[Federal Register Volume 63, Number 161 (Thursday, August 20, 1998)]
[Notices]
[Pages 44659-44662]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-22412]
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NUCLEAR REGULATORY COMMISSION
Use of PRA in Plant-Specific Reactor Regulatory Activities: Final
Regulatory Guide and Standard Review Plan Section; Availability
The Nuclear Regulatory Commission has issued a new guide in its
Regulatory Guide Series, along with its conforming section of the
Standard Review Plan. Regulatory Guide 1.174, ``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' describes a method
acceptable to the NRC staff for assessing the nature and impact of
changes to a plant's licensing basis when the licensee chooses to
support these changes with risk information. The accompanying Standard
Review Plan Chapter 19, ``Use of Probabilistic Risk Assessment in
Plant-Specific, Risk-Informed Decisionmaking: General Guidance,''
conforms to the guide to provide guidance to the NRC staff in reviewing
such changes.
In June 1997, the Nuclear Regulatory Commission issued for public
comment a series of draft regulatory guides and Standard Review Plan
sections and a draft NUREG document addressing the use of PRA in
support of risk-informed regulatory activities. The preparation of
these documents followed from the Commission's Policy Statement of
August 16, 1995, on the use of PRA methods in nuclear regulatory
activities (60 FR 42622). The draft guidance documents were being
developed to provide acceptable approaches for using probabilistic risk
assessment (PRA) information in support of plant-specific changes to
plant licensing bases. The use of such PRA information and guidance by
power reactor licensees is voluntary, and alternative approaches may be
proposed.
The Commission conducted a workshop on August 11-13, 1997, during
the comment period, to provide an overview of the draft documents, to
answer questions regarding their intended application, and to solicit
comments and suggestions. Comments received from the workshop have been
considered in preparing this final general regulatory guide (1.174) and
its accompanying Standard Review Plan (Chapter 19) for risk-informed
applications. Comments received from the workshop on application-
specific guidance documents for technical specifications, inservice
testing, and graded quality assurance are currently being considered.
These guidance documents will be issued at a later date.
Comments and suggestions in connection with items for inclusion in
guides currently being developed or improvements in all published
guides are encouraged at any time. Written comments may be submitted to
the Rules and Directives Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555.
Single copies of regulatory guides, both active and draft, and
draft NUREG documents may be obtained free of charge by writing the
Reproduction and Distribution Services Section, OCIO, USNRC,
Washington, DC 20555-0001; or by fax to (301) 415-2289; or by email to
[email protected]. Active guides may also be purchased from the National
Technical Information Service on a standing order basis. Details on
this service may be obtained by writing NTIS, 5285 Port Royal Road,
Springfield, VA 22161. Copies of active and draft guides and the
Standard Review Plan are available for inspection or copying for a fee
from the NRC Public Document Room at 2120 L Street NW., Washington, DC;
the PDR's mailing address is Mail Stop LL-6, Washington, DC 20555;
telephone (202) 634-3273; fax (202) 634-3343. Regulatory guides are not
copyrighted, and Commission approval is not required to reproduce them.
I. Background
On August 16, 1995, the Commission published in the Federal
Register a final policy statement on the Use of Probabilistic Risk
Assessment Methods in Nuclear Regulatory Activities (60 FR 42622). The
policy statement included the following policy regarding NRC's expanded
use of PRA:
1. The use of PRA technology should be increased in all regulatory
matters to the extent supported by the state-of-the-art in PRA methods
and data and in a manner that complements the NRC's deterministic
approach and supports the NRC's traditional defense-in-depth
philosophy.
2. PRA and associated analyses (e.g., sensitivity studies,
uncertainty analyses, and importance measures) should be used in
regulatory matters, where practical within the bounds of the state-of-
the-art, to reduce unnecessary conservatism associated with current
regulatory requirements, regulatory guides, license commitments, and
staff practices. Where appropriate, PRA should be used to support
proposals for additional regulatory requirements in accordance with 10
CFR 50.109 (Backfit Rule). Appropriate procedures for including PRA in
the process for changing regulatory requirements should be developed
and followed. It is, of course, understood that the intent of this
policy is that existing rules and regulations shall be complied with
unless these rules and regulations are revised.
