ML20033E996

ACRS Committee & Consultants Repts - 174

ML20033E996
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Issue date:	02/28/1990
From:	Advisory Committee on Reactor Safeguards
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ACRS-GENERAL, NUDOCS 9003150236
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SUMMARY

LETTER 02/28/90 II) ACRS REPORTS TO CHAIRMAN, NRC 1. Michelson ltr to Carr, ACRS Review and Evaluation 02/15/90 of Nuclear Power Plant Operating Experience 2. Michelson ltr to Carr, Proposed Power Level 02/15/90 Increase for Indian Point Nuclear Generating Station Unit 2 3. Lewis ltr to Carr, Coherence in the Regulatory 02/15/90 Process III) ACRS REPORTS TO OTHERS 1. Michelson 1trs to Quayle SENATE and Foley USHR 02/15/90 re Report to Congress on the Safety Research Program'of the Nuclear Regulatory Conunission r IV) ACRS CONSULTANTS REPORTS 1. CT-1949A Page ltr to Siess re NRC/PG8E Meeting 08/11/89 on Diablo Canyon held 8/8-10/89 2. CT-1950 Bender ltr to Igne re Comments on the 08/22/89 .Seabrook Subete Meeting held 8/17/89 .3. CT-1951 Thompson ltr to Siess re NRC/PGLE Meeting 09/06/89 on Seismic Source Characterization for Diablo Canyon held-8/8-10/89 4 4. CT-1952 Stevenson 1tr to Siess tsmtg Paper presented 10/10/89 at American Institute of Professional Geologist, "New Nuclear Power? - When and If" 5. CT-1953 Lee ltr to Boehnert re Thermal Hydraulic 11/13/89 Phenomena Subete Meeting on 11/8-9/89 4 i

F - l '. g o" i C0NTENTS (cont'd) t. DATE IV) ACRS CONSULTANTS REPORTS (cont'd) 6. CT-1954 Bender ltr to Alderman re Comments on 11/17/89 Nine Mile Point Unit 1 Restart Subete sl' Meeting held 11/14/89 7. CT-1955 Schrock ltr to Catton/Boehnert re Thermal 11/17/89 Hydraulic Phenomena Subete Meeting held 11/8-9/89 8. CT-1956 Tien 1tr to Boehnert re Thermal Hydraulic 11/?2/89 Phenomena Subete Meetin9 held 11/8-9/89 9. CT-1957 Davis memo to Kerr re Interfacing Systems 12/15/89 LOCA and ORNL Precursor Study 9 r l ii _J

. 'pem maeu 8 'o UNITED STATES ) o ~,, I NUCLEAR REGULATORY COMMISSION g { p ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o wAsHINoTow,p.c. m es February 28, 1990 The Honorable Kenneth M. Carr l Chairman i U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dear Chairman Carr

SUMMARY

REPORT - THREE HUNDRED FIFTY-EIGHTH MEETING OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS, FEBRUARY 8-10, 1990 During its 358th meeting, February 8-10,

1990, the Advisory Committee on Reactor Safeguards discussed several matters and completed the reports and memorand noted below.

Copies of these reports and memoranda have already been provided to you. A summary of the activities of the Committee during this meeting is given below. FJitPRIt to the Conar3ss on the NRC Safety Research PIqgra (Report to Vice President Danforth Quayle, and Thomas S. Foley, Speaker of the House, dated February 15, 1900). Reoorts to_th;;pmmissilon Asocint109.at_RL.,3C33 Members (Report to Chairman Carr, dated e February 14, 1990). ACRS Review Gnd Evaluation,sf Nuclear Power _ P1pnt OneIfitAng fXI2Arience (Report tc Chairman Carr, dated February 15, 1990). pohereqqp in the Reaulatory Procgga (Report to Chairmr.n Carr, e dated February 15, 1990). o Pronosed Power Level Increase for Indian Point Nuclear I Generatina Station Unit 2 (Report to Chairman Carr, dated February 15, 1990). Memorandum to the Commission Division of Resconsibilities Between ACRS and ACNW (Memorandum e from R. F. Fraley for Chairman Carr, dated February 23, 1990). Consistent with the Commission's guidance, the ACRS and ACNW have arranged for an interface in areas of mutual interest and concern, particularly decommissioning of nuclear facilities. Mr. Fraley's memorandum describes this arrangement.

g, l 3., j-The. Honorable Kenneth M..Carr 2 February 28, 1990 Memorandum to the EDO Recuest for Postoonement of Comoletion of the Seismic Marcins e Evaluation Procram for the Perry Nuclear Power Plant (Memorandum from R. F. Fraley for J. M. Taylor, EDO, dated February 14, 1990). Consistent with the Committee's decision, Mr. Fraley has informed Mr. Taylor that the Committee has no objection to the Cleveland Electric Illuminating Company's proposal to conduct. -its Seismic Margins Evaluation Program within the context of resolving the severe accident issues under the Individual i Plant Examination for External Events- (IPEEE) Program. j Other Matters Considered by the Committee Meetina with the Defense Nuclear Facilities Safety Board e The. Committee met with the members of the Defense Nuclear Facilities Safety Board, at the request of the Board, and exchanged views with regard to providing technical assistance to the Board in its review of certain matters. In accordance with the provisions of the Public Law 100-456 of. September 29, 1988, " National Defense Authorization Act," which states that the Board may obtain the advice and recommendations of the ' ACRS on matters relating to the responsibility of the Board, the Board members proposed the following: The Board would-like to gain access to the ACRS reports, L files, and other information related to the previous ACRS l review of~the DOE's defense nuclear facilities. ~ The Board would like to get a list of those consultants, along with their areas of expertise, who assisted the ACRS in the 'past in its review of the DOE's defense nuclear. facilities so as to enable the Board to get assistance from these consultants, as necessary, in-its review of certain matters. 1 L In the time of need, the Board will look to the ACRS'as L-one of the most-prominent places to get technical assistance. The Board would like to work with the ACRS to develop a mechanism for use in obtaining technical assistance either from the ACRS full Committee, its Subcommittees, or fron. individual ACRS members.

' m, e .n. ,e .The Honorable Kenneth M. Carr 3 february 28, 1990 The Committee informed the Board members that: ACRS reports related to its review of DOE's defense nuclear facilities are being reviewed for declassifi-cation. Declassification would make them more readily available. Upon completion of this effort, the Board may obtain necessary information as needed.. The Committee will work with the Board as a collegial body rather than-individual experts, o Meetina with the AEOD Director Mr. Jordan, Director of the AEOD, presented to the Committee several items related to the activities of the AEOD, including Diagnostic Evaluation Program, Incident Investigation Program, bases for forming AIT and IIT for investigation of operating events, Emergency Response Data System, etc. ? e Meetina with the Reactor Safety Committee of the Federal Reoublic of Germany (FRG) 1 The Committee agreed to a proposed meeting with the Reactor Safety Committee of the FRG during June 1990 to exchange views i and information regarding issues related to LWRs and Modular High Temperature Gas Cooled Reactors. This meeting will be 1 held in the FRG and will include a visit to the nuclear power _ i station at Phillipsburg,.FRG. l' e Meetina with~Jacanese Reoresentatives E The Committee tentatively agreed to meet with representatives y of Hittachi, Toshiba, Tokyo Electric, and MITTI to discuss items of mutual interest, with emphasis on ABWR issues. 'This L meeting is-tentatively scheduled to=be held in Japan-during I the-fall of 1990. e-Prooosed Final Resolution of Generic Issue B-56. " Diesel ~ Generator hgliability." and Associated Reculatorv Guide 1.9. [ E.evision 3 l' The Committee heard presentations by and held discussions with l representatives of the NRC staff and - NUMARC regarding this i matter. Differing views in several areas were presented to the-Committee by representatives of the NRC staff and NUMARC.

The Honorable Kenneth M. Carr 4 February 28, 1990 Since Revision 3 to Regulatory Guide 1.9 endorses, with certain exceptions, the provisions of Appendix D to the NUMARC-8700 document it may be revised to

reflect, as 7

appropriate, the final version of the Appendix D that is expected-to be completed in the beginning of March 1990, the Committee decided to reconsider this matter after the final. version of Appendix D to NUMARC-8700 and the revised Regulatory Guide 1.9 are made available to the Committee. J e Dr. Catton's Visit to Northeast Utilities Dr. Catton briefed the Committee regarding his visit to j' several of the Northeast Utilities plants (Haddam Neck, and Millstone Units 1-3). e Evolutionary Licht-Water Reactor certification Issues The NRC staff briefed the Committee regarding the evolutionary light-water reactor certification issues and their relationship to current regulatory requirements. The staff has identified 15 significant issues in SECY-90-016," Evolutionary Light Water Reactor (LWR) Certification Issues and their Relationship to Current Regulatory Requirements," as fundamental to agency decisions on the acceptability of evolutionary ALWR designs. The Committee assigned these 15 issues to various Sub-committee chairmen and -individual members for review and comment. The Committee plans to continue discussion of this matter during the March 8-10, 1990 meeting. 1 i Subcommittee Meetina i Since the last summary report of ACRS activities, the following Subcommittee Meeting has been held: Safety Research Procram, FG rGary 7, 1990 - The Subcommittee e met with the EDO and representatives from RES and NRR and discussed the ongoing and proposed NRC Safety Research Program and Budget, the impact of the budget reductions on the NRC Safety Research Program, and other related matters. Schedule for the 359th Meetina The committee agreed-to the following tentative schedule for the 359th ACRS meeting, March 8-10, 1990: e Reactor Ooeratina Experience (Ocen/ Closed) Briefing and discussion of recent events at operating nuclear power plants.

. +, . ('.- g The Honorable Kenneth M. Carr 5 February 28, 1990 o AEOD Reoorts/ Activities (Ocen) Briefing and discussion regarding an overview of the NRC AEOD reports and=related activities, o Evolutionary Licht-Water Reactors Certification Issues (Ocen) Reports by and discussion of cognizant ACRS members and subcommittee chairmen regarding issues identified by the NRC staff in SECY-90-016, Evolutionary Light-Water Reactors (LWR) Certification Issues and their Relationship to Current Regulatory Requirements. e Containment Performance Imorovement Procram (Ocen) - Review and comment regarding the proposed NRC containment performance improvement program for all LWR containment-types except the BWR Mark I type containment. o ACRS Subcommittee Activities (Ocen) - Hear and discuss reports regarding the status of cognizant subcommittee. activities regarding safety related matters including, performance. of valves in nuclear power plants. o Occucational Radiation Excosure to Skin from Hot Particles l Briefing and discussion of proposed NRC inte ^' (Ocen) action regarding this matter. (tentative) e NRC Safety Research Procram (Ocen) Discuss impa?. budgeting. policies and practices regarding. the NRC se'"ty research program. e Accointment of ACRS Members (Ocen/ Closed) - Discuss the status of appointment and qualification of candidates proposed for appointment to the ACRS. Sincerely, 1 Abh b% 7 c Carlyle Michelson Chairman L e w-- m a-

_ ~.. _ - - 'k 3 3 1 utg( -} e.. UNITED STATES 4. 8- . NUCLEAR REGULATORY COMMISSION - i c- ,,H o K .i ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS l 0 g. WASHINGTON, D. C. 20666 d

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February 15, 1990 6 The Honorable Kenneth M. Carr Chairman-U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS REVIEW AND EVALUATION OF NUCLLAR POWER PLANT OPERATING EXPERIENCE In your letter dated January 12, 1990, you. requested a report discussing ACRS reevaluation of the Committee's role in the review 1 and evaluation of nuclear power plant operating experience. During our January 11-12, 1990 meeting, we reviewed the existing ACRS subcommittee structure with particular emphasis on identifying issues requiring increased ACRS involvement..Several new subcom-mittees.were established and the scopes of some existing subcommit-tees were modified to reflect the Committee's perception of its

future activities.

t Attached is'a list of the current and - recently. disbanded' ACRS subcommittees. You will note that many of the subcommittees'are involved with some facet of plant operating experience as part of their. charters. Five subcommittees have specific assignments for-review of operating experience: A Plant Operations Subcommittee was established to act as the o lead subcommittee in the area of plant operating experience. This subcommittee will review selected nuclear power.. plant operating events and the short-term actions associated with these events, including interfacing with the NRR Division'of Operational Events Assessment.- The subcommittee will con- . sider, as appropriate, activities such as - the LER, SALP, Diagnostic Evaluation Team, AIT, and IIT processes; opera-tional quality; operator performance; and Technical Specifica- . tion-issues.

2 The' Honorable Kenneth M. Carr 2 February 15, 1990 o A Systematic Assessment of Experience Subcommittee has the responsibility for the long-term and generic implications of operating experience, including interfacing with AEOD and INPO on activities regarding long-range assessment of operating experience, reviewing lessons learned from AIT and IIT reports, comparing PRA input /results to actual operating ex-

perience, and reviewing the development of operational performance indicators.

o An Adopted Plants Activities Subcommittee was established to handle procedural and resource issues and member assignments associated with. the ACRS adopted plants program. ' This program, which was established in 1987, involves the assign-ment of a number of operating plants to each ACRS member. The member receives licensing and inspection correspondence and SALP reports on his plants, visits the plants as appropriate, and reports items of interest to the ACRS. o An International Activities subcommittee was established to coordinate all activities on foreign reactors and interface with AEOD and GPA on international activities. Subcommittee. responsibilities include the review and evaluation of foreign nuclear power plant operating experience. P The Regional ~ Activities Subcommittee meets periodically with o each Regional Administrator and his staff to review-the various inspection and enforcement programs in the Regions. l A complete round of these meetings has taken place and a u second round will commence-this year. In addition to these subcommittee activities, the ACRS staff L receives and reviews Weekly Information Reports; NRC bulletins, notices, and generic letters; AEOD reports; performance indicator . reports; preliminary. notifications; the monthly compilation of LERs; and Headquarters and Regional Office daily reports. Com- . mittee members receive certain of this information depending on their interests. This information provides input to the Commit- } tee's process for establishing subcommittee work assignments and requesting staff briefings on operating experience issues. 1

' The Honorable Kenneth M. Carr 3 February 15, 1990 In summary, we believe that ACRS has the appropriate mechanisms in - place to review and evaluate U.S. and foreign nuclear plant oper- . ating experience and to-advise the Commission in this important-area. Sincerely, 4 Iw r_ Carlyle_Michelson Chairman

Attachment:

List of Current and Recently Disbanded ACRS Subcommittees s 's u

3 F'I ATTACHMENT LIST OF CURRENT AND RECENTLY DISBANDED ACRS SUBCOMMITTEES NEW SUBCOMMITTEES Adopted Plants Activities Defueling/ Fuel Pool Storage International Activities Plant License Renewal Plant Operations TVA Plant Licensing / Restart PREVIOUSLY ESTABLISHED-SUBCOMMITTEES Comanche' Peak Units 1 & 2 Computers in Nuclear Power Plant Operations Seabrook Nuclear Plant Unit 1 AC/DC Power. Systems Reliability Advanced BWRs (GE) Advanced PWRs (W & CE) Advanced Reactor. Designs l: ~ Auxiliary & Secondary Systems i Babcock &.Wilcox Reactor Plants Containment Systems l. Core Performance y Decay Heat Removal Systems Extreme External Phenomena - FTOL Conversions General Electric Reactor Plants Generic Items Human Factors Improved LWRs Instrumentation and Control Systems Integrated Safety Assessment Program Maintenance Practices and Procedures Materials & Metallurgy Mechanical components Naval Reactors Non-Power Reactors Occupational and Environmental Protection Systems Planning and Procedures Probabilistic Risk Assessment l Quality and Quality Assurance in Design and Construction Regional Programs Regulatory Activities Regulatory Policies and Practices Reliability Assurance Safeguards and Security Safety Philosophy, Technology and Criteria --.-- ?---

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..l

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e..,

lJ s - 2 Safety Research Program Severe Accidents Structural Engineering j Systematic Assessment of Experience Thermal ~ Hydraulic Phenomena l Westinghouse and Combustion Engineering Reactor Plants ?

