ML20205C633

Summary of ACRS Subcommittee on Safety Philosophy,Technology & Criteria 861210 Meeting in Washington,Dc Re Implications of Chernobyl Accident & Status of EDO Work on Development of Commission Safety Goal Policy.Supporting Documentation Encl

ML20205C633
Person / Time
Issue date:	12/18/1986
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2478, NUDOCS 8703300224
Download: ML20205C633 (62)

Text

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ggg D ff.99 hk H , i3 (!j)/ SNM7 E J DAlt ISSUED: 12/18/86 p nio Meeting Minutes on the December 10, 1986 Meeting of the ACRS Subcomittee on Safety Philosophy, Technology, and Criteria The ACRS Subcomittee on Safety Philosophy, Technology, and Criteria met on December 10, 1986 at 1717 H Street, N.W., Washington, DC. The )

purpose of this meeting was to: (1) continue the Subcommittee's dis-cussion of the implications of the Chernobyl accident, (2) discuss the status of the ED0's work on the development of the Commission Safety Goal Policy, and (3) review the NRC Staff's proposed policy statement on the reactivation of the licensing of deferred plants. ACRS action on (

n e c tt e d s s onbgn t90 a nd w re concluded at 3:45 p.m. All of the discussions were held in open ses-sion.

The attendees at this meeting were as follows:

ACRS NRC Staff F. Remick, Acting Subc. Chairman B. Sheron, NRR H. Lewis, Member S. Bryant, NRR D. Ward, Member F. Congel, NRR M. Carbon, Member Z. Rostoczy, NRR C. Michelson, Member S. Schwartz, IE J. Ebersole, Member P. Williams, NRR R. Savio, Staff M. Taylor, ED0 T. Michaels, NRR B. Boger, NRR Highlights

1. The NRC Staff briefed the Subcommittee on the status of their Chernobyl implications and fact-finding reports. The Staff has prepared and transmitted a second draft (December S,1986) of their Chernobyl implications report to the ACRS. The NRC Staff had

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SPTC Meeting Minutes December 10, 1986 planned to present their finished implications report to the Commission on December 16. This action has now been postponed to late January 1987. The Staff expects to have issued their fact-finding report by this time. Current schedule calls for finalizing the fact-finding report at a scheduled December 19 interagency meeting and completing the final text by the end of December,

2. The NRC Staff discussed the December 5, 1986 draft of their Chernobyl implications report and the chanaes which had been made since the discussions held at the November 5, 1986 Subcommittee meeting. (The minutes of the November 5 Subcommittee are included asAttachmentA). The major changes which have been made are:

(a) The section on " Economic Effects" has been deleted from this report. NRR plans to address this issue within the context of the Safety Goal Policy implementation plan work.

(b) A section has been added on " Graphite Moderated Reactors."

The NRC Staff is currently reviewing a petition for rulemaking (submitted by the Committee to Bridge the Gap) which will address the probability and consequences from graphite fires in Fort St. Vrain and NRC licensed non-power reactors.

Non-power reactors are not addressed in the Staff Chernobyl implications report but will be considered in the rulemaking petition. The Staff implications report makes recomendations on Fort St. Vrain and the Modular HTGR concept. The Staff believes that the Chernobyl accident does not raise any new concerns but will be giving consideration to a Fort St. Vrain PRA and some additional thermal stress experiments with structural graphite.

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SPTC Meeting Minutes December 10, 1986 (c) The NRC Staff is recommending that the benefits of a high-level'on-site nuclear safety manager, with no other duties or responsibilities, be considered.

(d) The NRC Staff is recommending that a program of preparedness for accident management, including training and procedures development for coping with severe core damage and for effec-tive management of the containment function, be examined. The Staff is also recommending that increased emphasis be given to the development and implementation of symptom-based E0Ps.

(e) The NRC Staff stated in their November draft that for multiple Unit sites there was some concern as to the ability of the operators in the undamaged unit to perform shutdown operations outside of the control room. The Staff now believes that this is not a concern.

(f) The NRC Staff does not believe that night shift operation is a concern and has deleted the section on this issue from their report.

3. The Subcommittee made the following recocinendations:

(a) It appears from what is known from Chernobyl that plant personnel knowingly violated safety guidelines in their efforts to accomplish an assigned task. Electrical utilities exist for the purpose of producing electricity and will develop attitudes and incentives which are oriented'toward managing the economical production of this commodity. It is important to establish safeguards which will assure that these

" production oriented" attitudes do not override safety consid-erations. This issue should be addressed in the NRC Staff's report and in the Staff's future regulatory actions.

F t-SPTC Meeting Minutes December 10, 1986 (b) The discussion in the " Operations" section of the NRC Staff's Chernobyl Implications Assessment deals most part with li-censed personnel. The Subcommittee stated that the subject was broader than this and recommended that more consideration be given in the NRC Staff's report to the role of utility nonlicensed personnel and utility management.

(c) The Chernobyl accident occurred during a test which was being performed as part of a an assessment of reactor behavior. The Subcommittee recommended that the entire subject of reactor tests, including surveillance testing be given more attention.

The philosophy now used in U.S. reactor may not be one that always reduces risk. An example would be the current practices used in establishing surveillance testing frequency.

4. Mr. Taylor of the NRC Staff presented a status report on the NRC Staff work on the implementation plan for the Commission Safety Goal Policy. The policy statement was issued in August 1986 and contained the two qualitative health goals, a performance guideline for the frequency of a large release, and a stated intent to take actions to prevent, with reasonable assurance, severe core damage accidents. The NRC Staff has developed a framework and schedule for a Safety Goal Policy implementation (See Attachment B) and will brief the Commission on January 8, 1987. Mr. Taylor proposed that the ACRS and the NRC Staff meet as necessary over the next several months to discuss the Staff proposal. A list of the Staff proposed milestones is given in Figure 1. The NRC Staff proposal for the implementation plan contains a modified version of the "EDO Decision Matrix." This matrix is shown on Figure 2.

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SPTC Meeting Minutes December 10, 1986

5. The NRC Staff has developed a proposed Commission policy statement on the procedures to be followed with deferred plants. The Staff's proposed policy statement addresses: (1) maintenance, preservation and documentation requirements, (2) the applicability of new regulatory requirements for reactivated plants, and (3) the proce-dures to be followed when reactivating deferred plants. The NRC Staff has recommended that the backfit rule be used for implement-ing (or not implementing) new requirements on reactivated plant construction projects irrespective of the duration of the deferral period. A number of plants are currently in the deferred, ter-minated, or cancelled status and it is not clear as to how many of these plants will eventually be completed. The number is expected to be small. (By way of an explanation, deferred and terminated plants have valid CP. Cancelled plants do not. If a plant is in a

" terminated" status, the plants owners have announced that con-struction has been terminated permanently.) A list of the de-ferred, terminated, and cancelled plants and a brief description of their status is given in Tables 1 and 2.

NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, N.W., Washington, DC, or can be purchased from ACE-Federal Reporters, 444 North Capitol Street, Washington, DC 20001, (202) 347-3700.

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FIGURE 1

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Safety Goal Implementation Tasks and Milestones ,

Office Tasks and Milestones Milestones Support Tasks RES Completion of NUREG-1150 Ref. Doc. (for Coment) Jan. 1987 RES Completion of Integrating Tools--SARA, IRRAS July 1987 RES/NRR Containment Performance /Large Release Study Apr. 1987

RES/ USERS Start Familiarization on SARA IRRAS Jan. 1987 RES/ USERS Start User Training on SARA, IRRAS Aug. 1987 Generic Issues / Severe Accidents / Source Tem -

D/ROGR Procedures for Review of New Generic Reg'mts Dec. 1986 NRR Procedures-Integrated Eval. of Generic Issues June 1987 USERS Implement SARA /IRRAS for Generic Issue Eval. Oct. 1987 -

NRR/RES Guidelines for Use in Future /Std.. Plant Rule- Sept. 1987 making NRR Generic Findings; Severe Accidents / Source Tenn Apr. 1987 Plant-Specific Issues NRR Evaluation of Plant-Specific Backfit Req'mts June 1987 NRR/0GC License Exemptions June 1987 .,

NRR Integrated Safety Assess. Program June 1987 5 NRR Plant-Specific Findings (Severe Acc.-Outliers) June 1987 Setting Regulatory Priorities (Resources /PRA Insights)

NRR Staff Resources--Generic Issues / Multi-Plant June 1987 RES Staff Resources--NRC Research Program June 1987 RES/NRR Risk Significance of Regs./ Reg. Guides Sept. 1987 NRR Standard Review Plan Resources Sept. 1987 IE Resources--Reactor Inspection (Ongoing PRAefforts)

Plant Operational Experience Assessments IE Performance Indicator Program (Decision req'd) Dec. 1987 IE/RES Risk-based / Safety Goal-Indicator Methods (0ngoing)

AEOD .

