ML20199E281

Reports Status of PRA Implementation Plan for Period of 971001-1231.Responds to Srm,Dtd 970528,which Related to Staff Plans for Using IPE Result to Assess Regulatory Effectiveness

ML20199E281
Person / Time
Issue date:	01/23/1998
From:	Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
References
SECY-98-012, SECY-98-012-01, SECY-98-012-R, SECY-98-12, SECY-98-12-1, SECY-98-12-R, NUDOCS 9802020120
Download: ML20199E281 (27)

See also: IR 07100001/2012031

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POLICY ISSUE

(Information)

Januan/ 23.1998

SECY-98-012

EQB:

The Commissioners

EROM.

L. Joseph Callar.

Executive Director for Operations

SUBJECT:

QUARTERLY STATUS REPORT ON THE PROBABILISTIC RISK

ASSESSMENT IMPLEMENTATION PLAN

PURPOSE:

To report the status of the Probabilistic Risk Assessment (PRA) Implementation Plan for the

period of October 1 to December 31,1997, and to respond to a Staff Requirements

Memorandum dated May 28,1997, which relates to staff plans for using Individual Plant

Examination (IPE) results to assess regulatory effectiveness.

SUMMARY:

This paper describes accomplishments and changes to the staffs PRA Implementation Plan for

the period of October 1,1997 to December 31,1997, The principal accomplishments include

preparation of the final versions of Regulatory Guide (RG) 1.174 (formerly draft guide DG 1061)

and Standard Review Plan (SRP) Chapter 19, which provide general guidance on the use of

PRA in risk informed decisions for changes in a reactor current licensing basis, completion of the

South Texas graded quality assurance pilot program, publication (for public comment) of the

draft RG and SRP on risk-informed inservice inspection, and the development of the staffs plan

to use IPE results to assess regulatory effectiveness in resolving major safety issues. The

principal change is the delay of the application-specific regulatory guides and Standard Review

Plan sections from December 1997 to March 1998, to permit the incorporation of the policy

decisions associated with the finalization of RG 1.174.

CONTACT:

NOTE:

TO BE MADE PUBLICLY AVAILABLE

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IN 5 WORKING DAYS FROM THE DATE OF

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Ashok Thaaani, OEDO

THIS PAPER

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BACKGROUND:

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In a memorandum dated January C .1996, from the t!xecutive Director for Operations to

Chairman Jackson, the staff committed to submitting quarterly reports on the status of its

development of risk informed standards and guidance. Previous quarterly reports were

provided to the Commission on March 26, June 20, and October 11,1996, and on January 13,

April 3, July 22, and October 14,1997. This quarterly report covers the period from October 1,

1997 to December 31,1997.

DISCUSSION:

The significant accomplishments and changes to the PRA Implementation Plan in the past

quarter are summarized below.' More detailed information is provided in Attachment 1.

Significant achievements during the past quarter include:

Section 1: Reactor Regulation (NRR)

1.1 Develop Standard Review Plans for Risk Informed Regulation

NRR and RES staff met with ACRS and CRGR to discuss the final versions of the general

guidance on use of PRA in risk informed decision making in changes to the plant specific

current licensing basis; Standard Review Plan Chapter 19 (NRR lead) and Regulatory Guide 1.174 (RES lead). A Commission paper providing the final versions of these documents will be

provided to the Commission in the near future.

Draft Standard Review Flan 3.9.8 (NRR lead) and Regulatory Guide DG-1063 (RES lead) on

risk informed inservice inspection of piping were published for public comment and the subjects

of a public workshop on November 20 and 21,1997, The workshop was well attended by

industry representatives who offered a number of constructive comments, some criticisms, and

some suggestions for changing the guidance. Overall, the comments indicated strong support

for pursuing risk informed inservice inspection (RI-ISI) but in a manner which would necessitate

some modifications to the draft guidance.

1.2 Pilot Applications for Risk informed Regulatory Initiatives

The staff evaluation of the South Texas Project risk informed graded quality assurance (QA)

implementation plan was transmitted to the Commission via SECY 97-229 on October 6,1997.

I The staff has modified the format of the PRA Implementation Plan to reduce

redundancy and improve readability. This revised format consists of the body of the

Commission paper, which now provides a summary of accomplishments and changes to the

plan for the past quarter, and the Plan's table (Attachment 1), modified to explicitly show where

milestones have been added, completed, or changed. Such changes are discussed and

additional information provided in endnotes to the table.

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By SRM dated October 30,1997, the staff was informed that the Commission had no objection

to the issuance of the South Texas graded QA safety evaluation report. By letter dated

November 6,1997, the licensee was informed that the staff had approved the graded QA

change and was provided with the associated staff safety evaluation.

,

The staff has received risk informed inservice inspection pilot submittals from Surry 1, ANO 2,

and Vermont Yankee which are currerstly being reviewed for completeness. The staff will

develop a review schedule if the submittals are determined to be complete and in conformance

with the DG 1061 and DG 1063 submittal guidance. The staff is also developing schedules and

priorities for the review of other RI ISI pilot submittals as well as submittals expected

subsequent to the pilot RI ISI program approvals.

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1.3 Inspections

The staff completed nine additional maintenance rule baseline inspections during this quarter,

which included inspection of licensee methods for using PRA in maintenance programs and

inspection of safety assessments performed by licensees when removing equipment from

service for maintenance in accordance with Paragraph (a)(3) of the Maintenance Rule. As of

December 31,1997, the staff has completed a total of 45 inspections.

Section 2: Reactor Safety Research (RES)

2.1 Develop Regulatory Guides

As discussed above, NRR and RES staff met with ACRS and CRGR to discuss the final

versions of Standard Review Plan Chapter 19 (NRR lead) and Regulatory Guide 1.174 (RES

lead). A Commission paper en policy issues was forwarded to the Commission as SECY-97-

287, dated December 12,1997. A Commission paper providing the final versions of these

documents will be provided in the near future.

2.4 Methods Development and Demonstration

A demonstration at the Seabrook nuclear power plant of the human reliability analysis method

ATHEANA (A Technique for Human Event Analys;s) has been completed. A medium break

LOCA scenario, including inappropriate termination of makeup (an error of commission), was

selected for analysis and simulator exercise. The ATHEANA demonstration helped plant

personnelidentify safety-related weaknesses in plant barriers and design. Specifically, the

exercise identified weaknesses in the use of well planned and tested emergency procedures as

well as identifying improvements needed in the draft ATHEANA documentation.

2.5 IPE and IPEEE Reviews

- The final version of NUREG-1560, *lPE Program: Perspectives on Reactor Saf9ty and Plant

Performance," has been submitted for publication. This report was initially issued in late 1996

for public comment. l Based on the comments received, the report was revised, with an

addrtional appendix written discussing the comments received and staff responses.

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The first IPEEE staff evaluation report, for the Diablo Canyon Power Plant, was completed and

issued to the licensee on December 4,1997, in addition, requests for additionalinformation on

fifteen IPEEE submittals were prepared to send to licensees.

An interim report has been developed that provides preliminary IPEEE perspectives and

summarizes the information presented in the first 24 IPEEE submittals reviewed by the staff.

This interim report will be sent to the Commission in the near future. (A summary of the

significant preliminary perspectives from the first 24 IPEEE reviews was provided to the

Commission in Attachment 7 to SECY 97 234.)