SUBCOMMITTEES WHICH HAVE RECENTLY BEEN DISBANDED (Tasks have been absorbed into either new or other previously established subcommittees).

L Bellefonte Plant Units 1 & 2 l Limerick Unit 2 ) L Watts Bar Units 1 &2 l ACRS Bylaws L Consideration of International Operating Experiences i Member Nominations On-Site Fuel Storage Plant Operating Procedures l t l l

x l' "

• %'o UNITED STATES NUCLEAR REGULATORY COMMISSION i

-{' ADVISORY COMMITTEE ON REACTOR SAFEGUARDS l A 0 WASHINGTON, D. C. 20666 February 15, 1990 1 The Honorable Kenneth M. Carr L Chairman .] l-

U.S. Nuclear Regulatory Commission L

Washington, D.C.'20555 l-l

SUBJECT:

- PROPOSED POWER LEVEL INCREASE FOR INDIAN POINT NUCLEAR GENERATING STATION UNIT 2 l During the 358th meeting of the Advisory Committee on Reactor I-Safeguards, February 8-10, 1990, we reviewed the application of L Consolidated Edison Company of New. York -(Licensee) for a license amendment, to permit ' it to operate the Indian Point Nuclear Generating. Station Unit 2 at a core thermal power level up to 3 071. '4 ' MWt. The current' core power level limit is 2758 MWt, so E this is approximately an 11 percent increase.' This matter was i discussed by our Subcommittee on the Systematic Assessment of Experience, on February 6,1990. During these meetings, we had the benefit ~of discussions;with representatives of both the NRC staff ll and the Licensee.- We also had the' benefit of the documents L referenced. The NRC staff recommends approval of this application. [ The plant was originally licensed in 1973, at a core thermal power l level up-to 2758 MWt, though the original' analyses and supporting-p environmental assessments, with the exception of the emergency core cooling system.(ECCS), were'made for.a core thermal power level of 3216'MWt. The ECCS was. evaluated <at 2758-MWt. There is nothing in the history to suggest that the lower' power level of the original license was based on anything other than a-(commendable) caution, since this was the first of the.large' Westinghouse 4-loop plants to seek a license. Since this is a license amendment, the' ~ staff review is based on the original license requirements, and our review-is confined to the implications of the proposed power level increase, not to a review of the original license decision. Since nearly all the original analyses were performed at the higher ' power,-the remaining need was to demonstrate ECCS operability at -the proposed' power, and.this was done in May of 1989. The analyses were reviewed by the NRC staff, and found to be in compliance with

\\ The Honorable Kenneth M. Carr 2 February 15, 1990 l 10 CFRL50.46 and Appendix K, with suitable conservatism. We have 1 -no reason to question these conclusions. 1 Many new requirements, not all due to the Three Mile Island accident, have been levied since the original license was issued in 1973. Some of these are. power related, and the staff should assure itself that those will be met at the new power level. The Licensee has assured the NRC, in a letter dated February 8, 199.0, that that is the case.- Subject to resolution of this matter to the satisf action of the NRC staff, we believe that the Indian Point Nuclear Generating Station Unit 2 can be operated at core power levels up to 3071.4 MWt without undue risk to the health and safety of the pilblic. Sincerely, Yb ab$u y, Carlyle Michelson Chairman

References:

1. Memorandum dated January 29, 1990 from S. A. Varga, Nuclear Regulatory Commission, to R. F.

Fraley, ACRS,

Subject:

. Transmittal of Revision to Draft Safety Evaluation to Increase Licensed Thermal Power Level of the Indian Point Nuclear Generating Unit No. 2 2. Letter dated September 30, 1988 from S. Bram, Consolidated Edison Company to U. S. Nuclear Regulatory Commission-transmitting. Application for Amendment to Operating License (Indian Point Station Unit 2) 3. Letter dated February 8, 1990 from S. Bram, Consolidated Edison Company to Donald S. Brinkman,~ Nuclear Regulatory Commission,

Subject:

Application for License-Amendment to. Increase Authorized Power Level

r,c 3 a o.sttoo ~8 ~ NUCLEAR REGULATORY COMMISSION -n

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E-ADVISORY COMMITTEE ON REACTOR SAFEGUARDS / - WASHINGTON, D. C. 20665 L o February 15, 1990 ) b 1 t The Honorable Kenneth M. Carr Chairman U.S.~ Nuclear Regulatory Commission 7 Washington, D.C. 20555

SUBJECT:

COHERENCE IN THE REGULATORY PROCESS In.our reports to you of November 24, 1989 (which also lists our-earlier reports) and December 21, 1989, we have discussed a variety. of aspects of the coherence problem -- the problem of-assuring that all. elements:of the NRC pull in..the same direction in the regula-tion? of nuclear ' power, a direction provided by the Commission itself. These ' reports have generally dealt with ' symptoms of. . incoherence:-- the most recent was about the internal use of SALP? ratings. Here we would like to take a more global view of 'the coherence. problem, leading in the end to a recommendation'for a-next step. It.is almost as-if the NRC were created to be. incoherent. There are five Commissioners and five statutory Offices. There are many . Branches and five Regional-Offices, with a kind of matrix manage-- ment' tieing--it all together. Regulatory-power is spread through-out, resulting in a melange. of technical positions, regulatory . guides, generic letters, policy statements, undocumented pressures, enforcement actions, etc.. The mechanisms for providing incentive .to the various elements of the-staff,to test-their actions in the light: of Commission objectives are inadequate.- Indeed - those objectives:are not-always easy to determine, for reasons that need no elaboration here. This is not to say that anyone-is deliber-ately misbehaving, only that too many are free to proceed in the light of their own best judgment. We-have long. argued that the best way to test the effectiveness of the regulatory process is to ma.asure the results in terms of the Commission's Safety Goals, and we do not depart from that position here, but a performance measure is not a coherence measure. The latter-has to do with efficiency, clarity, and ' ultimately, acceptability of the process. w y w -

The Honorable'Kenneth-M. Carr 2 February 15, 1990 In our November 24 report on this subject we emphasized that the coherence problem can be divided into many categories -- it is not a neat subject. The Commission itself can and should make its policy statements and other issuances as unambiguous as possible (we know that is not easy; we often fail ourselves), so_as to minimize opportunities for misinterpretation. Also, as mentioned in that report, many of the examples lie within the province of the EDO, and he should be aware of his responsibility to keep the various offices working toward the same ends. Perhaps his own staff needs expansion. But the real tests of coherence lie in the NRC's interactions with the outside world, and we doubt that only internal modifications can solve these problems, although we believe improvements could be made. We are not prepared to recommend reorganization of the NRC, though that is one of the options available to you. Certainly, incentives for lateral communication would be helpful. We do not believe coherence can be proclaimed from above. Not only is the effect of proclamations attenuated as they penetrate any organization, but high-level policies are necessarily imprecise. Not all ramifications or interpretations of a policy statement can be foreseen, and coherent policies have to be molded in use.- It .is,the body of regulatory practice that is in question here, much of it in the form of corporate memory and lore, and the job at each e level is to provide sufficient guidance and incentive to make it possible (and desirable) for the next level to function consistent-ly with the global policies. Above all, the governing policy guidance must be simple, clear, and understandable to both regula-tor and regulatee. How is coherence approached elsewhere? One necessary ingredient appears to be feedback, through which interpretations of policy are constantly tested against the policies themselves, not in every ' case but through a sampling process that, in the end, leads.to a m o r e ' c o h'e r e n t structure. The guiding law of the land is the Constitution, embodying our' principles of government. The real law of the land,

however, is the enormous body of case law generated by innumerable court decisions, each reviewable, and some in fact reviewed, by the next level of appellate court.

Thus the regulatees, in this case the population, have a set of recourses that can bring any rule or ruling to a test of its coherence with the guiding principles. Further, and most important, those who do the testing are not those who make the rules, so there is at least the perception that there is a genuinely unbiased feedback process. The founders were careful to include this in the system. In addition, feedback loops need not be end-to-end; intermediate loops are also helpful. .~

~~ si. i i ,e The Honorable Kenneth M. Carr 3 February 15, 1990 There are many examples of this process in other areas. A taxpayer ) who feels mistreated by the Internal Revenue Service can appeal 1 within the system, but can in the end go to the Tax Court, an j entirely independent forum. A pilot denied his or her license by i the Federal. Aviation Administration has the right to appeal to the National Transportation Safety Board, an independent agency, whose ruling is final. In each of these there is some risk, but the constant feedback provided by external review helps to create a body of case law that is under continuous testing for coherence. This is'not true in the nuclear business, where the only external review is in the courts, and their primary mission is not coherence in the regulatory process. The only appeal from a Regional decision (for example) is within the system, and we all learn early that it is unwise to complain about someone who has power over you, unless you're sure you'll win. All engineers recognize that complex systems are better controlled by feedback than by blind input one measures the errors and corrects the input accordingly. The key is the ability to make objective measurements through a separate sensing system. What appears to be needed in our case is a mechanism through which' frequent. testing of-the body of " case law" against-the guiding principles laid down by the Commission is made possible. To be credible and effective, that job cannot be assigned entirely to the Commission staff. The current situation is analogous to one in which there is a constitution (commission policies), a body of law L (letters, guides, enforcement actions, rules),-but no courts. In general, those with the most to gain by coherent regulation are L the regulatees-(and of course the rest of us, because safety will benefit), and they would be in a better position to seek coherence if they could do so without fear of retaliation. It is the fear-of--being taken to court that serves to constrain police forces -- the constraints in our case are entirely-internal. I This kind of feedback solution has been. used in many places. Governments and police forces have courts; factories have grievance committees; some agencies have ombudsmen for employee complaints, though these usually have no power. The NRC has nothing com-parable. t We-believe the ultimate solution to the coherence question must include the provision of an adequate feedback mechanism. To be sure, you have made any number of commendable requests to the regulated community to come forward with complaints, but less has come of it than might have been hoped. Even if more had happened, l this would still have been symptomatic treatment of the problem,

j: The' Honorable Kenneth M. Carr 4 February _ 15, 1990 and we believe that a mechanism in place is required. Some of us believe that, in the end, only an external Nuclear Safety Board can help, while others believe that great-strides can be made within the NRC itself. However, just as we are not prepared to recommend reorganization of the NRC, we do not suggest what form the feedback mechanism should take. We do recommend that possible means for achieving the objects stated above be explored, and doubt that it would be wise to simply ask the staff (or us) to do the job. We think it would be entirely appropriate, given the importance of the issues, to take a major initiative by asking some respected outside ' group to explore the subject, and to lay out the feedback options available to the country, even if they require legislation. Such a study group could be chartered by the NRC -- there are precedents -- and should include representation from the affected industry. The National Academy of Sciences has done such things, or it could be an_ entirely free-standing operation. The result should not be a specific recommendation, but a list of options and analyses, which could then be freely debated within the interested community. This is a complex subject, and we do not think it should be resolved by hip shot. We also do not think it should be neglected, since the effectiveness of the regulatory process is at-issue. Additional comments by ACRS Members Carlyle Michelson, Chester P. Siess, and Charles J. Wylie are presented below.- Sincerely, x Harold W. Lewis Acting Chairman Additional comments by ACRS Members Carlvle Michelson, Chester P. Siess, and Charles J. Wylie If there is a problem with coherence in the regulatory process, we-do not believe that it has been identified and characterized in this report with' sufficient clarity to support a recommendation that the NRC charge some outside group to explore it. We agree that there have been examples of inadequate integration of regula-tory staff activities, sometimes serious, but it should not require an outside panel to tell management how to correct such deficien-cies. If the ACRS believes that there is a coherence problem beyond the capability of the Commission to highlight and correct,

4. .) 4 The Honorable Kenneth M. Carr 5 February 15, 1990 then it should clearly articulate the problem before suggesting that the ultimate solution must include provisions for an adequate feedback mechanism and asking some outside group to lay out the feedback options. There are other portions of this letter to which we would take exception; but unless the ACRS can define the problem that needs to be fixed, they may not be worthy of mention. It is our observation that the agency knows its responsibilities and has been successful in carrying out its mission. l w w- -w

--aof*;p'ase( j l f l- 'cp . UNITED STATES ,, ( i- . NUCLEAR REGULATORY COMMISSION i

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, c,i ; ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g f. WASHINGTON, D. C. 20666 j g,, ,g J February 15, 1990 t i l. The Honorable J. Danforth Quayle President of the United States Senate Washington,.D.C. 20510

Dear Mr. President:

~ 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 Committee on Reactor Safeguards has reported each year to the - Congress on the Safety Research Program of the Nuclear Regulatory Commission. In our December 18, 1986 letter to the Congress, we proposed to provide more focused reports on specific research issues rather than one all-inclusive annual report. The Commission agreed with our suggestion, and-then NRC' Chairman Zech submitted a legislative-proposal' in the form of a draft bill to the 100th Congress. on December 2, 1987 to amend Section-29 of the Atomic Energy Act of 1954..to. accomplish this. Since the 100th-Congress did not act on this matter, he submitted. a similar, but somewhat modified, legislative proposal on February 2, 1989 to the 101st Congress for consideration. We expect that the Congress.will consider this matter-during.this year. In the past year we have reviewed the NRC safety rsearch program t and other closely related matters in the following areas: -' Accident Management Strategies

• Application of Leak-Before-Break Technology:
• Containment Performance
• Containment. Structural Integrity
• Embrittlement'of Reactor Pressure Vessel Supports
• Fire Risk Scoping' Study
• Human Factors Research Program Plan I

F v

.c The Honorable J. Danforth Quayle February 15, 1990 v '" Inservice Inspection of Boiling Water Reactor Pressure Vessels i

• Occupational Radiation Exposure to Skin from Hot Particles

'-Piping Integrity

• Severe Accident Research Program Plan
• Thermal-Hydraulic Phenomena.

