ML20213A536

Summary of ACRS 316th Meeting on 860807-09 in Washington,Dc.Viewgraphs Encl

ML20213A536
Person / Time
Issue date:	01/28/1987
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2454, NUDOCS 8702030332
Download: ML20213A536 (124)

Text

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TABLE OF CONTErlTS MINUTES OF THE 1:

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h 316TH ACRS MEETING '

AUGUST 7-9, 1986 h[ g g l'D21/MN1 I. Chairman's Report (0 pen)................................... 1 II. San Onofre Nuclear Generating Station, Unit 1, Water Review (0 pen)............................................... 2 III. Operating Experience at Nuclear Power Stations (0 pen)....... 8 IV. Management and Disposal of Radioactive Wastes (0 pen)....... 11 V. Human Factors Issues (0 pen)................................ 13 VI. Seismic Qualification of Equiprent (0 pen).................. 18 VII. Standardized Nuclear Plants (0 pen)......................... 22 VIII. Improved Light Water Reactors (0 pen)....................... 23 IX. NRC Regulatory Process (0 pen).............................. 27 X. Lono Range Plan (0 pen)..................................... 29 XI. Executive Sessions (0 pen).................................. 30 A. Subcommittee Assignments................................. 30

1. Conduct of ACRS Meetings............................ 30 B. Reports, Letters, and Memoranda.......................... 30

1. ACRS Comments on the Standardization Policy Statement........................................... 30

2. ACPS Report on the Tennessee Valley Authority's (TVA) Vanagement Reorganization and Shutdown of TVA's Nuclear Power Plants.......................... 30

3. ACRS Coments on the NRC Policy Statement on Fitness for Duty of Nuclear Power Plant Personnel... 31

4. ACRS Actions on Reaulatory Guide 1.114, Revision 2,

" Guidance to Operators at the Cortrols and to Senior Operators in the Control Room of a Nuclear Power Unit"......................................... 31

5. ACRS Comments on Various NMSS and RES Waste Manage-nent Topics......................................... 31 g20gg2 07012s ACRS-2454 PDR

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Q 316TH ACRS 11EETING 11

6. ACRS Comments on Loss of Power and Water Hanmer Event at San Onofre, Unit 1......................... 32 C. Generic Issues........................................... 32

1. ACRS Comments on the Resolution of USI A-46 (Seismic Qualification of Equipment in Operating Plants)............................................. 32

2. Improved Light Wate r Reactors . . . . . . . . . . . . . . . . . . . . . . . 32

3. Design Performance of Containments.................. 33 D. Future Agenda............................................ 33

1. Future Agenda....................................... 33

2. Future Sub :ccmi ttee Activi ties. . . . . . . . . . . . . . . . . . . . . . 33 E. Proposed Power Level Increase for the North Anna Nucl ear Station , Uni ts 1 and 2. . . . . . . . . . . . . . . . . . . . . . . . . . . 33 F. RSK Report on the Chernobyl Accident..................... 33

4 iii TABLE OF CONTEtlTS APPENDICES TO MINUTES OF THE 316TH ACRS MEETING AUGUST 7-9, 1986 Appendix I - Attendees........................................... A-1 Appendix II - Future Agenda...................................... A-6 Appendix III - ACRS Subcommittee Meetings........................ A-8 Appendix IV - NRR Presentation on SONGS-1 Water Hammer Event..... A-13 Appendix V - Recent Significant Events........................... A-45 Appendix VI - NRC Presentation - USI A-46 ACRS Presentation...... A-56 Appendix VII - Standardization Policy............................ A-64 Appendix VIII - Notes by W. Kerr on Improved Light Water Reactor. A-66 Appendix IX - Regulatory Process................................. A 68

) Appendix X - NRC Strategic Planning.............................. A-72 Appendix XI - Additional Documents Provided for ACRS Use......... A-80

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h& kYbNc Federal Regist:r / Vcl. 51, N::.149 / M:nd:y, August 4,1986 / Notices -

1 By the Commission. Paul S. Cross.

Administrative Law judge.

(202) 523-7894. Written comments / November 18,1986: Technological and received, as well as other informatio'n in

/

Noreta R. McGee. Docket H-040, will be availably for Economic Feasibility; December 9,1986: Permissible ExposurV

/

Secretary. inspection and copying at this address, Limits, Including Short Term Expoenn (FR Doc. 86-17400 Filed 8-1-88. 8.45 am) Monday through Friday Srl'5 a.m. to 4:45 Limits and ActionlavalDimena=fons, suas coes ren.ei m p.m.

POR FURTHER INFORefATION COtrrACT:

/ and January 13,1967: Review of fdmmittee

/

Mr. Tom Hall, Division of Consumer Recommendations for DEPARTMENT OF LABOR Affairs, Occupe'tional Safety and Health Register Publication. ):sideral Administr ifbn.U.S.Deparunent of Minutes of these mIeetings will be Occupational Safety and Health Labor, m N-3637,200 Constitution available for pu)lfc inspection at the Administration Aven NW., Washington, DC 20210; OSHA Docke} Office, U.S. Department Tel one (202) 523-8615. of Labor. Rnt' N-3670. 200 Constitution n'e7nt't,"2"y'#ase -A-- On tober 22,1985. OSHA announced its

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Telep ne (202) 523-7894. j/

AGENCY: Occupational Safety and intent to make use of mediated s ed in Washington. Dethis 30 day of Health Administration, Labor. rulemaking in developing a proposed i 1988.

ACTloN: Notice of meetings and agpddas. standard for MDA (50 FR 42790-42793). ohn A.Pendergrass'y The notice also set forth the basic AssistantSecsefaryoff., abor.

suMuany: Under the provisforpr,of the concepts of mediated rulemaking and Federal Advisory Committer'A' ct (Pub. (Ht Doc. 17532 Filed 7-31-46; 11:44 sm) outilned the participant selection 92-463, as amended). notice is he eby * ' ' ' "

criteria which OSHA expected to e in ,

given of the schedule of six Committee establishing an MDA Mediated -

meetings to be held from August 1986 Rulemaking Committee.

through January 1987. Notice is also OSHA established the co intee in "

given of the tentative topics of accordance with the Fed 1 Advisory discussions. It is anticipated that the Committee Act(FACA d section 7(b) Advfsory Committee on Reactor meetmas willlast from one to three days of the Occupational ety and Health Safeguards; Meeting but this ay vary as the work of the ate issues Com tee proceeds. For the purpose of Act (OSH Act) to associated with development of a in accordance with the purposes of this otice only the beginning dates will Notice of Proppsed Rulemaking on sections 29 and 182b.of the Atomic b given. Locations of the meetings are MDA. Energy Act (42 U.S.C. 2039,2232b) the so provided in the notice. Information Appointee s to/ the committee include Advisory Committee on Reactor on room numbers will be available in represe tives from labor, industry, Safeguards will hold a meeting on the lobbies of the designated buildings- heal nd safety groups, and August 7-e.1986, in Room 1046,1717 H carts:The meetings are scheduled to gov nment agencies. Street NW., Washington, DC. Notice of begin on: embers of the public wishing to this meeting was published in the August 5.1986 at 9.30 a.m. at the liyatt bmit written statements to the Federal Register on July 24,1986.