Licensee Events / Precursors July 1987 Final Environmental Statements (FES)

NRR Safety Goal Comperisons in FES Sept. 1987 i

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,l (GUIDELINE FOR COREMLT. LARGi MDIDACHTL RELEASE. HEALTH ttttCTS V .

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i{ Melt Frequency (Per RT)_ (Per RY)**

I Meet both objectives No furth e safety i-j4 <10-3 10-6 or improvement i! Don't meet one taprove($1.000/p-r)

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> 104 , health effects Don't meet one +1005AOSC) analyses likely required ,.

2 Meet both objectives improve ($1.000/p-r 10'3 -10~4 Presumed not to meet health .+ 10 7 11 A05C) effects uni.i1 detailed i analyses reveal otherwise improve ($1,000/p-r

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>10-3 effects until detailed analyses reveal otherwlse improve (cost no limit)

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Summary Data for Nuclear Power 8eactor facilities !ssued Construction Permits Whose Construction has been DeferrsJ  ;

Preservation Initial CP OL Program CP Extension Constr'n SE8 date subeltted-Empirat.

Reactor '% CP Issue Deferral Date X OL Docket (draft Approved Reactor Unit Licensee Vendor A/E Date Date Date. Sequest Complete Date or final) (dates)

<Orand Culf 2 Miss. P&L CE Bechtel 09/04/74 10/01/H 08/15/84 64/30/91 35 06/30/78 09/81f None Perry 2 Clev. Elect. CE *. Gilbert 05/03n 7 06/30/84 07/17/84 11/30/918 44 01/30/81 05/82f None WP-1 Washington 8&W \ U.E. 12/24n5 01/01/82 04/30/82 06/01/918 62 07/16/82 None 8 WP-3 Pubile Power C-E E8ASCO 04/11/78 01/01/85 07/08/83 07/01/99 75 08/27/82 . 11/19/85d 05/24/&5-is/Ja/65 U.E. - United Engineerina ,

8 CP has been extended.

2 Permittee plans to subelt.

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Summary Data for Nuclear Power Reactor f acilities issued Construction Permits

' Whose Construction has been Terminated pr Cancelled .

OL SER Constr'n OL SER date Site Stabili- NRC CP Current Licensee ration Plan Cancellation CP Empirat. Terminat. E Gocket (draft Reactor CP issue Complete Date or final) Submitted Letter A/E Date Date Date Reactor Unit Licensee vendor None None 08/82 11/02/82 S&L 05/02/74 05/06/82 08/81 1 08/17/82 03/16/83 No. Ill. P.S. GE None None Balliy 1 Bechtel 04/16/76 02/28/84 10/09/81 1 09/21/83 09/12/85 Callaway 2 Union El. We 18 None None Duke 12/30/77 05/31/84 04/29/83 09/21/83 09/12/85 Duke C-E 0 None None Cherokee 1 Duke 12/30/77 11/10/86 11/02/82 09/21/83 09/12/85 C-E 0 None None Cherokee 2 Duke 12/30/77 05/31/89 11/02/82 3* 09/11/85 Cherokee 3 C-E 10/14/83 1 09/08/80 02/17/82f GE S&L I 02/24/76 10/01/81 None None None None Clinton 2 lit. Pwr.

07/10/73 02/01/85 11/06/80 5 None *None Forked River Jersey C.P.L. C-E B&R 06/01/86 12/21/83 4 12/22/81 11/83f Carolina

• We E8ASCO 01/27/78 None None None None liarris 2 E8ASCO 01/27/78 06/01/90 12/18/81 1 None None liarris 3 Power & We None None 18ASCO 01/27/78 06/01/88 12/18/81 1 11/85 None Harris 4 light We 44 None None CfD 05/09/77 02/28/83 08/29/84 11/85 None Hartsville Al TVA Gf 34 None None 08/29/84 GE CI 8 05/09/77 02/28/84 None 03/22/83 Mone Hartsville A2 IVA 08/31/83 08/22/82 17 None GE Cf8 05/09/77 None 03/22/03 None Hartsville 81 IVA Cf8 -

05/09/77 08/31/84 08/22/82 1 None 3 None Itartsville 82 LVA CE 8echtel 11/04/74 12/31/89 12/29/81 18 None None 02/02/81 slope Creek 2 P.S.E&C, Del. GE 01/19/80 0 None None None Jamesport 1 Long Island We S&W 01/04/79 07/31/90 0 None None None 02/02/81 01/04/79 07/31/92 01/19/80 04/04/858 Hone Jamesport 2 Lighting We S&W 03/01/88 12/31/83 55 02/25/83 None pon, Public We S&L 04/04/78 04/04/85:

Marble flill 1 04/04/78 09/01/89 12/31/83 35 02/25/83 None None Narble Hill 2 Svcs., Ind. We S&L 85 11/18/77 05/82f 10/02/86 8echtel 12/15/72 12/01/84 07/01/86 10/02/86 None Nidland 1 Consumers B&W 12/15/12 07/01/84 07/01/86 85 11/18/77 05/82f Nidland 2 Power B&W 8echtel 8 None None Yes None 07/26/74 12/31/83 06/22/83 None North Anna 3 Va. Electric B&W S&W 12/31/84 11/25/80 0 None None Mone North Ar.na 4 & Power B&W S&W 07/26/74 None None 02/16/83 None Cf8 01/16/78 04/01/84 08/25/82 29 None Phipps Bend 1 TVA GE 08/25/82 5 None None 02/16/83 CE CIS 01/16/78 04/01/85 Phipps Bend 2 TVA l

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a ATTACHMENT A

a pu,seg'o,, UNITED STATES

5 n NUCLEAR REGULATORY COMMISSION

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' ~ l wasHWOTON, D. C. 20555 f

\a...**/ DEC 51986 .

MEPORANDUM FOR: Victor Stello, Jr.

Executive Director for Operations FROM: James H. Sniezek l Deputy Executive Director Regional Operations and Generic Requirements

SUBJECT:

COMMISSIONPRE-BRIEFING (DECEMBER 8,1986)ONSAFETYG0AL IMPLEMENTATION Enclosures 1 and 2 hereto are being provided for your consideration in advance of the planned pre-briefing by the offices on safety goal implementation.

These enclosures provide the heart of what I would plan to send to the Comission on this matter. These would be transmitted by a Comission paper that briefly captures their contents, Enclosure 1 presents the integrated framework established for implementing the safety goals. This would be sent to the Comission for infomation purposes. However, I would plan to seek Comission approval on Enclosure 2 (which addresses CRGR implementation procedures) in order to begin a timely incorporation of the safety goals into

. the CRGR deliberation process.

Regarding Enclosure 1. I have attempted to accommodate those office coments received in response to your October 15, 1986, memorandum to the offices on this matter. Most of the coments received were on target, helpful and positive. However, I was not able to accomodate some of the NRR views that the further implementation planning and further development of specific performance guidelines for use of the safety goals should be deferred for a number of months and that this effort be swept largely under the umbrella of the program to implement the Severe Accident Policy. I believe we should not wait further to do what can be done today on this matter. With this in mind, I have instructed my staff to work with RES, NRR, IE, and AEOD to develop an overall approach which we can at least generally agree upon.

Avv }

Jjmes H. Sniezek Deputy Executive Director Regional Operations and Generic Recuirements

Enclosures:

As stated cc: H. Denton

. E. Beckford J. Taylor C. Heltemes J. Murray

Enclosure 1 Framework for Safety Goal Implementation s -

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December 1986 I 1

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Framework for Safety Goal Implementation I. Objective .

i The objective of this document is to contribute an initial step toward the establishment of a framework for the systematic and structured implemen-tation of the Commission's Safety Goal Policy into the regulatory deci-sionmaking process and to effect the generation of detailed office and agencywide procedures for this purpose.

II. Background and Discussion Commencing in early 1981 and pursuant to recommendations of the Kemeny Commission on TMI-2, the Commission undertook the development of safety goals. This development involved a number of public meetings and work-shops. In February 1982 a proposed Policy St'atement on safety goals was issued for public comment and, in March 1983, a revised Statement was issued (also see NUREG-0880, Revision 1, May 1983) in response to comments received. A 2-year trial evaluation of the safety goal was then initiated a

and it concluded with an April 18, 1985, report to the EDO supporting the use of safety goals in the regulatory decisionmakir.g process as a factor In essence, that would augment the traditional safety review methods.

these trial evaluations revealed that the insights from use of probabil-isticriskassesiments(PRA),togetherwithsafetygoalsasameasuring ytrdstick, could serve to strengthen the traditional methods and add more l objectivity and predictability to a wide range of regulatory issues. This l

l 2-year trial evaluation also tested the use of safety goals against a )

range of individual regulatory issues and identified a number of NP.C program areas and applications that could potentially benefit by imple-mentation of the safety goals into the decisionmaking process.