We have provided reports to the Commission on several of the matters mentioned cbove'and copies of these reports are attached. We expect to continue to review various elements of the NRC Safety Research Program and provide reports to the Commission as war-ranted. ~ ' Sincerely, faW L m Carlyle Michelson Chairman Attachments: 1. Report from Forrest J. Remick, ACRS Chairman,-to Lando W. i Zech,-U.S. NRC Chairman,

Subject:

Additional Applications of Leak-Before-Break Technology, March 14, 1989 2.- Report from Forrest J. Remick, ACRS Chairman, to Lando W.

Zech, U.S. NRC Chairman,

Subject:

Proposed Severe Accident Research Program Plan, March 15, 1989 3. Report. from Forrest J. Remick, ACRS Chairman, to Lando W.

Zech, U.S.

NRC Chairman,

Subject:

NRC's Human Factors Programs and Initiatives, May 9, 1989 4.. Report =from Forrest J. Remick, ACRS Chairman, to Lando W.

Zech, U.S.

NRC Chairman,

Subject:

NUREG-1150, " Severe Accident-Risks: An Assessment for Five U.S. Nuclear Power Plants," May 9, 1989 5. Report from Forrest J. Remick, ACRS Chairman, to Lando W. Zech, U.S. NRC Chairman,

Subject:

Generic Letter Related to occupational Radiation Exposure of Skin from Hot Particles, May 9, 1989

_~ 9

• s..

n The Honorable J. Danforth Quayle. February 15, 1990 6. Report from David A. Ward, Acting ACRS Chairman, to Lando W.

Zech, U.S.

NRC Chairman,

Subject:

NRC Thermal-Hydraulic Research Program, June 15, 1989 7-Report from Forrest J. Remick, ACRS Chairman, to Kenneth M.-

Carr, U.S.

NRC Chairman,

Subject:

Proposed Staff Actions Regarding the Fire Risk Scoping Study (NUREG/CR-5088), July. 18, 1989 8. Report from Forrest J. Remick, ACRS Chairman, to Kenneth M.

Carr, U.S.

NRC Chairman,

Subject:

Draf t Supplement 2 to Generic Letter 88-20, " Accident Management Strategies for Consideration in the Individual Plant Examination Process," November 20, 1989 l l l. l-l- 1. l l'

c.- 't, }.~.?4

• tog'of UNITED $TATES

47

/ f J'c 'n NUCLEAR REGULATORY COMMISSION ' ,1< J - i-ADVISORY COMMITTEE ON REACTOR SAFEGUARDS k%....f,/ j WASHINGTON, D. C. 20666 February 15, 1993 The Honorable Thomas S. Foley 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 Committee on Reactor Safeguards has reported each year to the. Congress on the Safety ' Research Program of the Nuclear Regulatory Commission. In our December 18, 1986 letter to the Congress, we proposed-to provide' more focused reports'on specific research issues rather than one all-inclusive annual report..The Commission agreed with our suggestion, and then NRC Chairman'Zech submitted.a legislative , proposal 'in the form of a draft bill to the 100th Congress. on December 2, 1987 to amend Section 29 of the Atomic Energy Act of -1954Lto-accomplish this. Since the 100th Congress did not act on this matter, he submitted a similar, but somewhat-modified, legislative proposal'on February 2, 1989 to-the 101st Congress for consideration. We expect that=the Congress will consider this matter during this year. In the past year we have reviewed the-NRC safety research program and other closely related matters in the following areas: --* Accident Management Strategies

• Application of' Leak-Before-Break Technology
• Containment. Performance
• Containment Structural Integrity
• Embrittlement of Reactor Pressure Vessel Supports
• Fire Risk Scoping Study
• Human Factors Research Program Plan

l'. The Honorable Thomas S. Foley February 15, 1990 b

• Inservice Inspection of Boiling Water Reactor Pressure Vessels
• Occupational Radiation Exposure to Skin from Hot Particles
• Piping Integrity
• Severe Accident Research Program Plan.
• Thermal-Hydraulic Phenomena.

We have provided reports to the Commission on several of the matters mentioned above and copies of these reports are attached. We expect to continue to review various elements of the NRC Safety Research Program and provide reports to the Commission as war-ranted. Sincerely, ^bw e Carlyle Michelson Chairman Attachments: 1. Report from Forrest J. Remick, ACRS Chairman, to Lando W. Zech, U.S. NRC Chairman,

Subject:

Additional Applications of Leak-Before-Break Technology, March 14, 1989 2. Report from Forrest J. Remick, ACRS Chairman, to Lando W. Zech, U.S.-NRC Chairman,

Subject:

Proposed Severe Accident Research Program Plan, March 15, 1989 3. Report from Forrest J. Remick, ACRS Chairman, to Lando W.

Zech, U.S.

NRC Chairman,

Subject:

NRC's Human Factors Programs and Initiatives, May 9, 1989 4. Report from Forrest J. Remick, ACRS Chairman, to Lando W. l

Zech, U.S.

NRC

Chairman,

Subject:

NUREG-1150, " Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants," May 9, 1989 5. Report from Forrest J. Remick, ACRS Chairman, to Lando W. Zech, U.S. NRC Chairman,

Subject:

Generic Letter Related to l'. Occupational' Radiation Exposure of Skin from Hot Particles, May 9, 1989 L 1 1

-.ef, - .o The Honorable Thomas S. Foley February 15, 1990 6. Report from David . Ward, Acting ACRS Chairman, to Lando W.

Zech, U.S.

NRC Chairman,

Subject:

NRC Thermal-Hfdraulic Research Program, June 15, 3989 7. Report from Forrest J. Remick, ACRS Chairman, to Kenneth M.

Carr, U.S.

NRC Chairman,

Subject:

Proposed Staff Actions 'Regarding the Fire Risk Scoping Study-(NUREG/CR-5088), July 18, 1989 8. Report from Forrest J. Remick, ACRS Chairman, to Kenneth M.

Carr, U.S.

NRC Chairman,

Subject:

Draft Supplement 2 to Generic Letter 88-20, " Accident Management Strategies for Consideration in the Individual Plant Examination Process,"- November 20, 1989

a .i .v. UNITED STATES o NUCLEAR REGULATORY COMMISSION g ADVlsORY COMMITTEE ON REACTOR SAFEGUARDS wAsmwoTow.o.c. mss t, March 14, 1989 L The Honorable Lando W Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ADDITIONAL APPLICATIONS OF LEAK-BEFORE-BREAK TECHNOLOGY During the 347th meeting of the Advisory Comittee on Reactor Safe-guards, March 9-11, 1989, we discussed the NRC staff's proposal on this subject embodied in a November 22,1988-draf t of SECY-88-325, " Policy i' Statement on Additional Applications of Leak-Before-8reak Technology.' .This matter was also discussed by our Subcomittee on Thermal Hydraulic Phenomena during a meeting on March 7, 1989. During these meetings, we had the benefit of discussions.with representatives of the NRC staff, several industry groups, and Brookhaven National Laboratory. We also had the benefit of the documents referenced. The central concept of -leak-before-break (LBB)' involves acceptance of the argument that, in a given piping system,-small leaks through cracks in pipe walls can be-detected before the cracks have grown to a size where they can cause a sudden gross failure of the pipe.; Further, the argument says that when the leak is detected, the damaged pipe will be taken out of service before the crack has had a chance to grow to a size the - NRC that.is on the threshold -of unstable propagation. _ In 1987,f the L88 revised General Design Criterion 4 (60C.4) to permit the use o concept for certain purposes -and under certain circumstances in both This revision made it possible existing and new nuclear power plants. for licensees to exclude the dynamic effects of hypothetical sudden pipe ruptures from consideration in the design of certain pipe ' support structures, if the piping systems in question met certain conditions. In granting its approval for the GDC 4 revision, the Commission recog-nized that there is nothing inherent in the LBB concept that limits the application to the use specified and stated that, 'There are possibly other areas which could benefit from expanding the leak-before-break concept and simplification of requirements such as environmental quali-fication and ECCS." In response, the staff solicited public comments on this subject through a notice in the Federal Register dated April 6, A range of opinions was cited in 23 comment letters.. After 1988. considering these coments,.the staff recomended that no rulemaking be undertaken to apply the LBB concept to either ECCS or environmental qualification. They pointed out that any safety benefits associated with the application of the LBB concept to ECCS can be more readily ~

z..,

c The Honorable Lando W. Zech, Jr. March 14,1989 J l obtained under the recently revised ECCS rule. In addition, the broad scope revision to GDC 4 permitted the use of exemptions for applying L88 to environmental qualification. ' In our discussions with the NRC staff, it became apparent that they j believe the potential safety enhancements that might result from ex-tending the LBB concept would not be great enough to justify the large They_ expenditure of-resources needed to develop bases for rulemaking. seemed to feel that the industry's failure to use the exemption option in the existing rule indicated-a lack of industry interest. The staff indicated that requests for exemptions, suitably documented and su >- ported, might eventually provide the basis for a rule extending the LIB approach to environmental-qualification. In presentations to the ACRS, some representatives of the industry expressed their belief that there was a real potential for suostantial . safety and/or economic benefits in applying the LBB concept to both-ECCS' and environmental qualification. However, they were reluctant to expend their own resources on activities that they felt would not lead to changes in the rules.- We' agree with the staff's conclusions to the extent that rulemaking at this time would be premature. However, we believe an avenue for -con-sideration of further extension of the LBB concept should exist. As a result of ~ our most recent discussions of this issue with the staff and with industry representatives, we believe that' the staff. is open to a serious consideration of industry proposals to extend the ccncept to; situations for which technical justification can be provided. We recom-mend that the policy _ statement contain language which makes it clear that this is the case. Sincerely, 1 orrest J. Remick Chairman

References:

U.S. Nuclear Regulatory Commission, SECY-88-325: " Policy Statement 1. on Additional Applications of Leak-Before-Break Technology" (Pre-decisional), received by ACRS on November 25,1988. Letter dated March 3, 1989 from Malcolm H. Philips, Jr., and 2. William A. Horin, representing the Nuclear Utility Group on Equip-ment Qualification, to David A. Ward, ACRS,

Subject:

Application of Leak-Before-Break Technology to Environmental Qualification of Electric Equipment.

' \\'. 65hlCO o UNITED STATES g NUCLEAR REGULATORY COMMISSION [ s, g ADVISORY COMMITTEE ON REACTOR SAFEGUARDS p, j WASHINGTON. D. C. 20666 0,, %,..... /c March 15, 1989 The Honorable Lando,W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

PROPOSED SEVERE ACCIDENT RESEARCH PROGRAM PLAN During the 347th meeting of the-Advisory Comittee on Reactor Safe-guards, March 9-11, 1989, we discussed with members of the NRC staff a draft Severe Accident Research Program Plan, dated February 1989. Our Subcomittee on Severe Accidents met with the staff on March 7, 1989 to discuss this matter. We also had the benefit of the document referenced.- l-Because of the staff's schedule for presentation of the plan to the L Comission, we were unable to perform - a detailed review before preparing this report. However, on the basis of a preliminary review, we make the following coments. The NRC began the Severe Accident Research Program shortly after the l THI-2 accident. The emphasis was said-to be on understanding severe l accident phenomena, and in developing a capability to calculate the risks of severe accidents. Computer codes were expected to play a - L key role in~ these calculations, and development of these codes and experiments related to their--validation have represented a signi-ficant part of the severe accident research. Our previous reviews of the program have frequently led us to question the relevance of this resenich to regulatory needs. As a result, we have written a number of reports to-the Comission recommending that there be -a closer correlation between the severe accident research proposed and the ( policy being formulated to ensure protection of the public from the risk of severe accidents. We saw much of the severe accident re- -search as not properly focused to provide the information needed. In contrast, the February 1989 program plan proposes a review of the information available from previous research to identify areas in which further information is needed for regulatory decisions. Existing and proposed research programs will be reviewed and, if necessary, redirected to make it more likely that the needed infonna-tion will be developed. It is also proposed that a method of evalua-tion, such as Code Scaling, Applicability, and Uncertainty recently developed by the staff for analysis of thermal-hydraelic codes, be used to evaluate a number of the severe ac.cident codes. Further, in I

-~

f'..

( The Honorable Lando W. Zech, Jr. Narch 15, 1989 light of the fact that there appears to be duplication among some of the severe accident codes under development, it is proJosed to examine which of these codes are needed for regulatory app 1" cations, and on the basis of the results, to decide which codes deserve further development. It is also proposed that documentation be required for both existing codes and those under development. On the basis of our preliminary review, we believe that this program plan represents a substantial change and is a very positive step. We endorse the staff's requirement that all contractors show that.their 4 proposed and continuing-work address analyses or phenomena important in the predictions of risk, and have clearly defined objectives. We recomend that the Commission encourage the staff to continue in the direction indicated. Because this represents a significant departure from previous practice, some parts of the program are likely to encounter opposition. It is important that this be monitored care-fully to ensure that it does not deter the positive aspects of the proposed program. We expect to continue our review. However, our initial examination leads to the following specific observations. The near-term program dedicates a major fraction of the total re-sources to studies of various phenomena associated with direct containment heating (DCH). We believe that as an alterative, a ' greater - priority should be given to studies that might very well demonstrate that risk from DCH is negligibly low, or could be made -low by readily achievable plant modifications or procedural changes, thus making much of the proposed DCH related research unnecessary. The draft plan we have does not indicate how results of previous work or expected results from existing research programs of U.S. industry or foreign organizations are to be factored into the NRC program. We expect to explore this further. . Sincerely, Forrest J. Remick Chairman

Reference:

Memorandum dated February 10, 1989, from Brian W. Sheron, Division Director, Office of. Nuclear Regulatory Research, to Forrest J. Remick, Chairman, ACRS,

Subject:

" Revised Severe Accident Research Program Plan" (Draft plan predecisional). >e.- ___-r.