Regency Washington. 400 New Jersey Conimittee that are germane to the nursday, August 7.1986 Avenue NW., Washington, DC 20001 agenda may do so. Such statements (202) 7J7-1234; should be in reproducible form and &J0 AAf.-Ac AJf. Report ofACRS September 9,1986 at 9:30 a.m. at the should be submitted to the OSHA hairman (Open)-%e ACRS Chairman 1.ittle American Hotel. 500 South ain Division of Consumer Affairs at least 5 will report briefly regarding items of Street. Salt Lake City. Utah 841 . days before the meeting. In addition, the current interest to the Committee.

(801) 363-6781: Mediatcr or Chairman of the Committee

&c AJf.-Jac AJf.: Standardized October 7,1966 at 9:30 a.m. In e has the authority to decide to what / Nuclear Plants (Open)--De members of Frances Perkins Departme of Labor extent oral presentations by members of the Committee will discuss proposed Building. 200 Constitutio Avenue i ACRS comments and recommendations NW., Washington. DC 210: the public may be permitted at the/

meeting. to the NRC regarding a proposed pohey November 18.1986 at 9 a.m in the At the first meeting held in / statement by the Nuclear Regulatory Department of Lab Du!! dins Washington on July 22-23.1989' the Commission on standardized nuclear indicated above: topics and the order of their dfscussion plants. Members of the NRC Staff will December 9.1966. 9.30 a.m. at the were established. For the p pose of 29 participate as appropriate.

Phoenix Park i tel. 520 North Capitol CFR 1912.28 these constitu the 112 AJf.-Jm PAf.t /mprovedlight Street NW.. \ ashington DC 20001. Agendar for the meeting nd are as WaterReactors (Open)-The members r

(202)638-epo and topaws: will discuse proposed Committee January 13. 087 at 9:30 a.m. In the comments and recommendations to the Depart nt of Labor Building Au ' pe d pplication.

D fin ijn co NRC regarding proposed characteristics Indicat d abose.

Emergencies. Hygtene Facilities and forimproved light water react, ors.

Stat : These meetings will be open to 2m PJf.-Jm PJf.:Sessnue

the p lic. ifousekeeping

/

September 9,1986: Personal Protective Qualification ofEquipment(Open/

AO ess: Submissions presented in Equipment. Fjposure Monitoring, and Closed}-The members will hear re onse to this notice should be sent in site visit: presentations as appropriate and will adruplicate to the Docket Officer. October 7.  : Health Effects. Risk discuss proposed ACRS comments and ocket No. H-040. Room N3670. Assess t Medical Surveillance, recommendations regarding the Occupational Safety and Health MedicgAppendices. Biological proposed program to seismically qualify Administration. 200 Constitution Monl% ring, and Removal and Rate safefy related equipment in operating Avenue NW., Washington, DC 20:10: Retention; nuclear plants. Representatives of the

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. ) 8 27920 Federal Register / Vol. 51. No.149 / Mondry. August 4.1986 / N:tices 1215P.AL-1230P.Af.: Future ACRS Procedures for the conduct of and NRC Staff will take part in this session participation in ACRS meetings were as appropriate.

Activities (Open/ Closed) *Ihe members will discuss anticipated published in the Federal Register on Portions of this session willbe closed subcommittee activities, items proposed October 2.1985 (50 FR 191). In as necessary to discuss Proprietary for consideration by the full Committee, accordance with these procedures, oral l Information applicable to this matter. or written statements may be presented J:00P.Af.-4:fMPJf. NRCRegulatory and proposed activities of Individual ACRS members.The proposed schedule by members of the public, recordings Process (Open}-ne members will for full Committee rneetings during CY will be permitted only during those discuss proposed ACRS comments and recommendations to the NRC regarding 1987 will also be discussed. portions of the meeting when a reevaluation of the nuclear regulatory Portions of this session willbe closed transcript is being kept. and questions as necessary to discuss information the may be asked only by members of the process. Committee. Its consultants, and Staff.

4:1d PJL-d 30PJf.:Long Range release of which would represent a clearly unwarranted invasion of Persons desiring to make oral Planning (Open}-The members will hear and discuss the report ofits personal privacy. statements should notify the ACRS subcommittee on a proposed guide for 1:30 PJf-J 00 PJf. Numan Factors Executive Director as far in advance as the preparation of a long range plan for Issues (Open}-ne members will hear practicable so that appropriate NRC activities, and discuss reports from its arrangements can be made to allow the subcomrnittee and representatives of the necessary time during the meeting for Friday. August 8.1988, NRC Staff as appropriate regarding such statements. Use of still, motion 8:30 AJf.-0:30 AJf.r Activi, ties of the proposed activities related to p cture and television cameras during A RC Office of Nuclear Afaterial Safety consideration of human factors, this meeting may be limited to seIected andSofeguards (Open/ Closed}-The including fitness for duty requirements, p rtions of the meeting as determined members will hear a briefing by the guidance for nuclear power plant by the Chairman. Information regarding Director. NMSS.regarding NMSS operators and senior reactor operators, the time to be set aside for this purpose activities of mutualinterest. Includin8 and educational requirements for senior may be obtained by a prepaid telephone safeguards and security at nuclear reactor operators. call to the ACRS Executive Director, power plants, fuel cycle facilities, and 3:15 PJf.-5:15 P.Af.: Operating R.F. Fraley, prior to the meeting. In view nuclear waste processing, storage and Erperience (Open)-The members will of the possibility that the schedule for repository facilities. hear and discuss the reports of its ACRS meetings may be adjusted by the Portions of this session willbe closed subcommittee and representatives of the Chairman as necessary to facilitate the as necessary to discuss a licensee's or NRC Staff regarding recent operating conduct of the meeting persons applicant's detailed security provisions experience and incidents at nuclear planning to attend should check with the at facilities of the types bein8 facilities.

considered when the public disclosure ACRS Executive Director if such 5:15PJf.-6:30PJE 'I'VA rescheduling would result in malor of such information could reasonably be Reorganization (Open)-The members expected to have a significant adverse of the Committee will discuss proposed inconvenience.

effect on the health and safety of the I ha.e determined in accordance with ACRS comments and recommendations public or the common defense and regarding the proposed reorganization of subsection 10(d) Pub. L 92-163 that it is

~* *'** # *' "'*'

Af.-10:15 AJf.: Afanagement meet g oe a todsc s andDisposalofRadioactive Wastes Saturday, August 9,1986 Proprietary Information [5 U.S.C.