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Following the 2-year evaluation report, a number of ACRS and Comission meetings were held with the aim of finalizing a Comission Policy State-ment on the matter of safety goals. This effort culminated in the August 4,1986, Policy Statement being ' issued by the Comission. The Policy authorizes the use of safety goals in the regulatory process. In route to this final Policy Statement, during late 1985 and early 1986, a number of important contributions and recomendations were provided to the Comission by the ACRS and by the NRC senior management. More focused guidelines for staff implementation and usage of the safety goals were provided for Comission consideration in mid-February 1986. An important feature within these guidelines was the development of a decision frame-work supported by the EDO offices which would guide the integrated imple-mentation of the various quantitative elements involved within the Comission's Policy Statement. This decision framework set forth an inte-

- grated approach for such factors as the core melt accident frequency, the risk of health effects and the safety-cost tradeoffs as a function of the plant performance and risk states. The Comission did not act on this proposed integrated approach. Rather, the Comission elected to review and

' approve implementation guidance later and apart from the final Policy Statemen't. However, the Comission and the ACRS both went a step further and offered added quantitative guidance for the staff to examine as part

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of the basis for implementation guidance. This general performance guideline proposed by the Commission is as follows:

" Consistent with the traditional defense-in-depth approach and the j accident mitigation philosophy requiring reliable perfomance of containment systems, the overall mean frequency of a large release of l l

radioactive materials to the environment from a reactor accident should be ]

1ess than 1 in 1,000,000 per year of reactor operation."

1 The implementation planning anticipates the need for further staff exami-  !

nation of this general performance guideline--primarily this will be led by the Office of Research, in coordination with NRR. This further examination is expected to consider the complete spectrum and ensemble of severe core damage accidents and releases as identified by current risk assessments that, with an appropriate consideration of the uncertainties involved, would either threaten, or cause to be exceeded, those mortality risk objectives in the Policy Statement. It is also anticipated that the key criteria governing " reliable perfomance of containment systems" over a spectrum of severe core damage accidents will be better delineated as l part of this further examination.

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To enable the staff to proceed with its detailed implementation planning

( for the use of the safety goals, a systeratic decision framework reflect-ing the provisions of the Safety Goal Policy Statement in its final form l

has been developed for interim use. The details of this framework and the accompanying general guidelines will be subject to review and approval by the Comission in accordance with its Policy Statement.Section III and V below present and discuss this integrated framework,

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l III. General Implementation Concept A. General Considerations Consideration of those various regulatory issues and program areas where use of the safety goals could strengthen the regulatory process and decisionmaking has resulted in seven broad areas being identified. These are: -

Severe Accidents Regulatory use of New Source Terms .

Resolution of Generic Issues

• Decisions about plant-specific requirements

• Setting Pegulatory Priorities

• Assessing Operational Events

• Implementation in staff risk assessments made pursuant to FEPA ,

(Environmental Statements)

- Within these seven broad areas are many subareas that the identified offices are expected to flesh-out in their development of detailed guid-ance and procedures for the safety goal implementation. The guidance and milestones set forth below address what is thought conceptually to be

' required for the-safe.ty goal implementation in the office having lead responsibility for the generation of detailed plans and implementing pro-cedures and the milestones. It is recognized that not all of the general guidelines herein may be specific enough or entirely applicable to the

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generation of the detailed office implementing procedures. "The respon-sible office should, therefore, add to or delete from these general .

guidelines as is determined to be appropriate. It should be noted that implementation of the safety goals and the quantitative elements involved is a complex matter. Results of safety goal comparisons must be balanced by other factors having decision <alue such as the results from the con-ventional technical and safety analyses. Sometimes these rescits must be supplemented by other arguments, either qualitative or quantitative in nature. Until further experience is gained with applying those quanti-tative elements of the safety goals in the decisionmaking process, it is recognized that the staff may, in the interim, have to rely on qualitative arguments to show the relationships of various issues to the safety goals.

The implementation of the safety geals will not by itself create a re-quirement for applicants or licensees to perfom PRAs when these have not been or are not otherwise required. Where an applicant or licensee may offer a peer-reviewed, plant-specific PRA for use in estimating the net safety benefits of proposed staff safety requirements and relates these to the Comission's safety goals, such a use is to be encouraged.

The staff has embarked on many new improvements in risk analyses and in the analysis tools, (e.g., NUREG-1150 and new source tems). It has em-barked upon a severe accident mission, for both existing plants and the future standard plant designs pursuant to the Comission's Severe Accident Policy Ctatement. It has embarked upon a more disciplined approach to prescribe requirements which increase overall protection to the public health and safety pursuant to the Backfit Rule (10 CFR 50.109) and in

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accord with similar procedures required by CRGR Charter and'with the i

plant-specificbackfitprocedures(MC-0514). It has embarked upon a more systematic review of operating experiences and of monitoring of plant per-formance to detect degraded operational performance that alert against ad-verse changes to plant risk states. It is developing a more disciplined approach for projecting risk significance from licensee event reports so as to gauge important accident precursors and those regulatory lessons that may be derived from such. Comparative evaluations of the risk sig-nificance of the new source term work has commenced in connection with severe accid'ent analyses presented in the Final Environmental Statements (FES)--see the South Texas analyses, NUREG-1171. This latter FES initia-tive could be readily extended over the short tern to include proiected safety goal comparisons that would be visible and open for further public comment. Staff views are being sought on the desirability of also using the FES to illustrate safety goal comparisons. All of these important agency initiatives and any related safety decisionmaking logically fall under the umbrella of the Commission's Safety Goal Policy Statement. All of these important initiatives should be reviewed by the responsible of-fices to determine the earliest practicable strategy for addressing the l

relationships to and achievement of the safety goal policy.

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B. Discussion on' Implementation Framework and New Assessment Tools

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By themselves, the safety goals represent a useful yardstick to use in gauging the acceptability of the overall state of risk being sought for a l

plant (or group of plants). On the other hand, the nature of the 1

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l regulatory safety decision process is such that it more frequently than not involves many individual issues that are not usually evaluated in a

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collective way. Each of these individual issues may, if implemented, have some piecemeal contribution to the overall state of risk. The real difficulty, however, lies in relating the absolute safety significance of a particular issue to the overall " bottom line" state of risk so as to determine whether resolution of the particular issue would be expected to make a meaningful contribution to the overall level of protection of the public health and safety. In implementing the safet,v goal policy, this difficulty With relating individual issues to the " bottom line" state of risk must now be squarely faced. Experience to date has shown that im-provements in this area of the regulatory safety decision process are needed. The tools to be used to relate the value of various safety issues (or a collection of issues) to an overall " bottom line" state of risk are in need of further sharpening and should be of benefit in the conduct of RegulatoryImpsetAnalyses(RIAs)andthedevelopmentofthesafetycon-clusion specified by 10 CFR 50.109. Use of quantitative analysis as the sole determining factor is not foreseen. Part of the process toward in-

' troducing and implementing safety goals as one of the authoritative fac-tors in the decisionmaking process will include the use of such tools and improvements.1 This work includes the System Analysis and Risk Assessment 1 The two. principal tools are System Analysis and Risk Assessment (SARA) pro,iect and the Integrated Reliability and Risk Assessment System IIRPAS).

SARA is a self-contained pc/xt software package of 6 reference plant de-l signs that allows the net effects of variable input by the user to be assessed against a baselire plant risk state (NUREG-1150) and it accom-modates an overlay of the safety goals for comparison purposes. SARA accommodates variable input on such items as initiating event frequencies, i

o (SARA) and Integrated Reliability and Risk Assessment System (IRRAS) pro-jects . Development of these tools are now well underway by the Office of Research with availability for trial use by the staff expected by mid-1987. An interim familiarization with use of these tools by the staff could, however, begin by early 1987. Thece analytical tools, together with the proposed integral approach and specific performance guidelines herein should also be of assistance in implementing safety goals as part of the Severe Accident Policy implementation. As now structured, the implementation of the Severe Accident Policy envisions issuance of a re-quest to utilities to examine their individual plants for severe accident vulnerabilities. Guidance to facilitate these plant-specific examinations will be issued soon (scheduled for issue in early 1987) and this guidance will also reflect the risk ob.iectives and defense-in-depth approach set forth in the Commission's Safety Goal Policy along with all further ap-proved guidelines as is being proposed herein. Regarding safety goal implementation for future plant applications, and in accord with the Severe Accident Policy, the staff is now preparing a NUREG report (for -

public comment with issue expected in early 1987) that will describe the system unavailability and sequence probabilities and can relate these changes to plant damage states, containment failure modes and offsite con-sequences. It displays the risk importances of various sequences and will allow the user to estimate the safety significance of proposed plant changes in terms of core melt frequency and overall public risks. Key systems-and event tree sequences are also displayed. IRRAS permits the user through main-frame linkage to perform detailed model analyses and i reconstruction at the system level (fault tree) and sequence level (event t-ee). IRRAS provides a means to determine the system reliability change oue to to design changes or due to degradations in design or operations that may be observed at plant component or system levels.

m e < - - -

l

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role of probabilistic risk assessments in plant licensing. "It is also  ;

)

expected that any guidance such as that herein concerning the evaluation of future plants for severe core damage / core melt accidents, for large radioactive releases and for the achievement of the safety goal risk f

objectives, will be appropriately reflected before this NUREG document for Gegulatory Guide, as appropriate) is to be published in a final form.

