ML20147E634
| ML20147E634 | |
| Person / Time | |
|---|---|
| Issue date: | 02/17/1988 |
| From: | Kerr W Advisory Committee on Reactor Safeguards |
| To: | Jason Wright HOUSE OF REP., SPEAKER OF THE HOUSE |
| Shared Package | |
| ML20147B148 | List: |
| References | |
| ACRS-R-1286, NUDOCS 8803070116 | |
| Download: ML20147E634 (2) | |
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p ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20565 g .
February 17, 1988 i
The Honorable James C. Wright . Jr.
Speaker of the United States House of Representatives Washington, D.C. 20515-
Dear Mr. Speaker:
In accordance with the requirements of .Section 29 of the Atomic Energy Act of 1954, as amended by Section 5 of Public Law 95-209, the Advisory Comittee on Reactor Safeguards has reported each year to the Congress on the Safety Research Program of the Nuclear Regulatory Comission.
In our December 18, 1986 letter to you, we proposed te prrvida more focused reports ~ devoted to fewer significant research issue: that we believe merit Congressional attention rather than one allainclusive annual report. The Comission has agreed with our suggestion, and NRC Chairman Zech has submitted a legislative proposal in the form of a '
draft bill on December 2, 1987 to amend Section 29 of the Atomic Energy Act of 1954 to accomplish this.
In our December 18th letter we noted also that, in the event that a year passes without the appearance of any significant research issue, we '
, would report that fact to you in writing. Although in the past year we have not encountered any major research issues that we believe merit Congressional attention, we have provided coments and recomendations -
to the Comission and the NRC Staff on several NRC research efforts. 'i Copies of these reports are attached for your infomation and use. ,
We are in the process of reviewing several elements of the NRC research program such as Thermal-Hydraulic Phenomena Severe Accident Manage.nent, Plant Aging, Containment Perforinance, and Seismic Design Margins. Also, l -we expect to review proposed research associated with Human Factors, i
Equipment Qualification, and Fire Protection. Upon completion of our ,
reviews, we will provide reports to you on those specific items found to ,
be sufficiently substantive to warrant Congressional attention. l Sincerely.
8903070116 080217 PDR ACRS PDR William Kerr R-1286 Chairman
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The Honorable James C. Wright, Jr. February.17, 1988 Attachments:
- 1. Letter from William Kerr, ACRS Chairman, to Lando W. Zech, U.S. NRC Chainnan,
Subject:
ACRS Report on Proposed Research to Reduce Source Term Uncertainty, dated May 13, 1987.
- 2. Letter from William Kerr, ACRS Chairman, to Victor Stello, Execu-tive Director for Operations, U.S. NRC,
Subject:
ACRS Coments on the Embrittlement of Structural Steel, dated July 15, 1987.
- 3. Letter from William Kerr, ACRS Chairman, to Victor Stello, Execu-tive Director for Operations, U.S. NRC,
Subject:
ACRS Coments on International Cooperation on Research Related to Radiation Protec-tion, dated July 15, 1987.
- 4. Letter from William Kerr, ACRS Chairman, to Lando W. Zech, I!.S. NRC Chairman,
Subject:
ACRS Coments on Draf t NUREG-1150, "Reactor Risk Reference Document," dated July 15, 1987.
- 5. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chainnan,
Subject:
ACRS Coments on Research into Continuous Containment Leakage Monitoring, dated July 16, 1987.
- 6. Letter from William Kerr, ACRS Chairman, to Lando W. Zech, U.S. NRC Chairman,
Subject:
Preliminary ACRS Views on Fire Risk Research Scoping Study, dated August 10, 1987.
- 7. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chaionan,
Subject:
ACRS Coments on Code Scaling, Applicability and Uncertainty Methodology for Detennination of Uncertainty Associated With the Use of Realistic ECCS E. valuation Models, dated September 16, 1987.
.8. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chainnan,
Subject:
ACRS Coments on Radioactive Waste Management Research and Other Activities, dated November 10, 1987.
