ML19354D689

Summarizes 355th ACRS Meeting on 891116-18 Re Draft Suppl 2 to Generic Ltr 88-20, Accident Mgt Strategies for Consideration in Individual Plant Exam Process & Proposed Resolution to Generic Issue 87

ML19354D689
Person / Time
Issue date:	12/11/1989
From:	Michelson C Advisory Committee on Reactor Safeguards
To:	Carr K NRC COMMISSION (OCM)
Shared Package
ML19354D688	List: ML19354D687 ML19354D689
References
REF-GTECI-087, REF-GTECI-NI, TASK-087, TASK-87, TASK-OR ACRS-GENERAL, GL-88-20, NUDOCS 8912290081
Download: ML19354D689 (6)

Text

,.

)

,s

1. g UNITO STATES

,,[

NUCLEAR REGULATORY COMMISSION q-

• ' g ADVISORY o0MMITTEE ON REACTOR SAFEOUARD6 WASHINGTON, D. C. NeD6 December 11, 1989 The Honorable Kenneth M. Carr Chairman U.S. Nuclear Regulatory Commissior.

Washington, D.C.

20555 Dear Chairman carrt

SUMMARY

REPORT - THREE HUNDRED FIFTY-FIFTH MEETING OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS, NOVEMBER 16-18, 1989 During its 355th meeting, November ' 16-18, 1989, the Committee completed the activities noted below and prepared the following reports and letters.

Copies of these reports and 2atters have been i

provided to your REPORTS TO THE COMMISSION 0-Draft Sunclement No.

2 to Generic Letter 88-20.

" Accident Management Strateales for Consideration in the Individual Plant Examination Process" (Report to Chairman Carr, dated i

l November 20, 1989) l o

Proposed Resolution of Generic Issue 87.

"HPCI Steam Line L

Break Without Isolation" (Report to Chairman Carr, dated

. November 20, 1989) o Annointment of ACRS Memberg (Report to Chairman Carr, dated November 20, 1989) l l

o Examole of NRC Emoloyees Inventina or Imoosinu New Reauirements (Letter to Chairman Carr, dated November 20, 1989)

Coherence in the Reculatory Process (Recort to Chairman Carr, o

dated November 24, 1989)

LETTERS TO THE ACTING EDO o

Relationship of the Ouantitative Safety Goal to the concept of Adecuate Protection (Letter to J.

Taylor, dated November 20, 1989) 8912290081 891227 FDR ADVCMNACG{lk

,. s The Honorable Kenneth M. Carr 2

December 11, 1989 o

Module 1 of the Draft Safety Evaluation Report for the Advanced Boilina Water Reactor Desian (Letter to J.

Taylor, dated November 24, 1989)

OTHER ACTIONS. AGREEMENTS. ASSIGNMENTS. AND REOUESTS o

Nine Mile Point Unit 1 Restart The Committee decided to continue its review of the restart of Nine Mile Point Unit 1 subsequent to the staff's approval of the restart.

Since the Committee's review may take place during the startup phase, the Committee decided that there is no need to delay the restart of the plant.

(See memorandum from R. Fraley for J. Taylor, dated November 24, 1989.)

o ProDosed 10 CFR Part 55. Operat,grs' Licenses The Committee decided to review the proposed revision of 10 CFR Part 55 to require compliance with fitness for duty programs and conforming modifications to the Commission's enforcement policy.

(Memorandum from R.

Fraley for J.

Roe dated November 24, 1989) o NUMARC/NUPLEX Industry License Renewal Toolcal Reports The Committee decided that the ACRS should be given an opportunity to review and comment on the industry topical l

reports related to license renewal that are being prepared by NUMARC.

Some of these reports have already been prepared and submitted to the NRC staff for review.

The NRC staff has been requested to provide copies of these reports to the ACRS and their evaluation of these reports when they are availabic.

(See memorandum from R. Fraley for T. Murley, dated December 5, 1989.)

o ACRS/ACNW Responsibilities The Committee acknowledged the division of responsibilities between the ACRS and the ACNW specified by the Commission in the memorandum from Chairman Carr dated November 6, 1989, and discussed mechanisms to provide for joint input in areas of mutual interest.

o Standardized Advanced LWRs The NRC staff briefed the committee regarding the status of its review of the following evolutionary and advanced light water reactor designs:

I.

+

The Honorable Kenneth M. Carr 3

December 11, 1989 i

Westinghouse WAPWR SP/90 Westinghouse AP 600 i

Combustion Engineering System 80+

The tentative schedule proposed by the staff for the ACRS full Committee consideration of these designs is given below:

Westinghouse WAPWR SP/90 (PDA)

- April 1990 Westinghouse AP 600

- April 1990 (Initial discussion)

Combustion Engineering System 80+ - February 1990 (Licensing Review Basis Document) o Three Mile Island Nuclear Station Unit 2 The NRC staff briefed the Committee regarding the status of the recovery efforts, including the investigation of the reactor '.;ssel lower head indications at TMI-2.