Today there exists a significant number of generic safety issues to be resolved. Additions to the list of generic safety issues have been outpacing the final safety decisions and the issue resolutions for some time now. This factor alone strongly suggests that the Comission's

! policy aims toward a stable and predictable regulatory process and toward a plant standardization policy can be effectively enhanced when the safety The goals are implemented into this generic issue resolution process.

safety goals-can provide a yardstick to better gauge what are the most meaningful generic safety issues and risk contributors.

The prioritization process for generic issues that the staff implemented several years ago (NUREG-0933) has gone a long way toward discerning those j

potentially meaningful safety issues on which further staff resources l could be devoted. This process of prioritization should be continued, but The it is in need of updating by the implementation of the safety goals.

current prioritization process also relies on an outdated and unchanged base of plant risk that is now over a decade old. Furthermore, this process does not currently cumulatively account for the safety benefits and costs attendant to issues already imposed and it is not now taking i

into account this changed status of plant risk. The cumulative (or inte-grated) effects (risk reduction benefits and the costs) to be expected from resolution of all remaining generic safety issues are also not now being taken into account. Further, the secuence of implementation (or schedular resolution) of generic issues is not usually well integrated and can result in various generic issues causing the same system or safety function to be " fixed" in an uncoordinated way that renders successive

" fixes" ineffective from' the star.dpoint of both safety and costs. This l

process potentially results in large costs to both the NRC and the in-dustry. It is also clear that the successful implementation of safety goals and the better integration of generic issues will effect a con-siderableimprovementtotheexistingprioritizationprocessandhothose analytical tools currently being relied upon in this process.

The risk rebaselining work that is ongoing under the direction of RES (i.'e., NUREG-1150) is expected to provide the NRC with the latest state-of-art display of plant risk states factoring in the THI-? safety im-provements as well as those cumulative improvements that may have been brought about to date through resolution of a variety of other generic safety issues. This risk rebaselining work by RES will incorporate the latest insights on PRA techniques and will include the risk significance of the new source term results. It will also update the severe accident consequence analises to include new health effects modeling as those effects are best understood today. This work will also provide the most comprehensive display of the phenomenological data, and modeling uncer-tainties involved in such risk pro,4ections that is available today to the

r ;

. _ 1 i

NRC. Further, this work will explicitly display the relationships of the Consnission's safety goals to the plant risk states found and will project l those overall uncertainties believed to exist about these plant risk states. This work represents a very important resource for safety goal implementation and for helping to more accurately estimate the value to be associated with further resolution of those generic safety issues now out-standing. It is intended that the NRC generic issue prioritization and resolution process and the requisite analyses for the 10 CFR 50.109 back-fit determinations be updated (when appropriate) by the use of the NUREG-1150 results in an integrated framework as illustrated by Figure 1.2 To this end, the integrated safety goal decision matrix (illustrated by Figure ?) is intended for use along with the NUREG-1150 results.

This decision matrix (Figure 2) is intended to help gauge the degree to which outstanding generic safety issues should be further pursued to im-2 Here it should be noted that the NUREG-1150 rebaselining will not expli-i citly address risks associated with external events (e.g., earthquakes,

' hurricanes,etc.) Rather, NUREG-1150 will at this time, focus on those core melt accidents and risks resulting from internally-caused events.

Other RES work is currently ongoing in regard to external events and the findings could, if judged necessary, be factored into the safety goal implementation framework. This is not viewed as any particular hindrance to implementation of the safety goal policy since insights on the risk from external events can be taken from other existing plant-specific PRAs as necessary for the decisionmaking process. Further, a review of the outstanding generic safety issues and multiplant actions indicate that a vast majority of'these would not require inputs on the external events to assist issue resolutions. It is also expected that any impacts from the severe accident at Chernobyl, if these are found significant to the NRC generic safety issue decisionmaking and resolution process, would be accomodated by the framework for safety goal implementation.

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_ 17

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prove the overdi plant risk states. As noted, the severa1' integrating software tools that will enable the staff to have a direct, hands-on i l

capability to use the NUREG-1150 risk baseline for reference analysis are .

now in the later stages of development. These tools will assist staff analyses of the probable effects of resolution of the remaining outstand-inggenericsafetyissues(individuallyorcumulatively).Theseintegrat-ing tools allow the staff to use micro-computers to propagate the net (or overall) risk effects of a proposed plant backfit or safety issue resolu-tion (individuallyandcollectively)regardlessofwhethertheseactions would involve plant changes at the component, system or severe accident sequence level. These tools will also provide the staff with a means to

' measure the effects of proposed backfits or generic issue resolutions against the Comission's safety goals--ergo, safety goals can indeed be implemented as a factor in the regulatory decision process for generic safety issues.

The several integrating tools (SARA and IRRAS) that are to be founded on i

! the NUREG-1150 results are expected to have an applicability that goes somewhat beyond the areas involving plant backfitting and generic safety issue resolutions. For example, the SARA and IRRAS tools can be poten-j tially useful to assist the assessment of a variety of operational ex-l periences and to estimate the risk significance of observed plant precursor events. These tools can potentially provide a useful means to help gauge the net risk importance and/or be of help in diagnosing trends of operations that may represent a degrading situation relative to a pre-viously predicted plant risk state (i.e., it may prove helpful to i

.  : l l

establishing the relationship of certain plant perfomance indicators to l the safety goals). It is anticipated that the implementation of these

~

tools into the process of assessing safety significance of the reported operational events and into the process of assessing plant perfomance indicators will foster implementation of the Comission's safety goals in these areas.

The framework provided by the NUREG-1150 results and the integrating tools (SARA and IRRAS) is intended to be a central means to be used in the rela-tively near term for implementing the Comission's safety goals to deal with existing generic issues. This framework, when taken together with the integrated safety goal decision matrix and the specific performance t

guidelines herein, should enable the staff to quantitatively assess the overall risk increase (or decrease) for a majority of its proposed safety actions and plant backfits. It will enhance regulatory decisionmaking by improving the perspective as to where and when these proposals may be supportable and justifiable in view of the safety benefits and cost trade-offs involved. This fraraework for safety goal implementation can also

' accomodate decisions about the risk vulnerabilities of existing plants, as these vulnerabilities (if they exist) become better defined through the implementation of the Comission's Severe Accident Policy. The safety goal decision matrix is also intended to apply to the risk states sought for the future and advanced reactor plant designs.

After the existing generic issues are resolved, the central means for implementing safety goals is expected to shift toward the evaluation of

_ _ _ ~__ . _ , _ - _ _ _ . . _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ . _ _ _ - . _ _ _ _ . . . _ _ _

plant safety performance (including the possible use of risk-based per-formance indicators) against certain safety performance targets estab-lished consistent with the safety goals. At this time, the staff recog-nizes a need to acouire further data and experience within the IE plant performance indicator program before it is a position to recomend fimly in favor of detailed guidance for safety goal implementation. This holds true also for any firm recomendations about establishing risk-based tar-gets that might enhance the regulatory decisionmaking (and any potential enforcement actions) deriving from this recently approved IE program.

It is clear, however, that risk-based indicators developed consistent with safety goals could play on important future role as part of this IE moni-toring and trending program. It is expected that by the end of CY 87 with further plant data and experience at hand, the staff will be in an im-proved position to evaluate the need for, and merits of, incorporating this additional detailed guidance and plant performance targets into this program. Recomendations to the EDO and discussions with the Comission on this matter can be expected at that time. In the interim, RES and IE are expected to continue with the examination of available methods for use to measure plant performance relative to risk-based indicators and safety goals as targets for detecting degrading plant performance and plant risk l states.