3. PRA evaluations in support of regulatory decisions should be as
realistic as practicable and appropriate supporting data should be
publicly available for review.
4. The Commission's safety goals for nuclear power plants and
subsidiary numerical objectives are to be used with appropriate
consideration of uncertainties in making regulatory judgments on the
need for proposing and backfitting new generic requirements on nuclear
power plant licensees.
It was the Commission's intent that implementation of this policy
statement would improve the regulatory process in three areas:
1. Enhancement of safety decision making by the use of PRA
insights.
2. More efficient use of agency resources, and
3. Reduction in unnecessary burdens on licensees.
[[Page 44660]]
In parallel with the development of Commission policy on uses of
risk assessment methods, the NRC developed an agency-wide
implementation plan for applying PRA insights within the regulatory
process (SECY-95-079). This implementation plan included tasks to
develop a series of regulatory guides and standard review plans (SRPs)
on general guidance, inservice inspection (ISI), inservice testing
(IST), technical specifications (TS), and graded quality assurance
(GQA).
The general regulatory guide, Regulatory Guide 1.174, and its
accompanying SRP section, Chapter 19, are intended to help implement
the Commission's August 1995 policy on the use of risk information in
the regulatory process. These two general documents are the first in
the series of risk-informed guidance documents. Together, they provide
the basic framework for an approach acceptable to the NRC staff for use
by power reactor licensees in preparing proposals for plant-specific
changes to their licensing bases using risk information. Alternative
approaches may be proposed. Application-specific guidance documents for
risk-informed technical specifications, inservice testing, and graded
quality assurance are currently being revised to address the public
comments that were received; these documents are scheduled to be issued
later in 1998. Guidance for inservice inspection is also being
developed on a later schedule.
II. Public Comment Summary and Resolution
The public comments on the draft regulatory guidance documents on
risk-informed applications were due by September 30, 1997. In addition
to comments received at the workshop, the NRC staff received
approximately 40 sets of written comments. Some of the more extensive
comments were provided by the Nuclear Energy Institute (NEI), in a
letter dated September 29, 1997, which provided comments on behalf of
the nuclear industry. In its letter, NEI commended the NRC staff for
its efforts in developing the draft documents, stating that the
industry recognized the significance of the drafts in articulating a
framework for the use of risk information in regulatory decisionmaking
and that the documents represent a milestone in the evolution of the
regulatory process. In addition, the NEI letter expressed concern
regarding four policy issues; NEI believes the resolution of these
issues is essential to the continued viability and the expansion of
risk-informed regulation. The issues cited by NEI were overall cost
benefit, use of numerical acceptance guidelines, treatment of
uncertainty, and PRA attributes and quality considerations. Each of
these areas highlighted by NEI will be addressed in the following
discussion of the principal issues.
Comment letters were also received from the Electric Power Research
Institute (EPRI), the American Society of Mechanical Engineers (ASME),
the owners groups for the four reactor vendors (General Electric,
Westinghouse, Combustion Engineering, and Babcock and Wilcox), one
vendor (Westinghouse), 18 electric utilities, one national laboratory
(Oak Ridge), five technical organizations, five other private industry
organizations or individuals, and two anonymous commenters. The
following discussion addresses the resolution of the principal issues
raised by the commenters. A more complete discussion of the comments
received overall is given in the attachment to a memorandum from Mr.
Mark A. Cunningham (Chief, Probabilistic Risk Analysis Branch, Division
of Systems Technology, Office of Nuclear Regulatory Research) to Mr. M.
Wayne Hodges (Director, Division of Systems Technology, Office of
Nuclear Regulatory Research) dated January 7, 1998, which is available
in the NRC's Public Document Room. The discussion in the attachment
covers the resolution of the NRC's specific requests for comments
included in the Federal Register notice for the workshop (62 FR 34321),
other issues raised by the commenters, and the principal issues
discussed in this announcement.
Principal Issues
1. Use of 10-4 Per Reactor-Year Core Damage Frequency as an
Acceptance Guideline
Issue: Comments were received indicating that the use of
10-4 per reactor-year core damage frequency
(10-4/RY CDF) as an acceptance guideline was overly
conservative, that the Commission's Safety Goal Policy quantitative
health objectives (QHOs) would be more appropriate for use as goals,
and that it was not clear how closely staff reviewers would hold
applications to this numerical criteria.