(Open)-The members will hear and a:30 A.Af.-12:30 PJf.: Preparation of 552b(cl(4)] app!! cable to the facilities ACRS Reports to the Nuclear being discussed. detailed information n ic re t d to r los vwae RegulatoWommission(O en/ related to the security arrangements at a management and disposal. including Closed)-ne members wil discuss, nuclear power plant [5 U.S.C.

residual radiation limits for the pr p 5ed reports to the NRC regardmg 552b(c)(3)] and information the relea:e disposition ofland, buildings, matters considered during this meeting. of which would represent a clearly equipment, and metals resulting from in addition. the members will discuss unwarranted invasion of personal the decontamination and proposed ACRS reports on safety

• privacy [5 U.S.C. 552b(c)(6)].

decommissiening of nuclear power related matters such as aptitude testing Further information regarding topics plants and fuel facilities, salvaging of f melear power plant personnel, to be discussed, whether the meeting contaminated smelted alloys, and the has been cancelled or rescheduled, the NRC radioactive waste management Portions of this session willbe closed as required to discuss Proprietary Chairman's ruling on requests for the program. Information applicable to the matters 1015 AJf.-12:15 P.Af.r San Onoff, opportunity to present oral statements being discussed. and the time allotted can be obtained by NuclearPbwerStation Unit 2 (Open/

2:30 P.Af.-J:00 PJf.: Activities of a prepaid telephone call to the ACRS Closed)-%e members will hear and ACRS Subsommittees (Open}-ne Executive Director. Mr. Raymond F.

discuss reports ofits subcommittee and NRC Staff representatives as members will hear and discuss the Fraley (telephone 202/634-3205),

appropriate regarding changes in the reports of ACRS subcommittees between B.15 A.M. and 5:00 P.M.

San Onofre Nuclear Station resulting regarding assigned activities on Dated: July 29.1986.

• • from the November 21.1985 loss of radioactive waste management and g

feedwater at this facility. disposal in geologic repositories, nuclear Samuell. Chilk.

Representatives of the licensee will power plant scram system reliability. Acting Advisory Committee Manesement participate as appropriate. degraded primary system piping. off,c,f, Portions of this session will be closed management of ACRS activities, and

[FR Doc. 86-1750t Filed 6-1-86. 8 45 am) as necessary to discuss Proprietary procedures for conduct of ACRS sw.o coot rimei-u Information applicable to this facility. activities.

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. t r 8 citruq% UNITED STATES E i, NUCLEAR REGULATORY COMMISSION

, $ ,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o, c WASHtNGTON, D. C. 20555 n ,,.

Revised: August 1, 1986 .

SCHEDULE AND OUTLINE FOR DISCUSSION 316TH ACRS MEETING AUGUST 7-9, 1986 .

WASHINGTON, D. C.

Thursday, August 7, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.

1) 8:30 - 8:45 A.M. Report of ACRS Chairman Open) 1.1) Opening Statement DAW) 1.2) Itemsofcurrentinterest(DAW /RFF)

2) 8:45 - 10:45 A.M. Standardized Nuclear Plants (0 pen) 2.1) Discuss proposed ACRS report on proposed NRCPolicyStatementregarding) ardized Nuclear Plants (CJW/HA Stand-10:45 - 11:00 A.M. BREAK

3) 11:00 - 1:00 P.M. Improved Light Water Reactors (0 pen) 3.1) Discuss proposed ACRS comments regarding characteristics of improved LWRs (D0/RKM) 1:00 - 2:00 P.M. LUNCH

4) 2:00 - 3:15 P.M. SeismicQualificationofEquipment(0 pen).

4.1) Report of ACRS Subcommittee regarding seismic qualification of safety-related equipment in nuclear power plants (CJW/RKM) 4.2) Discussion with NRC Staff, as appropriate

5) 3:15 - 4:15 P.M. NRCRegulatoryProcess(0 pen) 5.1) Discuss proposed ACRS report on the NRC regulatory process (WK/GRQ) 4:15 - 4:30 P.M. BREAK

6) 4:30 - 6:30 P.M. LongRangePlan(0 pen) 6.1) Discuss proposed ACRS coments regarding a guide for the preparation of a long range plan for NRC activities ..

(MWC/RKM) v./ae b'S

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. l 316th ACRS Meeting Agenda '

Friday, August 8,1986, Room 1046,1717 H Street, NW, Washington, D.C.

7) 8:30 - 9:30 A.M. Improved Light Water Reactors (0 pen) 7.1) Continue discussion of proposed ACRS 1 coments on characteristics of improved 1 LWRs .

8) 9:30 - 10:15 A.M. Management and Disposal of Radioactive Wastes (0 pen) 8.1) Report of ACRS Subcommittee regarding Sub-comittee meeting on July 21-23, 1986 on proposed residual radiation limits for disposition of land, buildings, equipment, and metals from the decontamination and decommissioning of nuclear facilities; salvaging of contaminated smelted alloys; and the NRC radioactive waste management program (DWM/OSM) 8.2) Discussion with representatives of the NRC Staff, as appropriate 10:15 - 10:30 A.M. BREAK

9) 10:30 - 12:30 P.M. San Onofre Nuclear Station Unit 1 (0 pen / Closed) 9.1) Report of ACRS Subcommittee regarding corrective action taken as a result of the 11/21/85 loss of feedwater incident (GAR /MDH) 9.2) Meeting with NRC Staff and Licensee, as appropriate (Portions of this session will be closed as necessary to discuss Proprietary Information applicable to this matter.)

10) 12:30 - 12:45 P.M. ACRS Future Activities (0 pen) 10.1) Anticipated ACRS Subcommittee activity (MWL) 10.2) Proposed items for ACRS consideration (DAW /RFF) 10.3) Proposed dates for CY 1987 ACRS full Comittee meetings 12:45 - 1:45 P.M. LUNCH

11) 1:45 - 3:15 P.M. Human Factors Issues (0 pen) 11.1) Report of ACRS Subcommittee regarding proposed Fitness for Duty Policy Statement; proposed NRC rule on Educational Requirements for SR0s; and R.G. 1.114, Guidance for Operators at i

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1 316th ACRS Meeting Agenda ,

the Controls and to SR0s in the Control Room (FJR/HA) 11.2) Meeting with representatives from the NRC Staff 3:15 . 3:30 P.M. BREAK

12) 3:30 - 5:30 P.M. OperatingExperience(0 pen / Closed) 12.1) Report of ACRS Subcommittee re recent events at nuclear power stations (JCE/HA) 12.2) Meeting with representatives from the NRC Staff (Portions of this session will be closed as necessary to discuss Proprietary Information and detailed Security provisions applicable to the facilities being discussed.)

13) 5:30 - 6:30 P.M. TVA Reorganization (0 pen) 13.1) Discuss proposed ACRS report on reorganization of TVA nuclear activities (CJW/RPS) f l

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_m.._ .,_.__,._r_________._ __ _ ___ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ , _ _ , . . _ _ , _ . - _ _ _ . _ . . _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ . . , _ _ _ _ _ - _ _ . .

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316th ACRS Meeting Agenda ~

Saturday, August 9, 1986,. Room 1046, 1717 H Street, NW, Washington, D.C.