,( ? 4 " pSttcy / 'o UNITED STATES g NUCLEAR REGULATORY COMMISSION o- ]. ADVISORY COMMITTEE ON REACTOR SAFEGUARDS msHWGTON, D. C. MS6 ,p May 9, 1989 The Honorele Lando W. Zech, Jr. Chairun (U.S. Nuclear Regulatory Commission Washington,_D.C. 20555 i

SUBJECT:

NRC'S HUMAN FACTORS PROGRAMS AND-1NITIATIVES During the 349th meeting of.the Advisory Comittee on Reactor Safeguards, May 3-6, 1989, wel discussed the draft Comission paper related to the NRC's human, factors programs and initiatives. Our Subcomittee on Human Factors dis-cussed this matter with the staff during a meeting-held on April 19. 1989. The subcommittee previously discussed. draft Revision 1 of the Human Factors Regulatory Research Program Plan with the staff on January 26, 1989. We also had the benefit of the document referenced. We are pleased that the NRC again is devoting a portion of its research program to human factors issues. The list 'of topic areas being worked on or > planned is extensive. This will require dedicated research program manage-ment attention to help ensure that the research progresses-in. a - timely fashion and the results are provided in a form-for possible use by the o agency. During the January 26,-1989 meeting of our. Human Factors Subcomittee, it concluded that the Human Factors Regulatory Research Program Plan be' expanded into a human factors plan for the entire agency.-i.e., to include the human factors programs and initiatives of all of the-NRC offices. We are pleased to see _ that the : staff has ' subsequently reached the same conclusion. We, .believe that the more comprehensive document will be of greater use to the Comission and. to the interested individuals. We recomend that the' dis z cussion of the other office programs and initiatives be retained in the NUREG document:when issued. We believe that the Office of Nuclear Materials-Safety and Safeguards' human l factors initiative, _ addressing material and fuel cycle activities, is a welcome and needed addition to the hRC human factors efforts. Because few human factors considerations have been-included in these activities in the past, : much effort will be - required. It is likely that additional human factors personnel will be needed by NMSS to carry out these activities in an effective manner. The utilization 'of a number of diverse institutions and organizations as L human factors research providers 1.s comendable. This is particularly

t ., i. - The Honorable,Lando W. Zech, Jr. May 9, 1989 noteworthy in the organization and management and in the reliability assess-ment. program elements of the research plan. The use of diverse research providers has already generated new input to, as well as interest in, the human factors research program. Finally, we have recomended to the staff that a human factors research effort be initiated to develop improved methodology for the selection and training of resident inspectors. These individuals play a significant role in the. regulatory program for operating nuclear power plants. Effective resident inspectors can have an extremely positive impact on nuclear safety through their interfacing role between the NRC and licensees. Conversely, inspectors who are poorly qualified either technically or in their approach l to regulation or their interpersonal skills can have a detrimental impact on nuclear plant safety performance. We believe that appropriate human factors research could develop aptitude testing to assist in the selection of resi-dent inspectors and develop training material relating to their work assign-i ments and their relationship to licensee personnel. l We recomend proceeding with the proposed human factors research program and initiatives. We would like to be briefed by the staff on the results of the research and any proposed implementation into the regulatory process at appropriate times. l Sincere i e Forrest J. Remick ~ Chairman

Reference:

Letter dated March 31, 1989 from F. D. Coffman, Jr., Office of Nuclear Regulatory Research to Herman Aldennan, ACRS, transmitting the Comission Information Pa on NRC's Human Factors Programs and Initiatives (PREDECISIONAL) per

'g. ./'.@ **?%'c; UMTE0 STATES a - y / ^h NUCLEAR REGULATORY COMMISSION s,g i [ f ADVISORY C0hWITTEE ON REACTOR SAFEGUARDS wAsMisvGTosv. D. c. setts May 9, 1989 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

NUREG-1150, ' SEVERE ACCIDENT RISKS: AN ASSES $ MENT FOR FIVE U.S. NUCLEAR POWER PLANTS" During the 349th meeting of the Advisory Comittee on Reactor Safeguards, May 3-6, 1989, we discussed the second draf t of NUREG-1150, *$evere Accident Risks: An Assessment for Five U.S. Nuclear Power Plants," with members of. the staff. We also had the benefit of the documents referenced. Although we have not had an opportunity for more than a brief look at this. second draft, we have been asked to recomend uses to which it could be put before the completion of the peer review as organized by the NRC staff. At this time, on the basis of a cursory examination, we can recommend only that, if its conclusions are used, they should be examined very carefully in light of the criticisms leveled at the initial draft. For the most part, criticism of the initial draft focused on what has come to be called the Level !! portionoftheprobabilisticViskassessments(PRAs)discussedinthereport. It would appear on this basis that prior to peer myiew of this second draft, infomation and insights that may come from the Level I portion of. the report can be given more credence than those from other parts of the PRAs. We observe, however, that the core-damage frequencies reported do not take into account a number of external accident initiators that in other contemporary PRAs have appeared as major contributors to the risk calculated. Of some interest to us, in connection with staff usage, are comments from some segments of the staff that might be expected to use either the results or the insights derived from the report. During the past month we have observed the following: During our April 6-8, 1989 meeting, the Director. of the Office of Nuclear Reactor Regulation reported on a major effort being con-sidered to reduce the risk that he believes is associated with the interfacing-systems LOCA. We observed that the draft NUREG-1150 report did not identify this as a major risk contributor. He responded that he was skeptical of the results of.PRAs. He felt

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i The Honorable Lando W. Zech, Jr. May 9, 1989 ) that, if his current co$c' erns are borne out by further investiga-tion, this issue is important enough that it should be resolved before the individual plant examination (IPE) program is completed. Also during our April 6-8, 1989 meeting, we discussed with members of the staff from the Office of Nuclear Reg)ulatory Research the per'fomance of motor-operated valves (MOVs in nuclear power plants. They presented a study, perfonned at Brookhaven National Laboratory, which they are using as partial justification for requiring a major program of testing, maintenance, and repair of MOVs in operating plants. The report concludes that the core-damage frequency for boiling water reactors (BWRs), taking into account what they now believe to be the perfonnance of MOVs, is more than an order of magnitude greater than the core-damage frequency for BWRs reported in the draft NUREG-1150. On the basis of the staff's conclusion regarding this matter, they are recom- ) mending an extensive program which they believe will enhance valve perfomance. They consider this problem so important that it too should not wait for the IPE program. They are convinced that NUREG-1150 does not represent properly what they view as a major risk contributor. We conclude from these experiences that it may be worthwhile, in the review j process, for those responsible for NUREG-1150 to solicit connents from other elements of the staff which might be expected to use the results of the report. l In sunnary, on the basis of a very preliminary review, the insights and the results of. the second draft of NUREG-1150 should be used with considerable caution before the planned peer review has been concluded. We expect that more credence might be given to the Level I parts of the PRAs than to Levels 11 and 111. However, we repeat that some of the Level I results have already been called into question by other parts of the staff. Sincerel Forrest J. Remick Chainnan i

The Honorable Lando k'. Zech, Jr. -3 May 9,1989

References:

1. U.S. Nuclear Regulatory Commission, NUREG-1150, ' Reactor Risk Reference Document," Volumes 1, 2 and 3. Draf t issued for comment, dated February 1987 2. U.S. Nuclear Regulatory Commission, NUREG-1150, ' Severe Accident Risks: 1 An Assessment for Five U.S. Nuclear Power Plants," Volumes 1 and 2 (SecondDraftforPeerReview),datedApril 17,1989(PreDecisional) 3. Memorandum dated April 18, 1989, for the Commissioners from V. Stello, i Or., Executive Director for Operations. SECY-89-121,

Subject:

Transmit-tal of NUREG-1150, Second Draft for Peer Review 4. Memorandum dated February 17, 1989, for the Commissioners from V. Stello, Jr., Executive Director for Operations, SECY-89-058

Subject:

Status Report and Preliminary Results of NUREG-1150 5. Memorandum dated December 8,1988, foi the Commissioners from V. Stello, Jr., Executive Director for Operations, SECY-88 337.

Subject:

Plans for Future Review of NUREG-1150 t I l l 1 t

o j i $.[. : UNITED 4TATES i [ 'n NUCLEAR REGULATORY COMMIS$10N AovisoRY cohahalTTEE ON REACTOR SAFEGUARDS 3 l waamwovow.o.c.senes

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May 9, 1989 The Honorable Lando W. Zech, Jr. 1 Chaiman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

Dear Chairman Zech:

GENERIC LETTER RELATED TO OCCUPATIONAL RADIATION EXPOSUR

SUBJECT:

FROM HOT PARTICLES During the 349th meeting of the Adn sory Committee on Reactor Safeguards, May ~ we reviewed the referenced draf t generic letter, Radiation Emittedincludin 3-6,1989, Interim $tandard on Occupational Dose for Skin from lete from a Hot Particle. Our Subcomittee on Occupational and Environmental Protection Systems, its consultants, and invited expert Dr. Dade W. Moeller, discussed this matter during a meeting held on April 20, 1989 with represen. i tatives of the NRC staff, the National Council on Radiation Protection and (NCRP), and the Nuclear Management and Resources Counct1 Measurements (NUMARC). We also had the benefit of the documents referenced. t During the past few years, high sensitivity personnel contamination monitor. ing equipment has been installed in most nuclear power plants to improve This has resulted in the occasional their radiation protection programs. discovery of microscopic hot particles on workers' skin and clothing at many (Fragments from Stellite faced components containing nuclear power plants. cobalt-60 and irradiated fuel fragments are the most common het particles.) J 1 It is clear that hot particles have always been around nuclear power plants We have been told that there is no evidence but generally were not detected. The that these hot particles have caused workers any adverse health effects. staff has concluded that the existing 10 CFR Part 20 limits intended for i exposures of large areas of skin (7.5 rem per quarter for skin of the whole body and 18.76 res per quarter for the extremities) are overly restrictive The staff plans when highly localized exposure results from a hot particle.to am Until this amendment to 10 CFR Part 20 becomes I the skin by hot particles.the staff proposes to use the interim standard, that is en effective in (raft form with the generic letter, in taking enforcement actions, j Industry representatives have been expressing concern since 1987 that, level as a an unduly high result of the current interpretation of the regulation,icle doses at nuclear j of attention and emphasis is being given to hot part These representatives have indicated that this situation is j power plants. We causing unnecessary fear and concern among nuclear power plant workers j l believe this to be a very serious issue. that workers could be exposed to substantially less whole-body j showing l Attachment $

e L The Honorable Lando W. Zech, Jr. May 9, 1989 radiation (from sources other than hot particles) by setting a more realistic In order to avoid what the staff is considering hot particle exposure limit. as 'overerposures' from hot; piarticles, licensee radiation protection programs require that workers be monitored frequently for hot particles during work in areas that have the potential for hot particle exposures. This more frequent monitoring increases the time workers spend in radiation areas to complete a The results of given task and thus increases whole-body radiation exposures. an industry survey reported by NUMARC indicate that implementation of a more realistic limit (discussed below) for hot particle exposure wovid result in an estimated reduction in whole-body dose of 5 to 45 person-rem per year per nuclear power plant unit. (For 1987, the average total collective dose per unit was 420 person-rem.) Other concerns expressed by industry are cost related (reduced worker pro-ductivity and the need for more health physics technicians), increased radwaste volume, impact on $ Alp ratings, and potential insurance and legal considerations. Industry representatives have emphasized that a change in the NRC position would not result in a decrease in the protection of workers or the general public nor in the controls that have been established to prevent het parti-cles from being transported off-site. The staff, in March 1987, asked the National Council on Radiation Protection and Measurements (NCRP) to study the health significance of exposure from het particles on the skin and to provide recommendations based on the findings of this study. (NCRP has an international reputation for excellence in the field of radiation protection and has been chartered by Congress to work with d i t ti matters.)gencies and others in developing guidance in ra iat on pro e federal,a A five-person NCRP subcommittee made this study, and the NCRP provided a report entitled, ' Recommendations on Limits of Exposure to 'Not Particles' on the skin' to the staff on June 17, 1988. This report was subsequently reviewed and approved by the full 75-member NCRP. The NCRP recommendations are ' based on ensuring that ulceration of minute The risk of radiation-induced skin cancer areas of the skin' does not occur. from exposure to a hot particle was not considered to be significant er NCRP's reconnended exposure limit for particles less controlling by NCRP. than 1 en in diameter is 1E+10 beta particles emitted from the surface of the (This limit is expressed as 75 microcurie hours where one beta particle. They reconsnend that any overexposed particle is emitted per disintegration.) individual be provided with follow-u Depending on particle size and isotopic composition, this ulceration. results in a dose limit ranging from 300 to 800 rad. To place this dose in perspective, a 2000 rad dose is the accepted limit for radiotherapy treatment involving large areas of the skin. This limit is also based on avoiding skin ulceration. In its June 17,1988 transmittal letter, NCRP stated that its recommendations may be considered " fire' (subject to final editorial changes) and 'may be l