- 9. Letter from William Kerr, ACRS Chairman, to Victor Stello, Execu-tive Director for Operations, U.S. NRC,
Subject:
ACRS Coments on Memorandum from Victor Stello, Jr., EDO, Dated October 7, 1987 Regarding the Embrittlement of Structural Steel, dated December 8, 1987.
- 10. Letter from William Kerr, ACRS Chainnan, to Lando W. Zech, U.S. NRC Chainnan,
Subject:
ACRS Coments on Development of a Method to
, Establish Priorities for Research Activities, dated February 16, l 1988.
l 11. Letter from William Kerr, ACRS Chairman, to Lando W. Zech, U.S. NRC l Chairman,
Subject:
ACRS Coments on Selected FY 1988 NRC Radio-active Waste Management Research Programs, dated February 17, 1988.
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ADVlsORY COMMITTEE ON REACTOR SAFEGUARDS
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, gw....+/y May 13, 1987 The Honorable Lando W Zech, Jr.
Chaiman U.S. Nuclear Regulatory Comission Washington, D.C. 20555
Dear Chaiman Zech:
SUBJECT:
ACP.S REPORT ON PP.0 POSED RESEARCH TO REDUCE SOURCE TERM UNCERTAINTY During the 325th meeting of the Advisory Comittee on Reactor Safe-guards , May 7-9, 1987, we discussed a proposed research program for resolution of source term uncertainty areas as described in SECY 86-369, "Plan To Address Source Term Technical Uncertainty Areas." We also con-sidered BNL report NUP.EG/CR-4883, an evaluation of this program by panels of experts sponsored by NRC. The ACRS Subcomittee on Severe Accidents considered this matter during a meeting on April- 22, 1987. In our review, we had the benefit of discussions with the NRC Staff and the documents referenced.
We commend the expert panels for their expedit'el review and for their coments concerning some very complex phenomena. We agree generally with their findings and recomend that the Staff give careful consid-eration to their suggestions in planning the proposed research program.
We make the following additional observations:
(1) In our report dated June 10, 1986 in which we comented on NUREG-0956, "Reassessment of the Technical Bases for Estimating Source Terms ," we recomended that the Staff attempt to quantify the j uncertainties that were identified. The expert panels also noted
! that there are no quantitative estimates of the magnitude of the identified uncertainties. We agree with the panels that those planning the research programs need guidance as to which contribu-tors to uncertainty are most important. To provide this guidance, the Staff should attempt not on'y to specify uncertainties in the descriptions of particular phenomena, but should also estimate l their contribution to risk. There is also a need for an estimate of the level of uncertainty that is acceptable in making regulatory l decisions. Although SECY 86-369 identifies areas of uncertainty, I it does not indicate what level of uncertainty would be acceptable,
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, The Honorable Lando W Zech, Jr. May 13, 1987 nor does it indicate how likely it is that the proposed research will reduce the uncertainty to an acceptable level.
(2) 'In the areas of steam explosions and hydrogen combustion, one of i the panels recomended a reduction in research activities. For l steam explosions within the vessel that lead'to early containment failure, the consensus is that the conditional probability for such an event is very small (0.01), and thus need not be considered fur-ther. This panel further concluded that hydrogen combustion is reasonably well understood and that uncertainty in its understand-ing contributes relatively little uncertainty to estimates of source terms and risk. However, significant uncertainties do remain in regard to the generation of hydrogen during *.n accident.
With the evidence now available to us, we agree with the panel's recommendation.
(3) A panel concluded that information needed to reduce the uncertainty in risk estirrates for direct containment heating (DCH) will not be available within the next four or five years, even if a crash orogram is implemented. In light of this estimate, the panel recomended the exploration of plant changes (hardware or proce-dures) which would eliminate the sequence. The panel also recom-mended that the DCH experimental program be reorganized to show the effects of water and structural failure on DCH. We concur in both recomrendations. In general, we conclude that the existing program is too narrowly focused. The program should be redirected to
, encompass a broader range of possible scenarios, including esti-mates of realistic mass flows from the vessel and possible vessel failure modes. The question of what is credible in the various situations trust not be submerged in some large computer code, but i
should initially be sorted out by more straightforward and trans-parent physical arguments concerning the range of possibilities.