Sectinn 3: Analysis and Evaluation of Operating Experience and Training (AtiOD)

3.1 Risk Based Trends and Patterns Analysis

Letters are in the concurret.ce process to distribute the common cause failure (CCF) database

and associated technical report to all U.S. nuclear utilities for their use. The database contains

CCF events from 1980 through 1995.

3.2 Accident Sequence Precursor (ASP) Program

All 1996 precursor analyses have been finalized, with the 1996 ASP report now in publication.

Three preliminary analyses of 1997 events are being reviewed. The annual Commission paper

describing the ASP program in more detail was sent to the Commission on December 23,1997

(SECY 97 296).

3.6 Staff Training

Development activities for the PRA Technology and Regulatory Perspectives (P 111) course

were comploted during this quarter. The first course presentation will be January 26 - February

6,1998. The staff has established a goal of having one Resident inspector at each she

complete the course by the end of 1998.

Significant changes made to the implementation Plan during the last quarter include:

Section 1: Reac'or Regulation (NRR)

1.1 Develop Standard Review Plans for Risk-Informed Regulation

As discussed above, the general regulatory guide and Standard Review Plan for use of PRA in

plant-specific current licensing basis changes will be transmitted to the Commission in the near

future. To permit efficient incorporation of the resolution of policy issues contained in these

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documents into the application-specific SRP sections on inservice testing and technical

specifications, completion of these SRP sections has been delayed until March 31,1998, a

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change from their previous completion date of December 31,1997.

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1.2 Pilot Applications for Risk Informed Regulatory initiatives

The staff is currently developing a draft safety evaluation report (SER) for the Comanche Peak

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risk informed inservice testing program (Rl IST) program. The licensee (TU Electric) is

currently developing a program that is sufficiently detailed and consistent with DG 1062. TU

Electric has indicated that it intends to complete a draft revision to their program description by

the end of January 1998. Assuming that the program is finalized by mid February, the staff

anticipates having a completed Comanche Peak Rl IST SER to the Commission in March 1998,

rather than December 31,1997.

The completion date for the graded quality assurance (GQA) pilot interactions has been revised

from March 1998 to July 1998 to reflect the anticipated issuance date of the final GOA

inspection guidance.

The staff received a supplemental amendment request from the San Onofre Nuclear

Generating Station (SONGS) in early January 1998 to put the configuration risk management

program description into the SONGS technical specif cations. SONGS has recently become

the lead plant for this Combustion Engineering Owner's Group (CEOG) activity, when the

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originallead plant decided not to pursue risk-informed TS changes at this time. With receipt of

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the SONGS supplen etal request, the staff anticipates completing the SONGS review as the

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lead pilot plant and isung the license amendment by March 31,1998. This is a change from

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the previous date of December 31,1997.

1.3 Inspections

The NRR Inspection Program Branch (P'PB) proposals for revising core inspection procedures

have been transmitted to the appropriate NRR technical branches having responsibility for

specific core inspection procedures. Due to the large number of branches involved, completing

all individual branch concurrences is anticipated to take an additional two months. The revised

completion date for th!s task is February 1998, a chan9e from the previous date of October

1997.

1.6 Evaluate Use of PRA in Resolution of Generic issues

As part of the IPE follow-up program, the staff is in the process of identifying generic issues to

be audited. These issues are those which have been explicitly identified and addressed by the

licensee as part of the IPE process.

A report that identifies the above generic issues and staff views on the adequacy of the

proposed resolution k. In preparation. The report will provide the basis for the selection of

generic safety issues to be audited. The staff has moved the completion date for this milestone

to March 1998, in order to utilize the report in the audit process.

In addition to the above issues, RCP seal LOCA had been identified as a dominant contributor

to core damage in many PWR IPEs. The staff has a separate ongoing activity in RES to

address this issue under Generic issue 23, and will utilize IPti insights in the proposed

resolution.

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1.7 Regulatory Effectiveness Evaluation

in a Staff Requirements Memorandum dated May 28,1997 (Attachment 2), the Commission

requested that the staff provide the scope and schedule of activities relatad to using IPE results

to assess regulatory effectiveness in resolving major safety issues. With respect to scope, the

staff identified three major safety issues for assessment. The selection had been based on

both the potential risk significance of the issue, and the fact that probabilistic techniques were

used extensively in the resolution process. These issues include:

1, Resolution of USl A-44 Station Blackout at Nuclear Power Plants

2. Resolution of USl A-45 Decay Heat Removal Reliability

3. Resolution of USl A 09 Anticipated Transient Without Scram

To evaluate the three major issues, the staff will utilizJ both representative plants, and

information contained in NUREG 1560, to audit and draw conclusions regarding regulatory

effectiveness. Information generated under Task 1.6, as deteribed above, and Task 1.10, as

described below, will also be integrated into the assessment process. These tasks may expar!d

the staff's consideration of other cafety issues and effectiveness of the regulatory process. The

staff will inform the Commission of any additional safety issues that come under consideration.

The staff plans to complete Task 1.7 by the end of December 1998, and will recommend at that

time any additional staff action.

1.8 Advanced Reactor Reviews

Doe to personnel being assigned to higher priority activities, such as risk-informed pilot

initiatives and IPE followup activities, the staff is reassessing their position regarding the

development of an SRP, especially since there are no new advanced design certification

submittals anticipated. We will provide the results of this reassessment in a future update of

the PRA implementation Plan.

1.10 Evaluation of IPE Insights

The staff has developed an IPE followup plan (Attachment 3) which describes those actions to

be taken to ensure that plant improvements warranted by 'he IPE results are, in fact, made.

This plan consists of a number of items and its implementation involves NRR, RES, and the

Regions, as described in the plan.

Section 2: Reactor Safety Research (RES)

2,1 Develop Regulatory Guides

As discussed above, the general regulatory guide and Standard Review Plan for use of PRA in

risk informed decision making for plant specific current licensing basis charges will be

transmitted to the Commission in the near future. To permit efficient incorporation of the

resolution of policy issues contained in these documents into the application-specific regulatory

guides on inservice testing, graded quality ascurance, and technical specifications, completion

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of these guides has been delayed until March 31,1998, a change from their previous

completion date of December 31,1997.

2.5 IPE and IPEEE Reviews

The staff has reviewed all the 76 IPE submittals and issued staff evaluatiun reports (SERs) on

their findings to each licensee, in three of the SERs, it is indicated to the licensees that the

staff was not able to conclude that the licensee met the intent of Generic Letter 88 20 for their

plant (s). These three IPEs include Crystal River 3, Susquehanna 1&2, and Browns Ferry 3.

The licensee for Crystal River 3 has indicated their intention to submit an updated analysis

(February 1998) addressing the staff's concerns. It is anticipated that the review of this new

IPE submittal will be concluded in June 1998. Discussions are still ongoing with licensees

regarding Susquehanna 182 and Browns Ferry 3.

Section 3: Analysis and Evaluation of Operating Experience and Training (AEOD)

3.6 Staff Training

Eight PRA for Regulatory Applications courses are now planned for FY 1998 and FY 1999 to

meet the needs of the technical staff. Funding for these courses was obtained by reducing the

number of SRA series from two to one per year. Modifications to the PRA Basics for

Regulatory Applications, PRA for Technical Managers, and PRA Technology and Regulatory

Perspectives courses have been made to 'alude the final draft R.G.1.174 and SRP, Chapter

19. Seven PRA for Technical Managers courses are planned for FY 1998, which will allow two -

thirds of agency technical managers to attend.

