ML20214W970
| ML20214W970 | |
| Person / Time | |
|---|---|
| Issue date: | 10/31/1983 |
| From: | Malaro J NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| Shared Package | |
| ML20213E209 | List:
|
| References | |
| FOIA-87-113, FOIA-87-60 NUDOCS 8706160275 | |
| Download: ML20214W970 (47) | |
Text
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SEVERE ACCIDENT DECISIONS FOR EXISTING NJCLEAR POWER PLANTS s5 DRAFT TWO OCTOBER 1983 t
U.S. NUCLEAR REGULATORY COMMISSION 1
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o SEVERE ACCIDENT DECISIONS FOR EXISTING MjCLEAR POWER PLANTS -
Contents
- 1. Introduction and Background
- 2. Identification and Organization of Regulatory Questions
- 3. Approach to Answering the Regulatory Questions
- 4. Information Needs ,
- 5. Schedule Considerations S
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- 1. Introduction and Background The NRC program for dealing with severe accidents has been described -in ___
NUREG-0900, " Nuclear Power Plant Severe Accident Research Plan," January 1983, and " Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regu'lation," April 13,1983 (48 FR 16014).
In a letter of June 16, 1983, commenting on the NRC safety research program budget for fiscal yeart 1985 and 1986, the ACRS noted its position on the severe accident program, advanced a list of issues that need to be addressed, and urged continuing effort toward coordination and prioritization of the various activities. On July 12, 1983, the staff briefed the Commission on the status and content of the severe accident research program and discussed the need to define the process and the regulatory issues for the severe accident decision.
In a Staff Requirements Memorandum dated July 18, 1983 to W. J. Dircks from S. J. Chilk, the Commission requested "...an expanded and refined...
list of Regulatory Issues related to severe accidents...which would serve as the basis for further Commission consideration of the content and orienta-tion of SARP and of the regulatory process to be used in addressing severe accident issues." The Commission asked that the ACRS review the list of Regulatory Issues prior to its submission to the Commission.
In accordance with the Commission's request, the Senior SARP Review Group drafted an initial effort toward a systematic identification of regulatory issues and a description of the approach (primarily deteministic, supple-mented by probabilistic and systems assurance analysis) the Commission might -
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use to arrive at severe accident decisions for existing nuclear power plants. The draf t was transmitted to the ACRS by a letter dated August 5, .g. ,
1983, that asked for ACRS review so that issues and the decision process could be refined, and so that any needed changes in SARP could be . identified.
The draft was the focus for discussions with the ACRS Subcommittee on Class 9 1
Accidents on August 23, 1983, and with the full ACRS on September 1,1983.
By letter dated September 2,1983, the ACRS stated, "We encourage' the NRC Staff to continue its development of this general approach. It appears to be an improvement over what has been proposed up to this time." The ACRS also provided several specific comments, which are addressed in this document. In particular, the ACRS commented that, although the systems assurance analysis approach may be useful for some purposes, the ACRS did not recommend its application to the severe accident decision problem.
Accordingly; we are proposing a combined detenninistic-probabilistic approach in Chapter 3, below, and have eliminated any expansion of systems assurance analysis techniques for severe accident decisionmaking.
We have limited this document to severe accident decisions for existing nuclear power plants because the Commission has set conditions for standard l
l designs for reference in future CP applications, namely:
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- 1. Demonstration of compliance with current Commission regulations;
- 2. Completion of a PRA before standard design approval;
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- 3. Completion of a staff review of the standard in accordance with the Standard Review Plan; . g.
- 4. Consideration of all applicable Unresolved Safety Issues;
- 5. Adherence to the requirements coming from th'e experience at TMI and set out in 10 CFR 50.34(f), 'the "CP/ML rule"; and
- 6. Approval through rulemaking under 10 CFR Part 50, Appendix 0, to certify standard designs for 10 years.