(4) There has been considerable discussion of the uncertainty associ-
- ated with the chemical fenn of iodine, either volatile (elemental) t or non-volatile (chemically bound as in CsI). Af ter the TMI-2 accident, the absence of elemental iodine led some to conclude that the estimated risk should be reduced by a factor of as much as 100 from risks reported in WASH-1400, where it was assumed that all of the iodine was in elemental forin. It is now reported that in l studies conducted in the preparation of NUREG-0956, the difference l in risk for volatile vs. non-volatile iodine is only about a factor of 3. A lesser priority should be assigned to research in this area.
l (5) We observe that estimates of accident progression at key points in l the ccre melt iequence depend on the prediction, using inadeouately i based cunputer codes, of such parameters as melt temperature and l
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The Honorable Lando W. Zech, Jr. May 13, 1987 time required for vessel melt-through. There appear to be signifi-cant uncertainties in the predictions of a number of these key parameters that tend to be masked by the codes. Since vessel penetration, core-concrete interactions, and the concurrent release
, of fission products, for example, are all very sensitive to melt tenperature, we urge that efforts, including both experiments and independent calculations, be made to provide some independent and nore transparent assessment of the behavior of key parameters.
Comparison with another code embodying the same underlying as-sumptions is not sufficient.
(6) In light of the importance of containment behavior in detemining the magnitude of the source term, we recomend that more attention be given to the identification and evaluation of other scenarios having the potential for leading to a large release of radioactive material.
Additional coments by ACRS Mer.ber Glenn A. Reed are presented below.
Sincerely, William Kerr ,
Chaiman Additional Corrents by ACPS Member Glenn A. Reed While I agree with the ACRS letter to reduce the research in described areas, I wish to focus on the panels' observation made as a "first suggestion" in the general conclusions that a prevention technique of "depressurization" (procedures and design) was iirportant "to make the problem go away."
I recomend that research be increased and accelerated on the depres-surization idea and that the research include application of depres-surization as an alternative technique for core decay heat removal.
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. The Henorable Lanao W. Zech, Jr. May 13, 1987
References:
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- 1. U.S. Nuclear Regulatory Commission Staff Document, "Plan to Address )
Scurce Term Technical Uncertainty Areas " SECY-86-369, dated December 11, 1986
- 2. Brookhaven National Laboratory Report, "Review of Research on Uncertainties in Estimates of Source Terms From Severe Accidents in Nuclear Power Plants," NUREG/CR-4883, dated April 1987
- 3. U.S. Nuclear Regulatory Comission Report, "Reassessment of the Technical Bases for Estimating Source Terms," NUREG-0956, Draft Report for Coment, dated July 1985
- 4. U.S. Nuclear Regulatory Comission Report, "Reactor Safety. Study --
An Assessment of Accident Risks in U.S. Comrrercial Nuclear Power Plants,"WASH-1400(NUREG-75/014),datedOctober1975
- 5. National Research Council Report, "Technical Aspects of Hydrogen Control and Combustion in Severe Light-Water Reactor Accidents,"
dated 1987
- 6. U.S. Nuclear Regulatory Comission Report, "A Review of the Current Understanding of the Potential for Containment Failure from In-Vessel Steam Explosions," NUREG-1116, dated June 1985.
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{o r ADVISORY COMMITTEE ON REA? TOR SAFEGUARDS w Asm NGTON, D. C. 20555 g ,4, July 15, 1987 Mr. Victor Stello, Jr.
Executive Director for Operations U.S. Nuclear Regu17 tory Comission Washington, D.C. 20sS5
Dear Mr. Stello:
'SUPJECT: ACRS COMMENTS ON THE EMBRITTLEMENT OF STRUCTURAL STEEL Surveillance samples of steel used in the pressure vessel of the High F'ux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory re-cently have shown that the nil-dgetility transition (NDT) temperature of steel irradiated slowly at 120 F can rise much more rapidly with ex-posure to fast neutrons than would be expected from the available experimental work obtained in test reactors. This appears to be due to two causes:
a flux rate effect (A lower fast neutron flux embrittles more than the same' fluence accumulated at a much higher flux in tesc re-actors.)
the difference in tgmperature (550 F for c'omercial reactor pres-sure. vessels vs. 120 F for the HFIR)
This has led to a significant shift in the NDT of the steel at a fast neutron fluence lower by roughly a factor of 20 than that predicted by the correlations used in the past for low temperature irradiations.