O ACRS Meetina Dates for Calendar Year 1990 The Committee approved the following meeting dates for calendar year 1990:

357th Meeting January 11-13, 1990 358th Meeting February 8-10, 1990 359th Meeting March 8-10, 1990 360th Meeting April 5-7, 1990 361st Meeting May 10-12, 1990 362nd Meeting June 7-9, 1990 363rd Meeting July 12-14, 1990 364th Meeting August 9-11, 1990 365th Meeting September 6-8, 1990 366th Meeting October 4-6, 1990 367th Meeting November 8-10, 1990 368th Meeting December 6-8, 1990 (The Jura 1990 meeting dates may change to accommodate the Second Int 00 national Conference of Advisory Committees.)

o Meetina with the Commission The committee p'roposed a meeting with the Commissioners during the 356th ACRs meeting, December 14-15, 1989.

Items tentatively scheduled for discussion are:

Status of ACRS activities related to the development of containment design criteria for future plants.

o

,k 4

The Honorable Kenneth M. Carr 4

December 11, 1989 NRC Safety Research Program budget (Tentative)

" Management" of Licensee activities by NRC Regional Staff.

(This meeting has been deferred to January 1990.)

SUBCOMMITTEE. MEETINGS Since the last report of ACRS activities, the following subcommittee meetings have been holdt o

Joint Containment Systems and Structural Encineerina, October 17, 1989 The Subcommittees discussed containment design criteria for future plants with invited speakers from the industry and r.ationa3 laboratories, o

Astyanced Boilina Water Reactors (GE ABWR), October 31, 1989 - The Subcommittee reviewed the NRC staff's Draft Safety Evaluation Report related to Module 1 of GE ABWR.

Advanced Pressurized Water Reactors, November 3, 1989 -

o The Subcommittee discussed the Westinghouse Advanced Pressuriced Water Reactor Design (RESAR-SP/90).

o Thermal-Hydraulic Phenomena, November 8-9, 1969 The Subcommittee discussed the capability of the thermal-hydraulic codes to model EWR core power instability, and the key thermal-hydraulic design aspects of the GE ABWR related to the ECCS and LOCA analyses, o

Thermal-Hydraulic Phenomena, November 14, 1989 The Subcommittee discussed selected portions of the NRC thermal-hydraulic research program, including future research needs in this area and the recent ACRS report commenting on thermal-hydraulic research, o

General Electric Reactor Plants, November 14, 1989 - The Subcommittee reviewed the restart of Nine Mile Point Unit 1.

o Reaulatory Policies and Practices, November 15, 1989 The Subcommittee continued its discussion on the integration of the regulatory process, o

Thermal-Hydraulic Phenomena, December 7,

1989 The Subcommittee discussed:

the proposed NRR and RES programs for the resolution of the issue on Interfacing

O

.p The Honorable Kenneth M. Carr 5

December 11, 1989 Systems IOCA; the status of the RES' Technical Program Group's efforts to apply the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology to calculation of a small-break LOCA; and the status of development of the Westinghouse best-estimate ECCS/LOCA model.

SPECIAL MEETING o

ACRS Meetina with the Canadian Advisory Committee on Nuclear Safety, November 1-2, 1989 - The Committees discussed several safety-related matters applicable to nuclear power plants including institutional quality assurance and

attitude, nuclear power plant personnel selection and training, severe accident analysis, and containment performance criteria.

ECHEDULE FOR THE 356TH MERIJjiQ The committee agreed to the following tentative schedulo for the 356th LCRS meeting, December 14-15, 1989:

Nuclear Power Plant Access Authorization - Review and report o

on proposed final rule on Personnel Access Authorization Requirements for Nuclear Power Plants (10 CFR Part 73.56).

l o

Containment Performance Imorovement Proaram Review and report on proposed NRC program to improve containment performance-during severe accident conditions for all containment types except Mark I containment.

Technical Trainina and Oualification Procram for NRC Emoloyees o

Briefing by NRC staff representatives regarding training courses at the NRC Technical Training Center at Chattanooga, TN.

Fitness for Dutv - Review and report on proposed revision of o

10 CFR Part 55 to require operator compliance with NRC fitness-for-duty programs and conforming modification to the Commission's enforcement policy.

o Meetina with Actina EDO Discussion regarding lack of coherence in the NRC regulatory process.

o ACRS Subcommittee Activities - Hear and discuss reports of ACRS Subcommittees regarding the status of assigned activities regarding safety-related matters, including the activities related to thermal hydraulic phenomena, etc.

(

.,s The Honorable Kenneth M. Carr 6

December 11, 1989 o

Evaluation of Ocerational Data Briefing and discussion regarding use of SALP ratings in the regulatory process and elsewhere.

o Meetina with the commissioners - Discussion of the following

^

items:

Status of ACRS activities related to the development of containment design criteria for future plants NRC Safety Research Program budget (Tentative)

" Management" of Licensee activities by NRC Regicnal staff.

(This meeting has been deferred to January 1990.)

Sincerely,

<fs G

,,g Carlyle Michelson Acting chairman l

l l

l l

1 1

1

l

• {

UNITED STATES NUCLEAR REGULATORY COMMISSION

(

)

ADVIDORY CDanslTTEE ON REACTOR SAFEOUARDS l

wAeNessetoev,0. c.seems November 20, 1989 l

I The Honorable Kenneth M. Carr Chairman U.S. Nuclear Regulatory Commission Washington, D.C.