14) 8:30 - 12:30 P.M. ACRS Reports to NRC (0 pen / Closed) 14.1) Discuss proposed ACRS reports to NRC regarding:

. 14.1-1) 8:30 - 9:00 A.M. -- Policy State-ment on Standardiled Nuclear Plants (CJW/HA) 14.1-2) 9:00 - 10:00 A.M. -- ACRS comments on TVA Reorganization (CJW/RPS) 14.1-3) 10:00 - 11:00 A.M. -- NRC Regulatory Process (WK/GRQ) 14.1-4) 11:00 - 11:30 A.M. -- Human Factors (FJR/HA) 14.1-5) 12:00 - 12:30 P.M.: Seismic Qualification of Equipment (CJW/RKM) 14.1-6) 11:30 - 12:00 Noon: SONGS Unit 1 (tentative) (GAR /MDH)

(Note: Portions of this session will be l

closed as necessary to discuss Proprietary Information and detailed Securit for the plants being discussed.)y provisions 12:30 - 1:30 P.M. LUNCH

15) 1:30 - 3:00 P.M. Complete Preparation of ACRS Reports (0 pen) 15.1) Discuss proposed ACR5 reports to NRC, as needed, regarding:

15.1.1) 1:30 - 2:00 P.M. -- Aptitude Testing (GAR /HA) 15.1.2) 2:00 - 3:00 P.M. -- Long Range Plan (MWC/RKM)

16) 1:30 - 3:30 P.M. ACRS Subcomittee Activities (0 pen)

(Tentative,iftima 16.1) 1:30 - 2:00 P.M. -- Scram System Reli-permits) abilitySubete. meeting 7/31/86(WK/PAB) 16.2) 2:00 - 2:30 P.M. -- Degraded Primary System Piping (Subcomittee meeting on July 1-2,1986)(PGS/EGI) 16.3) 2:30 - 3:00 P.M. -- ACRS Management Subcomittee meeting on July 9,1986 (DAW /RFF) 16.4) 3:00 - 3:30 P.M. -- ACRS Procedures and Administration Subcomittee meetingonAugust9,1986(DAW /RFF) "

FINUTES OF THE 316TH ACRS MEETING AUGUST 7-9, 1986 The 316th meeting of the Advisory Committee on Reactor Safeguards, held at 1717 H Street, N.W., Washington, D.C., was convened by Chairman D. A. Ward at 8:30 a.m., Thursday, August 7, 1986.

[ Note: For a list of attendees, see Appendix I. H. W. Lewis did not attend the meeting. J. C. Mark did not attend the meeting on Thursday, August 7.?

Chairman D. A. Ward noted the existence of the published agenda for the meeting, and identified the items to be discussed. He noted that the meeting was being held in conformance with the Federal Advisory Committee Act and the Government in the Sunshine Act, Public Laws92-463 and 94-409, respectively.

He also noted that a transcript of some of the public portions of the meeting was being taken, and would be available in the NRC's Public Document Room at 1717 H Street, N.W., Washington, D.C.

[ Note: Copies of the Transcript taken at this meeting are also available for purchase from ACE-Federal Reporters, Inc., 444 North Capital Street, Washington,D.C.20001.]

I. Chairman's Report (0 pen)

[ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.]

Chairran D. A. Ward briefly discussed arrangements being made for the Wingspread International meeting scheduled for October 20-22, 1986 in Racine, Wis, and requested that written drafts of technical papers to be presented by ACRS members be submitted to T. G. McCreless, ACRS Assistant Executive Director, by September 1,1986. He noted that the GPR (France) and RSK (Federal Republic of Germany) plan to send six-to-eight represen-tatives. The Japanese now plan to participate, and the U9SR has acknowl-edged participation but provided no details of their attendance.

Chairman Ward indicated that a plan is in the works to obtain a profes-sional critique of ACRS full Committee reports. The Committee intends to crgage the services of a professional reviewer. He also noted that the impediments to the first meeting of the Regional Operations Subcommittee, which is to be held in Bethesda, Md., have now been settled.

Chairman Ward indicated that a preliminary response has been received from the General Counsel (0GC) regarding the May 27, 1986 ACPS request to the OGC on whether the new backfit rule is applicable to the resolution of USI A-17, Systens Interactions in Nuclear Power Plants. The Comittee expects to get a formal written response at a later date. He also mentioned an August 6,1986 memorandun frem the NRC Chairman recardina guidance to the ACRS on the types of activities the Comission believes the ACRS should involve itself in during the next year to enhance the value of its work to the Commission's safety responsibilities.

316TH ACRS MEETING MINUTES 2 Chairman Ward indicated that a full power operating license was granted to the Hope Creek Generating Station, Units 1 and 2 on July 25, 1986.

The Staff has given its approval to the restart of the Enrico Fermi Atomic Power Plant, Unit 2. The San Onofre Nuclear Generating Station, Unit 1, has recently restarted. Also noted was the appointment of former NRC Commissioner Peter J. Bradford as Chairman of the New York State Public Service Commission.

The Cormiittee discussed W. Kerr's attendance in Vienna at the IAEA Meeting on the Chernobyl Accident, August 25-29, 1986. W. Kerr solicited specific questions regarding Chernobyl from the other nembers. D. Okrent requested that W. Kerr secure copies, in English, of all technical papers presented at the conference for distribution to ACRS members.

II. San Onofre Nuclear Generating Station, Unit 1, Water Hammer Review (0 pen)

[ Note: M. D. Houston was the Designated Federal Official for this portionofthemeeting.]

G. A. Reed explained that the Westinghouse Reactor Subcommittee met on August 8,1986 to review the loss-of-power and water-hammer event at the San Onofre Nuclear Generating Station, Unit 1 (SONGS-1) on November 21, 1985 (including root causes and corrective actions), and to review the generic implications for other operating plants. He stated that on July 24, 1986 the NPC Staff authorized restart of SONGS-1. S0 HGS-1 is a 1963-vintage plant which has the unioue feature of very laroe diesel generators to run the boiler feed pumps which in a LOCA provide safety injection water. The plant does not have automatic loading of the diesels onto the safety busses. Pegarding the loss-of-power portion of the November 21 event, although the loss of power set off the event, electrical aspects of the event proceeded accordinq to design. The only issue of concern was cable aging, a potentially significant iten, which should be referred to the Subcommittee on Reliability Assurance. A major issue involved the failure of five check valves in five individual lines (three in auxiliary feed, two in main feed lines). This set the stage for the serious water-harmer portion of the event. He suggested that the ACRS may wish to pursue aspects of the issue of valve design. He noted that the valve industry remains fragmented. A comprehensive program on the part of an owners group regarding valve design is missing as well as guidance on appropriate application for the various services that valves have to perform. Of concern might be the role of licensees in identifying valve design flaws and operating problems and the issue of monitoring and maintenance of valves. There is evidence that vendors and designers are not accepting responsibility for their products.

G. Reed indicated that a fourth major item is the fact that the main steam lines on San Onofre I are quite different from other plants of that vintage. The main steam lines tie all three stean generators together inside containment without intercepting valves and without non-return valves. He noted that the San Onofre 1 plant has an absence of valves

4 316TH ACRS MEETING MINUTES 3 comparable to other plants of its genre. Part of the modifications made prior to restart was the addition of three check valves. He suggested that a PRA might be useful in assessing the worth of these check valves.