This acceleration is independent of the copper content of the material.
l This suggests that steel structures outside the pressure vessel in
! commercial nuclear power plants may have embrittled where such behavior wn not expected. We believe it would be prudent for the NRC to do the following:
- 1. Detemine if the brittle failure of any structural steel component near the outside of the primary pressure boundary weuld have safety significance.
- 2. Detemine, using the low temperature irradiation data now available from test reactors, whether an increase in the fast neutron fluence by a factor of 10-100 would be predicted to give brittle behavior in these components.
- 3. Implement a research program which would assemble better infoma-tion on the rate of shift of the NDT of structural steels in Attactinent 2
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Mr. Victor Stello, Jr. July 15, 1987 comercial nuclear power plants at these lower rates and tempera-tures.
- 4. Include consideration of the accelerated shift in NDT as part of the evaluation of structures in the program on plant aging.
Sincerely, 1
William Kerr Chairman 4$W' 4
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July 15, 1987 Mr. Victor Stello, Jr.
E Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C. 20555
Dear Mr. Stello:
-E SUBJErT: ACRS COMMENTS ON INTERNATIONAL COOPc. RATION ON RESEARCH RELATED TO RADIATION PROTECTION During the 327th meeting of the ACRS, July 9-11, 1987, it was brought to our attention that members of the NRC Office of Nuclear Regulatory Research -
are attempting to develop a system for the coordination of research being j conducted in various countries on the biological effects and control of imizing rauiation.
These efforts hold promise for assu,ing that key problems are effectively addressed and for reducing unn1cessary d"plication and the wasting of ra-sources. We endorst these effo"ts and ent.aurage their support. '-
Sincerely, y
William Kerr Chairman g
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' . E -E ADVlsORY COMMITTEE ON REACTCR SAFEGUARDS "Q, ,8 C'ASHINGTcN,0, C 20555 July 15,1987 The Honorable Lando W. Zech, Jr.
Chairman U.S. Nuclear Regulatory Comission Washington, D.C3 20555
Dear Chairman Zech:
SUBJECT:
ACRS COMMENTS ON DRAFT NUREG-1150, ' REACTOR RISK REFERENCE DOCUMENT" During the 327th meeting of the ACRS, July 9-11, 1987, we discussed the draf t report NUREG-1150, "Reactor Risk Reference Document," which was issued for corr.. nt in February 1987. The ACRS Subcomittee on Severe Accidents col.sidered this report during meetings on January 29 and May 1,1986 and the ACRS Subcomittees on Severe Accidents and Probabilistic Risk Assessment continued the review on June 3 and July 8, 1987. In our review we had the benefit of discussions with the NRC Staff and its consultants from Sandia National Laboratory (SNL). We also had the benefit of the documents referenced.
NUREG-1150 describes probabilistic risk analyses (PRAs) of several operating nuclear power plants. Results of PRAs for two of these were previously reported in WASH-?.400. The plants analyzed had different containment types and included both PWRs and BWRs. The analyses are said to be ' risk re-baselining"; i.e , the methods used are current, the data used in the analyses include both generic and pl'nt-specific information, the computations make use of codes that have been developed -
since the publication of WASH-1400, and the risk calculations make use of the so-called Source Tenn Code Package (STCP) that includes much of the informatiJn developed by the NRC research program on severe acci-dents (althou published STCP)gh.
that was Containment used wasisatreated perfonnance slightly codified ir, much version more detail of the in NUREG-1150 than it was in WASH-1400.
In addition to calculations of risk attributed to internal . initiators, l
this report describes the results of studies which attempted to predict the uncertainties in tht predictions of a number of relevant quantities, including core melt frequency and the probabilities of early and delayed fatalities.
In assessing public risk, the current version of NUREG-1150 is incom-plete, since external accident initiators are not treated and, based
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. The Honorable Lando W. Zech, Jr. July 15, 1987 on results from other PRAs, these may produce significant contributions to risk. Work has begun on external initiators, and later versions of the report will centain the results.