Procurement actions for acquisition of risk monitor software are in process. The EPRI Risk and

Reliability (R&R) Workstation is the current industry standard for risk monitors Current plans

are to integrate the R&R workstation into the reactor technology and PRA technology curricula

to improve student understanding of configuration management, the importance of plant

operations to the risk profile of the plants, and use of the tool to provide insights regarding the

use of risk informed applications by the industry. The workstation will also be used to

demonstrato the capabilities and limits of this and simi!ar tools as they are being used by the

industry.

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Section 4: Nuclear Materials and Low Level Waste Safety and Safeguards Regulation (NMSS)

44

Risk Assessment of Material Una

The target schedule for the work to develop and demonstrate a risk assessment for industrial -

gauges containing cesium 137 and cobalt-60 using PRA (and other related techniques) has

been extended from July 1998 to September 1998. The extension is due to difficulties in

obtaining data from non-licensees related to actual and potential doses to the public resulting

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from gauges which enter the scrap metal cycle.

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The tar 9et schedule for the work to develop and demonstrate risk assessment methods for

application to medical and industrial licensee activities has been determined to be September

1998 based on scheduling of a planned Commission paper on the topic.

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Framework for Use of PRA in'Reauintina Nuclear Materials

The target schedule for providing a p

for developing a framework has been extended from

October 1997 to January 1998 to permit interoffice coordination.

COORDINATION:

The Office of the General Counsel has reviewed this paper and has no legal objections to its

issuance.

M G (/ *

L. Jclseph Callan

Executive Dire

VOperations -

Attachments:

As stated

DISTRIBUTION:

Commissioners

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OCA

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ACRS

ACNW

CIO

CFO

EDO

REGIONS

SECY

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ATTACHMENT 1

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PRA IYPLEMENTATION PLAN TASK TABLE (December 1997)

1.0 REACTOR REGULATION

Cegulatory Activity

Objectives

Methods

Target

iend

Status (tNs

Schedule

Offee(s)

queder)

1.1

DEVELOP

Standard revew plans for NRC

' Evaluate available industry

NRR

STANDAHL

staff to use in rak 6nformed

guidance.

/RES

REV!EW PLANS

regulatory decison making

FOR RISK.

• Develop a broad sco

INF ORM* D

standard rewew plan ( ,RP)

REGULAtlON

chapters and a senes of

applicaton specife standard

revew plan chapters that

correspond to industry

initiatives.

• These SRPs will be

consistent with the Regulatory

Guides developed for the

industry.

• Draft SRPs transmitted to

Commisson to issue for

puble comment

G,neral

497C'

IS T

497C

ISI

E97C

TS

497C

• Final SRP transmitted to

Commmagin for approval

General

1/98

in final renew

IST

393

Changed (Note 1)

ISt

498

TS

398

Changed (Note 1)

la

PILOT

• Evaluate the PRA methodology

• Interface with Industry

NRR/RES

APPLICATIONS

ond deveksp staff positons on

groups

FOR RISK.

emerging, rak informed

INFORMED

Initiatives enciuding those

• Evaluaten of appropnate

REGUL ATORY

essociated with

documentaton te g .10 CFR.

INITIATIVES

SRP. Reg Guides, mspecten

1. Motor operated vatves

procedures. and industry

1. 2,96C

codes) to edentify elements

2 IST requirements

entcal to acheving the entent

2a Comanche Peak

of extsting requrements.

2a 398

Changed (Note 2)

2b Paio Verde

20.TDD

' Evaluaton of industry

3 ISirequirements

proposain.

3 TBD

4 Graded quality assurance.

• Evaluaton of industry pilot

4. 7/98

Changed (Note 3)

program implementaten.

5 Maintenance Rule

5 E95C

' As appropnate, complete

6 Techncalspecifcatons

pilot revews and tasue staff

6a Commisstun Approval

findings on regulatory

6a 597C

6b. Pilot Amendments issued

requests.

6b.198

Changed (Note 4)

7. Other appleatens to be

identifed later (apphcations

related to desel generator start

times and hydrogen control are

expected)

1 C = Tash prewousy comp 6eted

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Cegulatory ActMty

Obrectrves

Methods

Target

lead

Status (this

Schedule

Offee(s)

quener)

_,

1.3

lNSl%',TIONS

• Provide guidance on the use of

• Develop IC #00 technical

697C

NRR

plant-specife and genene

guidance on the use of PRAs

enformaton from IFEs and other

in the power reactor inspecten

plant specife PRAs.

program.

• Revse IC 25t$ Appendix C

7/g7 C

on the use of PRAs in the

power reactor inspecton

program.

• Propose guklance options

t097

Completed

forint ton procedures

totot to 60 59 t,stuntons

and regular rnanntenance

observatens.

• Review core inspecten

1697

Completed

procedures and propose PRA

guidance where needed

• Complete revision to

propened core .nspecten

2/98

Changed (Note 5)

procedures

• losue dran Graded OA

Inspecten Procedure

498

Changed (Note 6)

• lasue final Graded QA

inspecton Procedure

7/98

C%nged (Note 6)

• Provide PRA training for

• Identify inspector functons

7/96C

NRR

inspen. tors

whch should utilize PRA

methods as input to

AEOD/TTO for their

development and refinement

of PRA trairung for inspectors

'

• Develop consolidated and

comprehensive 2 3 week PRA

1697

NRR/

Completed

for regulatory ap

AEOD

tranng course. picatens

' Provide PRA training for Senior

Reactor Analysts (SRA)

• Conduct training for

Mantenance Rule baseline

&960

NRR

inspectons

' Conduct training courses

according to SRA training

Ongoing

AEOD

programs

• Rotatonal assignments for

SRAs to gain working

Ongoing

NRR/RES

exponence

' Continue to provide expertise in

• Monitor the use of nsk in

Ongoing

NTIR

risk assessment to support

inspecten reports.

regionalinspection activites and

to communcate inspecten

• Develop new methodologes

program guidance and

and communcate appropnate

examples of its implementatort

uses of nsk insights to

regional offces

• Update inspecten

procedures as needed

• Asset regional offces as

needed

• Conduct Maintenance Rule

baseline inspectons

7/98

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Cogulatory ActNi'y

Objectwos

Methods

TarDet

Lead

Status (this

Scheoute

Offce(s)

quarter)

14

OPERATOR

Monstry insf.;s from HRAs and

• Reese the Knowiedge and

&95C

NRR

UCENSING

PRAs (including IPEs and

Atwides (K/A) Cataicos

IPEEEs) and operating

(NUREGs 1122 and 1123) to

1

e ronce to ioentify possible

encorporate operating .

e ncements for inclusaon an

exponence and nok 6nsghts

planned revisions to guidance for

operator licensing actuttes (rvtal

• Revise the Examiner

197C

and requahfcation)

Standards (NUREG-1021), as

needed to reflect PRA

Insights

19

EVENT

' Continue to conduct quantitative

' Continue to evaluate 50 72

Ongoing

NRR

ASSESSMENT

event assessments of reactor

events using ASP models

events while at. power and dunng

low power and shutdown

conddions

• Assess the desirab6ldy and

• Define the current use of risk

TBD

NRR

feasibihty of conducting

analysis methods and insights

quantitafive nok assessments on

in current event assessments

non-power reactor events

• Assess the easibility of

r

developing appropnate not;

assessment models

• Develop recommendations

an the feasitulity and

desirabihty of conducting

quantitative risk asses!,ments.

