result in more than a minimal increase in the probability of
occurrence of an accident previously evaluated in either the final
safety analysis report (as updated), 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;
(ii) Result in more than a minimal increase in the probability of
occurrence of a malfunction of equipment important to safety previously
evaluated in either the final safety analysis report (as updated), 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;
(iii) Result in more than a minimal increase in the consequences of
an accident previously evaluated in either the final safety analysis
report (as updated), 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;
(iv) Result in more than a minimal increase in the consequences of
a malfunction of equipment important to safety previously evaluated in
either the final safety analysis report (as updated), 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;
(v) Create a possibility for a design basis accident of a different
type than any previously evaluated in either the final safety analysis
report (as updated), or in evaluations performed pursuant to this
section and safety analyses performed pursuant to Sec. 50.90 with
respect to design basis accidents after the last final safety analysis
report was updated pursuant to Sec. 50.71 of this part;
(vi) Create a possibility for a malfunction of equipment important
to safety with a different result than any previously evaluated in
either the final safety analysis report (as updated), 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;
(vii) Result in a reduction in the margin of safety associated with
any Technical Specification.
(d)(1) The licensee shall maintain records of changes in the
facility and of changes in procedures made pursuant to this section, to
the extent that these changes constitute changes in the facility as
described in the final safety analysis report (as updated) or to the
extent that they constitute changes in procedures as described in the
final

[[Page 56121]]

safety analysis report (as updated). The licensee shall also maintain
records of tests and experiments carried out 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. The report may be
submitted annually or along with the FSAR updates as specified by
Sec. 50.71(e), or at such shorter intervals as may be specified in the
license.
(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 five 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 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 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) 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 must 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
submittal must 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 facility or procedures as described in
the FSAR;
(2) All safety analyses and evaluations performed by the licensee
either in support of requested 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;
(3) All analyses of new safety issues performed by or on behalf of
the licensee at Commission request; and
(4) The net effect of all changes made since the last update on the
safety analyses, including probabilities, consequences, calculated
values, system or component performance, that are in the FSAR (as
updated). The updated information shall be appropriately located within
the update to the FSAR.
* * * * *
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.

[[Page 56122]]

PART 52--EARLY SITE PERMITS, STANDARD DESIGN CERTIFICATIONS; AND
COMBINED LICENSES FOR NUCLEAR POWER PLANTS

6. The authority citation for part 52 continues to read as follows:

Authority: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat.
936, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244,
as amended (42 U.S.C. 2133, 2201, 2232, 2233, 2236, 2239, 2282);
secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended (42
U.S.C. 5841, 5842, 5546).

7. Appendix A to Part 52 is amended by revising Section VIII.B,
paragraphs 5.a,b,d, and Section X.A.3 as follows:

Appendix A--Design Certification Rule for the U.S. Advanced Boiling
Water Reactor

VIII. Processes for Changes and Departures

* * * * *

B. Tier 2 information

5. * * *
a. An applicant or licensee who references this appendix may
depart from Tier 2 information, without prior NRC approval, unless
the proposed departure involves a change to or departure from Tier 1
information, Tier 2* information, or the technical specifications,
or otherwise requires a license amendment as defined in paragraphs
B.5.b and B.5.c of this section. When evaluating the proposed
departure, an applicant or licensee shall consider all matters
described in the plant-specific DCD.
b. A proposed departure from Tier 2, other than one affecting
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would--
(1) Result in more than a minimal increase in the probability of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the probability of
occurrence of a malfunction of equipment important to safety
previously evaluated in the plant-specific DCD;
(3) Result in more than a minimal increase in the consequences
of an accident previously evaluated in the plant-specific DCD;
(4) Result in more than a minimal increase in the consequences
of a malfunction of equipment important to safety previously
evaluated in the plant-specific DCD;
(5) Create a possibility for a design basis accident of a
different type than any evaluated previously in the plant-specific
DCD;
(6) Create a possibility for a malfunction of equipment
important to safety with a different result than any evaluated
previously in the plant-specific DCD; or
(7) Result in a reduction in the margin of safety associated
with any Technical Specification for an application or license
referencing this design certification.
* * * * *
d. If a departure requires a license amendment pursuant to
paragraphs B.5.b or B.5.c of this section, it is governed by 10 CFR
50.90.
* * * * *