IV. Imple' mentation Tasks _and Milestones t

As indicated above, RES and NRR each have a significant number of tasks to complete under the proposed safety goal implementation framework. The l

- s -

various tasks, milestones and responsible offices are outlined in Table 1 below. As can be seen, RES has the lead responsibility for completing a j number of the support tasks that will enable safety goal implementation to take place across many of the regulatory decisionmaking areas. These involve the completion of the NUREG-1150 risk rebaselining work, comple-tion of the related software tools, SAPA and IRRAS (including validation of the input data supporting these tools), commencing staff familiariza-tion and training on SAPA/IRRAS, the further examination and development of the general performance guideline on large releases. The examination of available methodology for use as risk-based indicators and, lastly, the development of guidelines that reflect use of safety goals in prioritiza-tion of RES work and resources. Similarly, NRR has a number of tasks that will rely on the RES support work, but these NRR tasks by in-large will involve the development of detaile:i procedures for implementing the safety goals in accord with the guidelines as proposed herein. It is clear, however, that a close coordination of efforts between the assigned RES and NRR staff will be needed to achieve the desired task completions and mile-stones. Similarly, the areas tasked to IE and AEOD will also likely re-quire a close coordination and further interaction with RES to become more familiar with the available risk assessment and performance monitoring tools available to assist implementation of the safety. goals. Those tasks assigned IE and AEOD in Table 1 below largly involve the development of detailed implementing procedures. Likewise, principal tasks assigned D/ROGR also involves detailed implementing procedures expected to be of assistance in CPGR deliberations on matters such as proposed new generic f

e

, Table 1 9

Safety Goal Implementation Tasks and Milestones Office Tasks and Filestones Milestones Support Tasks RES CompletionofNUREG-1150Ref. Doc.(forComment) Jan. 1987 RES Completion of Integrating Tools--SARA, IRRAS July 1987 RES/NRR Containment Performance /Large Release Study Apr. 1987 RES/ USERS Start Familiarization on SARA IRRAS Jan. 1987 RES/ USERS Start User Training on SARA, IRRAS Aug. 1987 Generic Issues / Severe Accidents / Source Term D/ROGR Procedures for Review of New Generic Req'mts Dec. 1986 NRR Procedures-Integrated Eval. of Generic Issues June 1987 USERS Implement SARA /IRRAS for Generic Issue Eval. Oct. 1987 NRR/RES Guidelines for Use in Future /Std. Plant Rule- Sept. 1987 making NRR Generic Findings; Severe Accidents / Source Term Apr. 1987 Plant-Specific Issues NRR Evaluation of Plant-Specific Backfit Req'mts June 1987 NRR/0GC License Exemptions June 1987 NRR Integrated Safety Assess. Program June 1987 NRR Plant-Specific Findings (Severe Acc.-Outliers) June 1987 Settina Regulatory Priorities (Resources /PRA Insights)

NRR Staff Resour es--Generic Issues / Multi-Plant June 1987 RES Staff Resources--NRC Research Program June 1987 Risk Significance of Regs./ Reg. Guides Sept. 1987 RES/NRR

'NRR Standard Review Plan Resources Sept. 1987 Resources--Reactor Inspection (Ongoing IE PRAefforts) i j

Plant Operational Experience Assessments IE Performance Indicator Program (Decision req'd) Dec. 1987 Risk-based / Safety Goal-Indicator Methods (Ongoing)

IE/RES July 1987 AEOD. License,e Events / Precursors j Final Environmenta1' Statements (FES)

NRR Safety Goal Comparisons in FES Sept. 1987 1

requirements. These latter procedures by D/ROGR are essentially completed and are awaiting trial use.

It is expected that the majority of the detailed implementing procedures will be in place within roughly one year. Thereafter, use of safety goals to augment regulatory decisionmaking is expected to become a more routine part of the NRC decision process.

The work plan in Table I has been structured in accordance with the pres-ent NRC organization. When pertinent details of the anticipated reorgani-zation are settled, the plan will be reviewed and ad,iusted accordingly.

V. General Guidelines for Safety Goal Implementation Set forth below are some general guidelines intended to assist the staff in its development of detailed implementation guidance for safety goal use in the regulatory safety decision process. This guidance will be subject to further Commission review and approval. At this time, the staff may find a need to flesh-out in specificity, add to or delete certain of these I

guidelines below depending on their safety decision needs and Office areas of responsibility. These guidelines are intended to be applicable for use in connection with the decisionmaking areas involving Severe Accident l

Policy Implementakion', Resolution of Generic Issues Decisions on Plant-specific Requirements Setting Regulatory Priorities, and Assessing the Risk Significance of Operational Events. As noted in III above, the inte-grating tools (SARA and IRRAS) and the integrated safety goal decision matrix are also intended to be put to use for the safety goal

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_ 17 implementation. This total framework should enable the staff to improve upon its quantitative assessment of the overall effects of a decision regardless of whether this regulatory decision involves additional safety requirements or the relaxation of existing ones. This integrated frame-work specifically incorporates use of the safety goal as a yardstick to be used for decisionmaking, and it provides sharper tools that should enhance staff decisionmaking for the generic safety issues. The general guidance provided below is also considered to be compatible with the existing back-fit rule requirements (10 CFR 50.109):

1. A phased implementation of the safety goal policy into the regulatory decision process is anticipated. That is, the staff should begin as soon as practicable to use it in those areas ready for such usage (e.g., the evaluation of proposed backfits and new generic safety requirements). Initially, the safety goal usage and comparisons may be less quantitative than is ultimately expected, but the staff's implementation process should strive to accurately and meaningfully i

address issue resolutions in terms of the relationships to the vari-ous objectives displayed in the integrated safety goal decision matrix. This should include a characterization of the degree to which accident prevention, accident mitigation and the overall pro-tection of the public health and safety is being changed by the issue iesolution ahd should include those cost tradeoffs and balances involved.

2. For those areas of regulation where the staff uses the safety goal, it will be used in conjunction with traditional safety review methods for making regulatory decisions. Nuclear power plant licensees will still be expected to meet NRC's regulations.

3. In using the results of PRAs, the staff will ensure that each PRA receives a peer review and will address and allow for estimated un-certainties by using judgment in applying the results in regulatory decisions. To the extent these uncertainties and the mean, median and confidence range values can be addressed in the use of results of l PRAs, they should be displayed. The staff will use mean values for implementing the safety goal and it is expected that these will de-rive from NUREG-1150 for purposes of generic safety issue resolutions

) and generic backfit proposals.

l 4. The staff will remain mindful that PRA results found to be initially acceptable from the standpoint of the state-of-the-art probabilistic i

techniques and from peer review can either improve or deteriorate l

- depending on the quality of performance of plant personnel and oper-i ations(e.g., testing, maintenance,managementpractices). The staff will strive to develop and use techniques that could monitor plant safety performance or would give precursor evidence to deteriorating

' performance..(or precursor events) that could adversely affect a plant's predicted state of risk. The staff will also encourage the

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r adoption of sound risk management and accident prevention practices F

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that emphasize maintaining the defense.-in-depth sought by the N t" s Comission's Safety Goal Policy. s>j

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5. The use of PRA results and the safety goal should not dirgi ish the ,

centinued importance of the defense-in-depth safety philosophy or the , f;

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traditional safety review methods used by the staff in making regula-tory decisions, no'r should they diminish NRC diligence in assuring licemee bpnagement attentionNo safe constru'ction and operation of nuclear p$wre plants.

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6. For plants where a PRA exists or reasonable judgments can be drawn i

from other PRAs,. safety goal objectives (as embodied % the inte- .

1,. . Y grated decision matrix) may be used as one factor in considering. [

s plant-specific issues like licensee exemptions, backfits.)and inte-grated schedules / for the implementation of, backfits. The3 use of ,)

these safety goal objectives is snlikely to replace theIn0mLSeIS-(' .

ods for reviewing licensing exemptions or backfits, but t!hese ob- [ (

jectives should be taken as authoritative guidarce to the st'aff as to <1 what constitutes an acceptable level of radiological risk that might '

ao .

- be imposed on the public--primrily due to accidents at a p yuclear '

power plant. The staff will not, however, require a plant-specific PRA for the sole purpose of safety goal evaluations, t

7. For plants where a PRA exists and it is found to be o'f benefit to compare these plants for conformance to the safety goal, such com-parisons may have to be judged on the basis of the type a. d frequency 1 .,

p:

of severe accidents, on the capability of containment for these acci-

- dents, and in some cases on plant location. It is recognized that the adoption of certain plant performance guidelines (as embodied in theint;gratedsafetydecisionmatrix)couldserveasfullyaccept-(g g able surrogates for the safety goal health effects objectives if it

< i, '

can be' determined with a reasonable degree of confidence that these

surrogate guidelines are met. Such plant performance guidelines may also prove more useful as a regulatory tool (or figure of merit) to rely upon in judging achievement of the safety goal. If it cannot be shown that these surrogate guidelines are met, explicit analysis to determine the achievernent of the public health effects objectives in the safety goal policy should be carried out.

8. The Comission has stated that, as a matter of policy, it will con-Y tinue to pursue a regulatory program that has as its objective pro-1 I l

- viding reasonable assurance, while giving appropriate consideration n

to the uncertainties involved, that a severe core damage accident p.

(; will not occur at,r U.S. nuclear power plant. Accordingly, the f following specific plant performance guidelines are set forth for l staff use:

l

a. The PRA predicted mean frequency of a nuclear reactor accident

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that resNts in a large-scale melt of the reactor core should not be more than 1 in 10,000 per year of reacto'r operation.

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t i y I l' This guideline should be applicable to a_11 but a few, small, I  ;

~, e,tisting nuciear plants. In working toward this guideline on f li V .

