[Federal Register Volume 62, Number 199 (Wednesday, October 15, 1997)]
[Notices]
[Pages 53663-53667]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-27235]


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NUCLEAR REGULATORY COMMISSION


Draft Regulatory Guide and Standard Review Plan Section; 
Issuance, Availability, and Notice of Workshop

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice of availability and workshop.

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SUMMARY: The Nuclear Regulatory Commission has issued for public 
comment drafts of a regulatory guide and a Standard Review Plan 
Section. These issuances follow the Commission's August 16, 1995 (60 FR 
42622) policy statement on the ``Use of PRA Methods in Nuclear 
Regulatory

[[Page 53664]]

Activities.'' In June 1997, the NRC published for public comment (62 FR 
34321) four draft guides, 3 standard review plans and a NUREG series 
document on the use of PRA in nuclear power reactor licensing. The NRC 
is developing guidance for power reactor licensees on acceptable 
methods for using probabilistic risk assessment (PRA) information and 
insights in support of plant-specific applications to change the 
current licensing basis (CLB) for inservice inspection of piping, known 
as risk-informed inservice inspection (RI-ISI) programs. The use of 
such PRA information and guidance will be voluntary. To facilitate 
comment, the Commission will conduct, a workshop to explain the draft 
documents and answer questions. Section VI of this notice provides 
additional information on the scope, purpose and topics for discussion 
at the workshop.

DATES: The workshop will be held on November 19-20, 1997, Registration 
begins on November 18 at 3:00 p.m. The comment period expires January 
13, 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. Mail written 
comments to: Rules and Directives Branch, Office of Administration, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. Please 
(1) attach a diskette containing your comments, in either ASC11 text or 
Wordperfect format (Version 5.1 or 6.1), (2) or submit your comments 
electronically via the NRC Electronic Bulletin Board on FedWorld or the 
NRC's interactive rulemaking Website.
    Deliver comments to 11545 Rockville Pike, Rockville, Maryland, 
between 7:30 a.m. and 4:15 p.m. on Federal workdays.
    Requests for free single copies of draft regulatory guide and 
standard review plan, to the extent of supply, may be made in writing 
to the Printing, Graphics and Distribution Branch, U.S. Nuclear 
Regulatory Commission, Washington, D.C. 20555-0001, or by fax to (301) 
415-5272. Copies of draft regulatory guide and the standard review plan 
section are available for inspection and copying for a fee at the NRC 
Public Document Room, 2120 L street N.W. (Lower Level), Washington, 
D.C. 20555-0001. Electronic copies of the draft document are also 
accessible on the NRC's interactive rulemaking web site through the NRC 
home page (http://www.nrc.gov). This site includes a facility to upload 
comments as files (any format), if your web browser supports the 
function.
    For more information on the NRC bulletin boards call Mr. Arthur 
Davis, Systems Integration and Development Branch, NRC, Washington, 
D.C. 20555-0001, telephone (301) 415-5780; e-mail [email protected]. For 
information about the interactive rulemaking site, contact Ms. Carol 
Gallagher, (301) 415-5905; e-mail [email protected].

ADDRESSES: The public workshop will be held at the Bethesda Marriott, 
5151 Pooks Hill Road, Bethesda, Maryland; telephone (301) 897-9400.

FOR FURTHER INFORMATION CONTACT: Jack Guttmann, Office of Nuclear 
Regulatory Research, MS: T10-E50, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, (301) 415-7732, E-mail [email protected].

SUPPLEMENTARY INFORMATION:

I. Background

    On August 16, 1995, the Commission published in the Federal 
Register (60 FR 42622) a final policy statement on the use of 
probabilistic risk assessment methods in nuclear regulatory activities. 
The policy statement included the following 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 decisionmaking by the use of PRA insights,
2. More efficient use of agency resources, and
3. Reduction in unnecessary burdens on licensees.

    To help implement the Commission's PRA Policy Statement, draft 
regulatory guides and Standard Review Plans (SRP) were developed in the 
areas of:

--General guidance,
--Inservice inspection (ISI),
--Inservice testing (IST),
--Technical specification (TS), and
--Graded quality assurance (GQA).