Another item of particular note was the fact that the separate STA was particularly valuable .in identifying some of the faults that occurred during this event. Similarly, one of the root causes cited by the NRC's Incident Investigation Team (IIT) was the quality of maintenance, the fact that electrical and mechanical maintenance were not properly coupled. He indicated that a topic at the Subcommittee meeting was the difference of opinion between NRR and IE and the RES Staff as to whether the waterhammer issue is resolved. He also pointed out that, as with the Davis-Besse incident in June 1985, the auxiliary feedwater system was again challenged.

J. C. Ebersole indicated that there is an absence of steam, as with the Davis-Besse plant, and this invites thermal shock. This should be considered by the Committee. He noted that another issue of concern is the dynamics of check valves which has not been adequately examined.-

W. Kerr, another Subcommittee member, thought that all due credit should be given to the operators in their handling of the San Onofre water-hammer incident.

R. Dudley, NRC, explained the sequence of events that occurred on Novem-ber 21, 1985 at San Onofre, Unit 1 (see Appendix IV). He noted that the transient initiator was a phase-to-phase fault of auxiliary transformer C and the automatic deenergizing of 4 KV bus 2C. A very significant preexisting condition to this event was the failure, undetected at the time, of five check valves in the feedwater system.

R. Dudley indicated that the east feedwater pump was deenergized with the loss of the 4 KV bus 2C and began to coast down. The west feedwater pump continued to run, but, because of the failure of a check valve in the east feedwater pump line, fee.1 water took a backward path through the failed check valve and then through the pump and into the condensate system, overpressuring the condensate system, causing east flash evaporator tube ruptures. In accordance with procedures, the operators manually scramred the reactor. Due to the alignment that the plant was in when the reactor was scrammed, there was a turbine trip / generator trip and all power was lost to the remaining three KV busses. The result was a total loss of AC power (station blackout) of four minutes duration. He mentioned a spurious annunciation of safety injection signal, to which the operators properly responded and ignored. It was noted that, while a steam-driven auxiliary feedwater pump started about 31 minutes after the loss of AC power, the flow was driven out of the steam generator lines through the rupture into the condensate system. The feedwater was not effective because of the failed check valves. J. C. Ebersole asked if automatic closure of check valves is commonplace. J. C. Ebersole pointed out the lack of automatic closure of motor-driven valves to prevent backflow in the auxiliary feedwater lines. R. Cudley agreed that many

l 316TH ACRS NEETING MINUTES 4 other plants have automatic closure of valves. E. Merschoff, IE, explained that the feedwater design in SONGS-1 is very commonplace, in that most plants have the feedwater control valve, a check valve, and an isolation valve downstream. J. C. Ebersole asked if there is an automatic backup on closure to prevent backflow of auxiliary feed. E.

Merschoff indicated that there is typically an isolation valve that can be operated, whether by control room personnel or switched automatically.

R. Dudley explained that, when power was restored, the operator shut the feedwater isolation valves manually, in accordance with procedure. This stopped the further voiding of the steam generators. At this point reactor coolant system pressurizer level was somewhat low and the plant was approaching the safety injection initiation setpoint on low pressure.  !

To minimize the reactor coolant system cooldown, the operators terminated auxiliary feedwater flow to all steam generators. A control room super-visor, upon noting this action, later instructed the operators to restore ,

auxiliary feedwater flow to a minimum flow rate of about 25 gallons per minute. This was done approximately 10 seconds after flow was terminat- t ed. An unusual event was declared and, 5 minutes after this feedwater flow manipulation, a loud bang was heard in the turbine mezzanine area. A water hammer had occurred in the feedwater line, causing a lesk in a check valve bonnet.

R. Dudley explained that the initiating event, which was caused by an electrical cable fault, was later determined to be due to a localized overheating. This overheating was caused by the proximity of this cable to a feedwater line. Sometime in the past, a repair of a gasket or ,

flange was made to that feedwater line. The thermal insulation was l removed from the line and was not replaced. As a result, the Staff investigated the relationship between mechanical and electrical mainte-nance. It was found, from an examination of both the mechanical and electrical groups, that accountability does exist within the organization. The Staff believes that the removal and nonreplacement of the thermal insulation was just an error by the maintenance man.

. G. A. Reed was not convinced that there was proper coupling and i

! interaction between the electrical and mechanical foremen. R. Dudley i noted that the Licensee has administered aptitude tests to all of the

maintenance people.

4 E. Merschoff focused on check valves as they relate to the plant-specific restart of SONGS-1, as well as potential generic issues that should be

, pursued further by the Staf f. He explained that the Staff's approach to l l bound the problems was to inspect several vendors of safety-related check j valves. The Staff visited five licensees to survey maintenance and IST activities, procurement, and design activities at those licensees. One potential peneric issue identified was that the licensee-vendor interface

with check valve suppliers is largely ineffective. Check valves were j challenged b location, and design application (see '

j Appendix IV)yJ.nonoptimal

. C. Ebersole sizing, asked if the dynanic loadings are routinely l supplied to the vendor by the users of these valves. E. Merschoff l

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l l

316TH ACRS MEETING MINUTES 5 indicated that they were not. Licensees informed the Staff that, in general, there is no preventive maintenance performed on check valves, as well as very little corrective maintenance. Inservice testing programs implemented at the surveyed plants was really only as good as the proce-

, dures in the implementation, and may not have been rigorously pursued.

E. Merschoff explained that the Staff reviewed failure data on check valves in an effort to compile data and understand the magnitude and types of problems. The failure data were obtained from the NPRDS, LERs, 50.55s reports and compliance reports with Part 21. Trends from the data showed that leak testing appears to find failures but degradation of the valve does not seem to be a function of leak rate. It was found that the most effective method for detecting the degraded valve was to open it up and look at it. Failures tended to indicate that they were application-oriented. About half of the total failures were found in the main steam, main feed, and auxiliary feedwater systems, which tend to be challenging applications with pressure fluctuations and cycling. J. C. Ebersole asked if some of these check valve designs have an external actuator.

E. Herschoff indicated that many do. J. C. Ebersole pointed cut that this illustrates the fact that there is a randomness in the design of these valves, some of which accommodate the dynamic problem unwittingly and others that do not. E. Merschoff also noted that a survey of the effectiveness of inservice testing programs indicated that only about one-third of the valves are tested quarterly. The remaining two-thirds are tested with greater periodicity. J. C. Ebersole asked if the dynamic loading problem was addressed by the vendors surveyed. E. Merschoff indicated that it was not. Vendors complained about the lack of comu-nication with licensees and were unaware of these failures. E. Herschoff indicated that the Staff has some concerns regarding the fact that inservice testing dces not pick up degraded valves, and, in some cases, failed valves.