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21 -

accident prevention, the staff will keep in mind the range of uncertainties involved and will rely on results from NUREG-1150 in judging'these absent an otherwise acceptab?e plant-specific PRA for such purposes.

b. Consistent with the Comission's Policy Statement on severe j accident issues for the future plants n'nd advanced designs and i in accord with the integrated safety goal decision matrix, the l staff should seek to achieve an even higher standard of severe accident safety perfomance by improvements in the plant design i

l and operations to ensure that a high degree of severe accident

' prevention exists consistent with an enhancement of the plant l, ,

operational reliability. For these future plant designs, the Comission's severe adcident policy makes clear that a PRA is to be provided that should enable a staf 7 detemination as to whether greater margins in safety are fact being. realized. To

! this end, the systematic and strategic adoption of a higher level of functional and/or system redundancy may be warranted relative to that currently required by Appendix A to 10 CFR 50.

t:

c. Consistent with the traditional defense-in-depth approach and the accident mitigation philosophy requiring reliable perfor-

- .mance of containment systems, the overall mean frequency of a large release of radioactive materials to the environment from a reactor accident should normally be less than 1 in 1,000,000 per ytar of reactor operaticn. This general performance guideline

for large releases should be applied to existing and future plant designs. The staff should also continue its current ef-forts to refine this general performance guideline for large releases so as to enhance the degree of understanding about the extent to which tradeoffs between accident prevention and  !

consequence mitigation may be permissible. The staff should also remain mindful in these current efforts, that the Safety Goal Policy addresses acceptable risks not just exposure levels, and that any surrogate guidelines to be used should remain roughly consonant with the Policy aims and objectives. In par-ticular, this staff effort should result in consideration of the total ensemble of severe cor. damage and large-scale core melt .

accidents and their release frequencies and magnitudes, to determine that the accident spectrum and the cumulative health l l

effects do not importantly threaten to exceed the Policy ob-  !

jectives on mortality risks (i.e., the 0.1% risk increment). I

9. The safety goal implementation is not intended for staff evaluations of security and sabotage issues, because there is at present no basis )

on which to quantitatively estimate the risk of the threat. The safety goal will apply to reactor accidents resulting from eith'er internally initisted or externally initiated causes. The staff will

,be particularly cognizant of the large uncertainties that appear to be involved in describing the plant core-melt risk from external events (e.g., earthquakes, fires, floods, etc.) and will ensure that these uncertainties do not mask the risk of other important core-melt

,__ ._ _ _ . _ _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ . _ _ . . - . _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _a

initiators or contributors to the frequency of large releases of radioactivity.

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10. In using the safety goal with matters such as generic issues involv-ing a particular accident sequence (where apportionment and frac-tional allocations of the safety goals may be at question), the staff will strive toward a balanced (i.e., no overwhelmly dominanti con-tribution to core melt or to public risk from such sequences. Addi-tionally, the staff will strive to minimize the situations whereby the mean core-melt frequency, the mean frecuency for large releases, and the mean health effects results (or their attendant uncertain-ties) are allowed to be driven or dominated by single accident se-quences. Some have suggested that apportioning a 1/10 contribution of individual sequences to the overall core melt frequency may be acceptable; and although this could be a suitable approach, the staff should remain mindful that the need for such balanced contributions may strongly depend on whether a plant is already at an acceptably low overall core-melt frequency and risk level. The staff should

< also remain mindful of the fact that sequences found to dominate the core melt frequency may not be the sequences that are the dominant contributors to the mortality risks. The guideline for large re-leases is intended to recognize this fact and to ensure that these

" sequences contributing most to the large releases and to mortality risks are of very low frequency.

..- -. - .. = ._. .. -_- . _ _ _ _ _ _ _ - . ..- _ - . .

11. In conducting benefit-cost analyses, the staff will include the cost of occupational exposures incurred as a result of implementing a

~

proposed safety requirement, as well as the benefit due to occupa-tional exposures averted as a result of the accident risk reduction.

Both will be evaluated on the basis of $1,000 per-person rem. The staff will strive to display all important safety benefits and costs attendant to its proposed regulatory safety decisions as has been its customary practice. Little decision weight should continue to be given to those economic benefits that may accrue to licensees

• rom

. proposed regulatory safety decisions. Greater decision weight is, however, warranted for proposed plant modifications and requirements which will reduce a very high core melt frequency or a high frequency for the large accident releases that may threaten to exceed the safety goal mortality risk objectives; whereas, a lesser weight is warranted when the frequencies are relatively low. This decision-weighting for safety improvement costs is displayed by the integrated safety goal decision matrix (Figure 2, fourth column) and it is based l

on a sliding scale; decreasing in the percentage of the decision l

< weight to be given to improvements as the core-melt frequency decreases.

12. The safety goal implementation will use the integrated safety goal

" decision matrix displayed by Figure 2. Where the staff considers this matrix to be not applicable to the safety decision or regulatory I

issue at hand, then the reasons for inapplicability should be clearly stated; for example, in matters involving enforcement and compliance

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- a u ,ai.es are tenen es seen seises *-

• • vme eversii ,etseitae for the fewgemeer of a terve end Itte throetente, reteese less see gr8 per af say serve as

== acceptable serve, ate (for health effects emelyses) that proefdes a high degree of esguyence that De CouplSstee's safety E<als are echteved. Otherwise. Supreveneet to acctdent attigetten er preventles any he destreble for added ,

defense-le-depth.

• Preset effects lategrated to I elle free site boundseyt latent effects to le elles.

• • (1) AOSC = Averted eastte costs (See NUstE/tR4568. A IIpeeesh_for Telse-Smmett Assessment)

(2) p-r = persen-eve. Sategrated to SS elles e

• o to existing regulations. The integrated decision matrix is believed to contain the essential elements needed to support implementation of the Commission's Safety Goal Policy Statement. It is also considered to be highly compatible with other Policy initiatives that bear on the public risks from severe reactor plant accidents, and with the traditional defense-in-depth approach regarding accident prevention and mitigation. This integrated decision matrix shall be placed in use by the staff to guide its decisions regarding the need for plant safety improvements and backfitting and to assess changes in the overall" risk states that may be brought about through relaxations of existing regulatory requirements. The matrix is intended to accom- ,

l modate a wide range of plant risk states that could potentially be identified to exist within today's population of existing reactors and thereby accommodates the implementation of the Severe Accident Policy for existing plants (i.e., should outliers be identified at a plant to indicate that safety improvements may be needed). Other features embodied in this matrix include the following:

a. A sliding scale on benefits and incentives is set forth to em-phasize that safety improvements will be sought by NRC where l

i needed for either existing or future plants. The top of the matrix is intended to set the expectations for future plants and

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advanced designs. In essence, the very top of the matrix re-flects a very low risk level for severe accidents below which no further safety improvements in plant design and operation would be necessary and that there exists a high degree of confidence l

(e.g., 95%) that Comission's safety goals are achieved. The  :

staff is expected to reflect such expectations for an improved .

level of safety in their decisionmaking and rulemaking processes for the future plants and advanced designs.

b. The matrix can impose significant costs (e.g., many millions of dollars) for safety improvements where the projected risk state involves a high core-melt frequency and the failure to achieve either the desired frequency for large releases or the mortality risk objectives. In these situations, it is considered to be necessary to restore the plant to a risk state where an adequate level of protection to the public health and safety is reason-ably assured. The very lowest part of the matrix reflects that the total dollar costs of safety improvements will not be a factor where the public health and safety cannot be reasonably assured by t?,e NRC.

c. An ALARA approach is embodied in this integrated decision ma-trix. It is intended to emphasize that, as a priority, the quantitative health effects objectives (0.1 percent) should be achieved through plant design and operation. However, the ma-trix yields a degree of flexibility to the staff to reach deci-sions on whether risk reduction can be best achieved through l

I improvements in the overall core-melt frequency, by reducing the

! probability of one or more of the dominant core-melt sequences, or by further improvements to mitigate consequences or reduce

27 the frequency of a large release of radioactive materials. This approach is considered to be consistent with the traditional NRC approach adopted for the radiological risks associated with routine plant operations.

d. The integrated decision matrix specifies a general perfomance 9uideline that the overall mean frequency for a large release of radioactive materials to the environment from a spectrum of severe reactor accidents should nomally be less than 10-6 per reactor-year of operation. This guidance may (as was suggested by the ACRS in mid-1986) usefully serve as an acceptable surro-gate in determining that the individual mortality risk objec-tives have been achieved. It is known; for example, that the very largest of PRA predicted release magnitudes (e.g., SSTI

~

magnitudes) occurring at a mean frequency level of about 10-5 /RY could potentially achieve the Connission's safety goal objec-tives at most plant sites. There may not, however, be very much margin present to account for uncertainties in the risk projec-tions. For the interim, and until further study of this matter is completed, this "large release" guideline can be viewed as being the more restrictive of that guidance presented in the Comission's Safety Goal Policy Statement. Here it should be noted that-it is customary in PRA's (dating since the 1975 ReactorSafetyStudy)todisplaythespectrumofpublicrisksby i

use of CCDFs.3 The area under this CCDF display is representa-tive of the mean risk outcome of interest. In the safety goal case, the individual mortality risks are the outcomes to use in gauging the achievement of the Comission's safety goals. For U.S. plant designs, PRA results available to date indicate that the controlling mortality risk of interest would be the risk of early fatality to an individual and this was stated in the Comission's Policy Statement.4 The CCDF for early fatality could thus be used to represent the mean frequency that would c6ver the aggregate risk of one or more early fatalities result-ing from the total ensemble of severe accidents, whether these might result from high frequency core damage events (having modest to very small releases) or the low frequency core damage /

core melt events (having intermediate to larger releases). In other words, all significant release magnitudes and release pathways at the plant affecting the mortality risk of interest would have been captured. The use of a CCDF cs is considered 3 Complementary Cumulative Distribution Function (CCDF) l 4 Fere it needs to be clearly understood that the focus on early fatality risks (as found by PRA to be controlling relative to the safety goals) does not imply that the risk of latent cancer fatality is being ignored.