These conditions and the time needed by rulemaking for the standard design approvals will afford ample opportunity to factor into reviews of future standard designs the lessons learned from severe accident decisions for existing nuclear power plants, as recommended by the ACRS.
Other ommnents from the ACRS meeting and from the ACRS letter of September 2, .1983, have been factored into the changes in this document that we have made since the August 5,1983, first draf t. The objective of this document is to invite comments and feedback from the ACRS and the Commission on the framework for the severe accident decision for existing plants so that issues and the decision process can be refined and so that any needed changes in SARP and the Severe Accident Policy Statement can be identified. We intend to include this redefinition in SARP, obtaining ACRS comments and agreements by the end of 1983, and in our rewrite of the Severe Accident Policy Statement for consideration by the Commission in January 1984.
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2.- Identification and Organization of Regulatory Questions Primary Question ._
The primary question to be addressed by the NRC's severe accident research program (SARP) and the Commission's severe accident decision is as.follows:
What changes, if any, should be made in nuclear reactor regula-tion to account -for accidents involving core damage greater than the present design basis, including core meltdown acci-dents?
There is general agreement by NRC staff that the primary question accurately characterizes the severe accident issue. The ACRS notes that, although this' is probably not the only way to define the issue, it is a reasonable approach.
The question applies to existing reactors (in operation or under construction) and to future reactors. For future reactors the Commission has proposed to follow an approach involving standard plant design certification through rule-making. This paper considers the issues and the process for answering the severe accident question for existing plants. There will be feedback from the existing plant decision process to the standard plant certification process, and vice versa.
Organization of Questions and Issues The primary question can be broken down into a few component questions. We call these the Regulatory Questions. The Regulatory Questions in turn serve as general headings for organizing more detailed questions. We have also identified a set of Technical Issues that are to be addressed in order to answer the Regulatory Questions. -
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-4 To provide a logical framework for the severe accident decision, we propose to agree upon the statement of Regulatory Questions and Technical Issues ,_
- - j with the ACRS and the Consnission and to agree on the approach (decision process) for answering the questions. The next step, after agreeing on the questions, issues, and decision process, will be to correlate the elements of SARP with the information needed to accomplish the decision process. ,
' Chapter 4, below, describes the process now underway to make that l
correlation. It has begun in parallel with discussions of the questions, issues, and decision process in the interest of saving time. It can be adjusted if needed. Finally, any gaps in knowledge or any unnecessary l elements of the present SARP thus identified by the overall process will be corrected.
The Regulatory Questions l
l The Regulatory Questions that are components of the primary question are as follows:
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- 1. What criteria should be used to determine whether additional protection for severe accidents is needed or desirable?
- 2. How should the Commission decide the severe accident questions?
- 3. How safe are the existing plants with respect to severe accidents?
- 4. How can the level of protection for severe accidents be increased?
- 5. What additional research or information is needed?
- 6. Is additional protection for severe accidents needed or desirable?
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These questions have been revised to reflect ACRS' reviews. Also the ACRS urges that priority be given, first to the method for answering these questions, and second to information that may be needed to provide the answers. _.
Underlying these questions are more detailed questions, as shown in Table 1.
Technical Issues .
A number of technical issues will affect the answers to the Regulatory Questions. Most of the technical issues are independent of the approach taken to severe accident decisionmaking, whether it is based primarily on deterministic or on probabilistic considerations. For example, a well-recognized issue is the magnitude af the source terms of fission products released from the containment to the environment. This particular issue plays a major role in our perception of the safety of existing plants (the third regulatory question in Table 1), whether we measure safety in terms of risk (probabilities and consequences) or in terms of the consequences of selected accident sequences in a deterministic evaluation of safety.
The answers to the Regulatory Questions posed in Table 1 are dependent upon the positions adopted on the technical issues. For example, if the results of experiments and analysis indicate that uncontrolled hydrogen combustion following a severe accident like TMI-2 results in containment failure (the technical issue), then the safety of the plant is probably inadequate (the 2-3
. . . , . _ .= _ _ .-. - .. - _ - . _ _ _ _ _. m . _ m TABLE 1 Organization of the Regulatory Questions
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- 1. -What decision criteria should be used to determine whether additional protection for severe accidents is needed or desirable?