    The draft regulatory guides provide a proposed acceptable approach 
for power reactor licensees to prepare and submit applications for 
plant-specific changes to the current licensing basis that utilize risk 
information. The draft standard review plans provide guidance to the 
NRC staff on the review of such applications. On June 25, 1997, all but 
the ISI draft regulatory guide and SRP were published for public 
comment (62 FR 34321).
    This notice specifically seeks public comment on Draft Regulatory 
Guide DG-1063, ``An Approach for Plant-Specific Decisionmaking: 
Inservice Inspection of Piping,'' and the accompanying draft Standard 
Review Plan Section 3.9.8, ``Standard Review Plan for the Review of 
Risk-Informed Inservice Inspection of Piping.'' These documents are 
discussed in more detail below.
    The draft guide and SRP are being developed to provide guidance to 
power reactor licensees and NRC staff reviewers on integrating risk 
information to support requests for changes in a plant's CLB for 
inservice inspection of piping. The regulatory guide describes a means 
by which licensees can propose plant-specific CLB changes under 10 CFR 
50.55a(a)(3)(I). Adopting the approach in this regulatory guide would 
be voluntary. Licensees submitting applications for changes to their 
CLB may use this approach or an equivalent approach. To encourage the 
use of risk

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information in inservice inspection programs of piping, the staff 
intends to give priority to applications for burden reduction that use 
risk information as a supplement to traditional engineering analyses, 
consistent with the intent of the Commission's policy. All applications 
that improve safety will continue to receive high priority.
    DG-1061, ``An Approach for Using Probabilistic Risk Assessment in 
Risk-Informed Decisions on Plant-Specific Changes to the Current 
Licensing Basis,'' and the draft SRP of Chapter 19 were developed to 
provide an overall framework and guidance that is applicable to any 
proposed CLB change when risk insights are used to support the change 
(62 FR 34321). The application-specific regulatory guide (RG) and SRP 
for ISI would build upon and supplement the general guidance contained 
in DG-1061 and provide additional guidance specific to inservice 
inspection programs of piping.
    The guidance provided in these documents is designed to encourage 
licensees to use risk information by defining an acceptable framework 
for the use and integration of risk information on a plant-specific 
basis, while promoting consistency in PRA applications. It is expected 
that the long-term use of risk information in plant-specific licensing 
actions will result in improved safety by focusing attention on the 
more risk-significant aspects of plant design and operation. The draft 
guidance highlights to licensees acceptable methods and scope of 
analysis required to support the proposed changes to the plant's CLB.

II. Policy Issues

    On May 15, 1996, the Commission requested the staff to recommend 
resolution of the following four policy issues associated with risk-
informed changes to a plant's CLB:

 The role of performance-based regulation,
 Plant-specific application of safety goals,
 Risk neutral vs. increases in risk,
 Implementation of changes to risk-informed IST and ISI 
requirements.

    These issues are applicable to RI-ISI programs. Public comments on 
these issues were requested in the June 25, 1997 FRN (62 FR 34321) 
under the heading, ``Use of PRA in Plant Specific Reactor Regulatory 
Activities: Proposed Regulatory Guides, Standard Review Plan Sections, 
and Supporting NUREG.'' Comments provided on these issues in response 
to the June 25 FRN on related guides will be used by the staff in 
finalizing this guide as well. Comments on these issues as they 
specifically apply to this guide are also requested.

III. Structure, Guidelines and Rationale for RG/SRP

    The approach described in the DG-1063 and the draft SRP has four 
basic steps. These are:

--Define the proposed change;
--Perform an integrated engineering analysis (which includes both 
traditional engineering and risk analysis) and use an integrated 
decision process;
--Perform monitoring and feedback to verify assumptions and analysis; 
and
--Document and submit proposed change.

    Five fundamental safety principles are described that should be met 
in each application for a change in the CLB. These principles are.

--The proposed change meets the current regulation. This principle 
applies unless the proposed change is explicitly related to a requested 
exemption or rule change (i.e., a 10 CFR 50.12 ``specific exemption'' 
or a 10 CFR 2.802 ``petition for rulemaking'');
--Defense-in-depth is maintained;
--Sufficient safety margins are maintained;
--Proposed increases in risk, and their cumulative effect, are small 
and do not cause the NRC safety goals to be exceeded;
--Performance-based implementation and monitoring strategies are 
proposed that address uncertainties in analysis models and data and 
provide for timely feedback and corrective action.