X. Records and Reporting

A. Records.

* * * * *
3. An applicant or licensee who references this appendix shall
prepare and maintain written evaluations which provide the bases for
the determinations required by Section VIII of this appendix. These
evaluations must be retained throughout the period of application
and for the term of the license (including any period of renewal).

8. Appendix B to part 52 is amended by revising Section VIII.B,
paragraphs 5.a,b,d, and Section X.A.3 to read as follows:

Appendix B--Design Certification Rule for the System 80+ Design

VIII. Processes for Changes and Departures

* * * * *

B. Tier 2 information.

* * * * *
a. An applicant or licensee who references this appendix may
depart from Tier 2 information, without prior NRC approval, unless
the proposed departure involves a change to or departure from Tier 1
information, Tier 2* information, or the technical specifications,
or otherwise requires a license amendment as defined in paragraphs
B.5.b and B.5.c of this section. When evaluating the proposed
departure, an applicant or licensee shall consider all matters
described in the plant-specific DCD.
b. A proposed departure from Tier 2, other than one affecting
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would--
(1) Result in more than a minimal increase in the probability of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the probability of
occurrence of a malfunction of equipment important to safety
previously evaluated in the plant-specific DCD;
(3) Result in more than a minimal increase in the consequences
of an accident previously evaluated in the plant-specific DCD;
(4) Result in more than a minimal increase in the consequences
of a malfunction of equipment important to safety previously
evaluated in the plant-specific DCD;
(5) Create a possibility for a design basis accident of a
different type than any evaluated previously in the plant-specific
DCD;
(6) Create a possibility for a malfunction of equipment
important to safety with a different result than any evaluated
previously in the plant-specific DCD; or
(7) Result in a reduction in the margin of safety associated
with any Technical Specification for an application or license
referencing this design certification.
* * * * *
d. If a departure requires a license amendment pursuant to
paragraphs B.5.b or B.5.c of this section, it is governed by 10 CFR
50.90.
* * * * *

X. Records and Reporting

A. Records.
* * * * *
3. An applicant or licensee who references this appendix shall
prepare and maintain written evaluations which provide the bases for
the determinations required by Section VIII of this appendix. These
evaluations must be retained throughout the period of application
and for the term of the license (including any period of renewal).

PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE

9. 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).

10. 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

[[Page 56123]]

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 facility may still be considered independent.
* * * * *
11. 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.
12. 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.
* * * * *
13. Section 72.48 is revised to read as follows:

Sec. 72.48 Changes, tests and experiments.