Resolution: Revised Section 2.2.4, ``Acceptance Guidelines,'' of
Regulatory Guide 1.174 addresses the use of 10-4/RY CDF as a
guideline in evaluating the acceptability of risk-informed
applications. The use of 10-4/RY CDF as a subsidiary goal is
consistent with past Commission guidance. The guidelines for assessing
risk, contained in the regulatory guide and SRP, are based upon the
QHOs in the Commission's Safety Goal Policy and upon previous
Commission guidance related to implementation of the Safety Goal Policy
and regulatory analysis guidelines (Revision 2 of NUREG/BR-0058,
``Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory
Commission,'' USNRC, November 1995). Specifically, the guideline value
of 10-4/RY for CDF is based upon a June 15, 1990, memorandum
from the Commission to the NRC staff on implementation of the Safety
Goal Policy, which established a 10-5/RY CDF as a benchmark
objective for accident prevention. The guideline value on CDF
of 10-5/RY is based upon the guidance in the Commission's
regulatory analysis guidelines, which establish 10-5/RY
CDF as a cutoff below which the significance of safety issues
is not large enough to warrant backfit analysis, assuming a reasonable
accident mitigation capability.
Accident mitigation capability is addressed via guidelines on large
early release frequency (LERF). The guideline value of 10-5/
RY for LERF contained in Regulatory Guide 1.174 is based upon risk
analysis results presented in NUREG-1150 (``Severe Accident Risks: An
Assessment for Five U.S. Nuclear Power Plants,'' Vol. 3, USNRC, January
1991), which calculated offsite health risks for five nuclear power
plants and compared them to the Safety Goal QHOs. Analyses for all five
plants calculated health risks well below the QHOs. However, if the
results of this analysis were adjusted so that the offsite health risks
just met the early fatality QHO (the most limiting QHO), with allowance
for the unanalyzed modes of operation (shutdown), and in some cases
external events, a corresponding LERF value of 10-5/RY would
result for those plants whose calculated offsite health risks are
closest to the QHOs.
Site-to-site variations in LERF were judged to be not a large
factor (this was also confirmed in a study reported by the Advisory
Committee for Reactor Safeguards in a September 19, 1997, letter to
Chairman Jackson), and thus a single value for all plants is used. The
guideline value of 10-6/RY for ``LERF is based upon the
regulatory analysis guidelines that, when used in conjunction with the
CDF guidelines discussed above, establish a cutoff below
which the significance of safety issue is not large enough to warrant
backfit analysis.
Figures 3 and 4 of Section 2.2.4 in Regulatory Guide 1.174
illustrate acceptance guidelines for CDF and
[[Page 44661]]
LERF and indicate that for each of these metrics, three regions have
been identified for use in screening the acceptability of proposed
changes in licensing bases. Region III, shown in the figures and
discussed in the text of Regulatory Guide 1.174, has been identified as
representing a sufficiently low CDF or LERF increase that, in general,
program changes associated with this region may be permitted without a
detailed assessment of the baseline CDF/LERF. As discussed in
Regulatory Guide 1.174, if there are indications that the baseline CDF
or LERF is above the guideline values, additional evaluation would be
needed even though the calculated changes in CDF or LERF were small and
in Region III. In Section 2.2.5, ``Comparison of PRA Results with the
Acceptance Guidelines,'' it is stated that the acceptance guidelines
(lines separating the regions) are not to be interpreted in an overly
prescriptive manner and that they are intended to provide an
indication, in numerical terms, of what is considered acceptable.
Graduated shading has been added to the guideline figures to indicate
regions in which proposed changes will be subject to gradually more
intensive NRC technical and management review. Regarding the use of the
QHOs, it is stated that the use of the QHOs in lieu of LERF in support
of risk-informed applications is an acceptable approach provided that
appropriate consideration is given to the methods and assumptions used
in the analysis and in the treatment of uncertainties. Also, in Section
2.2.6, ``Integrated Decisionmaking,'' it is noted that Level 3 PRA
information can be submitted and will be considered in support of those
cases in which increased NRC management attention is needed during the
review (e.g., when the calculated CDF/LERF changes and baseline values
are close to the acceptance guidelines).