R0555

SUBJECT:

DRAFT SUFPLE*4ENT NO. 2 TO GENERIC LETTER 88-20, ' ACCIDENT l

MANAGEMENT STRATEGIES FOR CONSIDERATION IN THE IN01VIDUAl.

PLANT EXAMINAYION PROCESS" Dcring the 355th meeting of the Advidury Committee on Reactor Safeguards, November 1618, 1989, we discussed the subject document with the NRC staff. We also reviewed a draft NUREG/CR report entitled, " Assessment of Candidate Accident Management Strategies,' that the staff proposes to send as an enclosure with the supplement to the generic letter.

We had the l

benefit of these documents which are referenced.

Our Subcommittee on Severe Accidents met on September 20, 1989 to discuss this matter.

We conclude that the information in these two documents will be useful to licensees in the process of performing Individual Plant Examinations, and we agree that the documents should be issued.

1 The draft NUREG/CR report referred to describes strategies for accident management that are said to be PRA based.

However, the report does not include information on the risk reduction that might be attributed to the strategies.

This information would be useful to. hose considering the strategies.

We recommend that this information be added if it is rea-sonably retrievable from existing sources.

We observe that a number of the strategies described in the draft NURES/CR report either overlap or are very similar to the content of the emergency operating procedures that are either being developed or are already in place in many plants.

We believe that labelling these procedures as accident management strategies where others label them as emergency operating procedures is likely to lead to confusion on the part of both the NRC staff and the industry.

Sincerely Forrest J. Remick Chairman g-i%Q p

~

i The Honorable Kenneth M. Carr November 20, 1989 References 1.

U.5. Nuclear Regulatory Commission, " Accident Management Strategies for Consideration in the Individual Plant Examination Process " Draft Supplement No. 2 to Generic Letter 88-20, dated November 8, 1989 (Predecisior,al) 2.

U.S. Nuclear Regulatory Connission, " Assessment of Candidate Accident Management Strategies," Draf t NURE6/CR Report (Unnumbered)

Prepared by BNL, October 1989 l

~_

g UNITE 3 STATES

... p l

ADVISORY COMMITTEE ON REACTOR SAFEOUARDS NUCLEAR RECULATORY COMMISSION n

g wasmwovon.o. c, sones November 20, 1989 The Honorable Kenneth M. Carr Chairman U.S. Nuclear Regulatory Connission Washington, D.C. 20555

SUBJECT:

PROPOSEDRESOLUTIONOFGEkERICISSUE-87(GI-67),"HPCISTEAM LINE BREAK WITHOUT ISOLATION" Octug the %5th meeting of the Advisory Comittee on P.eactor Safe-svards November 16-18, 1989 we discussed the EC st6ff's propcsed resolution of 61-87. This sub, ject was also considered during our 358th 1

meeting on Octeber ti-6,1989. Our Subcomittee on Mechanical Components met with members of the NRC staff on October 3,1989 te discuss this matter. He also had the benefit of the document referenced.

In spite of its title, the scope of GI-87 includes all safety-related t

l valves that ray be required to close against high differential pressures and/or high flows such as experienced during a large downstraam pipe break.

Possible deficiencies in the performance of such valves have been recognized recently as the result of operating experience and of tests and have been addressed by the staff in Generic Letter 89-10,

• Safety-Related Motor-0perated Valve Testing and Surveillance."

This generic letter requires each licensee to undertake a program to identify safety-related valves that may not perform adequately under design-basis conditions.

Unfortenately, as pointed out in our May 9,1989 report related to Generic Letter 89-10, the design basis for the HPCI steam line and other valves in some plants may not specify the type of heavy duty that is of concern.

This means simply that the heavy duty loads for these valves do not have to be considered in the program required by Generic Letter 89-10, and the deficiencies addressed by GI-87 will not be remedied.

Unless Generic Letter 89-10 is amended to require updating or revision of the design basis for valves of this type we do not consider the requirementsofGenericLetter89-10asacompleteresolutionofGI-87.

Sincerely.

Forrest J. Remick Chairman

Reference:

Memorandum dated July 17, 1989 from R. Wayne Houston, Office of Nuclear Resolution Regulatory Research,(NRC, for R. F. Fraley, ACRS,

Subject:

GI-87), "HPCI Stea of Generic Issue-87 l

k% Y)$@fS f) '

\\

UNITED 8TATES NUCLEAR REGULATORY COMMISS60N J

ADVie0RY COtahAITTEE ON REACTOR SAFEOUARDS wAnnmotom.o.c.somes November 20, 1989 The Honorable Kenneth M. Carr Chaiman U.S. Nuclear Regulatory Coenissier.

Washington, D.C.

20$$5 Dear Chairman Carrt You heve recently indicated that you have an interest in receiving examples of MRC esp oyees inventing or imposing new requireseats that are not part of the egittestely constituted body of regulations.

We share your concern in this area and feel that the attached letter is such an example.

Sincerely.