1.1 What are the possible reasons for additional protection?
1.2 What considerations apply to them?
1.3 What criteria can be formulated?
1.4 'What should be the role of the Commission's Safety Goal?
- 2. How should the Commission decide the severe accident question?
2.1 Should' the decisions be generic or plant specific?
2.2 Should other regulatory issues and programs be integrated or excluded from the severe accident decision?
2.3 How will we assure that the requisite level of protection for severe accidents, once decided upon, is maintained throughout the life of a plant? -
2.4 Should rulemaking be used or some other method for obtaining public comment?
- 3. How safe are the existing plants wi.th respect to severe accidents?
3.1 What should be considered in this measurement of safety?.
3.2 How do the terms of measurement compare, including uncertainties?
3.3 Which accidents are to be considered and which ones can be ruled out?
3.4 Using these measurements, how safe are the existing plants? -
- 4. .How can the level of protection for severe accidents be increased?
4.1 What types of improvements are available?
4.2 How effective are they?
4.3 What are their costs and side effects?
- 5. What additional research or information is needed?
5.1 What are the information gaps bearing on the severe accident decision?
5.2 Are additional data necessary for implementation of the decision?
5.3 Are additional data necessary for confirmation of the decision?
5.4 Are there specific issues that require more data before they are decided?
- 6. Is . additional protection for severe accidents needed or desirable?
6.1 What conclusions do we reach when we assess available information in light of severe accident decision criter14 (from question 1)?
6.2 What is the likelihood that performance could be improved by the alternatives available?
6.3 Are the likely-improvements worth the cost? -
6.4 What is an appropriate be. lance of prevention and mitigation measures? -
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regulatory question) and a hydrogen control system should be backfit into the plant (the regulatory decision). The technical issues .should be the , l initial focus of the. severe accident decision process.
The NRC staff and IDCOR have developed a joint list of technical i~ssues.
An earlier version of that list was provided to the ACRS on June 3,1983.
Although there have been some discussions of the list with ACRS, further subcommittee reviews are needed and planned for the October-November time frame. In upcoming NRC/IDCOR technical exchange meetings, discussions will be held to identify issues for which there is general agreement, issues where there is disagreement, and issues where the state of information is inadequate to support decisions. The positions taken on the technical issues will form the basis for the regulatory decisions that will be made later. Agreement will also be soug'ht on descriptions of the magnitude of the uncertainties associated with the issues. If the uncertainties are too large, it may not be possible to base a decision on an accurate under-standing of the issue. If further research can narrow the uncertainty, it may be better to defer the decision. Otherwise, it may be necessary to treat the issue conservatively and to proceed with the decision.
In Chapter 3, below, the recommended approach for answering the Regulatory Questions is described. A list of Technical Issues is discussed in Chapter 4, along with our recommended approach for stating staff positions on the issues, obtaining ACRS and IDCOR input, and revising as necessary.