The Honorable Lando W. Zech, Jr. May 9, 1989 used and quoted as appropriate." This letter indicated that the NCRP report The staff subsequently would be published in final form in the fall of-1988. raised a number of technical and philosophical questions with respect to the NCRP recommendations that are currently in the process of being answered. 4 MCRP also requested that NUMARC provide coments on the NCRP report. MUMARC's coments supported NCRP's approach to the hot particle problem but pointed out what NUMARC believed to be considerable conservatism used in the As a result of the staff and NUMARC coments, there is NCRP re'comendations. no firm schedule for final publication of the NCRP report. The staff plans to revise appropriate sections of 10 CFR Part 20 to limit hot particle exposure of the skin and will consider the final NCRP recomenda-tions and recent research results. However, the staff recognizes that it will be at least two years until this revision can become effective and believes that it is appropriate to use an interim standard in the exercise of its enforcement discretion regarding hot particle exposures. The staff considered implementing the recommendations in the NCRP report in its interim standard for skin exposures to hot particles. However, the staff decided, for a variety of reasons cited in the draft generic letter, that it would be inappropriate to implement these NCRP recomendations at this time. the interim standard enclosed with the draft generic letter, in Instead, hanges the limit for exposure of the skin to radiation from hot effect c particles from 7.5 rem (skin of the whole body) or 18.75 rem (skin of the hands and forearms, and feet and ankles) per calendar quarter to 50 rad per hot particle exposure. Recommendations We do not endorse the staff's proposal to issue the generic letter and interim standard in its present form. Industry, in its presentation to us, has made a strong case that the proposed interim standard for hot particle exposure would provide very little relief in addressing the hot particle problem and believes that the interim standard should be based on the NCRP recomendations. The staff, on the other hand, has obvious difficulty in basing an interim Accordingly, we recomend that staff standard on an unpublished NCRP report. senior management take an active role in effecting a timely resolution of remaining outstanding issues with NCRP so that its report may be published. The staff should then develop on an expedited basis an interim standard based Based on what we have been told, we believe on the NCRP recomendations. To the that this interim standard could be completed by September 1989.recomendatio the NRCP extent the standard differs from reasons for such modifications should be clearly and completely documented. Also, the staff concurrently should move ahead with its planned revision of i 10 CFR Part 20 rulemaking on this subject. a,-,- e-m ww- _ __ - _--=, _. - -__ m ,__._____m .. w w i--+--w---

1 The Hor:orable Lando W. Zech, Jr. May 9, 1989 ) There are two additional items concerning the draft generic letter and interim standard that we beljeve should be corrected in the final interim I standard. First, the draft interim standard fails to define a hot particle with respect i to size for purposes of regulatory control. This is a very important issue, i since the size of the exposed area of skin is central to the detemination as to whether the exposure limits for large areas of skin or hot particles should be used. NCRP uses 1 millimeter as the maximum size that should be used in implementing its reconnendations. We believe that this issue needs to be clarified in the final version of the interim standard and in the planned revision of 10 CFR Part 20 on hot particles. Second, we recomend that the regulatory concept contained in Section 4, Occupational Exposure Limit, of the draf t interim standard be reconsidered. The section states that the NRC will not issue a notice of violation (N0V) for a single hot particle exposure (less than the proposed limit) to an individual during a calendar quarter. It further states that the staff may issue an NOV if any individual is exposed to two or more hot particles during a single event or to hot particles in two or more separate events during a calendar quarter. This policy appears to be an unnecessary and complicating feature of the draft interim standard given the existing regulatory require-ments of 10 CFR Part 20.201, Surveys, which requires that licensees must perform ' adequate surveys.' It is also inconsistent with the staff's posi-tion that hot particle exposures are not to be added to skin dose for record-keeping purposes and are not themselves additive unless they occur in the same location on the skin. We intend to follow the progress of the interim and final resolutions of this difficult and controversial issue and will provide you with further comments as appropriate. Sincere orrest J. Remick Chaitsan

References:

Letter dated February 9,1989 from J. H. Sniezek, Office of Nuclear 1. Reactor Regulation, to E. L. Jordan, Committee to Review Generic Re-quirements,

Subject:

Generic Letter and Interim Standard Concerning Not Particle Exposures of Skin 2. Letter dated June 17, 1988 from W. R. Ney, National Council on Radiation Protection and Measurements, to R. E. Alexander, Office of Nuclear Regulatory Research, transmitting NCRP Report 80-1, ' Recommendations on Limits of Exposure to ' Hot Particles' on the Skin' (draft of June 1988/Rev.3) ,m ..._...-~..,_.....

*[ UNITE) STATES 4 NUCLEAR RECULATCRY COMMISSION e, t ADVISORY contatlTTEE ON REACTOR sAFEOUARDS - g WAsMineatoev,0.c.senes June 15, 1989 1 The Honorable Lando W. Zech, Jr. Chairman U.S. Nuclear Regulatory Comission l Washington, D.C. 20555-

SUBJECT:

NRC THERMAL-HYDRAULIC RESEARCH PROGRAM During the 350th meeting of the Advisory Comittee on Reactor Safeguards. June 8-10, 1989, we reviewed the NRC's plan for continuing thermal-hydraulic research as related to the design and operation of nuclear power plants. This matter was also considered by our Subcomittee on Themel Hydraulic Phenomena at a meeting on May 23, 1989. During these meetings, we had the benefit of presentations by representatives of the Office of Nuclear Regula-tory Research (RES). We also had the benefit of the documents referenced. The Comittee last comented to you on this subject in our report of June 7. 1988. Thermal-hydraulic research has always been a central and major part of the NRC's research program. Much of the work was inspired by the perceived need to better understand hypothetical large-break loss-of-coolant-accidents (LB LOCAs) and the performance of emergency core cooling s stems (ECCS). Experiments and analytical models, such as the RELAP and TRfC codes, have confimed compliance with the ECCS rule. Continuing research on LB-LOCAs culminated with a 1988 revision to the ECCS rule which pemits licensees to use more accurate means of analysis and makes possible certain safety and operational improvements in existing plants. MRC contractors have demon-strated a methodology that can be. used to estimate the magnitude of uncer-tainty associated with code predictions. 1 In addition, the experimental infomation base and the codes have been found useful in assessing and predicting the consequences of transients and small-break loss-of-coolant-accidents (SB-LOCAs) which are now recognized to-be much more risk significant than the L8-LOCAs. The codes are also being used to analyze the early stages of severe accident scenarios. Proposed Research Procram We understand the continuing NRC program in thermal-hydraulic research to have two principal purposes:

• Bring development of the major computer codes to a successful comple-tion.

l t + . + - - - - - m

The Honorable Lando W. Zech, Jr. Jun 15,1989 9 O Maintain, within the NRC and its contractors, a capability for themal-hydraulic analysis sufficient to deal with safety and regulatory con-cerns that might arise in the future. This includes the continuing availability of a cadre of experts. RES representatives indicated these general purposes would be realized through achievement of several specific objectives: 0 The major codes will be maintained indefinitely and some further devel-opment will be carried out. The scope and depth of further development seems not to have been decided. Apparently, it will include appropriate reactions to new data from foreign experimental programs and assessments which are expected to continue for some time. It may also include a review and redevelopment of the important constitutive equations in the codes. O The current experimental programs related to specifics of the Babcock and Wilcox (B&W) nuclear steam supply (NSS) system will be completed. Beyond this, any further experimental programs will be carried out at universities, rather than by the creation or operation of any major facilities at national laboratories. Relatively inexpensive " integral" facilities, of scope similar to the facility now operating at the University of Maryland, are being considered as contrasted with what I have been called " separate effects" facilities. These would be mockups of specific NSS systems and of an advanced LWR (600 MWe size) design. An expanded program of applications research is planned. Apurently, O much of this activity is expected to be in response to issues tsat arise from experiences with operating plants. But, it will include prepara-tion of input data for several more plant types than are now available to the NRC. This will pemit more rapid analysis than would otherwise be possible in response to future safety or regulatory issues. This program may also include exploratory, in-depth studies of a range of possible transients for a variety of plants. In addition, two other specific program elements were mentioned: 0 A further demonstration of the ' Code Scaling, Applicability, and Uncer-i tainty" methodology will be carried out for an SB-LOCA with RELAPS/ MOD 2, similar to that recently completed for an LB-LOCA. Improvements will be made to the NSS system process models now incor-O porated in training simulators at the NRC Technical Training Center. This will permit more accurate simulation of off-normal scenarios for the study of emergency and accident management procedures. .---..-2..,o,.- e-- s.

The Honorable Lando W. Zech, Jr. 3 June 15, 1989 Before comenting on these; r' search proposals, it is pertinent to consider e two statements made by the NRC staff at the May 23, 1989 Thermal Hydraulic Phenomena Subcomittee meeting, because the ideas expressed have an infNence on our recomendations: A representative of the Office of Nuclear Reactor Regulation said, 'NRR is not relying extensively on the codes to address current licensing issues." A representative of RES said, ' Codes have now reached an accept-able level of accuracy and maturity... further development is not likely to produce major changes in our understanding of [ plant] performance or [ accident] consequences." ~ ACRS Recomendations We agree with the general objective of the research program to maintain, within the NRC and its contractors, a capability for themal-hydraulic analysis sufficient to deal with safety and regulatory concerns that might arise in the future. Also, we agree with the general level of funding pro-jected for the next several years. However, we believe there is too much emphasis on further development of the existing codes in the planned program. Maintenance of the needed NRC capability is more a matter of ensuring the availability of a cadre of experienced and expert analysts and access to the general body of experimental data, than it is of improving or even ensuring the availability of large systems codes. The Cosoittee reiterates its comments in the report of June 7,1988, that " marginal improvements that could be made [in the codes) over the next few years by extrapolating the recent levels of development work will not be sufficient to attain a signifi-cantly higher plateau of code accuracy and validation.' To accomplish this general purpose, we recomend a program of four primary elements: (1) Code Development Maintain the present large system codes, TRAC-PF1/M001. RELAP5/M002 TRAC-BWR, and RAMONA-3B, for an indefinite period. Limit improvements only to those required by: (a)thediscoveryofimportanterrorsor(b) crucial new information from the foreign experimental and assessment programs or the B&W testing program. Do not undertake major new re-structuring or *2ero-based" improvements to the constitutive equations or numerical algorithms in these codes. We are not convinced by the arguments given for the need to develop TRAC-PF1/M002 and RELAPS/M003. It is our view that the proposed modifications will not substantially improve the codes. I

The Honorable Lando W. Zech, Jr. June 15, 1989 Instead, consideration'should be given to the development of a new type of systems code that will be more useful for analysis of extended plant transients involving interactions of plant systems. The Committee also made this recomendation in its June 7,1988 report. TRAC and RELAp were originally designed to analyze the LB-LOCA, a rapid and severe reactor transient, in great detail. There is a need for a more empir-ical and efficient analytical tool. We envision a code that would be able, for example, to make a rapid and sufficiently accurate analysis of the power oscillations observed last year at the LaSalle County Station, Unit 2 plant. Such a code would be more akin to advanced simulator codes than to TRAC and RELAP. The BWR code (HIPA) now in use at Brook-haven National Laboratory is an example of the type of code we are suggesting. (2) Experimentation The staff proposal to develop relatively inexpensive ' integral" test facilities at universities is sound. We see this as consistent with our recomendation for a new type of systems code. We agree that it would be inappropriate to build several such facilities at one time. A gradual approach is warranted. The first such new facility might be one that would incorporate features of the advanced LWR designs. Also, it will be better to completely assess the benefit that has been obtained from tests with the University of Maryland facility mentioned above. In addition, a small program to deal with more fundamental research should be maintained. These are experiments of the sort that have been previously called ' separate effects" tests. An effort should be made to develop a consensus among experts as to which particular phenomenon should be investigated. At this time, we suggest consideration be given to the investigation of:

• fluid-elastic instability related to vibration of tubes in U-tube steam generators.

departure from nucleate boiling with oscillating flow and power in O

BWRs, O dynamic instabilities and loads on valves.

(3) Data Analysis A major effort is needed to organize data from test pro 1 rams into a useful fo m other than the large systems codes. In particu' ar, with the 2D/3D. ROSA-IV, and the B&W test programs all coming to closure, mea-sures are needed to ensure that these expensive and valuable bodies of data are preserved and used. In addition, older data from, for example,

y.., ..c, -The Hercrable Lando W. Zech, Jr. -S-June 15, 1989 the FIST and FLE06 programs can be of greater value if they are effectively organized into more useful forms. (4) Applications Research A program in this area should include three elements: O Analysis of transients indicated to be of interest as a result of plant operating experience. O Preparation of input data decks for several classes of plants so that turnaround time for analyses in response to experience is shortened. O Analysis of transients that are indicated by PRA or other sources of information to be of particular irterest, but which are not presently well understood. We suggest the following for considera-tion - feed and bleed scenarios - secondary depressurization scenarios finally, we suggest that RES broaden its perspective as to what other re-search in the thermal sciences should be included in its program, rather than being limited to the traditional scope of concerns in thermal-hydraulic We suggest that it include studies of a broad range of thermal and areas. fluid transport issues related to reactor safety. ACRS Members William Kerr and Forrest Remick did not participate in the review of this matter. Sincerely, M 0-a David A. Ward Acting Chaiman i

References:

1. U.S. Nuclear Regulatory Comission, draft SECY Paper: " Status and Plans for Thermal Hydraulic Resear'ch conducted by the Office of Nuclear Regulatory Research," provided to the ACRS in May 1989 2. U.S. Nuclear Regulatory Comission NUREG-1252: ' Nuclear Power Plant Thermal-Hydraulic Performance Research Program Plan," Office of Nuclear l Regulatory Research, July 1988

j 'f,Uru A ~ o, UNITED STATES ^, NUCLEAR REGULATORY COMMISSION P ' aq n -{ f ADVlsORY COMMITTEE ON REACTOR SAFEGUARDS 1 wam.atow, o. c. uses g g % *... +.e July 18, 1989 i i The' Honorable Kenneth M. Carr Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

