ML20235A879
| ML20235A879 | |
| Person / Time | |
|---|---|
| Issue date: | 06/07/1988 |
| From: | Ward D Advisory Committee on Reactor Safeguards |
| To: | Zech L NRC COMMISSION (OCM) |
| Shared Package | |
| ML20235A877 | List: |
| References | |
| ACRS-R-1341, NUDOCS 8806290287 | |
| Download: ML20235A879 (7) | |
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION.
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ADVis0RY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20665 e,,
- ,4 e.,,e June 7, 1988 The Honorable Lando W. Zech, Jr.
Chairman U.S. Nuclear Regulatory Comission Washington, D.C.
20555
Dear Chairman.Zech:
SUBJECT:
NRC RESEARCH RELATED TO HEAT TRANSFER AND FLUID TRANS NUCLEAR POWER PLANTS During the 338th meeting of 'the Advisory Comittee. on Reactor Safe-guards, June 2-4, 1988, we considered a report from our Subcommittee on Thermal Hydraulic Phenomena pertaining to its review of research activities sponsored by the NRC on reactor thermal-hydraulic phenomena.
This subject had been considered during the 337th ACRS meeting, May 5-7, 1988 and a number of previous meetings of the ACRS and the Subcom-mittee. We also had the benefit of the documents referenced.
Background
The technical subjects of heat transfer and fluid transport are of importance when' considering the safety of nuclear power cardinal They are of chief concern in relation to LOCA and reactor plants.
transients, and the performance of ECCS, steam generators, secondary systems, and containment phenomena.
These issues have been studied extensively in experimental and analytical programs sponsored by NRC and the industry.
The earliest reactor safety research dealt principally with reactivity Later research, sponsored initially by the Atomic addition accidents.
Energy Comission and then by the NRC, was devoted almost exclusively to providing a better understanding of the hypothetical large-break and the related performance of ECCS and containment LOCA (LBLOCA)
It was soon perceived that the complexities of two-phase flow systems.
and the time sequences involved in a LBLOCA were such that straight-forward ' experimental representation was difficult.
It was further perceived that traditional methods for application of empirical data to plants were subject to challenge.
In response to this, the NRC spon-sored development of complex computer codes at the Department of Energy's(DOE's)nationallaboratories.
These codes were intended to provide consistent treatment of the relationships among various plant systems during rapid transients and
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4 The Honorable Lando W. Zech, Jr. June 7, 1988 to bridge gaps in data from the various test programs.
In time, as the physical representations of two-phase flow and heat-transfer phenomena and the plant systems were made more detailed, broad interpolation and extrapolation from experiments were attempted and came to be relied upon.
This general strategy, that is, primary dependence on detailed math-ematical models of physical phenomena coded for rapid analysis by j
computers, has been adopted by the NRC for studying other technical areas involving complex phenomena and interactions.
Another product of this era, in addition to the extensive base of experimental information and the codes, has been a skilled cadre of experts.
These experts can be found in the NRC, the national labora-tories, the universities, and the industry.
These people understand thermal-hydraulic phenomena associated with LBLOCA probably as exten-sively as almost any other similar subject in modern technology is understood.
These experts are also well-schooled in the general strategy described above.
It is important that a cohesive group of experts be maintained.
The next period of thermal-hydraulic research followed the accident at TMI-2 and the gradual assimilation of the perspectives provided by probabilistic risk assessments.
Interest shifted toward small-break LOCA and plant transients.
De-errphasis of LBLOCA began.
The large system codes, developed in the previous era, were available and, with modification, came to be the primary means by which these less dramatic reactor events were analyzed.
It was recognized that these codes, i
originally written to incorporate certain conservatism in an attempt to envelop uncertainties in LBLOCA analyses, would serve their new purposes only if they could more realistically track the evolution of The conversion to realistic or "best estimate" codes is transients.
now complete.
thermal-hydraulic research, scientific In each of these periods of interest was confined to phenomena and time sequences associated with nonnal plant conditions and with faulted conditions extending to, but not beyond, the point at which a coolable core geometry is lost; research activity also included consideration of single-and two-phase ficw, heat transfer, and nonequilibrium conditions.
General Recommendations The following comments include recommendations for future research in the traditional " thermal-hydraulic" area, including specific recom-mendations for the code development program.
The Honorable Lando W. Zech, Jr. June 7, 1988 4
For the sake of discussion we have posed our comments in this section as a series of questions, with our reconinendations following as an-swers:
- Is there a need to continue a program of experimental research in the traditional thermal-hydraulic area?