    These principles represent fundamental safety practices that the 
staff believes must be retained in any change to a plant's CLB to 
maintain reasonable assurance that there is no undue risk to public 
health and safety. Each of these principles is to be considered in the 
analysis and integrated decisionmaking process.
    The guidelines for assessing risk proposed in the draft guide and 
draft SRP are derived from the Commission's safety goal quantitative 
health objectives (QHOs). Specifically, the subsidiary objectives of 
core damage frequency (CDF) and large early release frequency (LERF) 
are used as the measures of risk against which changes in the CLB will 
be assessed, in lieu of the QHOs themselves, which require level 3 PRA 
information (offsite health effects). These measures were chosen to 
simplify the scope of PRA analysis needed, to avoid the large 
uncertainties associated with level 3 PRA analysis, and to be 
consistent with previous Commission direction to decouple siting from 
plant design. These values are described in the June 25, 1997 Federal 
Register Notice (62 FR 34321) on ``Use of PRA in Plant Specific Reactor 
Regulatory Activities: Proposed Regulatory Guides, Standard Review Plan 
Sections, and Supporting NUREG.''

IV. Comments

    The staff is soliciting comments related to the guidance described 
in the draft regulatory guide DG-1063 and SRP Section 3.9.8. Comments 
submitted by the readers of this FRN will help ensure that these draft 
documents have appropriate scope, depth, quality, and effectiveness. 
Alternative views, concerns, clarifications, and corrections expressed 
in public comments will be considered in developing the final 
documents.

V. Workshop

    The Commission will conduct a workshop on November 19 and 20, 1997, 
to discuss and explain the material contained in the draft guide and 
SRP, and to answer questions and receive comments and feedback on the 
proposed documents. The purpose of the workshop is to facilitate the 
comment process. In the workshop, the staff will describe each 
document, its basis, and solicit comment and feedback on its 
completeness, correctness and usefulness. Since these documents cover a 
wide range of technical areas, many topics will be discussed. Listed 
below are topics on which discussion and feedback are sought at the 
workshop:
    (A) Is the level of detail in the guidance contained in the 
proposed regulatory guide and SRP clear and sufficient, or is more 
detailed guidance necessary? What level of detail is needed.
    (B) Is it acceptable to use qualitative information (e.g., not 
quantifying the change in risk--CDF and LERF) to 
propose changes in ISI programs? If so, does DG-1063 provide adequate 
guidance in this regard? Can qualitative assessments be used to 
identify and categorize piping segments as high, medium and low safety 
significant? How? What are the limitations of such an approach?
    (C) Under the risk-informed approach, what is the appropriate size 
of the sample of welds or piping segment areas that should be 
inspected? What should the criteria be for selecting the sample size?
    (D) How should welds or piping segment areas in the inspection 
sample be selected for inspection: randomly,

[[Page 53666]]