Forrest J. Remick Chaiman

Attachment:

Letter (without enclosures) dated September 11, 1989 to W. F. Conway, Executive Vice President, Nuclear, Arizona Nuclear Power Project, from J. B. Martin, Regional Administrator, NRC, Region V,

Subject:

Report of Meeting with ANPP Management g4gS NM 1

.m.-

,i,.

.v ass t

n NUCtEAR REGULATORY COMMIS$10N

{

l RselDN V A

\\.*...+/.

1460 MARIA LANs. Suits 210 tWALWUT CRss K. CALIFORNIA 94996 SEP 11Itu Docket Nos. 50-528, 50-529 and 50-530 Acirona Nuclear Power Project P. O. Box 52034 Phoenix, Arizona 85072-2034 J,ttention:

W. F. Conway Executive Vice President, Nuclear 1

Gentlemen:

i g -*

SUBJECT:

REPORT OF MEETING WITH ANPP MANAGEMENT i

This refers to a meeting held with yourself, and members of your 'staf f and myself, and other siembers of the NRC staff, at the Arizona Fublic Service

+

i Company Offices in Phoenix, Arizona en Septetsber 1,1989.

Thesubjects L

l discussed are summarized in Meeting Report Nos. 50-528/89-42, 50-529/89-42, and 50-5.40/89-42, onclosed herewith.

During our meeting, I expressed to you sy extreme concern regarding the failure of ANPP_ managers to devote any significant amount of time to the observation of activities in important areas of the plant. Myfrustrationis' heightened due to the fact that this issue has been previously raised at past i-j management meetings with your staff.

Furthemore, the relatively large number l

of ANPP managers who are new to Palo Verde would seem to require an increased effort on their part to get out in the plant and learn first hand about the facility and the staff. As I stated to you during the meeting, I consider the

+

failure of the ANPP management team to spend time in the plant to be a major oversight in your efforts to implement positive changes at Ptle Verde.

I strongly recommend that you act in this area promptly and thoroughly.

In discussing the Unit 2 Main Feedwater System overpressurization event, and l

the event wherein an operator at Unit 2 failed to properly flash the main i

generator field, we identified to you that your staff had not exhibited the appropriate instincts when problems arise.

You should continue to reenforce to your staff the basic principles you stated at our June 5 meeting of stopping in the face of uncertainty and rea;; ting conservatively when faced with questionable situations.

Regarding your efforts to review long standing concerns, your actions as outlined in our meeting appear appropriate. We again retterate the need to perform this review thoroughly, particularly in light of the backlogs of various open issues.

In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosures I

)

will be placed in the NRC Public Document Room.

y p w v. W

• 6, i

.g.

Should you have any questions concerning our minutes of the meeting, documented in the enclosed meeting report, we will be pleased to discuss them with you.

Since ly, 8f f

J. B. hartin Regional Administrator

Enclosures:

2.

Report Nos. 50-528/89-42, 50-529/89-42, 50-550/80-47

2..ANPP Presentation Package cc w/ enclosures:

W. F. Quinn, ANPP B. E. Ballard, SR.. ANPP T. D. Shriver, ANPP C. N. Russo, ANPP D. Canady, ANPP A. C. Rogers, ANPP L. Bernabei, GAP J. R. Brown, ACC A. C. Gehr. Esq.. Snell & Wilmer t

l l

l l

l l

~

[k y u.

a0 O **%}'.

~

~

L**

UNITED STATES i

M1 NUCLEAR REGULATORY COMMISSION n

D {~

I.

- ADVis0RY 00MMITTEE ON REACTOR SAFEGUARDS waswsworow. o. c. noons

....+

November 24, 1989 L:

h I

L L-The Honorable Kenneth M. Carr Chairman l

U.S.' Nuclear Regulatory Comission L

Washington, D.C.

20555 I

SUBJECT:

C0HERENCE IN THE REGULATORY PROCESS l

During the 355th meeting of the Advisory Comittee on Reactor Safe-l guards, November 16-18, 1989, we discussed the need for a strategy for L

achieving coherence in the rer 1atory process.

Our Subcomittee on L

Regulatory Policies and Pra

- '3 also met on August 9 and November l-

15. 1989 to discuss this

% r.

This is in response to a Staff e the regulatory process.g for "ACRS Requirements Memorandum de

• August 18, 1989 askin thoughts on-how to best ir; c L

As we have observed ir

,er of the referenced reports, the NRC seems to suffer increa, from a lack of coherence in the formu-

-lation and. implementation of its regulatory strategy. This is hardly

-a subject of which the Comission:is unaware, and it is a problem that

=is perhaps unavoidable as the body of regulatory practice grows with time, and institutional memory fades correspondingly. Nonetheless, it poses problems for those who try both to understand the Comission's regulatory policies and-to construe the staff's actions in. the light of those policies.

It seems to us axiomatic that regulation will be most' effective in support of nuclear safety--our comon objective-if it is coherent and defensible, and thereby understood and respected by those who are regulated.