Yes, but not indefinitely nor without specific purpose. There are several matters that currently need attention and will require several more years of experimental work at a moderate rate.
Specific recommendations will be given below.
NRC has played a key role in the fundamental development of thermal sciences related to nuclear power plants.
It should continue to furnish leadership, perhaps by more clearly defining basic research needs or directions for the DOE and industry.
- Is the strategy of dependence on large system codes as primary tools for analysis valid?
This strategy has both strength and weakness.
As strength, the system codes have the ability to model, in a consistent ~ and reasonably accurate way, the dynamic relationships among the various elements in a plant heat transport system.
They are weaker in the accuracy with which they model the complex physical behavior of system subelements, especially in extreme off-normal This weakness becomes an important problem because conditions.
analysts and decisionmakers tend to overlook the inaccuracies and to behave as if the codes were revealing physically correct and validated information about the plants. These codes are also very expensive to use and require specialists to use them properly.
if they are regarded as simply one These codes can be useful albeit often an important one, to the understanding of
- input, The codes can be dangerously misleading if they plant transients.
are used without engineering judgment and to the exclusion of simpler but less comprehensive analyses.
We are concerned that those conducting research in severe accident phenomena have fallen into this trap.
I
- Should traditional (i.e., LOCA) code development be continued?
The codes are now adequate for the purposes for which they are and further development is unjustified.
First, they needed For this satisfy the regulatory need related to the ECCS rule.
the Code Scaling, Applicability, and Uncertainty (CSAU) Program is
The Honorable Lando W. Zech, Jr. June 7, 1988 helpful.
Second, they are adequate as general-purpose tools for exploring and gaining understanding of other plant transients, from a safety rather than a regulatory perspective.
In this use, analysts should be guided by the coments above.
In making this recommendation, we recognize that the codes are not without flaws.
However, we believe that not all of the imper-fections in the codes can possibly be corrected by any reasonable program of research and code development.
Marginal improvements that could be made over the next few years by extrapolating the recent levels of development work will not be sufficient to attain a significantly higher plateau of code accuracy and validation.
l The code development effort has been a substantial technical achievement and the codes have made an important contribution to l
nuclear power plant safety.
Further refinement is unnecessary.
The CSAU Program will provide a reasoned perspective on the accuracy of the existing codes.
With that perspective available, we endorse the general strategy proposed by the RES staff toward maintenance of existing codes.
This would provide for completion of RELAP-5 and TRAC-PWR development through the International Code Assessment Program Consortium in 1989.
A modest level of re-sources would be provided to maintain the codes overall (including j
TRAC-BWR, COBRA-NC, and RAMONA-3B), based on regulatory needs.
Nuclear power plants are complex machines, even in normal operat-i We ing modes; they have many interrelated systems and processes.
believe that computer codes can model nomal operating behavior accurately and usefully, if extreme physical phenomena are not involved and if the codes can be validated by comparing their results to measurements of plant operating parameters. There is a significant resource in code development expertise at the national laboratories.
Consideration should be given to using this re-source with an approach to code development that takes advantage of inherent strengths in the present codes.
Efforts should be 4
concentrated on including all of the plant systems, providing code versions validated for specific plants and providing modeling and is transparent and understandable for use by interfacing that those expert in plant operation rather than just those expert in analysis by computers.
- Is it essential that a cadre of experienced people be maintained?
It is essential to maintain such a cadre, because questions of fluid and heat transport will always be central to reactor safety.
The NRC should maintain a center of expertise in experimental and
4 The Honorable Lando W. Zech, Jr. June 7,1988 The Technical analytical research in thermal-hydraulic phenomena.
Support Center at the Idaho National Engineering Laboratory serves this purpose.
However, the NRC should limit the program to:
(1) confirming selected information supplied by industry and (2) exploring important issues that the industry is not addressing.
Involvement of universities and other nongovernment research organizations should be encouraged. There should be free exchange of information with industry and international experts.
Specific Recommendations (1) The CSAU method, or something similar, can be used in other areas of safety analysis, that is, beyond the currently conceived purpose of assessing uncertainty associated with calculations by thermal-hydraulic codes.
In particular, its application to severe accident studies and risk assessments could serve, not only to provide an improved perspective on uncertainty, but also as a This should be inves-guide to allocation of research resources.
tigated.
of research on B&W reactor systems and programs (2) The current steam generators should be continued only to the once-through point that the technical understanding of B&W systems is compara-ble to that of other nuclear steam supply systems.
In particular, it should be demonstrated that adequate capability for predicting B&W systen performance is in hand.