E. Merschoff indicated that the Staff, after identifying the problems, has decided to turn these problems over to the industry to resolve. The Staff has met with owners groups and INPO (April 7,1986; May 21,1986) to address the problems and work towards resolution. The industry cur-rently plans two actions:

INPO plans to issue an 50ER, or Safety Operating Event Report (about August 15, 1986) which will address and enhance inspection tech-niques, maintenance, and testing techniques for these valves Industry plans issuance of a design and application guide (spring of 1987) to address current accepted criteria for sizing, location, and design application for check valves The Staff expects to meet with the industry again in September 1986 to discuss their progress.

316TH ACRS NEETING MINUTES 6 G. A. Reed mentioned the Committee's concern about the fraomented indus-try and owners group activity on check valves. He wondered if the Staff believes that it has the check valve situation well-covered from a safety point of view. R. Hernan, NRC, indicated that each utility will have to respond to Generic Letter 83-28, which came out of the Salem event. That generic letter highlighted an ineffective interface between Salem and the vendor of their reactor trip breakers. Included in each utility's response will be vendors of their check valves.

R. Dudley discussed modifications made at San Onofre to reduce the probability of future feedline water hamers similar to that which occurred on November 21, 1985. The primary method of protection will be to prevent the feedlines from voiding. As a result, all five of the check valves that failed, in addition to three others, will be replaced with valves of a new design that have been reviewed by representatives from IE. IE has concluded, along - with the Licensee, that these new valves are not subject to the failure mechanism that occurred in the old valves. In addition, some new valves have been moved to less turbulent areas. Improved inservice testing will be provided for these valves along with periodic inspections. In some cases these inspections will go beyond the requirements of Section XI of the ASME Code to insure that the i valves do not degrade. The Comittee discussed the merits of the new valves in reducing the probability of the previous feedline failure and water hammer (see Appendix IV). R. Dudley noted that the Licensee has provided full automatic closure of the feedwater regulating valves. This will provide another, but not imediate, barrier against voiding of the feedline. These valves are somewhat fast-acting and should close in 10-15 seconds. Finally, the Licensee has replaced the feedline to steam generator B inside containment with new piping that slopes continuously away from the steam generator. The addition of the slope will change the refilling characteristics of this line. G. A. Reed indicated that he understood that many other upgrades and small modifications related to the event were made. R. Dudley confirmed this, mentioning that steam generator blowdown will be automatically isolated.

A. Serkiz, NRR, discussed a water-hamer review done by NRR as part of the SONGS-1 event. He discussed water-hamer occurrences during the period 1969-1980 (whose underlying causes were principally design- and operations-related) and water hamers which occurred during the time period 1981-1985. Unresolved Safety Issue (USI) A-1 on water hamer was resolved in March 1984, as reported in SECY-84-119. Technical sources for the resolution are in NUREG-0927. Rev.1, and the regulatory analysis to determine the safety significance of water hamers was reported in NUREG-0933, Rev. 1. He indicated that the Staff arrived at the conclu-sion, from analysis of water-hamer data between 1969 and 1980, that the water-hamer issue as defined under USI A-1 could be accommodated with plant fixes that were already implemented or planned, such as jockey pumps, and that the safety analysis did not warrant backfitting. A re-analysis of the problem, based upon the occurrences from 1981-1985, indicated similar underlying causes to the previous set of data.

316TH ACRS HEETING MINUTES 7 A. Serkiz indicated that the Staff has reviewed the information available from 1981-1985 and finds the underlying causes to be voided lines, steam-water entrainment, system switchovers, and operational errors--not inconsistent with the underlying causes that would derive from water-hamer occurrences in the 1969-1981 time frame that were used as the basis for resolution of USI A-1. A reinvestigation of the water-hamer issue was prompted by the SONGS-1 water-hamer event which the Staff recognizes was a result of gross check valve failures. The Staff's review of the 1981-1985 water-hammer occurrences has not revealed a new phenomenon. The reported number of events actually shows a frequency of occurrence that has decreased. The level of damage, which is principally to pipe hangers and snubbers, is similar for this period with previous data. New plants were not major contributors to the occurrences of water hammer in the 1981-1985 period. The Staff found that the San Onofre, Unit I water hamer was attibutable to grossly failed check valves in the feedwater system and is not a basis for reopening the water-hammer safety issue. D. Okrent expressed concern that analysis of the historical record will preclude the occurrences of unexpected and undefined water-hamer events. He suggested a program to investigate a postulated catastrophic water hamer, a scenario examination and not just a research project. He posed the possibility of other kinds of water hamers that can occur in transients that the Staff has not observed. A. Serkiz defined the spectrum of water hammers as spanning the range from steam-condensation-induced water hamers, which will continue to occur, to the classical water hammer which occurs during the startup of a pump, where liquid is pumped down a voided line and the water column hits a check valve. D. Okrent indicated that he was not talking about some new phenomenon that has never occurred. He spoke of circumstances that might arise during a transient which could lead to a very severe water hammer.

J. C. Ebersole asked if impact loading resulting from pipe failure and resulting fast closure of check valves is within the scope of water-hammer investigations. A. Serkiz indicated that that particular aspect, including water-hammer loads, was considered in the resolution but the safety significance was judged to be low, such that dynamic load should not preclude addition of water-hammer conditions in the design of piping.

J. C. Ebersole pointed out the fact that there is no consistent require-ment for looking at the dynamics of check valve closure for the critical set of valves in a steam line or feedwater line. A. Serkiz agreed with his characterization.

G. A. Reed suggested the need for an ACRS report on this issue. He indicated that vulnerabiity of the auxiliary feedwater system ought to be included in the report, as well as a call for alternative techniques for decay heat removal on PWRs. J. C. Ebersole thought the letter ought to be critical of the dependence of the industry on the line performance of check valves under dynamic loads. He suggested that it might be pointed out that this plant was part of the Systematic Evaluation Program, a comprehensive review which did not identify potential check valve failure scenarios. The Committee might want to question the effectiveness of the

l 316TH ACRS MEETING MINUTES 8 Systematic Evaluation Program evaluation. D. Okrent suggested that the Committee call attention to the need for the Staff to examine other water-hanner scenarios than those that have happened in a new systems study. W. Kerr suggested that this is truly a generic issue and he objected to using the SONGS-1 plant as the focus of the issue of removal of decay heat.

III. Operating Experience at Nuclear Power Stations (0 pen)

[ Note: H. Alderman was the Designated Federal Official for this portion ofthemeeting.]

J. C. Ebersole indicated that the Subcommittee on Reactor Operations met on August 4,1986 and discussed 10 operating events. Subjects discussed were feedwater transients without reactor scram at LaSalle 2, and main steam isolation valve spring failures and potential failure of all emergency diesel generators at Turkey Point 3 and 4. Also discussed were PWRs (four plants) with inadequate designs of ECCS minimum flow paths, fire in the massive charcoal off-gas beds at the Perry nuclear power plant, a problem with scram solenoid valves at Vermont Yankee, as well as the degradation of the auxiliary feedwater system due to fuse failures at Catawba 1. Also mentioned were pump wear and ring failures and reactor trip complications at the Palo Verde nuclear generating station.