1C

EVALUATE USE

• Audit the adequacy of hcensee

• Identrry genere safety issues

198

NRR/RES

Changed (Note 7)

OF PRA IN

analyses in IPEs and IPEEEs to

to be audited

RECOLUTION OF

identify plant-specifc apphcability

GENERIC ISSUES

of genene issues closed out

• Select plants to be sudded

398

Changed (Note 7)

i

based on IPE and IPEEE

for each issue.

programs.

• Desende and discuss

TBD

l

licensees * onetyses supporting

i

issue resolution.

• Evaluate results to determine

TBD

regulatory response; i e., no

action, additional acids, or

regulatory accon.

1.7

REGULATORY

• Assess the effectiveness of

• Develop process / guidance

ongoing

NRRIRES

EFFECTNENESS

maior safety resue resolution

for assessing regulatory

EVALUATION

afforts for reducing nsk to pubic

effectiveness.

health and safety.

• Apply method to essess

ongoing

reduction in nsk.

Note Work in this actuity will be

Integrated with broader agency

• Evaluate resulting

12,98

Changed (Note B)

efforts in response to DSI 23

effectmeness of station

blackout and ATWS rules and

Unresolved Safety issue A-45.

• Propose modifications to

TBD

Changed (Note 8)

resolution approaches, as

needed (SBO rule

imp' ament.h vi and RCP seal

assue).

• Identify othe Swes for

ongoing

Changed (Note 8)

assessment 9 4;y:gnate.

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Regulatory Activity

Objectrves

Methods

Target

Lead

Status (this

Schedule

Offee(s)

quader)

1.8

ADVANCED

' Contnue staff reviews of PRAs

• Continue to apph current

ongoing

NRR

REACTOR

for design certifcation

staff review process

REVIEWS

appicatens

-

• Devehp SRP to support revww

• Develop draft SRP to tech

TBD

NRR

Changed (Note g)

of PRAs for design certifcatic.a

staff for rev ew and

reviews of evolutonary reactors

corcurrence

(ABWR and System 80+)

• rrialize SRP.

TBD

• Develop indeper. dent technical

' Reevaluate nsk-based

12/96C

NRR/ RES

analyses and critena for

aspects of the techncal bases

evaluating tndustry initiatives and

for EP (NUREG4396) using

petitons regarding sirnphfcaten

Inssghts from NUREG-1150

of Emergency Preparedness

the new source term

(EP) regulations.

Informat.m from NUREG-

1465. and available plant

design and PRA informaton

for the passive and

evolutonary reactor designs.

1.g

ACCIDENT

• Develop generic and plant

• Deveioo plant-specife A!M

TBD

NRR/RES

MANAGEMENT

specife nsk ins s to support

insghtsinformaton for

staff auditsof

accident

selected plants to serve as a

management

programs at

basis for assessing

selected plants

cor p6eteness of utihty A/M

program elements (e g ,

severe accident training)

1,10

EVALUATING IPE

• Use insights from the staff

' Reyww the report *1PE

E97C

NRR/RES

INSIGHTS TO

review of IPEs to idert.N

Program Perspectives on

DETERMINF

potental safety,

Reactor Safet

technical ssues, policy, and

Performance *y and Plant

NECESSARf

to determine en

and identify the

FOLLOW UP

appropnate course of action to

initiallist of required staff and

ACTIVnIES

resolve tnese potentialissues,

industry actions (if an )

and to identify possible safety

including insghts on

_

enhancements.

• Revww IPE results and

6/99

NRR/RES

Changed (see

interact with heensees.

Attachment 3)

• Determine appropnate

• Complete backfit analysis

12<99

NRR

Changed (see

approach for tracks tne

and actions.

regulatory uses of 1

PEEE

Attachment 3)

results

• Followup on accident

SSB

NRR/

Changed (see

i

management programs and

regions

Attachment 3)

Iconsee-stated actions.

• If sopropriate develop

1298

NRR/ RES

a

h for hnking

i

PEEE data bases.

4

.

.

.

..

.

. - -

- - - - - _

,

. _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

.

.

20 REACTOR SAFE *Y RESEARCH

Cogulatory Adivty

Objectives

Methnos

Target

Lead

Status (this

Schedule

Offee(s)

quarter)

,

11

DEVELOP

Regulatory Guides for industry to

• Draft PRA Regulatory

RES/NRR

<

I

REGULATORY

use in rak informed regulaton

3uides transnuttedto

,

l

GUIDES

ommesson for aproval to

j

issue for pubic comment.

l

General

C

l

IST

C

t

ISI

C

GOA

C

TS

C

• Forwl PRA Regulatory

3uides transmitted to

l

Commission for approval.,

[

Genersi

1/98

in final revww

IST

398

Changed (Note 1)

isi

498

GOA

396

Changed (Note 1)

TS

398

Changed (Note 1)

33

TECHNICAL

• Provide techn. cal support to

• Continue to provide ad hoc

Continuing

RES

SUPPORT

agency users of risk

technical support to agency

assesernent in the form of

PRA users

support for nak based

}

regulaten activitwo, techncal

• Expand the database of PRA

Continuing

reviews, essue nok

nodels available for staff use,

assessments, statist. cal

expand the scope of avadabie

analyses, and aevelop

'nodels to include external event

guidance for agency uses of nok

and low power and shutdown

assessment

accidents, and refine the tools

weded to use these models, and

ontinue maintenance and user

w

for SAPHIRE and

CS computer codes

• Support agency efforts in

'eactor safety improvements in

Continuing

'ormer Soviet Union countnes.

3.3

SUPPORT FOR

' Modify 10 CFR 52 and develop

• Develop draft guidance and

5/98

RES

NRf1 STANDARD

. Mance on the use of updated

'une.

REACTOR PRA

". tAs beyond denen cenirmation

REVIEWS

(as desenbod in SECY 93487).

• Solett pubic comment.

11/98

• Finales staff guidance and

12,99

'ulo =

,

24

METHODS

• Devoiop, demonstrate, maintain.

• Develoo and demonstrate

9/98

RES

DEVELOPMENT

and ensure the quality of

Tiethods for including aging

AND

methods for performina,

effects in PRAs.

DEMONSTRATION

revww1ng, and ussng F4tAs

and related techniques for

• Develop and demonstrate

9<98

,

existing reactor designs.

Tethods for including human

errors of commissen in PRAs.

• Develop and comonstrate

TBD

netfods to incorporate

gnaational performance into

• Develop and demonstrate

9/96

methods for flru nsk analysis.

• Develop and demonstrate

6/99

nothods for assessing

'ehability/nsk of digital

systems

5

,.

.

. .. .

.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

.

.

.