PROPOSED STAFF ACTIONS REGARDING THE FIRE RISK SCOPING STUDY l (NUREG/CR-5088) During the 351st meeting of the Advisory Comittee on Reactor Safeguards. l July 13-14, 1989, we discussed with representatives of the NRC staff the proposed actions delineated in SECY-89-170, " Fire Risk Scoping Study: Sumary of Results and Proposed Staff Actions," for dealing with various recomendations resulting from the Fire Risk Scoping Study. Our Subcomittee l on Auxiliary and Secondary Systems met on July 12, 1989 with members of the NRC staff and the Sandia National Laboratories to discuss this matter. We also had the benefit of the documents referenced. One of the significant findings of the scoping study is that fire PRAs do not normally address fire vulnerabilities in several important areas, including): .(a) fire-induced alternate shutdown / control room panel interactions. (b smoke control and manual fire-fighting effectiveness, (c) adequacy of fire barriers, and (d) seismic / fire interactions. The staff agrees with this finding and is currently considering including an effort in the Individual Plant Examination for External Events (IPEEE) program to search for such vulnerabilities. Also, we understand that the staff's External Events. Fire 4 Subcomittee is developing appropriate guidance for dealing with ' these l issues. We consider these actions reasonable. l In SECY-89-170, the staff has concluded that no new fire-prctection research is needed at this time. The need for additional research will be recon-sidered following final definition of the fire-related parts of the IPEEE program later in 1989, the peer review of NUREG-1150 fire analyses, and further discussions with the Comittee. We plan to coment on the need for further research in the fire-protection area after receipt of the IPEEE guidance document for examination of fire-related effects. I

v_ n. 1 i The Honorable Kenneth M. Carr July 18, 1989 J Additional remarks by ACRS' bmbers William Kerr and Charles J. Wylie are presented below. Sincerely, Forrest J. Remick Chaiman ~ Additional Remarks by ACRS Members William Kerr and Charles J. Wylie We recommend that the staff require the use of amored electrical cable in advanced light water reactors. There are more than 20 years of U.S. electric utility experience which demonstrates its advantages in both nuclear and fossil electric generating plants. There is extensive experience with armored cable in naval and maritime vessels nd in chemical plants. The British are requiring its use in the Sizewell B plant. The armor makes it significantly more difficult for external. heat sources to kindle and-to propagate fires within the cables. It is practically impossi-ble to kindle and propagate a fire from internal short circuits and over-loads. - Annor provides a high degree of mechanical protection for the cable. It also provides shielding against external electromagnetic fields. This feature becomes more important as the application of solid-state components in power plants increases. It is particularly.important in providing protec-tion against electromagnetic pulses generated by lightning. References Results and Proposed Staff Actions" (Predecisional) g Study: SECY-89-170, dated June 7,1989, " Fire Risk Scopin Sunnary of 1. i. 2. U.S. Nuclear Regulatory Connission NUREG/CR-5088, " Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, Including Previously Unaddressed Issues " Sandia National Laboratories. January 1989-3. Memorandum, dated December 28, 1988, from Frank P. Gillespie, NRR, for l' Eric S. Beckjord, RES,

Subject:

" Fire Risk Scoping Study: Sunnary of Results and Proposed Research Action" l+

l / 'o, UNITED ETATEs s I NUCLEAR REGULATORY COMMISSION n ) l ADVISORY COMMITTEE ON REACTOR sAFEOUARDS j g a y wash:Norow. o. c.aoses ...e November 20, 1989 k The Honorable Kenneth M. Carr Chainnan U.S. Nuclear Regulatory Comission Washington, D.C. 20555

Dear Chairman Carr:

[

SUBJECT:

DRAFT SUPPLEMENT NO. 2 TO GENERIC LETTER 88-20, " ACCIDENT MANAGEMENT STRATEGIES FOR CONSIDERATION IN THE INDIVIDUAL PLANT EXAMINATION PROCESS" 5 During the 355th meeting of the Advisory Comittee on Reactor Safeguards, November 16-18. 1989, we discussed the subject document with the NRC staff. We also reviewed a draft NUREG/CR report entitled, " Assessment of Candidate Accident Management Strategies," that the staff proposes to send as an enclosure with the supplement to the generic letter. We had the benefit of these documents which are referenced. Our Subcosuittee on Severe Accidents met on September P0, 1989 to discuss this matter. We conclude that the infomation in these two documents will be useful to I licensees in the process of perfoming Individual Plant Examinations, and we agree that the documents should be issued. The draft NUREG/CR report referred to describes strategies for accident management that are said to be PRA based. However, the report does not include infomation on the risk reduction that might be attributed to the strategies. This infomation would be-useful to those considering the strategies. We recomend that this information be added if it is rea-sonably retrievable from existing sources. We observe that a number of the strategies described in the draft NUREG/CR report either overlap or are very similar to the content of the emergency ~ operating procedures that are either being developed or are already in place in many plants. We believe that labelling these procedures as 4 E accident management strategies where others label them as emergency-opfating procedures is likely to lead to confusion on the part of both tie NRC staff and the industry. Sincerely, t Forrest J. Remick Chairman , _ _. _ _ _ _. _ _ _ _ ~ _ _.. _ d

The Honorable Kenneth M. Carr November 20, 1989 References 1. U.S. Nuclear Regulatory Comission, ' Accident Management Strategies for Consideration in the Individual Plant Examination Process," Draft Supplement No. 2 to Generic Letter 88-20, dated November 8,1989 (Predecisional) 2. U.S. Nuclear Regulatory Comission, ' Assessment of Candidate Accident Management Strategies," Draf t NUREG/CR Report (Unnumbered), Prepared by BNL, October 1989 O

^~ ' khk, UNIVERSITY STANFORD

DLPARTMLNT OF GEOLOGY TELLPHONE: (4151 723-2537

$<lmot of [arth Sciences TELEX: 346402 STANI RD STNU Stani rd California 941052115 FAX: (4151 725 2199 11 August 1989:.,

1. .W N Dr. Chester P. Siess 3110 civil Engineering Building i:-

University of Illinois Urbana, IL 61801 'gp g W a !.. Dear Chet! I participated in the 3-day meeting August 8-10, 1989, held in Rockville/ White Flint, concerning the Diablo Canyon Nuclear power Plant. The meeting was attended by NRC staf f, pG&E representatives, and consultants. A partial roster is enclosed. Leon Reiter was the main spokesman and questioner on behalf of the staff. The meeting was informative and useful. Some of the highlights follow - not in chronological order. On the second day, the characterization of the Hosgri (as to fault type, slip type, and dip) was discussed again, at length, with bits of new data and new reaponing. All geologists present, including Bob Brown of the USGS, and all/ geophysicists who gave an opinion, now finally agreed _that the Hosgri is most probably a steep strike-slip fault, perhaps with some obliqueness in slip (i.e., a reverse component). The NRC staff appeared to be satisfied and relieved. This rather tardy consensus alone made the meeting a success, because the fault dip and predicted focal mechanism are prime parameters in the logic tree that leads from source to ground motion. The arguments supporting strike-slip included the familiar ones, but were bolstered by re-processed seismic reflection profiles, recognition of flaws in Jim Crouch's thrust fault interpretation, seismicity (a cloud of hypocenters roughly terminates near a hypothetical steeply dipping Hosgri fault), recent focal mechanisms of small earthquakes (includ-ing a strike-slip mechanism beneath the trace of the Hosgrt), and calculations of horizontal:vertica) slip ratio based on the San Simeon slip rate vs. vertical offset of Cenozoic strata in seismic profiles. Thrust faults certainly are present, and some adjoin the Hosgri, which itself may lean a bit. PG&E favors a dip of 700-900, which seems reasonable. I enclose a rough diagram which I sent you in the hey-day of the thrust hypothesis. Alternative possible dips and behaviors of the Hosgri fault are retained in branches of the. logic tree, but are not very heavily weighted. Parts of the logic tree were displayed and discussed at length. Leon asked how the weights were chosen for alternative possible parameters for the Hosgri fault. It seems that they were produced subjectively by a group including Clarence Allen, Bruce Bolt, Lloyd Cluff, Kevin Coppersmith, p Woody Savage, Cole McClure, Dill Lettis, Tim Hall, Doug Hamilton, and perhaps one or two others. The combined experience, judgment, and responsible reputation of most of these people is first-class. The group met for extended sessions, we were told, reviewed data, expressed opinions, argued, and arrived at numerical weights they could all live with. Kevin Coppersmith talked about segmentation. The Hosgri fault has been divided into 6 segments, using Knuepfer's empirical world-wide data. The end-points of segments were picked after statistical treatment, described by Bob Youngs,(?). Statistics purportedly show which features / changes

'cleng fcults retu:11y corrolcto with dif for:nces in brhtvior of various

• +"

parts. For example, double bends in the fault trace, fault junctions, large step-overs, differences in vertical slip rate, and other items are deemed to be statistically significant. A distinction is made between fault segments and rupture segments, as rupture in 1 event can involve 2 or more fault segments. Statistical analysis of empirical data has been more successful in identifying the kinds of end-points of segments that may be ignored (transected) by propagating rupture than in identifying points that will stop a rupture. This kind of study is worthwhile, but not yet very useful. Coppersmith reported on the use of moment magnitude (M ) rather y than surface wave magnitude (M ) in characterizing the results of faulting. s As I understand it, a given amount of slip gives the same magnitude (M ) y regardless of the mode of faulting (strike-slip, reverse, or oblique). However, the type of source mechanism for a given amount of slip might affect the ground motion, if not the magnitude. Burt Slenmons presented new plots and curves relating dimensions of surface faulting to EQ magnitude, based on world-wide data on all kinds of faults. He divided the faults irCto two groups, those in contractional domains, and those in extensional domains. In one study, he used average displacement along the fault, rather than maximum displacement. These procedures produced only small differences from earlier plots of dis-placement vs. magnitude. During the course of the meeting, it became evident that virtually all attendees were satisfied with the H, 7.2 magnitude chosen by PG6E for the maximum credible earthquake from the Hosgri f ault. Two small but bothersome geologic structures received much attention because of their proximity to the Diablo plant: the San Luis Bay fault and the Olson fault (?). These do not worry me because: they do not display the juxtaposition of unrelated rock units that would signify large displacement; although seen locally at the ground surface, they cannot be traced, and are therefore probably minor; their low slip rates indicate that tens of thousands of years would be required before a 1 m slip could occur. These faults could not produce an H 6 earthquake, much less an M 7. In the end-of-meeting caucus of the NRC staff and consultants, the usual demands for more work by PG&E were toned down a bit. Most of the staff suggested things that are still needed, but they seemed to agree that field work is reaching the point of diminishing returns. Some ampli-fications and clarifications of the written record were requested, inluding full explanation of the manner in which key decisions and judgments were made. It was proposed that a small group of USGS geologists / geophysicists should spend time with PG&E counterparts in going over certain critically located seismic profiles and other types of observations (mostly offshore) that have not yielded useful information hitherto, or that have yielded confusing results. I urged the staff to include George Thompson. The presentations and discussions were excellent. The staff asked some important questions. The geologic / seismic investigation is drawing to a close, and probably should not be prolonged much farther. With all good wishes, cc El Igne Benjamin M. Page

CT-/956 QUERYTECH ASSOCIATES, INC. 9040 EXECUTIVE PARK DRIVE, SUITE 217 KNOXVILI.E, TN 37923 Phone (615) 690 2728 f.tm 7'B Y August 22,1989 U S. Nuclear Regulatory Commission Advisory Committee on Reactor Safeguards Mailstop P-315 Washington, D. C. 20555 Attention: Mr. E. P. Igne,

Dear Al,

Enclosed is a copy of my comments on the Seabrook meeting of August 17,1989. If there is a need for any further assistance, please call me. Sincerely, M. Bender Copy to Professor Wm. Kerr l l

. ', f I COMMENTARY ON SEABROOK EMERGENCY PLAN PREPARED BY M. BENDER, QUERYTECH ASSOCIATES, INC. August 22,1989 ) GENERAL OBSERVATIONS The Seabrook emergency plan seems to be fully responsive to the NRC regulations. The Licensee has evidently bent every effort to be sure that the local authorities in both Massachusetts and New Hampshire are conscious of their obligations in an emergency and will respond appropriately if the need arises. Organization for emergency response planning appears to be more elaborate than provided for many other nuclear power sites. Local concern of the citizenry for nuclear accidents probably warrants the level of effort expended but in the longer term it might be more realistic to reduce the size of the response organization and rely on a smaller group to provide key actions if a real emergency arises, i EMERGENCY EVACUATION Emergencies requiring site evacuation are very low probability events. The licensee's conservative estimate is that at a time of peak occupancy the entire area within the emergency planning zone could be evacuated in less than 8 hours. The transient visitors to the beaches represent the principal load on the emergency routes and under nonemergency conditions, observed subsequent to rainstorm warnings, only about two j hours have been needed to clear the beaches. i The licensee's emergency actions are predicated on evacuation at an early stage of an accident. Since the PWR system installed at Seabrook has ample heat capacity and a strong reinforced concrete containment 1

) i l SEABROOK EMERGENCY PLAN BENDER COMMENTS,8-2249 structure, there is every reason to believe that even the lowest probability events would not result in potential for significant radionuclide release from i the reactor system and its containment before complete evacuation could j be implemented. i EMERGENCY RESOURCES The resources available to the Ucensee in the event of a serious accident L are not very clearly defined. Presumably, the operating staff understands I the basics of dealing with bulk radionuclide releases if they occur. The docu. mentation describing the training for such circumstances is not readily available but the Seabrook SER (NUREG 0896, Supplement 8), indicates that the regulatory staff has reviewed the Seabrook capabilities and found them adequate.

ll Obviously, there are uncertainties associated with any accident that might not be anticipated by the training program. Backup capabilities to provide I

advice in such circumstances ought to be identified. EMERGENCY RESPONSE TESTING Most of the emergency response tests performed, thus far, seem to involve actions related to a site emergency conditions defined by the 20 critical safety functions outlined in NUREG 0654. The accident scenarios for testing response are developed by the Seabrook emergency planning staff and used to challenge the response of the operating staff and the external emergency response organization. This seems to be an effective approach to making sure that the organization remains "on its toes". It would be useful to compare the test scenarios being used to real emergency circumstances that, in the past, raised questions about emergency response capabilities at nuclear power plant sites (e.g. TMI-1, 2

. t ', * '.. SEABROOK EMERGENCY PLAN BENDER COMMENTS,8-22-89 the Browns Ferry Fire, the Fermi simulatoi error initiated operating mistake,- the Davis-Besse feedwater malfunction, loss-of-power at Millstone) to see how they would be addressed in current emergency response planning. Such a review should show that all of the previous deficiencies in emergency response capability have been addressed. TOWN OF ESSE); SAFETY ISSUES The classical issues raised in the July 31, 1989 letter from the Town of Essex should be recognized in Seabrook emergency plans. Effectiveness of the containment in limiting the dispersal of radionuclides in the event of an accident, even if leakage occurs, should be established. The experience at TMI-1 shows that even with substantial core melting, only the noble gases are likely to escape and these will be quickly dispersed in a manner that results in negligible health threat to the public. The NRC staff should be able to provide documentation to support such a position in order to display appropriate interest in questions of this sort when raised by public intervenors. L 3

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STANFORD, CAllfORNIA 9430b2171 "*IO$7[E2E P"5'" s . September 6, 1989 r e Dr. Chester-P. Siess 3110 Civil Engineering Bldg. University of Illinois Urbana, IL ~ 61801 i

Dear Chet:

Ben Page and I represented the ACRS at the meeting on seismic-source characterization for the Diablo Canyon nuclear plant held 11n Rockville, Maryland, on August 8, 9, and 10 The object of the meeting was to discuss earthquake magnitudes, the resulting ground motion, and the risk analysis from both deterministic and probaba-listic viewpoints. Discussions were ' thorough and conctructive on all sides (PGLE, HRC, USGS consultants, etc.). Agreement seemed to converge on most major issues but there was incomplete agree-ment on the potential for earthquakes on two small faulte. Clearly the Hosgri fault dominates the hazard, and the moment-magnitude Mw (nearly equivalent to Ms in this range) chosen for the Hosgri is 7.2. - The previous Hosgri magnitude, an unspecified M of-7.5 was based on an Mc of 7 3 for the 1927 earthquake, now down-graded to 7.0 and not on the Hosgri. A thorough discussion of the H assigned was based-upon slip rate, probable maximum rupture length (70km), and maximum o f f set per event. Evidence continues .to accumulate that the Hosgri is primarily a near-vertical strike-l slip fault but that thrust splays and folds are (not surprisingly) associated with it. Careful examination of reflection seismic data that led Crouch to the " major thrust" hypothesis clarified reasons for this opinion, based partly on distortions due to vari-able rock velocities. Three smaller faults mapped on land are judged to be capable; they trend northwestward toward possible submarine intersections with the Hosgri. Of the three, the Los Osos fault is assigned-a potential Mw of - 6.8-7.2 With a 600 dip the fault plane is 8 km from the plant site. The magnitude would have to be 7.5 to rival' the'Hosgri.. The San Luis Bay and Olson faults constitute part of the_ southwest boundary of the block on which the plant rests. They offset marine terraces (The San Luis Bay (last 10,000 years) fault at a rate of l .02 mm/ year): but have no detectable Holocene L slip. A worst-case scenario would have one of these faults con-L ceivably extending toward the Hosgri and passing within 3 km of the plant. Such a projection can't be ruled out because of the difficulty of' obtaining seismic reflection data in the surf ~ zone. L This situation generated considerable discussion about near-field effects, etc. I conclude by pointing out that the Olson and San Luis Bay are small faults that have already been studied fairly exhaustively. In my opinion they pose little hazard com-pared to the Hosgri. I regret not being able to stay for the wrap-up caucus of the t )

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a structural-mechanical consulting engineering firm j . 9217 Midwest Avenue

• Cleveland, Ohio 44125 + (216) 587 3805 + Telex: 5106015834 + Fax: (216) 587 220$

99906.2A CSiess C.T-l95A October 10, 1989 0 01 ? '3 0[.12 9N ) Dr. Chet Siess 1 Advisory Comittee on Reactor Safeguards Nuclear Regulatory Comission . Washington, D.C. 20555 1

Dear Chet:

Enclosed herewith please find a copy of a talk I recently gave to the American Institute of Professional Geologist which may be of interest to you or your colleagues. Sincerely, el s 7.J W J n D. Stevenson President JDS/m Enclosure l l

l j i 99906.2A 1.wpf t New Nuclear Power? - When and If by J. D. Stevenson! This talk in the Government Affairs Session of the American Institute of Professional Geologists is going to be concerned with Nuclear Power. One might ask what connection does nuclear power have with Geologists in general and Government Affairs in particular? Siting of a nuclear power plant, with the 4 possible exception of a major dam or a super fund waste storage or dump site, employs more geologists than any other single project. USNRC Standard Review Plans 2.5.1 and 2.5.2 contain a summary of the various technical areas requiring geological-seismological inputs associated with the preparation of a Safety Analysis Report required to obtain a construction or operating permit for a nuclear power plant. It is estimated that more than 100 person years of geologist labor is required to prepare this input and define necessary geology-seismology related project design parameters. f Nuclear power plants are the most heavily regulated projects currently in existence in the U.S. Government and industry developed regulation for a nuclear power plant are estimated to have increased nuclear power plant direct manpower design and construct requirements by more than 100 percent and total costs (due primarily to additional interest, escalation, and fixed costs associated with lengthened schedules) more than 200 percent as compared to equivalent nuclear plants built in the U.S.15 years ago and in most other industrial countries today. It has been popular to site federal regulatory bodies for most of the regulatory induced cost increases in recent years. While such bodies should carry a. fair share of the responsibility for these cost 3Senior Consultant, Stevenson and Associates 9217 Midwest Avenue Cleveland Ohio, 44125.

l increases, equally to responsible have been changes to industry standards such l as the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. As an example, the ASME Code in 1977 began to require Design Report for Class 2 and 3 components. Prior to that time, the Code had only required such a report for Class 1 components. This single ASME Code change which was not propted by any identified difficiency in older plant designs reliability, or safety has added over 100,000 engineering hours to the design of each nuclear power plant. Nuclear power plants currently generate about 20 percent of the electric power produced in the U.S. and account for about 16 percent of the generating capacity. This differential is explained by the fact that nuclear plants tend to be based loaded, that is they operate continuously (between refueling periods which typically range from 9 to 18 months) and at a constant electrical output. Fluctuations in electric power demand are accommodated by variable output operation of fossil power plants most recently in the form of gas turbines which are usually much smaller, can be brought on or taken off line quickly and simply and can be operated economically at different power levels. Not withstanding this significant usage of nuclear power there has not been a new plant ordered since 1978 and since that time about 40 ordered plants have been either cancelled or delayed indefinitely. Currently there are 106 operating plants with 6 more expected to go on line within the next 2 or 3 years. After that there are no nuclear power plants in the pipeline. Since it currently takes about 14 years in the U.S. to put a nuclear plant into commercial operation after a decision is made to build it, there can be no new nuclear plants in operation in the U.S. until after the turn of the century using current design and construction practices. Since 1973 the U.S. has reduced per capita power consumption by about 25 percent primarily due to conservation and the loss of on shore industrial production. In addition cogeneration where industry uses its process waste heat to generate electricity has also undergone a dramatic change since 1980. Prior to 1978 government regulations forbade industry to sell power generated by the industry to the public. After 1978 government regulators reauired the local power company to purchase industry generated power at the highest differential rate that the

. _. -.-= t utility was generating power. This changed policy has resulted in a significant increase in cogeneration produced electricity which has had direct impact on the reduced need of electricity produced from electric power company owned central power stations. Over the past 10 - 15 years these two equivalent sources of new energy, (a) a reduced power demand made possible by conservation and reduced industrial l activity and (b) cogeneration by industry have largely run their course. The reason they have run their course can be determined by the fact that the easily implemented conservation projects have already taken place. In addition there is growing resistance to expanding industry off shore as well as a growing government requirement for foreign products to be manufactured in the U.S. For example, for a car to be counted in the required gas mileage base mandated by Congress it will have to have 45 percent U.S. content. Also power companies have essentially eliminated their inefficient high cost production facilities thereby greatly reducing the incentives for cogeneration. It is anticipated that excess power generating capacities of the 1980's based on planning of the late 1960's and early 1970's will give way to power shortages in the 1990's based on the planning (or lack thereof) in the 1980's. Power companies since the early 1980's in general have been burned by the second guessing of state Public Utility Commissions when it comes to putting new power plant generating facilities in the rate base hence have not taken a conservative, or lower bound view of future power demands. What will happen in this coming j environment of electric energy shortage is still unclear. Will we as a nation accept the dislocations caused by brownouts and possible rotating blackouts or will we embark on a crash power generation construction program. If I were a betting man, I would put my money on a crash program to provide the new power generating facilities both fossil and nuclear. Realistically as far as nuclear is concerned this would require a dramatic change in government policy and associated industry practices which currently dictate the 14 year project schedule. It is possible and is currently being routinely accomplished in Japan and Western Europe to complete a nuclear power plant project in 7 years or within half the time it typically takes in the U.S. This is being done by rationally addressing w -'w-- ae w a-

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the three areas which have the greatest impact on nuclear scheduling and cost:

(1) Licensing Process (2) Seismic (Earthquake) Design (3) Quality Assurance The current two step licensing process which permits public intervention at both the construction permit and operating permit stage and associated delays should be cut'to one. In France it is usual to permit only a 60 day public review and potential intervention at the construction permit stage and then only in matters related to the particular site in question. The overall operation and design safety of.the nuclear ~ plant is not subjected to public review or debate but is determined by the technical experts of regulatory authorities interacting with .the utility and nuclear steam system supplier. It is thought that a similar process in the U.S. coupled with a single permit process would cut-1 to 2 years off of current plant-schedules. The'second area of concern and unwarranted delay is in the area of seismic or earthquake design. Currently such efforts add 10 to 12 percent to typical total-plant costs and because the greatest effort is applied to plant piping design which normally is on the plant critical path this adds another 1 year to plant schedule. Based on recent evaluations of true earthquake hazard potentials it should be possible-to dramatically reduce the analytical efforts such that typical total plant costs due to seismic should be reduced to no more than 4 percent and schedule delay of one-year eliminated without any reduction in the hardware. needed to mitigate seismic effects. In fact relative to high temperature systems there may actually be an overall increase in plant safety and reliability. Table 1 is an example of observed seismic damage and failure. in power plant piping due to earthquake effects. Finally the current document intensive quality assurance procedures used in plant design and construction must give way to performance oriented quality assurance programs. Unfortunately the term " quality assurance" has been taken as synonymous - with " quality". As a result any negative criticism of quality assurance is seen as an attack on quality. To attack quality assurance as it is currently practiced in the U.S. is seen as the equivalent of an attack on s

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motherhood. n Strangely enough quality assurance as practiced in the U.S. today may actually lead to less quality. Recently there has been great publicity given to some cases where material certifications used in nuclear plant compcients ure'.e falsified. Unfortunately, the quantity of paper or documentation required to 4-assure quality in nuclear components typically increased.the cost ci.hp.c component several times the cost of the same component procured to normal commercial standards when in fact they are both manufactured in the same manner. . As-a result, there is a great profit potential insfalsifying documents. That such falsification is in fact occurring should come as no surprise and is the ~ direct result of irrational quality assurance policies and regulations.which handsomely reward the individual who falsifies the documentation. As a caso in point structural steel constructed to the requirements of the American Institute of Steel Construction used in all conventional building structures, typically costs 90 cents a pound in place. Structural steel ~ constructed to the requirements of the ASME Boiler and Pressure Vessel Code Subsection N-F which has essentially the same design requirements as AISC except that it costs more than $4.00 per pound in place. The differences in cost is due primarily to quality assurance documentation requirements for ASME Boiler and Pressure Vessel Code Section III, Subsection NF compared to the AISC. In the author's opinion by far the best quality assurance policy is that practiced in nuclear plant construction and design in W. Germany which practice performance oriented quality assurance in nuclear plant construction. In N. Germany an essentially independent audit is performed by the regulatory agencies prior to issuance of a construction permit to assure the design and construction has developed the necessary procedures and have qualified personnel capable of , performing and executing the design and following construction procedures. Periodically during the active design and construction period they audit in detail, on a small sample basis, the design and construction activity to insure correct procedures are available and are being followed. Finally and perhaps-most importantly, as any deviation from "as designed" to "as built" is detected it is resolved informally by competent engineering personnel in the field. This final aspect of quality assurance in the U.S. requires the preparation of ,s

extensive formal documentation in the form of Non-Conformance Reports which generally must go back to the engineering main office for resolution. It is estimated that approximately one third of the total engineering in a single nuclear power station (3,500,000 manhours) and at least an equal amount ing construction craft hours are spent in resolving these 10's of thousands of Non-Mi: b Conformance Reports. The vast majority of these non-conformance reports do not.' effect the design or construction adequacy of the plant. This preparation and resolution of Non-Conformance Reports are activities which essentially do not exist outside the U.S. In addition to the extra manhours spent these activities typically add 3-4 years to the plant construction schedule. CONCLUSION Nuclear power in the U.S. is presently 15-25 percent more expensive than equivalent coal generated power and takes 5-7 years longer to construct then coal generated power. In most of Western Europe and Japan the cost statistics are just the reverse with no significant difference ir. the hardware that is actually installed in the plants between the U.S., Western Europe and Japan. Nuclear power will again only become a viable new power source in the U.S. when regulators and industry standard writers address a) the streamlining of licensing process, b) simplification and rationalization of seismic design and c) substitution of performance for document quality assurance. The current emphasis on smaller plant sizes (i.e. 600 versus 1200 MWe), passive safety systems (i.e. containment spray gravity fed from the Refueling Water StorageTanklocatedinthecontainmentdome) and standardization featured in current and developing advanced power reactor design will help in achieving the goal of more reliable and economical nuclear power. However, these improvements without explicitly addressing the seismic and quality assurance issues identified. in this talk will not in themselves reach the goal of cost competitive reliable nuclear power in the foreseeable future. /'

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{nq N Department of Cooley Building, North Campus College of Engineenng theeleanSegintpdffspec Ann Artsor. Michigan 48109 2104 The University of Michigan 311764 4260 FAX: 311763 4540 November 13, 1989 ' ~ ~ Mr. Paul Boehnert Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission Washington, DC 20555 NOV 1 ne9

Dear Paul:

This is written to offer my comments and observations regarding the Thermal Hydraulic Phenomena Subcommittee meeting on November 8 and 9,_1989. I am generally encouraged by the approach that the NRC staff and industry have taken to tackle the boiling water reactor (BWR) stability issue, but feel that increased attention should be given to this issue within the framework of anticipated transients without scram (ATWS). Specific comments are offered in the following, essentially as reiteration of the items I touched upon at the meeting:

1. The accuracy in our prediction and analysis of the nuclear-coupled density-wave oscillation (NCDWO) phenomena, including estimation of-the peak power level, is still limited and will likely remain so for some time in the future.

This pessimistic opinion is based on two observations: (l')' The nonlinear NCDWO phenomena are highly sensitive to small variations in initial conditions or system configurations. This was made evident in the Leibstadt test. where a variation of 3% of rated power and 1% of flow rate resulted in a factor of four difference in the oscillation amplitude of local power range monitor readings. Similar sensitivities to void coefficient of reactivity and form loss coefficients of boiling channels have been noted. (ii) Additional effort will be required to validate physical models for complicated oscillatory phenomena involving reverse flow and boiling boundary movement. Numerical algorithms for transient nuclear-thermal-hydraulic calculations also require additional validation and optimization effort. For these reasons, NCDWO calculations involving peak power amplitudes greater than 200 ~ 300% of rated should be viewed with considerable skepticism.

2. Further effort will be required to study the effects of NCDWO events superimposed on some other ongoing transients and the pos.sibility of inadvertently inducing severe transients as a result of NCDWO control maneuvers.

