Federal Register / Vol. 62, No. 199 / Wednesday, October 15, 1997 / Notices 53663

of any changes in schedule, etc., that

may have occurred.

Dated: October 8, 1997.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 97–27239 Filed 10–14–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Advisory Committee on Reactor

Safeguards; Subcommittee Meeting on

Safety Research Program; Notice of

Meeting

The ACRS Subcommittee on Safety

Research Program will hold a meeting

on November 4–5, 1997, Room T–2B3,

11545 Rockville Pike, Rockville,

Maryland.

The entire meeting will be open to

public attendance.

The agenda for the subject meeting

shall be as follows:

Tuesday, November 4, 1997—8:30

a.m. until the conclusion of business.

Wednesday, November 5, 1997—8:30

a.m. until the conclusion of business.

The Subcommittee will discuss the

NRC Safety Research Program, industry

research activities, and related matters,

and gather information for preparing a

draft annual report to Congress. The

purpose of this meeting is to gather

information, analyze relevant issues and

facts, and to formulate proposed

positions and actions, as appropriate,

for deliberation by the full Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff engineer

named below five days prior to the

meeting, if possible, so that appropriate

arrangements can be made.

During the initial portion of the

meeting, the Subcommittee, along with

any of its consultants who may be

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittee will then hear

presentations by and hold discussions

with representatives of the NRC staff,

Nuclear Energy Institute, Electric Power

Research Institute, and other interested

persons regarding this review.

Further information regarding topics

to be discussed, whether the meeting

has been cancelled or rescheduled, the

Chairman’s ruling on requests for the

opportunity to present oral statements

and the time allotted therefor can be

obtained by contacting the cognizant

ACRS staff engineer, Dr. Medhat ElZeftawy (telephone 301/415–6889)

between 7:30 a.m. and 4:15 p.m. (EDT).

Persons planning to attend this meeting

are urged to contact the above named

individual one or two working days

prior to the meeting to be advised of any

potential changes in the proposed

agenda, etc., that may have occurred.

Dated: October 8, 1997.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 97–27240 Filed 10–14–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Sunshine Act Meeting

AGENCY HOLDING THE MEETING: Nuclear

Regulatory Commission.

DATE: Weeks of October 13, 20, 27, and

November 3, 1997.

PLACE: Commissioners’ Conference

Room, 11555 Rockville Pike, Rockville,

Maryland.

STATUS: Public and Closed.

MATTERS TO BE CONSIDERED:

Week of October 13

Tuesday, October 14

10:00 a.m.

Briefing on EEO Program (Public Meeting)

(Contact: Ed Tucker, 301–415–7382)

1:00 p.m.

Briefing on Severe Accident Master

Integration Plan (Public Meeting)

(Contact: Charles Ader, 301–415–5622)

Wednesday, October 15

10:00 a.m.

Briefing on PRA Implementation Plan

(Public Meeting)

(Contact: Tom King, 301–415–5790)

11:30 a.m.

Affirmation Session (Public Meeting) (if

needed)

Week of October 20—Tentative

There are no meetings the week of October

20.

Week of October 27—Tentative

Wednesday, October 29

11:30 a.m.

Affirmation Session (Public Meeting) (if

needed)

2:00 p.m.

Briefing on Site Decommissioning Plan

(SDMP) (Public Meeting)

Week of November 3

Tuesday, November 4

2:00 p.m.

Meeting with Commonwealth Edison

(Public Meeting)

Wednesday, November 5

9:30 a.m.

Briefing on Staff’s Plans for 50.59

Regulatory Process Improvements

(Public Meeting)

11:00 a.m.

Affirmation Session (Public Meeting) (if

needed)

The schedule for Commission

meetings is subject to change on short

notice. To verify the status of meetings

call: (Recording)—(301) 415–1292.

CONTACT PERSON FOR MORE INFORMATION:

Bill Hill (301) 415–1661.

* * * * *

The NRC Commission Meeting

Schedule can be found on the Internet

at: http://www.nrc.gov/SECY/smj/

schedule.htm

* * * * *

This notice is distributed by mail to

several hundred subscribers; if you no

longer wish to receive it, or would like

to be added to it, please contact the

Office of the Secretary, Attn: Operations

Branch, Washington, D.C. 20555 (301–

415–1661).