The Committee spent some time discussing the Turkey Point 3 and 4 loss of all emergency diesel event (plant doesn't meet single failure criterion),

as well as the Catawba 1 incident involving loss of both auxiliary feedwater trains due to loss of improperly designed fuses which failed at certain solder joints. C. J. Wylie suggested that this was a case of a manufacturer defect. Some time was spent discussing the fire in the Perry charcoal beds. A radiant heater was brought too close to the filter and a fire resulted in one canister in each of two four-canister trains. J. C. Ebersole indicated that cold nitrogen was used as a purge to bring the temperature down and starve entrained oxygen which was fueling the fire. D. W. Moeller suggested that it was similar to the Windscale fire in that a heat sink was needed to bring the fire under control.

A. Feedwater Transient Without Reactor Scram at the LaSalle County Station, Unit 2 A. Bournia, NRR, discussed the feedwater transient without scram that occurred at about 2:21 a.m. on June 1,1986 when the LaSalle Plant, Unit 2 was at 90 percent power. The problem was that feedwater controls induced a transient and three or four low reactor water-level scram system instruments did not trip when the water level dipped below the trip setpoint. After the Licensee determined what had happened an alert was declared and the Plant shut down. An NRC augmented inspection team was sent and a confirmatory letter issued to the Licensee not to start up Unit 2, or Unit I which was

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4 316TH ACRS MEETING MINUTES 9 in refueling at the time. The Licensee initiated a test program to further study this problem and recalibrated switches and found erratic behavior of the setpoint. J. C. Ebersole thought the Committee would be interested in the fact that these water-level scram switches were supposedly improved switches because they were environmentally qualified for the external environment. A. Bournia agreed that these switches were put in because they met the environ-mental qualification requirements. J. C. Ebersole expressed sur-prise that these instruments proved to be less safe than the ones they replaced. J. Rosenthal, NRC, explained that the real concern here is erratic and unrepeatable behavior of the instrument. The LaSalle Unit 2 Plant has about 60 of the same switches, a potential common mode failure situation. W. Kerr asked if all the switches behave this particular way. J. Rosenthal indicated that all of the switches examined are sticking regardless of the other level appli-cations to which they are put in the plant.

A. Bournia described the level switch of concern as essentially a diaphragm if it has water on both sides, or a piston if steam. It is a mechanical linkage actuated switch device. He mentioned the Licensee's complete testing program which found a significant pressure shift when these devices were calibrated from atmospheric pressure to operating pressure. It takes approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of aging until the pressure shift stabilizes. W. Kerr asked if they are still using the same switches. A. Bournia indicated affirma-tively in the short term. Another effect found by the Licensee was that when calibrating some of these switches and using others to cycle them it was found that cycling of the switches didn't rectify the pressure shifts. As a result, the Licensee changed its proce-dures in their calibration to assure that the calibration is as it is at first with no cycling. The Licensee concludes that, with set-pointed calibration methods and provisions, the switches will trip within the technical specifications limits.

A. Bournia indicated that on August 7 NRR concluded that Unit 2 can start up with all of the proposed actions and test results on the switches. They provide an adequate basis for the restart in short term operation of Unit 2 until the next refueling outage (scheduled for December 1986). D. W. Moeller suggested generic applicability to several other BWRs and PWRs which might be similarly impacted.

A. Bournia noted an IE Bulletin which attempts to locate these plants. D. Allison, IE, mentioned the Millstone 1 and Oyster Creek plants as operating plants the Staff knows about which have the same types of switches in some applications. He also roted that the Segouyah and Browns Ferry plants of TVA, which are not yet operat-ing, have these switches. J. Rosenthal indicated that the Staff has a long history of unsatisfactory behavior for these switches and, in the long term, plans to encourage a switchover to analog devices, such as a piezoelectric crystal or perhaps a membrane which provides a difference in a reactivity control circuit.

316TH ACRS MEETING MINUTES 10 B. Inadequate Design of PWR Safety Injection Systems H. Bailey, IE, indicated that a 10 CFR 21 report submitted on July 24, 1985 by the Point Beach Licensee described a design deficiency involving the minimum flow recirculation valves for the safety injection pumps at the Point Beach Nuclear Plant. He discussed this problem of a potential loss of safety injection pump minimum flow due to single failure vulnerability which the Staff has been inves-tigating since the Point Beach report. The deficiency is signifi-cant because it can involve the potential ' loss of all high-head safety injection pumps during a small break LOCA. Loss of electric-ity or loss of air can cause the air-operated valves to fail closed and also result in loss of indication of closure in the control room. He noted that the H. B. Robinson, Ginna, and Turkey Point plants have also reported that these plants have the Point Beach design deficiency. The short term solution to the problem has been to disable the recirculation valves and block them open. They would be manually closed before the start of LOCA accident recirculation.

The Staff has now drafted a bulletin to address this minimum flow I problem on PWR plants and has this bulletin under review at this time. J. C. Ebersole pointed out that this is a problem that has been known for many years and never noticed. G. A. Reed noted that instrument air is not considered a safety system. E. Jordan, IE, agreed that the Staff knows that this is a problem because loss of 7

instrument air on such a balance of plant system does cause prob-lems. It is a source of concern to the Staff and there is a con-tinuing review of loss-of-air issues to seek approaches to resolu-tion of the problems. G. A. Reed asked when the Staff is going to tighten up on instrument air requirements since there are many examples of the vulnerability of instrument air and its conse-quences. E. Jordan indicated that the Staff plans to look at instrument air in depth on a plant-by-plant basis.

C. Scram Solenoid Valve Problems at Vermont Yankee (0 pen)

E. Weiss spoke of the report on June la, 1986 by Varmont Yankee that

one control rod failed to move and five control rods hesitated for a

, period. of about five seconds because of prcblems with scram L solenoids. The event is significant because this particular Licensee permitted the plant to go critical with these rods inoperable. The root causes of these failures were manufacturing i defects in rebuild kits for the scram solenoid, an issue which may l affect other plants. There was apparently inadequate post-testing and a potential for at least a hypothetical core-damage accident if two rods stuck that were adjacent. The defective parts in the rebuilt kits were supplied by ASCO to General Electric as commercial-grade parts and then sold by General Electric as safety-grade. One of the saving graces involved with this event is

that back-up scram solenoids are alternate rod-insertion solenoids

(

316TH ACRS MEETING l11NUTES 11 and, if present, would cause rod movement even if the scram solenoids failed. The applicability of this problem is rather wide, applying to all BWR 2s, 3s, virtually all 4s, and some 5s. E. Weiss indicated that it is interesting to note that these valves make up the one out of two twice logic. One needs both the scram solenoid and the back-up scram solenoid valve to trip before one gets the scram.

E. Weiss indicated that IE has conducted a vendor inspection at ASCO and has gained some additional details regarding the types of failures and has found some additional failures. General Electric (GE) has issued a RICSIL or Rapid Information Communications Services Information Letter to notify its customers of the nearly 3000 of these kits out in the industry somewhere. G. A. Reed suggested that the Committee consider a letter to the EDO encourag-ing a more comprehensive look at the vulnerability of instrument air systems. E. Jordan indicated that on a plant-by-plant basis there always seems to be a way to upgrade air systems to avoid classifying them as safety-grade. One can add a larger accumulator or change a particular air-operated valve to a motor-operated valve on a case-by-case basis. C. P. Siess asked the Staff if they are looking at this incident regarding its significance to quality assurance (QA).