Analysis of industry experience with water hansner events suggests (3) thet water hammer is not a significant initiator of nuclear power plant accidents.
However, insufficient consideration has been given to whether water hamer, occurring as a consequence of other contribute to unexpected failures that could initiators, might compromise core cooling. This issue should be investigated.
The recent steam generator tube rupture (SGTR) at North Anna has (4) been explained as the result of a series of mechanisms which indicate that multiple SGTRs are no more probable than has been believed.
The licensee's technically complex explanation was The NRC should explore this based on poorly understood phenomena.
issue sufficiently to confinn the licensee's conclusions.
the feed-and-bleed cooling process is not directly Although (5) in assessing the required by regulations, it is given creditindividual plants and of the overall safety of The contribution made by feed-plants in the United States.
and-bleed cooling to the safety of plants needs to be better
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4 The Honorable Lando W. Zech, Jr. June 7, 1988 i
established.
It is regarded as a "last ditch" cooling mode that -
can be effective in some plants.
Risk assessments are ambiguous about its importance.
There is significant uncertainty about the reliability with which this process can actually be carried out in many plants, perhaps in most.
In particular, there are questions about the flow capacity and reliability of the valves (usually power operated relief valves) essential to provide bleed flow and blowdown quenching capacity.
In addition, the complex flow path and the effects of uncovering the core do not seem to be well understood for all plants.
Research should be directed toward resolving the key uncertainties related to providing assured feed l
and bleed at plants that depend on the process for a margin of improved safety.
(6) The LBLOCA, the design-basis accident for certain plant systems, should be reconsidered in view of the results of research on leak-before-break and the revision to General Design Criterion 4.
Thermal-Hydraulic research will be necessary in support of this effort.
(7) The designs for so-called evolutionary LWRs and especially the
" passive" LWR being developed by the Electric Power Research Institute and DOE, will require research by the NRC to confirm certain favorable characteristics being claimed. The DOE Advanced Reactor Severe Accident Program is not sufficient for this purpose.
The NRC should use existing codes to review these designs so there is sufficient lead time to conduct more experimental or code development work, if necessary.
(8) There is some uncertainty about applicability of the RELAP-5 code to BWRs and to LBLOCAs. This should be resolved.
(9) Full documentation should be completed for the NRC codes that are maintained for active use.
This should include not only user manuals but developmental assessment reports and "models and correlations" documents.
Ideally, these would be published as NRC documents in the NUREG series to ensure widespread availability.
(10) Thorough analyses have generally been made only for the initial as LOCAs.
Analyses of the period of reactor accidents such follow-on transition to stable long-term cooling have been less comprehensive.
We recomend that NRC determine whether a more systematic and complete study of the reliability of such tran-sitions should be undertaken.
The Honorable Lando W. Zech, Jr. ACRS Members William Kerr, Harold Lewis, and Forrest Remick did not participate in the review of this matter.
Sincerely, 00.aQ David A. Ward Acting Chairman
References:
U.S. Nuclear Pegulatory Commission, NUREG-1080, Volume 2:
"Long-1.
Range Research Plan FY 1986 - FY 1990," Office of Nuclear Regula-tory Research, August 1985.
2.
U.S. Nuclear Regulatory Commission, NUREG-1266, Volume 2:
"NRC 1987," Office of Safety Research in Support of Regulation Nuclear Regulatory Research, May 1988.
Nuclear Regulatory Commission, Draft NUREG-1252:
" Thermal U.S.
3.
hydraulic Research Program Plan," Office of Nuclear Regulatory Research, November 9, 1987.
4.
U.S.
Nuclear Regulatory Comission, NUREG-1236:
"NRC Thermal Hydraulic Research Plan for B&W Plants," February 1988.
R.
A.
Dimenna, et al, Idaho National Engineering Laboratory:
5.
"RELAP-5/ MOD 2 Models and Correlations" (Draft Report), December 31, 1987.
" TRAC-PF1/
D. R. Liles, et al, Los Alamos National Laboratory:
6.
MODICorrelationsandModels"(DraftReport),providedtotheACRS in December 1987.
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I W ASHINGTON, D. C. 20555 July 20, 1988 The Honorable Lando W. Zech, Jr.
Chairman U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Chairman Zech:
SUBJECT:
REPORT ON THE INTEGRATION PLAN FOR CLOSURE OF SEVERE ACCIDENTISSUES(SECY-88-147) i During the 339th meeting of the Advisory Committee on Reactor Safe-guards, July 14-16, 1988, we discussed with members of the NRC staff a plan for the integration of the various severe-accident-related programs as described in SECY-88-147, " Integration Plan for Closure of Severe Accident Issues." This plan was also considered by our Severe Accidents Subcommittee during a meeting held on July 13, 1988.