2. Definition of Risk Neutral
Issue: A number of comments were received indicating a need for a
definition of risk neutral applications, and indicating that increased
NRC management and technical review should not be required for risk
increases below some threshold.
Resolution: See responses to Issues Number 1 and 3 addressing very
small increases in risk.
3. Allowance for Very Small Increases in Risk
Issue: Comments stated that facilities with CDFs greater than
10-4/RY should be allowed small risk increases and that the
level of effort and information required in submittals was excessive
for small risk increases.
Resolution: Section 2.2.4, ``Acceptance Guidelines,'' addresses the
treatment of small increases in risk using the metrics of CDF and LERF.
As noted in the discussion for Issue Number 1, this section has been
revised and now includes a special category of application in which the
estimated level of CDF/LERF increase associated with the application is
sufficiently low that, in general, program changes associated with this
region may be permitted without a detailed assessment of the baseline
CDF/LERF. This category is displayed in Figures 3 and 4 of Section
2.2.4.
4. Treatment of Uncertainties
Issue: Comments stated that the inclusion of uncertainty could lead
to confusion regarding the decision criteria and that the use of PRA
inherently takes care of uncertainty.
Resolution: Several approaches were reconsidered for the treatment
of uncertainties, and it was concluded that the approach described in
Draft Regulatory Guide DG-1061 appeared to be the most practical and
useful approach at this time, although the text needed to be clarified.
Uncertainty is addressed in Section 2.2.5, ``Comparison of PRA Results
with the Acceptance Guidelines,'' in Regulatory Guide 1.174. In this
section, it is noted that it is important, when interpreting the
results of a PRA, to develop an understanding of the impact of a
specific assumption or choice of model on the prediction. PRA only
inherently takes care of those uncertainties modeled in the analysis.
Others must be qualitatively or quantitatively addressed. The impact of
using alternative assumptions and models may be reasonably evaluated
using appropriate sensitivity studies. The major sources of uncertainty
should be understood, but it is not necessary, in all cases, to perform
elaborate uncertainty evaluations (e.g, propagation of uncertainty
distributions).
5. Quality of PRA
Issue: Numerous comments were received indicating concern that the
PRA standards included in Draft NUREG-1602, ``The Use of PRA in Risk-
Informed Applications'' (USNRC, June 1997), were unnecessarily high for
many risk-informed applications. The comments also indicated that the
requirements for PRA quality were not clear and that graded levels of
PRA quality should be provided for different applications.
Resolution: The issue of PRA quality is addressed in the revised
Section 2.2.3, ``Scope, Level of Detail, and Quality of the PRA,'' of
Regulatory Guide 1.174. In this section it is stated that PRA quality
should be commensurate with the application for which it is intended
and with the role that PRA results would play in the integrated
decision process. A PRA used in a risk-informed application should be
performed in a manner that is consistent with accepted practices, and
it should be commensurate with the scope and level of detail, which are
also discussed in Section 2.2.3 of Regulatory Guide 1.174. The NRC has
not developed its own formal standard nor endorsed an industry standard
for PRA quality, but it supports such a standard and expects that one
will be available in the future. Draft NUREG-1602 was cited in Draft
Regulatory Guide DG-1061 as a potential reference for PRA methods that
could be used to support regulatory decisionmaking. There were a number
of comments indicating that the ``PRA standard'' represented by Draft
NUREG-1602 was excessive for many risk-informed applications that did
not require sophisticated or state-of-the-art methods. While Draft
NUREG-1602 was not intended to be used universally as a PRA standard,
it is acknowledged that it would be more useful to have a standard that
addresses the differing needs for PRA scope and detail depending on the
application. Accordingly, Draft NUREG-1602 is no longer referenced in
Regulatory Guide 1.174, and a separate discussion on PRA quality has
been added, including the use of peer reviews or PRA cross comparisons.
PRA peer review activities such as those presently being done under
various industry PRA certification programs are examples. Peer review,
PRA certification, or cross comparison do not replace a staff review in
its entirety, and licensees need to justify why the PRA is adequate for
the proposed application. In the interim, until a consensus PRA
standard is available, the NRC staff will evaluate PRAs submitted in
support of specific applications using the guidelines given in Chapter
19 of the Standard Review Plan.