Cegulatory Actrvity

Objectives

Methods

Target

Lead

Status (this

'

Schedule

Office (s)

quarter)

15

IPE ANDIPEEE

• To evaluate IPE/IEEE

• Complete revews of IPE

TBD

RES

Changed (Note 10)

REV?EWS

submittats to obtain reasonable

6ubmittals.

assurance that the hcensee has

adequately anatyzed the

• Complete reviews of IPEEE

699

piant desgn and operations to

bubmrttais

discovervulnersbees andto

document the sgnificant safety

• Continue regionat IPE

C

insghts resulting from

yesentations

IPEstPEEEs

'lasue IPE insghts report for

10/96C

xiblic comment.

• FinalIPE 6nsghts report

9S7

Completed

• losue prehminary IPEEE

1/98

in final review

insghts report

=,

• Initiate revew of eight

6/98

New milestone

additionalIPEEE submittals

• Comp 6ete contractor

6,98

New milestone

avsluations on twelve IPEEE

wbmittals

• lasue draft IPEEE insghts

699

New milestone

report for comment

• lasue final IPEEE insghts

1299

oport

2.6

GENERIC ISSUES

' To conduct genene safety issue

• Continue to pnortize and

Continuing

RES

PROGRAM

ma

ment activites,

eserve genenc issues

inct ing pnorit&tation, resolution,

and documentation, for issues

relating to currently

rating

reactors. for

need

reactors as appropnate, and for

development or revtsion of

associated r ulat and

standards 6n me a

17

NEl INITIATIVE TO

• Review NElinstetive to conduct

• Agree on ground rules for

198

RES/NRR

CONDUCT

three pilot "whole

plant"

study.

"WHOLE PLANT *

nsk intormed studes of

RISK STUDY

requirements vs nsk and cost

• Complete study

TBD

2.8

PRA STANDARDS

• work with industry to develop

• Initiate actrvity.

9S7C

RES

DEVELOPMENT

national consensus standard for

PRA scope and quality

• Finalize standard.

TBD

19

LOW POWER AND

• Collect studes of LP&S nsk as

• Collect and review existing

9S8

RES

SHUTDOWN

a benchmark for assessi the

.P&S nsk information (domestic

BENCHMARK

need for further staff

es

and foregn).

RISK STUDY

• Instete additional work.

10S8

l

l

110

SAFETY OOAL

'Assesa need to revtse

• Initete disct=sion with ACRS

2/98

RES

l

REVISION

Commissiorfs Safety Goalto

i

make core damage frequency a

• Recommendation to

198

l

fundame ital goal and make other Dmmission

changer,,

l

6

,

_ _ _ _ _ - _ _ - _ _ - _ _ _ _ - _ _ - _ - _ - - _ - _ _ _ - - - _

.-.

.

.. .

. . .

.

.

'

,

.

3 0 ANALYSIS AND EVALUATION OF OPERATING EXPERIENCE, AND TRAINING

Regulatory

Otyctives

Methods

Target

Lead

Status (t*us

Activity

Schedule

Offee

quarter)

3.9

RISK BASED

• Use reactor opeisting

• Trend performance of nsk.

1198

AEOD

TRENDS AND

exponence data to assess the

important components

PATTERNS

trends and pottems in ooiapment,

ANALYSIS

systems, inrtistire " .. hurrun

• Trend performance of nsk.

1Z98

performance, s.

ant

important systems.

.

accxlent sequence.

l

• Trend fr

of nsk.

3,98

!

6mportant

events

• Trend human performance

TBD

for rehatuity charactenstes

• Evaluate the effectiveness of

• Trend reactor operating

As Needed

AEOD

teensee actons taken to

exponence associated with

resolve nok signifcant safety

specife safety issues and

losues

assess nsk impications as a

measure of safety

performance

• Develop trending methods and

• Develop standard trending

C

AEOD

special databases for use in

and statistical anatysis

AEOD trending activttes and for

procedures for 6dentified areas

PRA ap

offees. pications in other NRC

for rehatulity and statistcal

apphcations.

• Develop special software

CCF.C

and databases (e g common

Penode

cause failure)for use in

updates

trending analys,s and PRA

studies

32

ACCIDENT

• Ide, tify and rank nsk

• Screen and analyze LERs,

Ongoing

AEOD

SEQUENCE

sgnifcance of operatonal

AITs, llTs. and events

PRECURSOR

events

identifed from etner sources

(ASP) PROGRAM

to obtain ASP events.

• Perform independent revew

Annual

AEOD

of each ASP analyses,

report,

Licensees and NRC staff peer

Ongoing

review of each analysis.

• Complete quauty assurance

of Rev. 2 simphred piant

3,97C

RES

specife models

• Complete feasitxl tudy for

low

and shut

11/96C

RES

• Complete initial containment

performance and

C

RES

consequence models.

• Compte development of the

Level 43 models

7/99

RES

• Comge the Rev. 3

s

plant specirc

11K)1

RES

• Complete extemal event

models for fire and earthquake

TBD

RES

• Complete low

power / shutdown models

TBD

RES

• Provide supplemental

• Share ASP analyses and

Annual rpt

AEOD

information on plant specife

insights with other NRC

performance.

offces and Regions.

,

7

e

x

.

.

Regulatory

Objectives

Aethods

Target

Lead

Status (this

ActMtf

Schedule

Offee

guarter)

33

tNDUSTRY 8t!SK

• Provide a rnessure of industry

• Develop program plan whch

C

AEOD

TRENDS

nsk that es as complete as

integrates NRR, RES and

possible to oetermine whether

AEOD actMties whch use

not is increasing. decreasing, or

desgn and operating

remaining constant over time

exponence to assess the

emphed level of rtsk and how ll

es changing

• Update plan for nsk based

Changed

ana!yss of reactor operating

(Note 11)

exponence

• Imp 6ement program plan

6/99

elements whch willinclude

piard. specifc models and

insghts from IPEs,

component and system

relaatulity data, and other nsa.

important :;esgn and

operational data in an

integrated frame work to

scally evaluate industry

30

RISK BASED

' Estabish a comprehenstve set

• Identify new or improved

C

AEOD

PERFORMANCE

of performance indicators and

nsk-based Pls whch use

INDICATORS

supplementary cerformance

component aN system

rneasures whch are more

rehabdity models & human and

clow*e related to nsk ar.d provide

organizational performance

bolo e. sty indcation and

evaluation methods

con 9mation of plant performance

problems.

• Develop and test candidate

Pis/ performance measurts.

900

• Implement nak-based Pts

with Commission approval

101

35

COMPILE

• Compile operating exponence

• Manage and maintain SCSS

Ongoing

AEOD

OPERATING

lnformation In database systems

and the PI data base, provice

EXPERIENCE

suitable for uantitative rehatxlity

oversjght and access to

DATA

and rak ana is appications.

NPRUS/ EPIX, obtain INPO's

Information

Id be scrutable

SSPl, compile IPE failure

to the source at the event level to

data, collect plant-specife

the exter.t practical and be

rehability and availatxiety data.

i

!

suffcient for estimatino reinatxlity

and avestatxhty parameters for

• Develop, manage, and

Ongoing

NRC appleations

maintain agency ostabases for

rahatxutyravailatulity data

(equipment performance

initiating events, CCF, AhP,

and human performance

data).