We also had the benefit of the documents listed as references to this letter.
he comend the NRC staff for its efforts to develop an integrated approach for dealing with the various severe accident issues and to centralize responsibility for resolving them.
SECY-88-147 describes the first step toward developing such a plan, namely, identifying the relevant issues.
However, it gives little information on how the various issues are to be integrated. Rather, it discusses the severe-accident-related issues and programs that should be integrated, but does not describe the process to be used.
i The need for additional in.tegration is illustrated in the discussion of external initiators.
In several recent PRAs, externally initiated sequences are major contributors to risk.
This fact appears not to have been considered in SECY-88-147.
Considering only internal initiators may well provide a distorted picture of the " major vulner-abilities" for a particular plant.
This may result in an inappro-l priate allocation of resources for plant-specif.ic fixes, unless all system changes are delayed until external events are treated.
This does not seem to be the procedure to be used.
Further, the statement is made, in support of delaying a consideration of external initi-ators, that no new sequences are likely to be initiated by seismic l
This seems to contradict the conclusions of a Brookhaven l
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The Honorable Lando W. Zech, Jr. July 20, 1988 I
study of the GESSAR PRA which concluded that relay chatter, produced by a seismic event, could be a major risk contributor.
Furthennore, it ignores the fact that a large seismic event has the capability (much less likely for other initiators) of simultaneously initiating a large number of risk-significant sequences.
The comments on severe accident management provide no indication of how the licensee is to proceed.
Although for this issue, immediate action is not required in connection with the Individual Plant Exami-nations (IPEs), the implication is that enough infonnation now exists to permit a licensee to formulate an appropriate program.
We note that on March 13, 1985, the ACRS sent a memorandum to the then-EDO, Mr. William J. Dircks, in which we asked if enough information existed to provide guidance to plant operators in a situation in which core melting had proceeded without a source of cooling.
Our question was whether a situation could develop in which, if coolant became avail-able after c. ore melt had begun, adding coolant to the in-vessel melt would exacerbate the accident.
We have yet to receive a response to our r.emorandum. This, we think, is a rather fundamental question. If the staff does not have the information to answer this question, how is a licensee to reach a decision?
Does : existing instrumentation provide the information needed? Does the ins'trumentation suggested in Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," suffice for the task?
For accident manage-ment, answers to such questions are required.
t We observe that, in the evaluation of containment performance, the licensee is to " consider" direct containment heating (DCH), a pos-tulated event about which there are major uncertainties. However, the proposed gereric letter on IPEs (Reference 3) states that no major changes in the containment are to be made until the NRC research l
program has produced information required to decide what, if any, system changes should be made.
Is anything to be done in the mean-time? What is the " consideration" by licensees to produce? We note that a Panel report on source term uncertainties (Reference 4) con-cluded that information needed to determine the effects of DCH is unlikely to be available for a long time. The Panel reconnended that, rather than wait for the results of the needed research, the probabil-ity of DCH should be made negligibly low by hardware changes or procedural measures.
Furthennore, in describing the resolution of some of these important issues, the process to be used is left so vague as to be uninterpret-able.
For example, from the discussion of the way in which it is proposed to deal with severe accidents for advanced light water reactors (LWRs), one gets the impression that if some as yet undefined
The Honorable Lando W. Zech, Jr. July 20, 1988 process, pessibly rulemaking, is put into place, the problem will somehow become resolved..
In the area of r.ontainmer.t performance criteria for advanced LWRs, it is especially important that some early decisions be made. The review process currently being considered appears to endorse the use of design criteria based on " design-basis accidents" formulated before the Reactor Safety Study (WASH-1400), which indicated a need to consider severe ' core damage accidents. This seems, at best, imprudent in light of all that has been learned since these criteria were first formulated.
Designs using these " obsolete" criteria are now being ccrsidered in the liter. sing process.
In our discussions with the staff, we explored how the Reactor Risk Reference Docurent (NUREG-1150) will be used in the resolution of the severe accident issues. Althcugh we were told that the infomation in this document will play a key role, we were unable to get a clear picture of just how.
If NUREG-1150 is to play a key role, it is irrpertant thtt its accuracy and credibility be established.
We believe that subjecting the final version of NUREG-1150 to a thorough peer review is required as part of the process of establishing credi-
- bility, j
We believe a glossery of terms used in SECY-88-147 would be helpful.