The staff has, on' occasion, been asked to describe its efforts to deal with-these problems, and has responded (e.g., SECY 88-178, " Policy

-Statement Integration," June 9, 1989; and memorandum for Chairman Carr from J. M. Taylor, Acting Executive Director for Operations (EDO),

" Integrated Approach on Regulatory Matters," October 18,1989) by describing those programs in place to achieve " integration," which are, in effect piling new programs on top of an assembly of un-affected and unintegrated parts.

Not only can integration not be accomplished by ordinance, but there-is a real and important distinc-tion between integration and coherence--the latter is the real objec-tive.

Coherence means that all the parts pull in the same direction, not that they are put in the same box.

It cannot be attained by 09%Mb

['

4 The Honorable Kenneth M. Carr November 24, 1989 i

re sackaging of existing programs; integration does not generate coterence.

As we have said, there are so many examples, and the )roblem is so well known, that it may seem like overkill to list examples, but it is l

useful to do so, if only to note that they differ in kind, so there is no one general sweeping solution.

There are some cases in which there is no problem of coordination among the various offices, but the problem is one of drawdown of the NRC and industry resources, with negative consequences that are clear but hard to identify.

This happens when any office acts, however worthily, on its own.

These are problems only the Comission can address.

L L

There are cases, like access authorization and fitness for duty, in which individub1 offices proceed, again however worthily, with closely related initiatives that arrive at the end stage before they finally come together in the Comittee to Review Generic Requirements (CRGR).

Those problems properly belong to the EDO, but there is something incongruous in having his influence felt only near the end of the process.

Indeed the CRGR was created to apply an end-game palliative to some of these same problems.

Such coordination would be more effective earlier.

4 l

There is the problem of the Regional Administrator:, who sometimes have practices that differ from each other, and from Headquarters.

In l

the end, it is the Regional Administrators with whom a licensee hM most contact, and who embody NRC in the field, and there are too Any cases in which their dicta go well beyond the policies yet by the Comission, l

l There are cases, like the initiatives on accident management and emergency operations, in which the hmission guidance is sufficiently unclear to permit separate tracks for different staff elements.

There are pervasive problems, like the applicability of the Safety Goal Policy and the Severe Accident Policy, in which the Comission seems to be playing a passive role, reacting to staff or ACRS initia-tives.

Again, neither the EDO nor we can help in such matters.

We all-can and do provide advice, but the Comission's safety philosophy ought to guide us.

The Comission has recognized these issues in the past and has pro-mulgated a number of important policy statements to, as we see it, provide an underlying coherence to its policies.

It has every reason to be proud of these efforts, but it remains necessary to find ways of

.,6

• The Honorable Kenneth M. Carr November 24, 1989 diffusing them into the fabric of a large and complex agency. The two principal policies that are relevant to this subject are the Safety Goal Policy and the Severe Accident Policy statements.

The Safety Goal Policy lays out the basic objective of the agency, to regulate in such a way as to provide reasonable assurance that a certain quantita-tive level of safety is achieved in the use of nuclear power. Nothing car be more fundamental, and we believe (and have seid before) that that policy should serve as a clear statement of your aims.

The Severe Accident Policy should, if there is any ambiguity, be applied in such a way that it conforms to and supports the safety goals.

Coherence in any of the NRC's activities should be sought through the litmus test of relevance to the safety goals.

That cannot be done by leaving every branch and every regulator to assess their actions by carrying out an analysis of the implications, to the point at which the ultimate effect on the health and safety of the public can be determined. These are complex assessments, replete with uncertainties, and it would be absurd for each member of the staff to measure their own activities in terms of the overall objec-tives of the agency.

No large organization functions that way, nor can it. People need to do more narrowly prescribed jobs that nonethe-less contribute to the strategy.

In our reports to the Connission, "ACRS Comments on An Implementation Plan for the Safety Goal Policy," dated May 13, 1987, and "Further ACRS Comments on Implementation of the Safety Goal Policy," dated February 16, 1989, we tried to face this problem by suggesting a hierarchical structure for safety goal implementation, in which each level of implementation becomes more precise and prescriptive than the one above it, and therefore easier to apply to real-life situations.

However, we cautioned, it is important that one not, in making the statement of each succeeding level more precise, introduce a new level of conservatism that makes it, in effect, a new safety goal.

The objective of our recommendation was to achieve coherence by mobilizing the so-called implementation in support of the policy, not as a substitute for it.

(We also urged that the policy statement be construed as a policy, and warned against using it too narrowly on individualcases,butthatisanothersubject.)

On top of all that, many of the issues of safety philosophy are not easily amenable to treatment under the Safety Goal Policy--fitness for duty, for example--and those will require guidance in another form.

All of the problems are complex and, as we have said, fall into different categories.

Certainly some fall under the management responsibilities of the EDO and we have not yet been able to schedule a meeting with him.

Since we hope to do so in the near future, and

The Honorable Kenneth M. Carr November 24, 1989 since we deem his input to be of importance in some of these areas, we feel it would be premature to make any explicit recomendations to you at this time. After we have met with the Acting EDO, and explored his views, we will be in a better position to provide sound advice to you.

What is clear to us from his memorandum to you, Integrated Approach on Regulatory Matters, dated October 18, 1989, is that we have not yet adequately comunicated our concern to him. We hope to do so soon.