In addition, distribution of this

meeting notice over the Internet system

is available. If you are interested in

receiving this Commission meeting

schedule electronically, please send an

electronic message to wmh@nrc.gov or

dkw@nrc.gov.

* * * * *

Dated: October 10, 1997.

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 97–27474 Filed 10–10–97; 3:05 pm]

BILLING CODE 7590–01–M

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.

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
53664 Federal Register / Vol. 62, No. 199 / Wednesday, October 15, 1997 / Notices

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

axd3@nrc.gov. For information about

the interactive rulemaking site, contact

Ms. Carol Gallagher, (301) 415–5905; email cag@nrc.gov.

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

jxg@nrc.gov.

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-theart 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 stateof-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 PlantSpecific 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
Federal Register / Vol. 62, No. 199 / Wednesday, October 15, 1997 / Notices 53665

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 RiskInformed 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 riskinformed 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,
53666 Federal Register / Vol. 62, No. 199 / Wednesday, October 15, 1997 / Notices

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 lowfailure-potential. The number of

inspections for the high-safetysignificant 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

jxg@nrc.gov. Comments on the

regulatory guidance and standard

review plan documents for discussion at

the workshop should be submitted in

writing and in electronic mail

(JXG@nrc.gov) 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 nearterm 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., preregistration) 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, email kessjones@aol.com.
Federal Register / Vol. 62, No. 199 / Wednesday, October 15, 1997 / Notices 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 riskinformed 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]

BILLING CODE 7590–01–P

NUCLEAR WASTE TECHNICAL

REVIEW BOARD

Panel Meeting: November 19–20,

1997—Arlington, Virginia: Spent

Nuclear Fuel Transportation Safety

Pursuant to its authority under

section 5051 of Public Law 100–203, the

Nuclear Waste Policy Amendments Act

of 1987, the Nuclear Waste Technical

Review Board’s Panel on the Waste

Management System will hold a

meeting November 19–20, 1997,

beginning at 8:30 a.m. The meeting,

which is open to the public, will focus

on transportation safety issues for spent

nuclear fuel.

The first day will include

presentations on the federal regulatory

framework under which transportation

will take place, transportation planning

and practices (using a specific example),

and risk analysis (with emphasis on

methodologies). Representatives from

the Department of Transportation, the

Nuclear Regulatory Commission, Sandia

National Laboratories, and Lawrence

Livermore National Laboratory have

been invited to make the presentations,

along with several private consultants.

On the second day, the presenters will

participate in a round-table discussion

of their topics. Representatives of the

state of Nevada, the environmental

community, the Department of Energy,

and other stakeholder groups also will

participate. The meeting will end

around noon. A detailed agenda will be

available approximately two weeks

prior to the meeting by fax or email, or

on the Board’s web site at

www.nwtrb.gov.

The meeting will be held at the

Sheraton national Hotel, Columbia Pike

& Washington Boulevard, Arlington,

Virginia 22204; (Tel) 703–521–1900;

(Fax) 703–521–0332. Reservations for

accommodations must be made by

October 17, 1997, and you must indicate

that you are attending the Nuclear

Waste Technical Review Board’s panel

meeting to receive the preferred rate.

Time has been set aside on the agenda

for comments and questions from the

public. Those wishing to speak are

encouraged to sign the Public Comment

Register at the check-in table. A time

limit may have to be set on the length

of individual remarks; however, written

comments of any length may be

submitted for the record.

Transcripts of this meeting will be

available on computer disk, via e-mail,

or on a library-loan basis in paper

format from Davonya Barnes, Board

staff, beginning December 18, 1997. For

further information, contact Frank

Randall, External Affairs, 2300

Clarendon Blvd., Suite 1300, Arlington,

Virginia 22201–3367; (Tel) 703–235–

4473; (Fax) 703–235–4495; (E-mail)

info@nwtrb.gov.

The Nuclear Waste Technical Review

Board was created by Congress in the

Nuclear Waste Policy Amendments Act

of 1987 to evaluate the technical and

scientific validity of activities

undertaken by the DOE in its program

to manage the disposal of the nation’s

high-level radioactive waste and

commercial spent nuclear fuel. In that

same legislation, Congress directed the

DOE to characterize a site at Yucca

Mountain, Nevada, for its suitability as