• Determine need to revise

698

LER rule to eliminate

unnecessary and less safety.

sgnifcant reporting

• Determine need to revise

6/98

reporting rules and to better

capture ASP, CCF, and

human performance events.

• Pubhsh revised LER rule.

10/9C

8

s

2

-,

-

_

____ ____ ___ _ _ _ __ _ _ _

.'

.

__

Regulatory

Obrectrves

Methods

Target

Lead

Status Ohis

Actrvey

Schedule

Offce

quarter)

t

,

i

36

STAFF TRAINING

' Present PRA cumculum as

• Continue current cWacts to

Ongoing

AEOD

presently scheduled for FY

present courses es

1998

scheduled.

• Maintain current reactor

technology courses that

include PMA ensghts and

apphcations

courses wa

• Rowew cunent PRA course

matenal to ensure consistency

wth Appendix C

• Develop and present Appendix

• Prepare course materal

C

RES/AEOD

C training courses

based on Appendix C.

• Present courses on

C

Appendir C

• Determine staff requirements

• Review JTAs performed to

C

AEOD

for traini} retuding

date.

ty t

a

• Perform re seentatrve JTAs

C

for staff

ens (JTA Pilot

Program .

• Evaluate staff trainino

C

requirements as identiTod in

the PRA im mentation Plan

and the Tec

I Training

Needs Survey (Phase 2) and

incorporate them into the

training requerernents analysis.

• Analyze the results of the

JTA Pilot Program and

C

determine requirements for

additional JTAs.

• Complete JTAs for other

staff posatons as needed.

C

• Sohcit a review of the

proposed training

C

requirements

• FAalize the requirements.

C

• Reese current PRA cumculum

• Prepare new courses to

Ongoing

AEOD

and develop new training

meet identified needs.

program to fulfill 6dentifed staff

neeos.

• Rewse current PRA courses

Ongoing

i

to meet identifed needs.

f

• Rewse current and New

PRA course to include Reg -

9/97C

Guide and SRP information

• Revise current reactor

technology courses as

Ongoing ,

necessary to include

additional PRA Insigtds and

applications

• Present revised PRA training

• Estabirah contracts for

Ongoing

AEOD

curnculum.

presentata:.of new PRA

curnculum.

• Present revised reactor

Ongoing

technoicgy courses.

rove courses based on

Ongoing

9

- - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - - _ _ -

.

.

4 0 NUCLEAR MATERIALS AND LOW. LEVEL WASTE SAFETY AND SAFEGUARDS REGULATION

Regulatory ActMty

Objectives

Methods

Target

Lead

Status (this

Schecale

Offee(s)

quarter)

4.1

VAllDATE RISK

• Validate nsk anatysis

• Hold a workshop consisting

BS4

NMSS

ANALYSIS

metxx3 ology developed to assess

of experte si PRA and HRA to

C

METHODOLOGY

the relative profile or most hkoty

examine existing work and to

DEVELOPEDTO

contnbutors to misedministration

provide recommendations for

)

I

ASSESS MOST

for the gamma stereotecte dence

further methodologea!

I

LIKELY FAILURE

(gemma knife)

development.

MODES AND

HUMAN

• Examine the use of Monte

9/95

PERFORMANCE

Carlo simulaten and its

C

IN THE USE OF

apphcation to relative nsk

INDUSTRIAL AND

profding

MEDICAL

RADIATION

• Examins the use of expert

SSS

DEVICES

judgement in developing error

C

rates and consequence

measures

• Continue the development of

• Develop functionalty based

TBD

RES/

the relative nsk methodology, with

genene event trees

NMSS

the additen of event tree

modehng of the trachvtherapy

remote after loader '

• Extend the appicaten of the

' Develop genere nsk

TBD

RES/

metnodology and as further

approaches.

NMSS

development into additional

devees. including teletherspy

and the pulsed high dose rate

after loader

!

48

CONTINUE USE

• Develop decisen enterna to

• Conduct enhanced

BS4 PR

RES/NMSS

OF RISK

support reculatory decision

partcipatory rulemaking to

C

ASSESSMENT OF

making tha't incorporates both

establish radmiogeal entena

Final Rule

ALLOWABLE

determaste end nsbbesed

for decon.missoning nuclear

Pubhshed

RADIATION

engineenng judgement.

sites; techncal e"prart for

7/97 C

RELEASES AND

rulemaking includtri

DOSES

comprehenstve nskbood

ASSOCIATED

assessment of resittual

WITH LOW LEVEL

contamination.

RADIOACTIVE

WASTE AND

• Develop uidance for

2/98

REstDUAL

I

ng the radologeal

ACTMTY.

er ena for iconse terminston..

• Work wtth DOE and EPA to

the extent practcable to

Ongoing

develop common approaches,

assumptons, and models for

evaluating nsks and attemative

remediaton methodologies

(nsk harmontraton)

4.3

DEVELOP

• Develop a Branch Techncal

• Solicit pubhc comments

5/97 C.

NMSS/RES

GUIDANCE FOR

Positen on conducting a

THE REVIEW OF

Performance Assessment of a

RISK

LLW disposal facdity

• Pubhsh final Branch

TBD,

ASSOCIATED

Techncal Positen

Dependent

WITH WASTE

REPOSITORIES

on

Resources

10

.

.

.

.

.

.

. . .

_.

. _ _ _ _ _

. _ _ _ _ _ _ _ _ _ _ - _ _ _ .

.

.

..

.

.

RegulWory ActMty

Obrectrves

Methods

Target

Lead

Status (tfus

Schedule

Offee(s)

cuarter)

44

RISK

• Develop and demonstrate a nsk

• Develop and demonstrate

B98

Changed (Note 12)

ASSESSMENT OF

assessment for Industnal gauges

methods for determining the

MATERIAL USES

containing cessum-137 and

nok associated with industnli

cobalt-60 using PRA and other

gauges containing cessum..J7

related techrwques

and cobalt 60

• The assessment should allow

• Fnal report as NUREG 12s96

Changed (Note 12)

for mcdfcaton based on

changes m regulatory

• Wcrkin0 Group with

&96

Changed (Note 13)

requrements

contractor assetance to

identify and document a

• Use empircal data as much as

t.achnical basis for a risk.

practcatde

informed approach to the

regulaton of nuclear

• Develop and demonstrate nsk

byproduct matenal, and to

assessment rnethods for

develop plans for a graded

appicaten to medcal and

approach to nuclear byproduct

industnel heensee actmtes

matenal regulation based on

nsk mformaton

45

FRAMEWORK

• develop a framework for

• Provxkt plan for developing

1/98

NMSS

Changed (Note 14)

FOR USE OF PRA

applying PRA to nuclear matenal

framework

IN REGULATING

uses, similar to the one

NUCLEAR

deve g'ed for reactor regulation

• Complete framework

TBD

MATERIALS

(SECY-95-280), where

appropnate.

!

'

1

11

.

_ _ _ - _ _ _ _ _ - - _ _ - - - _ - _ - - _ -

- - - -

.

.

5.0 HIGH LEVEL NUCLEAR WASTE REGULATION

,

Cogulatory ActNRy

ObjectNes

Methods

Target

Lead

Status (this

Schedule

Offee(s)

cuarter)

51

REGULATION OF

• Develop guidance for the NRC

• Assist the staff in pre-

Ongoing

NMSS

{

HIGH-LEVEL WASTE

and CNWRA staffs in the use of

hcensing actNtes and in

PA to evaluate the safety of HLW

Icense appicaten revews

programs

• Develop a techncal

assessment capattityin totab

system and subsystem PA for

use in hcensing and pre-

j

licensing revews.