We suggest that SEVERE ACCIDENT, DAMAGED CORE, CORE DAMAGE, CORE MELT, VULNERABILITIES, RAD 10ACTIVL RELEASE, LARGE RADIOACTIVE RELEASE, CON-
]
TAINMEh! PERFORMANCE, CONTAlhMENT FAILURE, and CONTAINMENT BYPASS be defined.
In addition, definitions for FRONT END, BACK END, LEVEL I I
PRA, FEEVENTION, and MITIGATION as used in this paper might be helpful.
Finally, we encourage the staff to continue its efforts toward inte-gration of the various programs being developed for resolution of the severe accidert issues. We believe that the most recent draft generic letter describing the IPE program (Reference 3) represents a move in the direction we have recommended in our letter to you of May 10, 1988.
We are convinced that further integration can conserve re-sources of both the staff and the licensees and can contribute to a more effective process for risk reduction in operating plants.
Sincerely,
- w. xer, Chairman l
i
The Honorabie Lende L'. Zech, Jr.
4-July 20, 1988
References:
1.
SECY-88-147, Memorandum dated May 25, 1988, for the Commissioner.
from V.
Stello. Executive Director for Operations,
Subject:
Integrat1cn Plan for Closure of Severe Accident Issues 2.
Brootnaven hational Laboratory Draft Report, "A Review of the GESSAR 11 BWR/6 Standard Plant Seismic Probabilistic Risk Assess-ment," September 1984(Unpublished-Predecisional) 3.
Memorandum dated April 1,1988, from T. Spets (NRC) to W. Kerr (ACRS), " Documentation Necessary for the Initiation of the Severe Accident Policy Implementation" (Draft Predecisional Attachments
- Portions Updated as of June 28,1988) 4.
Erookhaver National Laboratory Rt. port, NUREG/CR-4883, " Review of Research on Uncertainties in Estimates of Source Terms from Severe Accidents in Nuclear Power Plants," H. Kouts April 1987 5.
U.S.
Nuclear Regulatory Commission. WASH-1400 (NUREG-75/014)
" Reactor Sefety Study," October 1975 6.
U.S.
Nuclear Regula+.ory Commission, NUREG.1150, " Reactor Risk Reference Document," Draft for Comment, February 1987 l
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W ASHINGTON, D. C. 20S55 August 16, 1988 i
The Honerable Lando W. Zech, Jr.
Chairman U.S. Nuclear Regulatory Commission Washington, D.C.
'20555
Dear Chairman Zech:
SU5 JECT:
REPORT ON NUREG-1150, " REACTOR RISK REFERENCE DOCUMENT" During the 340th meeting of the Advisory Committee on Reactor Safe-l g u a rc's, August 11-13, 1988, we discussed the staff's plan for the development of the final version of NUREG-1150, " Reactor Risk Reference Document," with Mr. V. Stello, Jr., Executive Director for Operations, and members of his staff.
We also had the benefit of the documents reference 6.
In our July 20, 1988 letter to you on the Integration Plan for Closure l
of Severe Accident Issues, we stated, "We believe that subjecting the l
final version of NUREG-1150 to a thorough peer review is required as l
part of the process of establishing credibility."
Reviews by a number of individuals and groups were highly critical of the original draf t of NUPEG-1150.
In view of the extensive modifica-tions that have been made in response to this criticism, the current version must be regarded as a new document.
Also, since this document is intended to play a substantial role in the implementation of the Comission's severe accident policy, its quality and credibility are very important.
We recommend that before publication in final form, the final version of NUREG-1150 be subjected to a thorough peer review.
Sincerely, 1
W. Kerr Chainnan
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j The Honorable Lando W. Zech, Jr. August 16, 1988
References:
1.
Memorandum dated August 10, 1988 from V. Stello (EDO) to R. Fraley (ACRS), " Plans for Review of Final NUREG-1150" 2.
SECY-88-147, Memorandum dated May 25, 1988, for the Commissioners from V.
- Stello, Executive Director for Operations,
Subject:
Integration Plan for Closure of Severe Accident Issues 3.
Brookhaven National Laboratory Report, NUREG/CR-5000, " Methodology for Uncertainty Estimation in NUREG-1150 (Draft): Conclusions of a Review Panel," H. Kouts et al., December 1987 4
Lawrence Livermore National Laboratory Report, NUREG/CR-5113
" Findings of the Peer Review Panel on the Draft Reactor Risk Reference Document, NUREG-1150," W. Kastenberg et al., May 1988 5.
Anerican Nuclear Society, " Initial Report of the Special Committee on Reactor Risk Reference Document (NUREG-1150)," L. LeSage et al.,
Draft Report dated April 1988 l
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