Sincerely.

Forrest J. Remick Chainnan

References:

Wreport entitled,licy " dated May"ACRS Comments on An Implementation Plan 1.

for the Safety Goal Po 13, 1987 2.

ACRS report entitled, "ACRS Comments on the Integrated Safety Assessment Program," dated July 15, 1987 3.

ACRS report entitled, "ACRS Coments on the Need for Greater CoherenceAmongNewRegulatoryPolicies,"datedMarch 15, 1988 4.

ACRS report entitled, Proposed Rule on Fitness for Duty Program

-- ACRS Coments," dated April 12, 1988 5.

ACRS report entitled, " Proposed Generic Letter on Individual Plant Examinations and the Proposed Integrated Safety Assessment Program II," dated May 10, 1988 6.

ACRS report entitled, " Report on the Integration Plan for Closure ofSevereAccidentIssues(SECY-88-147),datedJuly 20, 1988 7.

ACRS report entitled, " Mark I Containment Performance Improvement Program," dated January 19, 1989 8.

ACRS report entitled, "Further ACRS Coments on Implementation of the Safety Goal Policy," dated February 16, 1989 9.

ACRS report entitled, " Proposed Final Rulemaking Related to Maintenance of Nuclear Power Plants," dated April 11, 1989

10. ACRS report. entitled, " Integrated Approach on Regulatory Mat-ters," dated April 17, 1989

11. ACRS report entitled, " Proposed Resolution of Generic Issue 128,

' Electrical Power Reliability,'" dated June 14, 1989

--q l

- ~.

_,.s Mig;

• • ?

'g%-

UNITED STATES 18 NUCLEAR REGULATORY COMMISSION

{

I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHWGTON, D. C. 20666 November 20, 1989 Mr. James M. Taylor Acting Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C.

20555

SUBJECT:

THE RELATIONSHIP OF THE QUANTITATIVE SAFETY GOAL TO THE CONCEPT OF ADEQUATE PROTECTION E

During the 355th meeting of the Advisory Comittee on Reactor Safe-L guards, November 16-18, 1989, we discussed the concept of " adequate l

protection" and its role in the plans for implementation of the NRC l

Safety Goal Policy.

We discussed this subject during several previous meetings of the Comittee and our Subcomittee on Safety Philosophy, Technology, and Criteria.

During this review, we had the benefit of discussions with members of the NRC staff and of the documents ref-erenced.

In a series of reports to the Comission culminating in the report of February 16, 1989, the ACRS has commented on the staff's proposals for

-implementing the Comission's Safety Goal Policy.

We also discussed this subject in a meeting between the ACRS and the Commission on May 3, 1989.

Following this meeting and a meeting with the staff on July 26, l

1989, the Commission asked for a clarification of the seemingly differ-L ent positions held by the staff and by the ACRS concerning the role of the concept of adequate protection in the staff's plan for implementing the Safety-Goal Policy (Staff Requirements Memorandum dated August 21, 1989). We provided an interim response in our report to Chairman Carr on October 11, 1989.

As an instrument for providing the requested clarification to the Comission, the staff prepared a draft paper entitled, " Adequate Pro-tection As It Relates to Safety Goals:

ACRS and Staff Positions," that was forwarded to us for review as an attachment to a memorandum from E.

S. Beckjord, Office of Nuclear Regulatory Research, to R. F. Fraley, ACRS, dated November 2,1989.

l We take exception to the description of the ACRS positions, as described in the draft paper, as follows:

(1) On page 2 of the draft paper, the staff provides a quotation from the Comittee's February 16, 1989 report (under Definition of

" Adequate Protection"), as follows:

'Qy%D&

J.

/

1 z

t.

c,g.

Mr. James M. Taylor

,2 -

November 20, 1989 l

We believe that the safety goal should play an impor-tant, but indirect, role in defining adequate protec-tion.

Ideally, compliance with-the Commission's 1

regulations is a suitable surrogate for defining adequate protection of the public. However, we believe L

that the adequacy of the regulations should be judged from the viewpoint of whether nuclear power plants, as i-a class, licensed under those regulations, meet the l

safety goals.

It is our understanding, following discussions with the staff, that the staff proposes the safety goal to be a sort of aspirational objective which would be sought but not necessarily reached.

To provide a better understanding of the ACRS position, the staff should also include the paragraph that precedes the above quota-tion, namely:

The term " adequate protection" has importance in the legal areas of safety regulation.

Although it is needed and used with apparent precision in legal instruments, its technical definition is not precise.

In general, it is accepted as equivalent to the term "with no undue risk to public health and safety" often used in other contexts. Another term, "in full compli-ance with the regulations" is used as a surrogate, on occasion, for either of these.

(2) Following this quotation, the staff's draft paper describes the ACRS position as, in effect, equating the concepts of " safe enough" and " adequate protection." This is not correct. The ACRS believes that the safety goal sets a standard of what is " safe enough," for L

the population of plants or a class of plants. As we have consis-L tently stated'in our previous reports on this subject, the quanti-tative safety goal should be used only to judge the adequacy of the NRC's body of regulations and should not be used to judge the adequacy of the design and perforinance of a particular individual plant.