(a) Definitions--As used in this section:
(1) Change means a modification, addition or removal.
(2) Final Safety Analysis Report (as updated) means:
(i) For site-specific licensees, the Safety Analysis Report for a
ISFSI, MRS or spent fuel storage cask, submitted in accordance with
Sec. 72.24, as modified as a result of changes made pursuant to
Sec. 72.48, and as updated in accordance with Sec. 72.70;
(ii) For general licensees, the Safety Analysis Report for a ISFSI,
MRS or spent fuel storage cask, as modified as a result of changes made
pursuant to Sec. 72.48, and as updated in accordance with Sec. 72.216;
and
(iii) For certificate holders, the Safety Analysis Report for an
approved cask, modified by as a result of changes made pursuant to
Sec. 72.48 and as updated in accordance with Sec. 72.248.
(3) The ISFSI, MRS, or spent fuel storage cask as described in the
Final Safety Analysis Report (as updated) means:
(i) The systems, structures, and components that are described in
the Final Safety Analysis Report as updated in accordance with
Secs. 72.70, 72.216 or Sec. 72.248,
(ii) The design, performance requirements and methods of operation
for such systems, structures, and components required to be included or
described in the Final Safety Analysis Report (as updated), and
(iii) The evaluations for such systems, structures, and components
required to be included in the Final Safety Analysis Report (as
updated) and which demonstrate that their intended function(s) will be
accomplished.
(4) Procedures as described in the Final Safety Analysis Report (as
updated) means information in the Final Safety Analysis Report (as
updated) regarding how structures, systems, and components are operated
or controlled and information describing conduct of operations.
(5) Reduction in margin of safety associated with any technical
specification means that the input assumptions, analytical methods,
acceptance conditions, criteria and limits of the safety analyses,
presented in the Final Safety Analysis Report (as updated), that
established any technical specification requirement, are altered in a
nonconservative manner.
(6) Tests or experiments not described in the Final Safety Analysis
Report (as updated) means any condition where the ISFSI, MRS or spent
fuel storage cask or any of its systems, structures, or components are
utilized or controlled in a manner which is either:
(i) Outside the controlling parameters of the design bases as
described in the Final Safety Analysis Report (as updated) or
(ii) Inconsistent with the analyses in the Final Safety Analysis
Report (as updated).
(b)(1) A licensee or certificate holder may make changes in the
ISFSI, MRS, or spent fuel storage cask 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 either a license amendment pursuant to
Sec. 72.56 (for licensees), if a change in the conditions incorporated
in the license is not required, and the change, test, or experiment
does not meet any of the criteria in paragraph (b)(2) of this section
or a Certificate of Compliance (CoC) amendment pursuant to Sec. 72.244
(for certificate holders), if a change in the terms, conditions or
specifications incorporated in the CoC is not required; and the change,
test, or experiment does not meet any of the criteria in paragraph
(b)(2) of this section. The provisions in this section do not apply to
changes in procedures when the applicable regulations establish more
specific criteria for accomplishing such changes.
(2) A licensee shall obtain a license amendment pursuant to
Sec. 72.56 and a certificate holder shall obtain a CoC amendment
pursuant to Sec. 72.244, prior to implementing a change, test, or
experiment if it would:
(i) Result in more than a minimal increase in the probability of
occurrence of an accident previously evaluated in either the Final
Safety Analysis Report (as updated), or in evaluations performed
pursuant to this section and safety analyses performed pursuant to
Secs. 72.56 or 72.244 after the last Final Safety Analysis Report was
updated pursuant to Secs. 72.70, 72.216 or Sec. 72.248, of this part,
as applicable;
(ii) Result in more than a minimal increase in the probability of
occurrence of a malfunction of structures, systems, and components
important to safety which were previously evaluated in either the Final
Safety Analysis Report (as updated), or in evaluations performed
pursuant to this section and safety analyses performed pursuant to
Secs. 72.56 or 72.244 after the last final safety analysis report was
updated pursuant to Secs. 72.70, 72.216 or Sec. 72.248, of this part,
as applicable;
(iii) Result in more than a minimal increase in the consequences of
an accident previously evaluated in either the Final Safety Analysis
Report (as updated), or in evaluations performed pursuant to this
section and safety analyses performed pursuant to Secs. 72.56 or 72.244
after the last final safety analysis report was updated pursuant to
section 72.70, 72.216 or Sec. 72.248, of this part, as applicable;
(iv) Result in more than a minimal increase in the consequences of
a

[[Page 56124]]