those most likely to experience degradation, or some combination of 
random and possible degradation? What would be the basis for the 
recommended selection process?
    (E) Once selected, should the same welds or piping segment areas be 
inspected at each inspection interval or should different welds or 
piping segment areas be included in the sample? What would be the 
basis?
    (F) DG-1063 proposes a method for meeting the criteria for 
acceptable safety and quality, as addressed in 10 CFR 50.55a(a)(3)(I). 
That method applies leak frequency target goals to maintain piping 
performance levels at or improved over the existing performance 
observed when implementing ASME Section XI requirements. Are there 
other acceptable risk-informed means by which to meet the criteria in 
10 CFR 50.55a(a)(3)(I)?
    (G) Should the scope of DG-1063 permit licensees to propose ISI 
changes to selected systems, in lieu of assessing the entire piping in 
the plant? For example, would it be acceptable for a licensee to limit 
its analysis to Class 1 piping (reactor coolant system piping) and not 
consider other piping in the plant? Such an analysis would not provide 
information required for categorizing piping in the plant and thereby 
grading the inspection based on plant risk. It would also discourage 
the use of risk-insights (e.g., PRA) to identify risk-significant 
piping within the plant. How can the concept of assessing risk in an 
integrated fashion be maintained if the scope were limited to one or a 
limited number of systems, such as Class 1 piping. What is gained by 
analyzing all the systems versus only selected systems? What is lost by 
minimizing the scope?
    (H) The decision metrics described in Attachment 2 to DG-1063 
identify a 2-by-2 matrix for identifying a graded approach to 
inspection based on risk and failure potential. Piping segments 
categorized as high-safety-significant and high-failure-potential 
receive more inspections than segments categorized as high-safety-
significant and low-failure-potential. The number of inspections for 
the high-safety-significant and low-failure-potential segments is based 
on meeting target leak frequency goals and incorporates uncertainties 
in the probability of detection. What other methods are available to 
provide a comparable level of quality and safety? What are the 
technical bases for those other methods?
    (I) How should the time dependence of degradation mechanisms be 
accounted for in selecting inspection intervals and categorizing the 
safety significance of pipe segments?
    (J) On what basis could the requirement for ISI be eliminated? For 
example, if a detailed engineering analysis identifies a Class 1 or 2 
piping segment as low-safety-significant and low-failure-potential, is 
it acceptable to eliminate the requirement for ISI or should a Class 1 
or a 2 pipe segment be considered part of the defense-in-depth 
consideration and be required to have some level of inspection 
regardless of its categorization as low-safety-significant and low-
failure potential? If yes, why? If not, why not?
    (K) Are data bases available on degradation mechanisms and 
consequences of piping failures? Is data available to identify the 
secondary effects that can result from a pipe break, such as high-
energy pipe whip damaging other piping and components in the vicinity 
of the break? What are the industry's plans for developing and 
maintaining an up-to-date data base on plant piping performance? Should 
a commitment to develop and maintain such a data base be required for a 
RI-ISI program? How could it be ensured that the data base is 
maintained?
    (L) Does the application of the Perdue-Abramson model (DG-1063, 
Attachment 4), with the use of the decision metrics and leak frequency 
goals (DG-1063, Attachment 2) provide an alternative acceptable level 
of quality and safety as required by 10 CFR 50.55a(a)(3)(I)? 
Alternatively, should there be a leak frequency goal independent of 
core damage frequency goal, as a measure of defense in depth?
    (M) Is the guidance proposed by the staff for finding a fracture 
mechanics computer model acceptable for use in RI-ISI programs clear 
and adequate? If not, what is missing?
    (N) Is the guidance on risk categorization clear and sufficient, or 
is additional guidance needed? What additional guidance is needed?
    (O) Table A5.1, in DG-1063, identifies a proposed checklist that 
could assist in identifying potential locations for various degradation 
mechanisms in a pipe. Is this checklist complete? What additional 
information could enhance the usefulness of such a check list?

Workshop Meeting Information

    A 2-day workshop will be held to obtain public comment on the 
subject draft Regulatory Guide (DG-1063) and the accompanying draft 
standard review plan (Section 3.9.8), and to respond to questions. 
Persons other than NRC staff and NRC contractors interested in making a 
presentation at the workshop should notify Jack Guttmann, U.S. Nuclear 
Regulatory Commission, MS T10E50, phone (301) 415-7732, e-mail 
[email protected]. Comments on the regulatory guidance and standard review 
plan documents for discussion at the workshop should be submitted in 
writing and in electronic mail ([email protected]) in WordPerfect 5 or 6.1 
compatible format.

Date: November 19-20, 1997.
Agenda: Preliminary agenda is as follows: (A final agenda will be 
available at the workshop.).

Tuesday, November 18, 1997

    Time--3:00 pm to 7:00 pm. Registration.

Wednesday, November 19, 1997

    Time--7:00 am to 4:00 pm. Registration.
Session 1: (Morning 11/19/97--8:00 am-11:30 am)
    Overview by NRC management of the draft regulatory guide and 
standard review plan, followed by NRC staff presentation on the draft 
documents (DG-1063 and SRP Section 3.9.8).
    Lunch: 11:30 am--1:00 pm.
    Session 2: (Afternoon 11/19/97--1:00 pm-5:00 pm)
    Public/Industry presentations on issues and recommendations for the 
general guidance documents, followed by open discussions.