We do not attempt to equate the safety goals to " adequate protection" in the sense in which the courts have recently con-sidered it.

Sincerely orrest J. Remick Chairman

s

+

3.

Mr. James M. Taylor November 20, 1989 References 1.

Memorandum dated November 2,1989 from Eric S. Beckjord, Office of Nuclear Regulatory Research, NRC, to Raymond F.

Fraley, ACRS, transmitting Draft Commission Paper for the Comissioners,

Subject:

Adequate Protection As It Relates To Safety Goals: ACRS and Staff Positions (Predecisional) 2.

Memorandum dsted August 21, 1989 from S. J. Chilk, Secretary, to J.

M. Taylor, Acting EDO, and R. F. Fraley,. ACRS,

Subject:

Staff Requirements - Briefing on Integration of Policy Statements For Severe Accidents, Advanced Reactors, Safety Goals, and Standardira-tion - July 26, 1989 i

lJ 1.

l I-

t

., - '1-n,

~

\\

UNITED STATES NUCLEAR REGULATORY COMMISSION 5

ADVISORY o0MMITTEE ON REACTOR sAFEOUARDs WAeHanoTON,D.C 20085

....+

November 24, 1989 Mr. James M. Taylor Acting Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

MODULE 1 0F THE DRAFT SAFETY EVALUATION REPORT FOR THE ADVANCED BOILING WATER REACTOR DESIGN During the 355th meeting of the Advisory Consnittee on Reactor Safeguards, November 16-18, 1989, we met with representatives of the Office of Nuclear Reactor Regulation (NRR) and the General Electric Company (GE) to discuss Module 1 of the staff's Draft Safety ) Evaluation Report (DSER) for the Advanced -Boiling Water Reactor (ABWR design.

This matter was also considered by our ABWR subcommittee during several meetings, the latest on October 31,.1989. We also had the benefit of the documents referenced.

The staff's DSER relates to the GE application for final design approval (FDA) and design certification of the ABWR design. The DSER is scheduled for completion in four modules.

Module 1-is the subject of this letter and addresses Chapters 4, 5, 6, and 17 of the ABWR Standard Safety Analy-sis Report (SSAR) and corresmnding chapters of the Standard Review Plan (SRP), NUREG-0800.

Our review of these chapters of the SSAR has been completed through Amendment 7.

A number of the S$AR and DSER sections included in the Module 1 chapters are presently. missing and will be issued as SSAR revisions and supplements to the DSER.

Even within the included sections, there are a number of open, unresolved, and confirmatory issues and incomplete interface re-quirements or other information that will delay completion of our review until.the revisions and supplements are issued. Comments on such missing or incomplete information will be included with our review of-future modules.

.Our consents should not be considered complete until we have prepared a report to the Commission concerning the final integrated DSER, which is presently scheduled for late 1990.

For now, we are providing the fol-lowing consnents and reconenendations concerning Module 1.

GENERAL 1.

The staff's ABWR licensing review bases letter to GE (Reference 2) states, "The degree of design detail necessary for providing an essentially complete design is to be that detail that is suitable for obtaining specific equipment or construction bids and to demonstrate

}_1.b(4AU T7 7 % fd

e 1,

j'

,p.?-

~

Mr. James M. Taylor November 24, 1989 l

b conformance to the design safety limits and criteria."

We believe that the level 'of design detail in Module 1 falls short of this p

requirement.

For example, we find that while GE has committed to follow applicable codes, standards, and regulatory guides, they have developed internal specifications for materials used in the fabrica-p tion of pressure boundary components that have not been submitted for NRC review.

We also find that a number of design details (such as L

those relating to design temperature and pressure and pipe size) are l

indicated on drawings in the SSAR as "to be established by others" or

(

similar statements.

Unless such information is included in the SSAR or other documents that are reviewed by the staff, it is clear that the level of design detail is inadequate.

We recommend that the staff revisit the issue of what constitutes an " essentially complete" design.

The staff should aise consider the question of form and depth of reporting differences between the ABWR being designed for construction in Japan and the ABWR design being proposed for certifi-cation.

2.

The SSAR chapters contain a number of sections for which there are no corresponding sections in the DSER or SRP, or the subjects of the DSER or SRP sections are different.

Also, there are cases wherein the SRP contains sections that do not appear in-the SSAR or DSER. We reconnend that the DSER sections be referenced by number and title to the corresponding SSAR sections they evaluate.

Differences, includ-ing the absence of any corresponding SRP sections, should be iden-tified in-the DSER.

CHAPTER 4 - REACTOR 3.

The fine motion control rod drive system (FMCRDS) materials list discussed in SSAR Section 4.5.1.1 shows Ste111te guide rollers and roller pins. Section 5.2.3.2.2.2 states that cobalt base alloys used for pins and rollers in the FMCRDS have been replaced with noncobalt alloys. The list of materials should be corrected.

4.

We were told by GE that the design of the integral rod ejection support system for the FMCRDS has been changed from that described in SSAR Section 4.6.1.

The staff should determine that their evaluation in the DSER is based on the revised design and the SSAR should be corrected.