malfunction of structures, systems, and components important to safety
which were previously evaluated in either the Final Safety Analysis
Report (as updated), or in evaluations performed pursuant to this
section and safety analyses performed pursuant to Sec. 72.56 or
Sec. 72.244 after the last final safety analysis report was updated
pursuant to Sec. 72.70, Sec. 72.216 or Sec. 72.248, of this part, as
applicable;
(v) Create the possibility for a design basis accident of a
different type than any evaluated previously in either the Final Safety
Analysis Report (as updated), or in evaluations performed pursuant to
this section and safety analyses performed pursuant to Secs. 72.56 or
Sec. 72.244 with respect to design basis accidents after the last final
safety analysis report was updated pursuant to Sec. 72.70, Sec. 72.216
or Sec. 72.248, of this part, as applicable;
(vi) Create the possibility for a malfunction of structures,
systems, and components important to safety with a different result
than any evaluated previously in either the Final Safety Analysis
Report (as updated), or in evaluations performed pursuant to this
section and safety analyses performed pursuant to Secs. 72.56 or
Sec. 72.244 after the last final safety analysis report was updated
pursuant to Sec. 72.70, Sec. 72.216 or Sec. 72.248, of this part, as
applicable;
(vii) Result in a reduction in the margin of safety associated with
any technical specification; (viii) Result in a significant increase in
occupational exposure;
(ix) Result in a significant unreviewed environmental impact.
(c)(1) Each licensee or certificate holder shall maintain records
of changes in the ISFSI, MRS, or spent fuel storage cask and of changes
in procedures it has made pursuant to this section if these changes
constitute changes in the ISFSI, MRS, or spent fuel storage cask or
procedures described in the Final Safety Analysis Report (as updated).
The licensee or certificate holder shall also maintain records of test
and experiments carried out pursuant to paragraph (b) of this section.
These records shall include a written evaluation that provides the
bases for the determination that the change, test, or experiment does
not require a license or CoC amendment pursuant to paragraph (b)(2) of
this section. The records of changes in the ISFSI, MRS, or spent fuel
storage cask and of changes in procedures and records of tests and
experiments shall be maintained until spent nuclear fuel is no longer
stored in the ISFSI, MRS or spent fuel storage cask, and the Commission
terminates the license or CoC. For a holder of cask Certificate of
Compliance who permanently ceases operation, any such records shall be
provided to the new holder of cask Certificate of Compliance or to the
Commission, as appropriate, in accordance with Sec. 72.234(d)(3).
(2) Annually, or at such shorter interval as may be specified in
the license or CoC, each holder of a license or cask Certificate of
Compliance shall submit a report containing a brief description of
changes, tests and experiments made by the license or certificate
holder under paragraph (b) of this section, including a summary of the
evaluation of each. Licensee and certificate holders shall submit their
reports in accordance with Sec. 72.4. Any report submitted by a
licensee or certificate holder pursuant to this paragraph will be made
a part of the public record pertaining to the license or CoC.
14. Section 72.56 is revised to read as follows:

Sec. 72.56 Application for amendment of license.

Whenever a holder of a 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.
15. In Sec. 72.70, paragraphs (a), (b), introductory text, and
(b)(2) are revised to read and a new paragraph (c) is added to read as
follows:

Sec. 72.70 Safety analysis report updating.

(a) The design, description of planned operations, and other
information submitted in the Safety Analysis Report for an ISFSI or MRS
shall be updated by the licensee and submitted to the Commission at
least once every six months after issuance of the license during final
design and construction, until preoperational testing is completed,
with a Final Safety Analysis Report (FSAR) completed and submitted to
the Commission at least 90 days prior to the planned receipt of spent
fuel or high-level radioactive waste. The FSAR shall include a final
analysis and evaluation of the design and performance of structures,
systems, and components that are important to safety taking into
account any pertinent information developed since the submittal of the
license application.
(b) After the first receipt of spent fuel or high-level radioactive
waste for storage, the FSAR shall be updated annually and submitted to
the Commission by the licensee. This submittal shall include the
following:
* * * * *
(2) A description and analysis of changes in procedures or in
structures, systems, and components of the ISFSI or MRS, as described
in the FSAR (as updated), with emphasis upon:
* * * * *
(c) The licensee shall submit revisions of the FSAR to the
Commission in accordance with Sec. 72.4, on a replacement-page basis
that is accompanied by a list which 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).
16. 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.
17. In Sec. 72.212, paragraph (b)(4) is revised to read as follows:

Sec. 72.212 Conditions of general license issued under Sec. 72.210.