-

• Combine specialized

techncal drsciplines (earth

scences and engineer )

with those of s,ystem__

i

,

_

rs

to improve ...u my.

• Idenhfy agnifcant events,

• Perform sensitNtty studes

Ongoing

NMSS

processes, and parameters

of key techncallasues ussng

effecting total system

iterative performance

performance

assessment (IPA)

• Use PA and PSA methods,

• Assist the staff to maintain

Ongoing

NMSS

results and insignts to evaluate

and to refine the regulatory

proposed change s to regulatons

structure in HLW disposal

gove

the ootential repository

regulaticas that pertain to PA.

at Yucca

ntain.

• Apply IPA analyses to advise

i

l

EPA in its development of a

Yucca Mountain regulaten

1

[

~

• Apply IPA enalyses to

develop a site-specife

regulation for a Yucca

Mountain site

• Continue PA activites dunng

• Provide guidance to the

Ongr ng

NMSS

interactions with DOE dunng the

DOE on site charactenzation

pre-heensing phase of repository

requirements ongoing oesign

development, site

work, and hcensing issues

charactenzaton, and repository

important to the DOE's

design.

development of a complete

t

and high-quality license

,

appleation.

• Compare results of NRC's

Iterstrve performance

assessment to DOE's VA to

identify major

differences /rssues.

3.3

APPLY PRA TO

• Demonstrate methods for PRA

• P opere user needs letter to

497C

RUS/NMSS

SPENT FUEL

of spent fuel storage facihtes.

RES.

STORAGE

FACILITIES

• Conduct PRA of dry cask

9<99

storage.

3.3

CONTINUE USE OF

Use PRA methods, results, and

• t,pdate the database on

End cf FY

NMSS

RISK ASSESSMENT

insights to evaluate regulations

transportation of raccactive

99

IN SUPPORT OF

governing the transportation of

matenals for future

RADICACTIVE

radcactive matenal.

MATERIAL

appications

TRANSPC9TATION

• Revalidate the results of

&99

NUREG-0170 for spent fuel

shipment nsk estimates.

12

v

'

A

_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

_

'

.

4

Notes

,

1.

The general regulatory guide and btandard Review Plan for use of PRA in plant-specific

current licensing basis changes will be transmitted to the Commission in the near future,

To permit efficient incorporation of the resolution of policy issues contained in these

documents into the application-specific regulatory guides and SRP sections, completion of

these guides and sections has been delayed until March 31,1998, a change from their

previous completion date of December 31,1997.

2.

The staffs RI IST team is currently working on a draft SE for the Comanche Peak RI-IST

program. The staff and TU Electric have been actively interacting through meetings and

discussions as the licensee develops a RI IST program description that is sufficiently

detailed and consistent with the draft Rl IST guidance provided in DG 1062. TU Electric

has indicated that it would be able to complete a daft revision to their RI IST Program

Description by the end of January 1998. The staff will continue to develop a draft SE

based on the licensee's responses to the staffs RAls and discussions with the licensee.

Assuming TU Electric finalizes its RI IST Program Description by mid February 1998, the

staff anticipates having a completed SE to the Commission on the proposed RI IST

program for Comanche Peak in March 1998.

3.

The completion date for the Graded Quality Assurance pilot application has been revised to

July 1998 to reflect the anticipated issuance date of the final GOA inspection guidance.

4.

With respect to the risk-informed TS pilot program, the staf' received a supplemental

amendment request from SONGS in early January 1998 tc put the configuration risk

management program (CRMP) description into the SONGS TS. The staff will review the

CRMP and, if acceptable, issue the risk-informed TS amendments for SONGS. Once

similar supplemental amendment requests are received frora the remaining pilot licensees,

the staff willissue those pilot plant amendments. Based on information from the CEOG,

the staff expects to receive the majority of the supplemental pilot amendment requests in

the first quarter of 1998. With receipt of the SONGS supplemental request, the staff

anticipates completing the SONGS review as the lead pilot plar:t and issue the amendment

by March 31,1998. This is a change from the previous date of December 31,1997, for

issuance of the lead pilot plant amendment, because of the decision by the originallead

plant not to oursue risk-informed TS changes at this time.

5.

The NRR Intipection Program Branch proposals for revising core inspection procedures

have been transmitted to the appropriate NRR technical branches having responsibility for

l

specific core inspection procedures. Due to the large number of branches involved,

completing all indivi :ual branch concunences is anticipated to take an additional two

months. The revised completion date for this task is February 1998.

6

A deci ion has been reached to generate the risk-informed regulatory documents in a

sequential manner, with the application specific guidance following the general regulatory

guide and standard review plan. Under this schedule, the regulatory guide for graded QA

will be finalized by the end of March 1998. Since the graded QA inspection procedure will

be dependent upon the technical content of the companion regulatory guide, the draft

graded QA Inspection Procedure will be prepared by April 1998 and finalized in July 1998

after having received appropriate NRC reviews.

13

,

a

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.

7.

As part of the IPE follow-up program, the staff is in the process of identifying generic issues

'

to be audited. These issues will be those which have been explic.itly identified and

addressed by the licensee as part of the IPE process.

A report that identifies the above generic issues and staff views on the adequar.,y of the

proposed resolution is under preparation. The report will provide the basis for the selection

of generic safety issues to be audited and selected plants. The staff has moved the

completion date for this milestone to March 1998, in order to utilize the report in the audit

process.

In addition to the .uove issues, RCP seal LOCA had been identified as a dominant

contributor to core damage frequency in many PWR IPEs. The staff has a separate

ongoing activity in RES to address this issue under Generic Safety issue 23, and will utilize

IPE insights in the proposed resolution.

8.

In at SRM (9700207) datsd May 28,1997, the Commission requested that the staff

proviie the scooe and Nhedule of activities related to using IPE results to assess

regulatory effectianc.s in resolving major safety issues. . lith respect to scope, the staff

identified three major safety issues for assessment. The selection had been based on both

the potential nsk significance of the issue, and the fact that probabilistic techniques were

!

used extensively in the resolution process. These issues include:

1. Resolution of USI A-44 Station Blackout at Nuclear Power Plants

2. Resolution of USI A-45 Decay Heat Removal Reliability

3. Resolution of USI A-09 Anticipated Transient Without Scram

To evaluate the three major issues, the staff will utilize both representative plants, and

information contained in NUREG-1560, to audit and draw conclusions regarding regulatory

effectiveness. Information generated under Task 1.6 and Task 1.10 will also be integrated

into the assessment process. In particular, the RCP seal LOCA and station blackout

issues are closely related; the station bisckout analysis in this activity willincorporate the

results of the RES seal LOCA analysis discussed in Note 7.

These tasks may expand the staff's consideration cf other safety issues and effectiveness

of the regulatory process. The staff willinform the Commission of any additional safety

issues that come under consideration. The staff plans to complete it analysin of the three

issues by the end of December 1998, and will recommend ai that time any additional staff

action.

9.