The report and its supporting documents are voluminous, and the amount of information reported is almost overwhelming. However, we believe the significance cf the results and the anticipated use of the infor1 nation in the regulatory process should be made explicit.
Among the conclusions reported, the following appear to be significant:
(1) The report concludes that, for the plants examined, the risk con-tributors are sufficiently disparate that no general conclusions can ta drawn concerning the risk of plants r.ot examined.
(2) The calculated risk of each plant analyzed was less than the quantitative realth effects objectives in the Safety Goal Policy Statement. However, as mentioned above, the calculated risk did not include contributions from external initiators.
(3) The calculated risks for Surry, Unit 1, and Peach Bottom, Unit 2, were not markedly different fccm those reported in WASH-1400. We were told that a number of risk-reducing improvements had been made for the5e plants since the original analysis, but that these were somewhat offset by newly discovered risk contributors.
One of the original aims of the work reported in NUREG-1150 was to determine if an analysis of thne selected plants would permit con-clusions to be drawn concerning thw risks of other operating plants not analyze 6 So far as we can determine from the report and from dis-cussions with the Staff, their conclusion is that these plants (and other plants that have been the subject of PRAs) are sufficiently different, and the rish contributors are sufficiently diverse, that little can be learned about one plant from the analysis of another plant, even when they are of the same general type.
This conclusion is both surprising and disturbing. If correct, it raises serious doubts about the breadth of application of these efforts.
The Staff has not provided convincing reasons for this conclusion. More effor+. is needed to determine why this conclusion should be accepted, because such a conclusion would have far-reaching consequences for several Comission policies.
We have the following additional comments:
(1) We are skeptical of the method by which' expert opinion was used in predicting uncertainties. Explanation of and justification for the
The Honorable Lando W. Zech, Jr. July 15, 1987 method are obscure. There is also reason to believe that the way in which the method is used can have a significant influence on the uncertainties prM4 3d. It is thus almost impossible to interpret the significana of the aported uncertainties or to subject them to peer review. l (2) Many of the codes used in ths calculation are relatively new. The validity of several of the codes is not well established. Further-more, many of then, have not been published and are not yet avail-able to people outside the national laboratories. Serious peer review of the results reported is thus almost impossible.
(3) It was emphasized by the Staff that a major contribution of the report was the "insights" provided. We recomend that these insights be better identified and that their significance for those who are not pRA practitioners be mode more clear.
(4) Human performance contributions to risk (both positive and nega-tive) are not well described by pRAs. This report dues not correct that deficiency.
(5) We were told by the Staf f that, in light of insights developed during the work reported, resolutions or proposed resolutions of a number of Unresolved Safety Issued are to be revisited. We recom-mend that, as an aid to understanding the report, these instances be identified. We recomend also that the interaction between those. respor.sible for the resolution of Safety Issues and those responsible for this report be improved.
Une might conclude, both from the report and from coments made by the Staff, that the NRC regulations are inadequate to determine plant equipment and procedures necessary to protect public health and safety.
If this is the Staff's conclusion, it is a dramatic find %g and should be emphasized, and the position developed mora effectively than it is in the present draft. If, however, regulations can t,e used at a mechanism to protect public health and sr.fety, cnd we believe they can, we re:om-menc that the Office of Nuclear Reactor Regulation begin early examina-tion of this report, both to apply its insights and to guide its further development.
Sincerely.
( L William Kerr Cnairinan i
,' The Honorable Lando W. Zech, Jr. July 15, 1987
References:
- 1. U.S. Nuclear Regulatory Commission, NUREG-1150, Reactor Risk Reference Document, Volumes 1, 2 and 3. Draft issued for comment.
dated February 1987
- 2. U.S. Nuclear Regulatory Commission, NURFG-75/104, "Reactor Safety Study - An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," dated October 1975 (formerly issued as AEC report WASH-1400).
- 3. U.S. Nuclear Regulatory Comission, NUREG/CR-4587, "Source Tenn Code Package: A User's Guide (MOD 1)," Battelle Columbus Labora-tory, dated July 1986.
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