In particular, we should understand better the importance of the 0.03-Hz feedwater transient on the 0.4-Hz NCDWOs in the LaSalle Unit 2 event.

i 9 ?. 2 In terms of a potential ATWS event, if-the LaSalle-operators had been successful in restarting the-recirculation pumps, i .thereby adding positive reactivityEto an ongoing power pulse, how serious would the resulting transient have been?

3. In light of the uncertainties and sensitivities inherent in our NCDWO analysis and the ATWS significance of NCDWO events, I suggest more effort should be made to enhance the-BWR stability.

I was encouraged by efforts under consideration in.this direction for the Advanced BWR design discussed at the meeting but would like to see similar effort made for operating BWRs to the extent possible. 4. If we were to account for the uncertainties-in the NCDWO phenomena properly, the exclusion region in a BWR power-flow map would increase substantially, resulting in frequent reactor scrams or control rod insertions. To avoid this reduction in the maneuverability of BWRs, one might consider adding another parameter to the power-flow map. One parameter that may be useful is the axial offset (AO) of power widely used for pressurized water reactors. The AO is defined as the normalized difference between the reactor power in the top half of the core and that in the bottom half, and provides a simple indication of the axial power distribution. Such an additional 1 parameter may account largely for the NCDWO sensitivity to axial power distribution and may help maintain the exclusion region to a reasonable size. This suggestion should be considered together with stability monitors currently under study.

5. To validate the coupled nuclear-thermal-hydraulic _models for NCDWO calculations, including the choice of spatial mesh and time steps, I wish to suggest that we make use of reactivity accident tests as benchmark cases.

One such test is the SPERT experiment (R. K. McCardell, D. I. Herborn, and J. E. Houghtaling, " Reactivity Accident Test Results and Analyses for the SPERT IIIE Core - A Small Oxide-Fueled, Pressurized Water Reactor," IDO-17281,-Phillips Petroleum Company (1969)}. I hope the above comments are useful to the Committee. Yours sincerely, n C. Lee Professor xc: I..Catton i

[ * : :, ' '" g 7.- / 9 5 9 QUERYTECH ASSOCIATES, INC. 9040 EXECUTIVE PARK DRIVE, SUITE 217 ~ KNOXVILLE, TN 37923 Phone (615) 690 2728 November _17,- 1989 U. S. Nuclear. Regulatory Commission Advisory Committee on Reactor Safeguards L Mallstop P-315 L Washington,. D. C. 20555 - y l Attention:. Mr. Herman Alderman

Dear Herman,

Enclosed is a copy of my comments on the Nine Mile Point Unit 1 restart b ~ meeting of November 14, 1989. Since I do not have most of the NRC ~ documents _' developed during the. staff review of Nine Mile Point Unit-1 -t some of the suggestions may have been covered by earlier material. If there is a need for any further assistance, please call me. Sincerely, M. Bender Copy to Professor Wm. Kerr. l o 2 l +

,;.,n.'.. - COMMENTARY ON NINE MILE POINT UNIT 1 RESTART PREPARED BY M. BENDER, QUERYTECH ASSOCIATES, INC. November 16,1989 GENEB&L. OBSERVATIONS The discussions at the November 14 meeting on the Nine Mile Point Unit 1 Restart primarily highlighted management issues. The Niagara Mohawk presentations largely emphasized new organizational arrangements for operating Nine Mile Point Units 1. The new management alignment puts more emphasis on organizational responsibility through intensive review and self assessment. These management enhancements should improve the Nine Mile Point success prospects. The Nine Mile Point Unit 1 presentation seemed to downplay technical aspects of organizations activities in favor of more attention to management style. Although management technique is important to the success of any activity, the Nine Mile Point Unit 1 posture was dangerously close to encouraging iess attention to technical matters as secondary to management techniques. This is clearly'an undesirable action since most of the concerns with Nine Mile Point Unit 1-stem from - failure to pay enough attention to technical details. ROOT CAUSE ANALYSIS. Root cause analysis is currently the pet appraisal technique for determining the reasons for deficiencies in licensee performance. The " root causes" are not directed to specific weaknesses in functional activities but rather to foibles commonly found in deficient management organizations e.g. poor. communications, poor attitude fostered by the cultural ~ environment, lack of teamwork, poor planning. In the general sense that a well managed and properly organized operation will display 1

~. '.~ NINE MILE POINT UNIT 1 RESTART BENDER, 11/16/89 fewer problems of this type, root cause analysis aimed at such behavioral matters may be helpful. Although there is nothing wrong with this type of evaluation, it does not really pinpoint the corrective actions needed to prevent recurrence of observed problems, it would not be helpful in attacking technological issues beyond the ken of the organization. For_ example, stress corrosion cracking -is a chronic problem in BWR primary systems. The corrective action has to deal with corrosion mechanisms active in the operational environment. Finding the " root cause of this type of problem" does not come within the scope of " root cause analysis" as applied to Nine Mile Point Unit 1. Neither would incomplete documentation or erroneous computations be corrected by eliminating these " root causes". TECHNICAL ISSUES The presentations at the Nine Mile Point Unit 1 meeting were so shal!ow that it was impossible to glean any substance about the problems resolved during the lengthy shutdown. Among the matters that deserve review are: stress corrosion and related repair or replacement of degraded primary coolant system boundary equipment (piping, valves, pumps) vibration sources of the type that initially led to the shutdown of Nine Mile Point Unit 1 corrective actions associated with containment deficiencies-including the current concern about torus thinning modifications in the reactivity controls including scram system modifications and liquid poison injection provisions a historical review of operational problems found in Nine Mile 1 2

, /..<. 0 NINE MILE POINT UNIT.1 RESTART BENDER, 11/16/89 Point Unit 1 since its startup to be sure that all significant safety problems have been addressed it_would be worthwhile to compare the status of Nine Mile Point Unit 1 to that of Millstone.1 which was examined by an Integrated Plant Safety Assessment (See NUREG 0824) prepared in the 1982/1983 era. MANAGEMENT ISSUES Of the matters covered in the discussion of new management initiatives only two deserve special comment: 1. The " restart review panel" is unusually experienced in 'its applied understanding of operational problems. It has the right kind of personnel balance for an overview function and should give effective guidance to the Nine Mile Point Unit 1 operational staff. 2. The engineering organization is -taking on more direct responsibility for engineering work associated with Nine Mile Point Unit 1 operational needs. While skill capabilities have to be' carefully monitored this should result in more responsible control of the technical problems requiring resolution.- Most nuclear utilities are too -dependent on.outside service organizations and are not sufficiently skilled to monitor the effectiveness of the services provided. Other licensees should be encouraged to follow this precedent. Other matters such as improved teamwork and more focused " problem solving" are " givens" in any successful management setup. Any new organization will promise more attention to these management skills, but only long term results provide any measure of their adequacy. 3

.,e, -,q' NINE MILE POINT UNIT 1 RESTART BENDER, 11/16/89 The present trend in nuclear power industrial organizations is to include - an extensive auditing and checking function that sometimes interferes with - organizational effectiveness. Usually the audit personnel are not sufficiently well versed in technology to determine quality deficiencies, but they can' provide a " tickler" service that will avoid some inadvertent oversights. 4

j CT-1955 0 UNIVERSITY OF CALIFOHNJA. BERKELEY n mr.v. mu wm. u, neraLwn m. moo sa macun um s.4=4u um cas COLLEGE OP f NOINERRIN0' BEREBLBY. CAUFORNIA 94730 IEL$ PHON 45642 0 Hoveniber 17,1989 FAX: (415) 643 Mas ' Dr. Ivan Catton NOV 2 0 ff# Chairman, Thermal Hydraulle Phenomena Subcommitee Advisory Cornmittee on Fleactor Safeguards U. 8. Nuclear Regulatory Commiss.lon Washington, D.C. 20555 ATTN: Paul Boehnert

Dear Ivan,

RE: Subecmmittee Meeting November 6 & 9 San Francisco Airport You requested rny comments on the meeting and the related background material. I have been ruled in confilet of interest for GE and INEL so my comments should be taken with that in mind. Also, I was able to attend the meeting on November 8 only. Regarding the TRAC G code capability to address the BWR stability issue, I have the following thoughts and questions. First, I thought that the GE presentation was exceptionally well planned and as a result this was the most informative and productive Interactions of a vendor with ACR8 that I can recall. I say this, not because I'm Interested in seeing GE get some brownle points, but because I have been quite concerned abo.pt the euality of technical content of presentations made to the committee both by the adustry and the NRC staff. I hope you understand why this happened and can entice othere to do as well. In his review of the background of TRAC-B codes, Dr. Shiralker made the point that at the conclusion of their joint NRC funded program with EG&Q (1984) several models were offered to EG&G for the NRC code, but they were not implemented. I believe these were listed as: Hot rod model 3 D Kinetics Numerloal efficiency My recollection of the 3 D kinetics issue in 1984 is that R was NRC-RES that made the decialon not to include 3 D kinetics in TRAC-8. I was rooommending to EG&G 1

c.E ELEY NUCLEAR TEL Nt.4156439685 Nov 20.89 11:54 No.006 P.03 2-t that it was needed if the code was to handle scenarios where space time kinetics are important. How CE appears to oun a better code for a small fraction of the cost, it is another exemple of what I call a " reactive mode" of planning at RES. Concerning the TRAC "non conservative momentum equations",I'm not sure that this is e defloiency peculiar to the stability problem. It is a problem in general for the codes and I'm not convinced that anyone really understands its impact. They are always valldeted against very global measures and these results are often dominated by comportssting errors in the constitutive models. An in depth study of this would be a good task for someone like W. Wultf. The assumption of quaal steady drift flux parameters for stability analysis I believe is questionablo. This was used to apply Ishil's drift flux oriented Interfaelal drag correlations to the two fluid TRAC models. I don't think this 18 very sound where 0.5 H, oscillations are present. I think there will be very signifloant profile distortions and local slip will deviate from the steady state. However, I recognize that there are no rollsble data to carefully test this premise. Uke it or not, the "translent" thermal hydraullos codes are quasi steady tools. To make them otherwise would require experimental research that RE8 has felt is unnecessary. I don't agree. I thought the data f rom a top blowdown experiment (PSTF Test 5901-15) was not a good choice for the assessment of the predictor-corrector method. That pressure history la strongly influenced by the details inside the vessel where the code appiloation uses very coarse noding. Jens Anderson thought exit temperature fluctuations seen in code results but not in the data are caused by a computational interaction between droplet concentration and host transfer. This seemed quite vague and I would agree with the comment made by John Lee that this difference ought to be carefully examined. B. Z. Rouhant's point about the experimental problem is correct but I don 1 think this provides a definite explanation, in the broad view the GE 3-D kinetles model as implemented is fine node axlat and course node radlal. The 700 plus channels are grouped into 20 transverse nodes. Thus, although GE concludes that ' TRAC-G prodlota regional osolliations observed under test conditions with no extemal forcing porturbation" they also acknowledge that it is necessary to use control rod pettoms to guide selection of noding (appropriate for large numbers of channels to be represented by a single neutronic and T H characterization) for suooessful almulation. This is something less than a full predictive capability. It would be interesting to know what the code could do if 730 transverse nodes could be run. The discussions on numerloal diffusion and the need for higher order numerica in the stability problem were informative but there la not a clear concensus and I gather that more will be done on this by both GE and EO&G. l

J C E. f:ERI.ELEY NUCLEAR TEL Nc.4156439685 Nov-20,89 11:54 No 006 P.04 L,y ?:? s tr

C 3-The TRAC BF1 capabilities (1 D kinetics) are clearly inferior to those of TRAC-G. Still, it appours that it will be of some help in meeting NRC's needs.

The conclusion Iri the presentation by Wilson that the data bate le Insufflolent for assessment of limit cycle amplitude is disturbing as is the conclusion attributed to Niarch LEUBA conoornig bifurcation and chaotto regimes. I presume we will revisit the stability problem at a future meeting, I am encouraged that there does appear to be a strong drive to get a good technical solution. Sincerely, i p v VWgli E. Schrock Professor VES/jmh

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SANTA BARBARA e 6ANTA CRUZ BERKELLY e DAYli e IRVINE e LOS ANGEL.tS e klVERb1D 'Ya"%I!M&'of"#i'#" November 22,1989 Mr. Paul A. Boehnert ACRS Nuclear Regulatory Commission Washington, D.C. 20555 RE: ACRS Thermal Hydraulic Phenomenn Subcommittee Meeting, November 8-9,1989, San Francisco -

Dear Paul:

This constitutes my report on the above-referenced meeting, which reviewed the capability of the thermal hydraulic codes to model the BWR core power instability phenomena. - Since I only attended the first-day (November 8) sessions and I am also refrained from commenting on General Electric work due to potential conflict ofinterest, my comments below are relatively brief and only on the INEL and BNL work.

1. The._ technical level of TRAC BWR is favorably compared with' the general level. for existing modeling capabilities of T/H codes.

The work in-validation / verification and assessment oflimitations appears to be competently carried out.

2. My only slight reservation lies in the lack of more definite knowledge on the use of existing steady two phase flow and heat transfer models and correlations in unstable / transient situations. It does not appear to pose any major problem, but more qualification work is needed.

._ 3. For limit cycle instability and bifurcation, more definitive data should be - available to validate existing code capabilities. Sincerely, Chang ,in Tien Executi re Vice Chancellor and UCI Distinguished Professor a s

} (, o 4 '- u CT-/951 P. R. Davis Dec.IS,1989 RE: Interf acing Systems LOCA and ORNL Precursor Study ( Bill,. Our susicions are apparently confirmed! You may recall that at the ISLOCA meeting on Dec. 7. Gary Burdick of the Staff made some comments to me regarding how the precursors were selected in the ORNL study. His description of the selection process and the same process described in one of the earlier precursor reports (II does not seem consistent. The report (Pg. XXIV) describes the process as: - If the event involved the f ailure of at least one function required to mitigate a loss of main feedwater, loss of offsite power, small break LOCA, or steam line break; - If the event involved the degradation of more than one function required to mitigate one of the above initiating events; - If the event involved an actual initiating event that required safety function response" It is not clear from this description, however, if the ORNL Precursor Study would pick up ISLOCA precursors. It seems perhaps not, since, for example, the discovery of a f ailed check valve in the LPIS injection line would apparently not meet any of the above requirements. (this will suffice for a christmas card) Pete n. ..........-_...... _ _}}