* * * * *
(b) * * *
(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.
18. In Sec. 72.216, new paragraph (d) is added to read as follows:

Sec. 72.216 Reports.

* * * * *
(d) The final safety analysis report (FSAR) for each approved cask
used by the general licensee shall be updated annually and submitted to
the Commission by the general licensee.

[[Page 56125]]

The submittal shall include the following:
(1) A description and analysis of changes in procedures or in
structures, systems, and components of the spent fuel storage cask, as
described in the FSAR (as updated), with emphasis upon:
(i) Performance requirements,
(ii) The bases, with technical justification therefor upon which
such requirements have been established, and
(iii) Evaluations showing that safety functions will be
accomplished.
(2) An analysis of the significance of any changes to codes,
standards, regulations, or regulatory guides which the general licensee
has committed to meeting the requirements of which are applicable to
the design, construction, or fabrication of the spent fuel storage
cask.
(3) The general licensee shall submit revisions containing updated
information to the Commission, in accordance with Sec. 72.4, on a
replacement-page basis that is accompanied by a list which identifies
the current pages of the FSAR following page replacement. The general
licensee shall also provide a copy of the submittal to the holder of
the certificate for the cask. 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).
Each replacement page shall also indicate the cask FSAR, including the
certificate holder's revision number, upon which the general licensee's
update is based.
19. 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.
20. 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.
21. Section 72.248 is added to read as follows:

Sec. 72.248 Safety analysis report updating.

(a) The design, description of planned operations, and other
information submitted in the Safety Analysis Report for a spent fuel
storage cask shall be updated by the certificate holder and submitted
to the Commission after the design of the spent fuel storage cask has
been approved pursuant to Sec. 72.238. This Final Safety Analysis
Report (FSAR) shall be completed and submitted to the Commission within
90 days after approval of the cask design. The FSAR shall incorporate
all changes and requirements contained in the CoC and the staff's
safety evaluation report (SER) associated with approval of the cask's
design.
(b) The FSAR shall be updated annually and submitted to the
Commission by the certificate holder. This submittal shall include the
following:
(1) A description and analysis of changes in procedures or in
structures, systems, and components of the spent fuel storage cask, as
described in the FSAR (as updated), with emphasis upon:
(i) Performance requirements,
(ii) The bases, with technical justification therefor upon which
such requirements have been established, and
(iii) Evaluations showing that safety functions will be
accomplished.
(2) An analysis of the significance of any changes to codes,
standards, regulations, or regulatory guides which the certificate
holder has committed to meeting the requirements of which are
applicable to the design, construction, or fabrication of the spent
fuel storage cask.
(c) The certificate holder shall submit revisions containing
updated information to the Commission, in accordance with Sec. 72.4, on
a replacement-page basis that is accompanied by a list which identifies
the current pages of the FSAR following page replacement. The
certificate holder shall also provide a copy of the submittal to each
general licensee using the spent fuel storage cask. 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).

Dated at Rockville, Maryland, this 14th day of October, 1998.

For the Nuclear Regulatory Commission.
John C. Hoyle,
Secretary of the Commission.
[FR Doc. 98-28066 Filed 10-20-98; 8:45 am]
BILLING CODE 7590-01-P

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Document Information

Published:: 10/21/1998
Department:: Nuclear Regulatory Commission
Entry Type:: Proposed Rule
Action:: Proposed rule.
Document Number:: 98-28066
Dates:: Submit comments by December 21, 1998. Comments received after this date will be considered if it is practical to do so, but the Commission is able to assure consideration only for comments received on or before this date.
Pages:: 56098-56125 (28 pages)
RINs:: 3150-AF94: Changes, Tests, and Experiments
RIN Links:: https://www.federalregister.gov/regulations/3150-AF94/changes-tests-and-experiments
PDF File:: 98-28066.pdf
CFR: (19): 10 CFR 50.82(a)(1); 10 CFR 50.59(c)(2); 10 CFR 50.71(e); 10 CFR 50.59; 10 CFR 50.66; More ...