Friday, November 20, 1997

Session 3: (Morning 11/20/97--8:00 am-11:30 pm)
    Open discussion of issues.
Session 4: (Afternoon 11/20/97--1:00 pm-3:00 pm)
    Overview of comments, issues and resolution options identified in 
the sessions. Concluding remarks and near-term plans will be covered by 
the staff.
    Location: Bethesda, Maryland.
    Hotel: Bethesda Marriott, 5151 Pooks Hill Road, Bethesda, Maryland, 
(301) 897-9400.
    Registration: There is no registration fee for this workshop. 
However, we request that interested parties register in writing to 
Kesselman-Jones, 8912 James Ave., NE., Albuquerque, New Mexico 87111 
their intent on participating in the workshop. Please include name, 
organization, address and phone number with your registration request. 
Notification of attendance (e.g., pre-registration) is requested so 
that adequate space, etc. for the workshop can be arranged. Questions 
regarding meeting registration or fees should be directed to Kesselman-
Jones, Phone (505) 271-0003, fax (505) 271-0482, e-mail 
[email protected].

[[Page 53667]]

VI. Regulatory Analysis

1. Statement of the Problem

    During the past several years, both the Commission and the nuclear 
industry have recognized that probabilistic risk assessment (PRA) has 
evolved to the point that it can be used increasingly as a tool in 
regulatory decisionmaking. In August 1995 the Commission published a 
policy statement that articulated the view that increased use of PRA 
technology would (1) enhance regulatory decisionmaking, (2) allow for a 
more efficient use of agency resources, and (3) allow a reduction in 
unnecessary burdens on licensees. In order for this change in 
regulatory approach to occur, guidance must be developed describing 
acceptable means for increasing the use of PRA information in the 
regulation of nuclear power reactors.

2. Objective

    To provide guidance to power reactor licensees and NRC staff 
reviewers on acceptable approaches for utilizing risk information (PRA) 
to support requests for changes in a plant's current licensing basis 
(CLB). It is intended that the changes in regulatory approach addressed 
by this guidance should allow a focussing of both industry and NRC 
staff resources on the most important regulatory areas while providing 
for a reduction in burden on the resources of licensees. Specifically, 
guidance is to be provided in several areas that have been identified 
as having potential for this application. This application includes 
risk-informed inservice inspection programs of piping.

3. Alternatives

    The increased use of PRA information as described in the draft 
regulatory guide being developed for this purpose is voluntary. 
Licensees can continue to operate their plants under the existing 
procedures defined in their CLB. It is expected that licensees will 
choose to make changes in their current licensing bases to use the new 
approaches described in the draft regulatory guide only if it is 
perceived to be to their benefit to do so.

4. Consequences

    Acceptance guidelines included in the draft regulatory guide state 
that only small increases in overall risk are to be allowed under the 
risk-informed program. Reducing the inspection frequency of piping 
identified to represent low risk and low failure potential as provided 
for under this program is an example of a potential contributor to a 
small increase in plant risk. However, the program also requires 
increased emphasis on piping categorized as high-safety-significant and 
high-failure-potential that may not be inspected under current 
programs. This is an example of a potential contributor to decreases in 
plant risk. An improved prioritization of industry and NRC staff 
resources, such that the most important areas associated with plant 
safety receive increased attention, should result in a corresponding 
contributor to a reduction in risk. Some of the possible impacts on 
plant risk cannot be readily quantified using present PRA techniques 
and must be evaluated qualitatively. The staff believes that the net 
effect of the risk changes associated with the risk-informed programs, 
as allowed using the guidelines in the draft regulatory guide, should 
result in a very small increase in risk, maintain a risk-neutral 
condition, or result in a net risk reduction in some cases.

5. Decision Rationale

    It is believed that the changes in regulatory approach provided for 
in the draft regulatory guide being developed will result in a 
significant improvement in the allocation of resources both for the NRC 
and for the industry. At the same time, it is believed that this 
program can be implemented while maintaining an adequate level of 
safety at the plants that choose to implement risk-informed programs.

6. Implementation

    It is intended that the risk-informed regulatory guide on inservice 
inspection of piping (DG-1063) be published by early to mid CY 1998.

    Dated at Rockville, Maryland, this 8th day of October 1997.

    For the Nuclear Regulatory Commission.
Mark A. Cunningham,
Chief, Probabilistic Risk Analysis Branch.
[FR Doc. 97-27235 Filed 10-14-97; 8:45 am]
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