6. Low Safety Significant Components Monitoring Needs
Issue: Comments indicated that the draft guidance placed too much
importance on monitoring low safety significant components (LSSCs). The
comments also indicated that monitoring performed under the Maintenance
Rule should be acceptable for risk-informed programs.
[[Page 44662]]
Resolution: Section 2.3, ``Element 3: Define Implementation and
Monitoring Program,'' has been revised to clarify the need for
monitoring LSSCs. While details for monitoring LSSCs will be provided
in the application-specific guidance documents, the following principal
needs should be satisfied for all applications. Monitoring programs
should be proposed that are capable of adequately tracking the
performance of equipment that, when degraded, could alter the
conclusions that were key to supporting the acceptance of the program.
It follows that monitoring programs should be structured such that SSCs
are monitored commensurate with their safety significance. Monitoring
that is performed as a part of the Maintenance Rule implementation can
be used when the monitoring performed under the Maintenance Rule is
sufficient for the SSCs affected by the risk-informed application.
7. Shutdown and Temporary Plant Condition
Issue: Several commenters noted that the guidelines proposed did
not distinguish between power operation and shutdown and did not
address temporary plant conditions. Separate guidelines for these
conditions were suggested.
Resolution: In response to these comments, Section 2.2.4 of
Regulatory Guide 1.174 has been expanded to address the shutdown
condition. Specific guidance for temporary plant conditions has not
been added, but will be considered in a future update of the guide.
8. Documentation Needs
Issue: Many commenters stated that the requirements in the drafts
for documentation were excessive and unmanageable, particularly for
proposals involving small changes in risk. It was also suggested that
certain items of documentation should not be required to be submitted
for the staff's initial review, provided that more complete
documentation was maintained at the utility for review as necessary.
Resolution: In response to the comments received, Section 3 of
Regulatory Guide 1.174 has been reevaluated to determine whether all
items listed in the draft were necessary. As a result, a number of
documentation items, particularly with regard to the PRA, have been
removed in the final regulatory guide, and the SRP has been revised to
be consistent.
9. Overall Cost Benefit
Issue: This issue was highlighted by NEI in its comment letter and
was also included in a number of other comment letters. A concern was
expressed that the resources required by licensees to prepare proposals
and to subsequently implement NRC-approved risk-informed changes to the
CLB would be too high considering the benefit in terms of burden
reduction.
Resolution: The question of how cost beneficial it would be for
utilities to prepare proposals for risk-informed changes to their
licensing bases and to implement such programs after review and
approval by the NRC will only be fully answered after the industry and
the NRC gain further experience in these types of programs. Certainly,
the pilot plant program proposals, which are currently being reviewed
for application to technical specifications, graded quality assurance,
and inservice testing and inspection, will provide useful insights into
the potential cost savings of these programs. While it is not the NRC's
responsibility to ensure that such risk-informed programs are cost
beneficial, it is believed that such programs can enhance safety by
better focusing utility and NRC resources on the most important safety
areas in reactors; this philosophy is consistent with the Commission's
Policy Statement on the use of PRA methods in nuclear regulatory
activities. During the preparation of this final regulatory guide and
standard review plan section, attention was paid to areas in which
needs for utility resources could be reduced, thus the cost beneficial
aspects of the risk-informed process were improved while still
maintaining an appropriate level of safety. Examples in Regulatory
Guide 1.174 are Section 2.2.3, ``Scope, Level of Detail, and Quality of
the PRA,'' which states that the level of detail required to support an
application can vary depending on the application, and not all
applications require an expensive, detailed PRA; Section 2.2.4,
``Acceptance Guidelines,'' identifies a special category of risk-
informed proposal as having a sufficiently low estimated risk increase
that, generally, the proposal would be considered without a detailed
assessment of baseline CDF/LERF (i.e., Region III of Figures 3 and 4 in
Regulatory Guide 1.174); and in Section 3, ``Documentation,'' where
some of the items that were identified in the draft guide and SRP as
being needed in program submittals have been removed since they were
not believed necessary.
(5 U.S.C. 552(a))
Dated at Rockville, MD, this 31st day of July 1998.
For the Nuclear Regulatory Commission.
Ashok C. Thadani,
Director, Office of Nuclear Regulatory Research.
[FR Doc. 98-22412 Filed 8-19-98; 8:45 am]
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