CHAPTER 5 - REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.

The SSAR states that the automatic depressurization system (ADS) utilizes safety relief valves (SRVs) each of which is equipped with an air accumulator and check valve arrangement designed to ensure two actuations following failure of the air supply. Although not stated in the SSAR, GE indicated that the accumulators are tacked up by the nitrogen supply system.

This backup arrangement needs to be des-cribed in the SSAR teether with how check valve operability will be ensured.

v

.?

+s x go.

Mr. James M. Taylor November 24, 1989 6.

The specifications given in the SSAR for the materials of the primary pressure boundary do not meet current " good practice," or the prac-tice GE says they would require in the construction of an ABWR--they should.

To clarify this issue, the SSAR should contain answers to the following questions:

(1) will the steel in the core beltline be forged rings or welded plate?; (2) will upper limits on sulfur content of the rolled plate in the pressure vessel be those given in the ASME Code SA-533, Specifications for Pressure Vessel Materials (0.04%) or lower values consistent with good modern practice (under 0.015% with shape control)?--an adequate level is sweified for forged segments (ASME Code SA-508 Class 3,

Specificatior, for Quenched and Tempered Vacuum-Treated Forgirigs) and is available as an option in SA-533 but not called out by GE; and (3) what will be the upper limit on delta ferrite for cast stainless steel components?

The Code's allowed value of-25% should be halved to substantially remove concern about long-ters aging.

7.

SSAR Section 5.3.3 states that design for vessel annealing is not required bgcause the predicted value of adjusted RT does not exceed 200 F. The DSER states that the integrity ofNDIhe reactor if necessary.

vessel is ensured because the vessel may be annealed, designed to be GE stated during our meeting that the vessel is not annealed. The DSER statement should be resolved with GE.

8.

We believe that potential safety hazards (e.g., excessive internal pressure) associated with an uncleared electrical fault inside a reactor internal pump (RIP) should be analyzed and documented in the SSAR.

9.

We were told by GE that motor restraint rods are provided to prevent ejection of an RIP. We believe that this important feature should be described in the SSAR and evaluated by the staff.

10. SSAR Section 5.4.6 states that the design basis for the. Reactor Core Isolation Cooling (RCIC) system is only 30-minutes of operation during a loss-of-ac power event.

We believe that a more complete discussion of the station blackout capability should be included in the SSAR.

The DSER should include an evaluation of the 30-minute capability as an acceptable design basis.

11. The DSER contains no specific references to SSAR Sections 5.4.4-5, 5.4.9, and 5.4.12-14.. These sections discuss feedwater piping, main steam line flow restrictors, isolation systems and piping, component supports, and valves.

There are no comparably numbered sections in the SRP.

It is not clear where the staff intends to report its evaluation of these important topics.

CHAPTER 6 - ENGINEERED SAFETY FEATURES

12. The design basis for the ECCS and the conclusions given gbout its l

performance do not include the ejection of an RIP (450 cm break).

m L.,.

j. g.

Mr. James M. Taylor November 24, 1989 b-The rationale for excluding such an event as a design basis break shon1d be discussed in the SSAR.

' 13. DSER Section 6.2.6 indicates that inflatable seals will be used for primary containment equipment and personnel air lock penetrations.

We believe that an appropriate description of the seals and the air supply arrangement.and reliability should appear in the SSAR.

The discussion should include the capability of the seals to function under elevated pressure and temperature conditions for prolonged periods of time following a design basis accident.

L

14. There is a new section 6.5.5 (Pressure Suppression Pools as Fission i

Product Clean-Up Systems) in the SRP which does not appear in the l

SSAR or DSER. Why is this SRP section not being used for the ABWR7 CHAPTER 17 - QUALITY ASSURANCE L

15. Cha >ter 17 of the SSAR is intended to describe how GE and its major tecnical associates.(not mentioned by name in the SSAR but we assume to be Toshiba Corporation and Mitachi Limited) engage in the joint development and engineering of the ABWR design.

The quality as-surance programs used by the technical associates are not described or referenced in the SSAR. We believe they should be.

In conclusion,' we believe that significant progress has been made by the staff in its review of the SSAR for the Advanced Boiling Water Reactor. A considerable amount of work remains to be completed before the FDA is issued as expected by the end of 1990.

We will continue to review this work as the documentation becomes available.

Sincerely.

i L

1 Forrest J. Remick Chairman

References:

1.

Letter dated August 17, 1989 from Charles L. Miller, Office of

^

Nuclear Reactor Regulation, NRC, to Mr. Patrick W. Marriott, General Electric Company, enclosing Draft Safety Evaluation Report Related to the Final Design Approval and Design Certification of the Advanced Boiling Water Reactor, dated August 1989 l

2.

Letter dated August 7,1987 from Thomas E. Murley, Office of Nuclear Reactor Regulation, NRC, to Ricardo Artigas, General Electric Com-pany, enclosing GE Advanced Boiling Water Reactor, Licensing Review Bases, dated August 1987 3.

GE Nuclear Energy, Standard Safety Analysis Report, Advanced Boiling Water Reactor, Chapters 4, 5, 6, and 17

-