Nor is the possibility of large, heated and~well-elevated releases such as those that resulted (unfortunately) from the USSR (Chernobyl) reactor design,'being ignored. The potential effects of elevated and heated re-leases from U.S. plant designs, have long(been studied (e.g., WASH-1400 (1975),NUREG/CR-2239(1982),NUREG-1062 1984),etc.). These effects were also considered in recomending the off-site distances applicable to the latent cancer fatality objective in the Comission's Safety Goal Policy Statement.

here, would also serve to constrain the mean value of all seri-

! ous health effects resulting from early failures of the con-

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tainment systems. This is so regardless whether these effects were to take place at a more or a less populated plant site.

This approach is consonant with the safety goal policy objec-tives on mortality risks and the treatment of uncertainties.

Further, the CCDF would also capture those likely risk-manage-ment strategies (e.g., controlled-filtered venting) that may be taken to control and further mitigate the potential severity of a large-scale core damage accident that may threaten uncontrol-led loss of containment integrity and, thus, challenge the safety goal mortality risk objectives. If the ordinate value on the CCDF for early mortality risks was to be limited to 110-6/RY at the aggregate value of 51 early fatality (integrated over all releases at all distances beyond the plant site boundary) then the mean value for all large releases affecting the mor-tality risk objectives would indeed be constrained below 10-6jpy as would all of the aggregated effects greater than 1 fatality.

, The staff, in its further examination of the Commission's gen-eral performance guideline, should consider the use of this CCDF i limit-line approach to define the aggregate early fatality ef-facts from all large releases over the ensemble of severe core damage / core-melt accidents.

e. The integrated safety goal decision matrix reflects the cost-tradeoffs (oraceiling)onsafetyimprovementcoststobe associated with a projected state of risk for a reactor plant.

Depending on this state of risk, cost-effective safety improve-ments should continue to be made until the safety goal is con-sidered to be met absent any other arguments to the contrary.

Alternatively, once the general performance guideline for the .

mean frequency of a large release is judged to have been achieved, the Comission's safety goal objectives are believed I to have been confidently achieved without the need for further safety improvements although the matrix suggests that some im-provements at small costs could still be warranted to further reduce the core melt frequency. Here it should also be noted that the range of cost-expenditures for safety improvements should be considered as a bounding of the costs; these reflect the total expenditures that are ju:;tifiable to drive the total risk state to an extremely small level. It is expected that in the absence of a plant-specific PRA to define the state of risk and plant-specific costs of safety improvements, surrogate de-signs will be used by the staff, including, where considered to be helpful, precursor impacts as projected from actual operating experiences. It is also intended that the staff will heavily rely on the new rebaselined risk results provided by NUREG-1150 for the generic safety issue resolutions absent any other equiv-alent state-of-art base-line for use.

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f. The question as to what constitutes a significant increase in the overall level of protection of the public health and safety (pursuant to 10 CFR 50.109) is, of course, dependent on where a plant risk state lies with respect to the Comisson's safety goal Policy. The integrated decision matrix and NUREG-1150 should help to answer this difficult question. For example, it is clear that the Comission's policy statement intends the pursuit of reasonable assurance of accident prevention against the occurrence of severe core damage accidents for both existing and future plants. It is also clear that the staff needs to work further to define with more specificity, for example, what particular collection of generic issues might represent a sig-nificant increase in the overall level of protection to public health and safety for those plants that may be above the risk state desired by the Commission's safety goal policy. This further staff effort is intended. It also need be recognized that once a plant risk state has confidently achieved a very low frequency for the large severe-accident caused releases (e.g.,

< 10-6/RY),thereappearstoexistfewreasonablepublichealth and safety arguments or cost-benefit algorithms that would sup-port significant further costs to be expended for safety improvements.

l l g. The health effect objectives on the incremental risk of early and latent cancer fatalities (i.e., the 0.1 percent objectivel is reflected in the matrix since these represent the ultimate

test that the Comission intended for use in gauging achievement of its safety goals. The Policy Statement is, however, silent i

on what the 0.1 percent increment means in tems of the absolute values to be taken for these individual mortality risks. NUREG-0880(1983) cited these absolute values to be at about 5x10-7/RY for the risk of an early fatality and 2x10-6/RY for the risk of a latent cancer fatality from a reactor accident occurring in any given year. Some(e.g.,ACRS)havepointedout thattherecould(anddoes)existyear-to-yearfluctuationsof the national average mortality risk statistics and questioned whether the absolute values on individual risk intended for use by the safety goal policy would likewise be allowed to fluc-tuate. It is intended that the safety goal implementation will use the constant values of 5x10~7/RY and 2x10~0/RY cited above ,

and that these will not fluctuate from year to year.

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ATTACHMENT B

.:[ 4 UNITED STATES 8' ~'^

n NUCLEAR REGULATORY COMMISSION

{ . ,$ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555

,,,,, December 15, 1986 MEMORANDUM T0: ACRS Members FROM: , Senior Staff Engineer -

SUBJECT:

CERTIFICATION OF THE ACRS SUBCOMMITTEE ON THE THE SAFETY PHILOSOPHY, TECHNOLOGY AND CRITERIA MEETING MINUTES, NOVEMBER 5, 1986, WASHINGTON, DC The minutes for the subject meeting have been certified as the official record of the proceedings for that meeting. Please destroy the Working Copy of the minutes issued November 14, 1986.

cc: R. F. Fraley M. W. Libarkin G. R. Quittschreiber J. C. McKinley T. G. McCreless M. N. Schwartz a

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DATE ISSUED: 11/25/86 ACRS Meeting Minutes on the November 5, 1986 Safety Philosophy, Technology and Criteria Subcomittee Meeting Washington, DC The ACRS Subcomittee on Safety Philosophy, Technology and Criteria met on November 5, 1986 at 1717 H Street, N.W., Washington, DC. The purpose of this meeting was to discuss the implications of the Chernobyl acci-dent for U.S. nuclear power plants. A NRC Staff draft report was avail-able prior to the Subcomittee meeting for the Subcomittee's review.

ACRS action on this matter is planned for the December 11-13, 1986 meeting with a second Subcomittee meeting scheduled for December 10, 1986. The Subcommittee meeting began at 9:00 a.m. and adjourned at 5:30 p.m. All discussions were held in open session. The principle attend-ees were as follows:

ACRS NRC D. Okrent, Subcommittee Chairman T. Speis D. Ward, Member B. Sheron M. Carbon, Member B. Boger H. Lewis, Member H. Richings C. Michelson, Member J. Stang J. C. Mark, Member S. Schwartz C. Wylie, Member L. Soffer R. Savio, Staff 0. Lynch S. Acharya FEMA B. Morris M. Sanders F. Congel ,

G. Sege W. Russell Highlights

1. The NRC Staff has produced a preliminary draft of a Chernobyl implications report. The report was transmitted to the Subcomit-tee under an October 30, 1986 and October 31, 1986 cover letters

o l SPT&C Meeting Minutes November 5, 1986  :

i for the Subcomittee's review. (See Meeting Briefing Book for the  :

November 6-8, 1986 ACRSMeeting.) This draft of the repcrt is currently also being reviewed and comented on by NRC management.

A revised report will be transmitted to ACRS in the later part of November. The NRC Staff has requested ACRS coment at the December 11-13, 1986 ACRS meeting and will present their recommendations to the Commission later in December. The NRC Staff's fact-finding report (reporting of the sequence of events) is expected to be completed by the end of November 1986. The information provided by the Soviets at the Vienna meeting will provide the primary basis for this report. The contents of the fact-finding and implications reports are sumarized on Figures 1 and 2.

2. The NRC Staff's objective in the implications report is to identify the candidate issues and assess the importance of these issues.