The major Technical Issues with respect to severe accidents are the following:
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t 1.0 Severe Accident Phenomenology 1.1 Progression of core melt in the reactor coolant system 1.2 Loading .of the containment ._ , -
1.3 Response of the containment and other essential equipment 1.4 Fission product release and transport 1.5 Ex-containment fission product transport and consequences 2.0 Safety Assessment 2.1 Characterization of plants and sequences 2.2 Assessment of existing plants 2.3 Assessment of plants with modifications 3.0 Decision Met,hodology Other issues are either subordinate to these major issues (see Table 3 of Chapter 4, below) or are significantly less important to the severe accident decision for existing plants. After agreeing on the Regulatory Questions, we will concentrate our efforts to reach agreement on the Technical Issues and the subissues 'and then to translate these into regulatory decisions about how much protection for severe accidents is enough. -
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- 3. Approach to Answering the Regulatory Questions The proposed severe accident policy statement published on April 13 1983 ,
suggested a three-step process for arriving at severe accident decisions for existing plants. First, quantitative risk assessment techniques were to be used to estimate the relative importance of potential nuclear power plant accident sequences where sufficient data exist to make comparisons. ,
Second, a range of possible design and operational changes to improve accident prevention avid consequence mitigation capabilities were to be studied to determine the costs and safety benefits of backfitting them to plants in operation or under construction. Finally, using engineering and policy judgment, supplemented by probabilistic risk assessment where appro-priate, the NRC was to make decisions on whether reductions in severe acci-dent risk are necessary. . If risk reductions are deemed necessary., research should tell how best to achieve them, whether by accident prevention or consequence mitigation, or by a balance of the two. The approach was cri-ticized for placing too much reliance on PRA. The main purpose of this present effort is to agre's on an approach to the severe accident decision which places primary reliance on deterministic engineering analysis and assigns a role to PRA which is consistent with its known strengths and weaknesses and the technical state of the art.
The principal decision path recomended by the staff is primarily determi-nistic in character. That is to say, it relies primarily upon engineering analysis of LWR safety and performance; of the response of existing plants to core melt accidents; and of potential performance objectives, hardware changes, and operational controls or procedures that could be backfit to .
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improve safety for severe accidents. This detenninistic method would also include a cataloging and assessment of the relevant considerations
.m for understanding the hazard that severe accidents pose, including some
. sense of their likelihood. Because of the primarily deterministic nature of the decision process and the primarily judgmental character of the final decision, the process would best be implemented by regulations. The ,
process would not place primary reliance upon the Commission's Safety Goal, but would use quantitative engineering analysis where possible and where supported by data.
Probabilistic risk assessment will be used to supplement the deterministic approach where justified by data and by a full consideration of uncertainties.
PRA will be of value in cataloging and arranging in order of significance the dominant accident sequences ind' associated containment response for the
" internal events" and in providing a convenient additional perspective on risk judgments. PRA has the advantages of most direct comparability with the Commission's Safety Goal and of direct utility in NRC's required system of Regulatory Analysis for all new generic requirements. It is difficult to a priori prescribe the weight to be given PRA in seve'e r accident decisionmaking. The weight will vary among the technical issues and the extent to which ongoing or past research has reduced uncertainty about them.
In sumary, we recommend a deterministic approach that most highly values engineering analysis and judgment, supplemented as appropriate by PRA, for the severe accident decision for existing plants.
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In the remainder of this chapter, we provide more details on the approach to decisionmaking and on the administrative considerations that apply. ,
regardless of what approach is finally selected. The details are provided in the fom of recommended technical and policy evaluations for each of the six Regulatory Questions.
It should also be noted that NRC has a program underway evaluating the usefulness of System Assurance Analysis (SAA) for nuclear power plant 1
applications. SAA is generally the methodology that NASA and D00 follow instead of PRA. The NRC program of SAA started only recently (1982) and is developing slowly due to budget limitations. Infonnation available from the SAA program will be coordinated with the SARP. The results of systems interaction analysis, failure modes and effects analyses, and system reliability assurance analyses perfonned under the SAA program, to the extent available, will be utilized in the severe accident decisionmaking process.
3.1. What Decision Criteria Should be Used?
Up to this time the safety of nuclear power plants in the U.S. has been primarily regulated on a deterministic basis using a defense-in-depth philo-sophy. That philosophy emphasizes the prevention of severe accidents. One element of the philosophy is to provide protection for a specified group of postulated accidents, the so-called " design basis accidents." Regulations and Regulatory Guides enumerate the accidents to be considered, the ground rules to be used for their evaluation and the criteria to measure the acceptability of a design. This approach has provided sufficient pro-e 3-3