[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 The draft guidance documents were being 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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