Judgements are to be then made as to: (1) whether the issue can be resolved on the basis of what is currently known, (2) whether it is likely to be resolved by ongoing work, or (3) whether new programs are needed. The NRC Staff intent is to focus this evaluation on the direct implications of the Chernobyl accident. The NRC Staff has concluded that no imediate regulatory action is needed for U.S. reactors and that existing U.S. practices and regulations

protect against Chernobyl-like events. The NRC Staff believes that I the lessons learned from Chernobyl can be used to reinforce and I

improve requirements which already exist or are being developed.

Studies will also be conducted to further evaluate the need for additional action. The NRC Staff's implications report discusses their recomendations in some detail. The major conclusions drawn l

are as follows:

l l (1) Reactor Operations - The emphasis given to human factors needs to be reinforced. Particular attention needs to be given to the adequacy of administrative controls, control of safety

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l SPT&C Meeting Minutes November 5, 1986 system status, and engineered safety features availability.

The NRC will be evaluating the mechanisms by which reactor modifications, tests, and experiments are controlled and the joint NSAC/AIF effort to develop criteria in this area, operator attitudes toward saf'ety and night shift operation were issues in the Chernobyl accident. The NRC Staff feels that the U.S. practices provide adequate safeguards against unacceptable operator attitudes and the problems associated with night shift operation. An assessment of the NRC require-ments on management systems will be conducted with emphasis on making determinations of the particular skills required for the various tasks. The NRC's current approach is to monitor licensee performance and to deal with problems on a case-by-case basis rather than to issue additional regulatory require-ments.

(2) Reactor Design - A wider range of reactivity insertion acci-dents than are currently considered under the SRP will be evaluated. The NRC Staff believes that these scenarios are less probable than those treated in the SRP. Accidents occurring when the reactor is shut down or at low power and the multiple unit aspects of reactor accidents will be reeval-uated. Current fire fighting procedures will be reviewed to determine if adequate consideration has been given to the problems associated with the presence of high levels of radiation.

(3) Containment Capabilities - The NRC Staff has concluded that the Chernobyl and Three Mile Island accidents graphically demonstrate the importance of containment performance in a severe accident. Containment performance in a severe accident and the use of containment venting or filtered venting as a

o SPT&C Meeting Minutes November 5, 1986 mitigation strategy are being evaluated as part of ongoing NRC programs. The NRC Staff has concluded that no additional evaluations are necessary and that the containment perfomance issue can be resolved by the ongoing NRC and Industry work.

(4) Emergency Planning - The lessons learned of the Chernobyl accident will be evaluated. Emphasis will be given to issues of sheltering vs evacuation and the effectiveness of the various protective actions. The NRC Staff believes that issues associated with long-term relocation, decontamination and use of contamination foodstuffs, the use of potassium pills, medical treatment, and accident recovery need to be carefully examined.

(5) Severe Accidcnt Phenomena - The NRC Staff believes that the current NRC treatment of steam explosions is adequate and will not initiate any new work in this area unless their examina-tion of reactivity insulation accidents indicates a risk-based need for examining energy depositions beyond 280 cal /gm-U02 . The NRC Staff has also concluded that the generation of combustible gases is adequately addressed in the current programs. The current work on accident source tems will be expanded to include more work on:

l (a) mechar.ical fuel fragmentation and dispersal by high

! energy processes.

(b) conversion of UO 2to loose / porous forms which could enhance aerosoliiation or radionuclide stripping of the fuel.

(c) effect of accident management on the release.

R -

SPT&C Meeting Minutes Npvember 5, 1986 (d) revaporization/resuspension of radionuclides.

(e) mechanisms for release of single element hot particles.

(f) hydrogen generation by dispersed fuel fragments.

(6) Economics Effects - The NRC Staff will use what has been learned on the economic effects of the Chernobyl accident to reevaluate the NRC's cost benefit methodology.

3. The Subcommittee members commented extensively of the NRC Staff's proposal. The summary of the comments made are as follows:

(1) The Chernobyl accident demonstrated the importance of human performance in safe operation of U.S. reactors. In particu-lar:

(a) The operators should have a good understanding of the principles of reactor safety, the technical basis for the operating procedures, and potential accident scenarios.

Their training should build on this basic understanding.

(b) The onsite plant personnel should include individuals with through understanding of the plant design and operational characteristics.

(c) A single competent individual whose primary concern is plant safety should exist and have the authority to order plant shutdown.

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SPT&C Meeting Minutes November 5, 1986 (2) The administrative controls which are intended to assure that procedures (in particular emergency procedures) are adequate and are followed should be very carefully evaluated.

(3) More attention needs to be paid to the fires which might be associated with a severe accident. Some issues to be con-sidered are overall fire fighting strategies (methods, fight-ing the fire vs letting it burn, providing better fire barri-ers, etc.) radiation protection, availability of equipment and personnel to fight multiple fires, and the impact of fires on the course of the accident, the accident release, and the safety of the potential for fires / explosions caused by leaks from process hydrogen stored in the plant should be considered adjacent units.

(4) Accidents which can cause reactivity insertion need to be examined. Multiple rod ejections, cold water insertions, void collapse, early-core-life positive temperature coefficients, and low power operation need to be examined.

(5) The effectiveness of U.S. reactor containments under severe accident condition should be carefully examined in view of the fact that such accidents were not part of the design basis.

It was noted that France, Germany and Sweden appear to be connitted to the use of filter-vented containment designs.

Mechanisms for containment failure (in particular early failure) such as missiles, degraded penetration, etc., such be carefully and objectively reevaluated.

(5) There was some disagreement with the NRC Staff's expressed satisfaction with U.S. administrative controls, operator training and attitude, and operational procedares. It was noted that the reasons for the failure of the system used by

W SPT&C Meeting Minutes November 5, 1986 the Soviets is not well understood. The vulnerabilities in the Chernobyl infrastructure need to be understood before it can be concluded that similar problems do not exist in the U.S.

(7) Technical Specifications are designed to control normal operations and may not adequately control tests, experiments, or other unusual reactor conditions.

(8) Methods for accident management need additional study and evaluation.

(9) Damage to large areas of land and other major societal re-sources need to be more carefully examined in NRC cost / benefit evaluations.

(10) The design process for advanced reactors should be directed as much as practicable to producing designs which will revert to a stable configuration when scrammed.

(11) It is essential that the important lessons learned from the Chernobyl accident be developed. A through and critical evaluation of the accident will be important.

(12) It is important that the significance of safety issues be understood and remembered even after resolutions have been found and implemented. These issues may need to be reevalu-ated in the future in the light of new information or changes which have been made in reactor design or operation.

o SPT&C Meeting Minutes November 5, 1986 NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, N.W., Washington, DC, or can be purchased from ACE-Federal Reporters, 444 North Capitol Street, Washington, DC 20001, (202) 347-3700.

jgyr--- l

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STATUS - FJY=A= 6 Y

_ CHAPTER TITLE LEAD AGENCY I

SUMMARY

9 NRC 2 INTRODUCTION NRC 3 PLANT DESIGN DESCRIPTION DOE 4 SAFETY ANALYSIS EPRI 5 ACCIDENT SCENARIO NRC 6 ROLE OF OPERATING PERSONNEL INP0 7 SOURCE TERM NRC 8 EMERGENCY PREPAREDNESS FEMA 9 ENVIRONMENTAL CONSEQUENCES EPA SCHEDULE ALL DRAFT CHAPTERS PREPARED AND SENT TO NRC (EXCEPT EPA)

NRC WILL ISSUE COMPLETE DRAFT FACTUAL REPORT TO PARTICIPANTS BY 11/7 MEETING TO REVIEW AND FINALIZE REPORT SCHEDULED FOR 11/18 EXPECT TO ISSUE FACTUAL REPORT BY END OF MONTH 9

9

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CHERNOBYL ACCIDENT CANDIDATE ISSUES I. OPERATIONS (AdministrativeControls)

I.1 Approval of Tests and Other Unusual Operations I.2 Controls to Assure Administrative Procedures are Followed 1 1.3 Bypassing Safety Systems ...  :* ,

I.4 Availability of Engineered Safety Features .

I.5 Operator Attitude Toward Safety I I.6 Night Shift Operations I.7 Hanagement Systems -

II. DESIGN -

II.1 Reactivity Accidents *

• II.2 Accidents at Low Power and When Shut Down II.3 Multiple Unit Protection II.4 Fires III. CONTAINMENT III.1 Beyond DBA Capabilities

, III.2 Venting IV. EMERGENCY PLANNING IV.1 Adequacy of Zone Distances IV.2 Long-Term Relocation, Decontamination Issues IV.3 Use of KI Pills IV.4 Data Acquisition & Reporting IV.5 Ingestion Pathway Monitoring & Ingestion of Foodstuffs IV.6 Emergency Medical Services IV.7 Onsite Response V. SEVERE ACCIDENT PHENOMEM Y.1 Source Terms '

Y.2 Steam Explosions V.3 Combustible Gas VI. ECONOMIC EFFECTS l

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