Due to personnel being assigned to higher priority activities, such as risk-informed pilot

initiatives and IPE followup activities, the staff is reassessing their position regarding the

development of an SRP, especially since there are no new advanced design certification

submittals anticipated.

10.

The staff has reviewed all the 76 IPE submittals and issued staff evaluation reports (SERs)

on their findings to each licensee. In three of the SERs, it is indicated to the licensees that

the staff was not able to conclude that the licensee met the intent of Generic Letter 88-20

for their plant (s). These three IPEs include Crystal River 3, Susquehanna 1&2, and

Browns Ferry 3. The licensee for Crystal River 3 nas indicated their intention to submit an

updated analysis (February ;998) addressing the staffs concerns, it is anticipated that the

14

___ - _ __- ____ _ ___ _____

.

t

review of this new IPE submittal will be concluded in June 1998. Discussions are still

ongoing wit! licensees regarding Susquehanna 1&2 and Browns Ferry 3.

.

11.

This 'ask has been subsumed into the office operating plan, which is periodically updated.

12.

The target schedule for the work to develop and demonstrate a risk assessment for

industrial gauges containing ces;um 137 and cobalt-60 using PRA (and other related

techniques) has been extended from July 1998 to September 1998. The extension is due

to difficulties in obtaining data from non-licensees related to actual and potential doses to

the public resulting from gauges which enter the scrap metal cycle.

13.

The target schedule for the work to develop and demonstrate risk assessment methods for

app!ication to medical and industrial licensee activities has been determined to be

September 1998 based on scheduling of a planned Commission paper on the topic.

14.

The target schedule for providing a plan for developing a framework has been extended

from October 1997 to December 1997 to permit interoffice coordination.

l

I

15

.

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4

Attachment 2

Sta# Requirements Memorandum

dated May 28,1997

i

i

P

/.ES/ DST

TEL 301-415-5062

Jur. 03'97

11:15 No.007 P.03

Acticn (Morrison, Nu/

.

Collins. NRR

.

• • • t

UNffED 5MES

Cygg ,Cg]]gn

g

NUCLEAR Rt0UL.AT0ftY C04mm8840N

Jon]en

-

y

• ApesGTon.or same

Thotpson

%

AN RESvuNse, FLk:AbE

(

May 28, 1997

REFER TO g g 70507

Blahe

-

Ross, AE0D

!

MEMORANDUM 70:

L. Joseph Callan

.

Director for Operationa

Exec # iv' k(%

FROM:

Johr v. Hoyle, Secretary

~

SUBJECT:

STAFF REQUIREMENTS - ERIEFING ON IPE INSIGHT

REPORT, 2:00 P.M

WEDNESn&V, M&V 7,

1997.

COMMISSIOhT.RS' CCNFERENCE ROOM, ONE WHITE

FLINT NORTH, ROCXV:LLE, MARYLAND (OPEN To

PUBLIC ATTENDANCE)

-

.-

The countiwmivu was briefed by the NRC stef f en the Individual

Flant Examination (IPE) insight report.

The Comission asked the

st a:l'

to expedite activhium in .Le !vilu lug areas:

(U using

IPE results to prioritize inspection activities: (2) improving

regional capabilities for the use of PRA and risk insights; and

(3) providin

ted inspecttr training,

fEDO)

MES)j

iSECT suspense:

wu'p3o

9700206

The Commission asked the stait to provice tne scopte anc scr.edule

of activities relar.ed to using IPE results to assess regulatory

effectiveness in resolving r.ager safety issues.

The comissien

stacifically reqv.2ted that the staff provide an estimate cf the

average cost to respond to the Staticn Blackout. rule per persen-

rem averted 1.n achieving an average reduction in core damage

frequency of 2E-5/RY.

These activities should be coordinated

with the regulatory effettiveness organization,

fEBO-)

(NRR)

(SECY suspense:

6/27/97)

9700207

After the IPE database has been placed on the Internet, the staff

should consider allowing licensees to update their IDES

voluntarily to reflect changer i.a plant configuration.

(RES)

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.

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_ _ _ _ _

_ - _ _ _ _ _ _ _ _ _ _

..

.

-d

Attachment 3

lPE Follo36222matarn

The IPE program was initiated to have licensees evaluato the;r plants for vulnerabilities to

severe accidents and to take actions to correct these vulnerabilities, where appropriate. In this

process it was recognized that licensees would gain an appreciation of their plant's overall

susceptibility to severe accidents which would help in developing accident management

tirategies and programs. In this regard the IPE pregam was principally for the benefit o'

licensees. Now, however, as a result of completion of the IPE reviews (except for the three

plants where comn!stion is still under discussion) and insights report (NUREG-1560), the staff is

now in a position to utilize these results to follow up and see if:

any additional plant specific improvements are warranted,

-

licensees have followed through on the actions they indicated they were taking as a

-

result of their IPE, and

any additional generic regulatory activities should be undertaken.

-

To accomplish this the staff has developed an IPE followup program which willinvolve the

efforts of RES, NRR and the Regions. The followup program will consist of the following

sctivities:

1

1)

reviewing the iPE results for risk significant items that may warrant further attention.

Examples of the screening criteria for selection of plants and items for additional

followup are as follows:

any contributor with a aCDF' >10-5 RY or

/

-

any contributor with a 6LERF2 >104/RY

-

2)

reviewing the IPE results for similar plants and whether or not actions taken by some

l'

plants are applicable to other plants of similar design,

{

3)

reviewing licensee responses to specific containment performance improvement items

identified in the IPE generic letter supplements to see if additional actions are

warranted,

4)

reviewing the basis for very low risk contributors that appear to be out of line with other

plants (i.e., was the analysis overly optimistic and should further action be taken?),

5)

assessing licensee stated actions (e.g., safety enhancements) resulting from their IPE to

see if, in fact, they have been completed,

' Core Damage Frequency

8 Large Early Release Frequency

1

. _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ .

,

9

'

2

6)

assessing licensee accident management programs to see if, in fact, they reflect the

results, assumptions and actions from the IPE. This action will be carried out through

the staff assessment of the licensee's Severe Accident Management Guidelines

(SAMG).

7)

assessing the results for their implications for the resolution of generic safety issues or

other major safety issues.

These activities are in addition to actions already underway to incorporate '.ne IPE insights into

the NRC inspection program.

Implementation of this program will consist of RES providing to NRR information related tc

activities 1 through 4 above with NRR then discussing with licensees the appropriateness of

additional actions. This will provide licensees an opportunity to provide updated information

related to these activities and ultimatcly for NRR to take regulatory action, if such action is

warranted and can be justified by the backfit rule. Activities 5 and 6 will be performed by NRR,

,

with Regional followup as necessary. Activity 7 is addressed by items 1.6 and 1.7 of the PRA

Implementation Plan.

High priority issues identified in the screening process will be pursued as they are identified.

Dates for accomplishing these activities relative to IPE followup are:

RES supply information to NRR on items 1-4

12/98

.

NRR interact with licensees on appropriateness

6/99

.

of additional actions for items 1-4

Backfit analysis and actions complete

12/99

.

Item 5, identify items for Regional followup

9/98

.

Item 6, identification of IPE insights for Se'. sre Accident

9/98

.

Management Guidelines

The specific IPEEE followup schedule will be developed following the completion of the IPEEE

4

reviews.

.

-