He wondered if there was a breakdown in GE's OA program. E. Jordan noted that these replacement parts from ASCO had a number of flaws in them, and GE's dedication is in cuestion; nevertheless, he suggested that tne problem appears to be a combination of failures of the manufacturer (ASCO), GE's dedication, and failures during installation by the Licensee. C. J. Wylie pointed out the similarity of this problem with Westinghouse's under-voltage devices on the DP 416 scram breakers. C. P. Siess suggested that QA is supposed to guarantee that everything that goes into the plant is perfect. He wondered whether QA in principle does not work.

IV. Management and Disposal of Radioactive Wastes (0 pen)

[ Note: 0. S. Merrill was the Designated Federal Official for this portion of the meeting.]

D. W. Moeller explained that during the Waste 11anagement Subcommittee meeting held on July 21-23, 1986, 14 NMSS and RES waste management topics were reviewed. The Subcomittee prepared six sets of comments as attach-ments to a draft transmittal letter to the EDO. The first two sets of comments dealt with the development of residual radiation limits and contaminateri smelted alloys. This was a continuation of the discussion on these related topics as contained in two May 13, 1986 ACRS reports to Chairman Palladino entitled "ACRS Coments on the Definition of Low-Level Radioactive Waste," and "ACRS Comments on Salvaging of Contaminated Smelted Alloys." Since a discussion of these topics was prompted by a specific reouest of Commissioner Bernthal, the Subcommittee's current comments on these items will be submitted to him. The third set of 1

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4 316TH ACRS MEETING MINUTES 12 comments was concerned with radioactive wastes below regulatory concern, which was addressed in SECY-86-204, " Policy Statement on Radioactive Waste Below Regulatory Concern," July 11, 1986. The fourth set of comments discusses two related generic technical positions on solubility and sorption as they apply to the media surrounding a geologic reposito-ry. Although the solubility generic technical position is already final (dated November 1984) the NRC Staff indicated that they are still open to comments. These positions are working documents that can be modified as required by subsequent need. The fifth set of comments addresses various waste management topics with brief comments on each: 1) the Division of Waste Management's five-year plan, 2) the establishment of a Federally Funded Research and Development Center, 3) rulemaking to bring key issues to closure, and 4) alternative methods to shallow land burial of low-level radioactive wastes. Subcommittee comnents are provided in the sixth set on the efforts of the Waste Management Branch of RES in their use of natural analogs to understand the behavior of radioactive materials in repositories and in their attempt to establish a system of peer review to assure the credibility of their research activities.

D. W. Moeller requested Committee approval of a draft letter for transmittal of the six sets of comments to the EDO, Comissioners, and the NRC Staff.

D. Okrent requested some discussion of the subject of the NRC generic technical positions on sorption and solubility. He asked what the role of these generic technical positions was. D. W. Moeller explained that they are guidelines for the U.S. Department of Energy (D0E) on what the Staff considers to be acceptable in terms of justifying DOE's position on specific items related to the repository. These generic technical positions are therefore somewhat comparable to a regulatory guide.

D. Okrent suggested that the Committee endorse the concept of rulemaking to bring key issues to closure, but indicated that care must be exercised in the selection of such issues to assure that they pertain to fundamental principles that must be established in order to expedite the licensing of a high-level waste repository. He indicated that he remains uninformed as to why the particular issues the Staff wishes to bring to rulemaking will in fact deal with the issues that the ACRS raises on the difficulty of meeting the quantitative aspects of the EPA's Standard with its qualitative caveats. He expressed the need to see an NRC " white l paper" explaining how and why this approach will accomplish the purpose.

S. J. S. Parry, ACRS Fellow, explained that such items as certain defini-tions within 10 CFR 60, i.e., " disturbed zone" and what constitutes "substantially complete containment," will also probably require a rulemaking to achieve closure. He stated that one must recognize that the NRC Staff is very much in a reactive mode, as they are controlled heavily by the activities of DOE, and the DOE program is in serious disarray with both legal and legislative questions. D. Okrent suggested that the ACRS press the Staff regarding how they intend to reach a position of confidence considering the uncertainties that exist in the l performance of the base material and with regard to methodologies to be I

used for analysis. He also expressed concern regarding the Staff's

316TH ACRS 11EETING MINUTES 13 ability to use analogs in a meaningful way for specific aspects of repository problems. The Committee discussed the issue of an adequate system of peer review of the NRC's repository activities. P. G. Shewmon thought that a truly independent review would not be possible without the resources of a professional society. D. Okrent noted the efforts by the U. S. Geological Survey (USGS) on their peer review of the Hanford site.

P. G. Shewmon asked, from an NRC viewpoint, what the USGS's involvement is in the review process. He wondered if the USGS was a consultant on specific concerns dealing with earthquake risks and siting.

S. J. S. Parry indicated that it was his understanding that the Staff does not actively consult with the USGS.

D. A. Ward made some specific suggestions for the draft letter and asked if they are Subcomittee recommendations or Committee recommendations.

D. W. Moeller indicated his suggestion was to forward them as Subcomittee comments. D. A. Ward thought that if these comments are in the nature of recommendations of the Subcomittee they cught to be pulled into the body of the letter to make then full Comittee recommendations.

D. W. Moeller said he would rewrite the letter and include key Committee recommendations drawn from the six sets of Subccmittee comments and reflectin [ Note: D. W. Moeller did revise the letter.] g the above discussion.

V. Human Factors Issues (0 pen)

[ Note: H. Alderman was the Designated Federal Official for this portion ofthemeeting.)

F. J. Remick indicated that the Human Factors Subcommittee met on July 15, 1986 to consider the Fitness for Duty Policy Statement, re-visions to Regulatory Guide 1.114, the Advance Notice of Rulemaking on Degree Requirements for Senior Reactor Operators, and the Staff evalu-ation of the INP0 accreditation program.

F. J. Remick explained that the Comission approved, on June 26, 1986, issuance of the " Commission Policy Statement on Fitness for Duty of Nuclear Power Plant Personnel." In answer to a question by P. G. Shewmon, he indicated that the Rule applies to everyone inside protected areas, including contractors but not including visitors or the NRC Staff. The question of fitness for duty has been before the Commis-sion for at least five to six years and the Comission was close to issuing the Rule several years ago. In August 1083 the Commission issued for public comment a proposed rule on fitness for duty. This became one of the early efforts of NUMARC regarding the question of a rule on training. NUMARC convinced the Commission to hold off and allow industry, through a cooperative effort of NUMARC working with the NRC Staff, to develop a program which would be an industry initiative. The Policy Statement just approved by the Commission is a result nf that effort and, as a result, the Commission has decided to defer

316TH ACRS MEETING MINUTES 14 i implementation of the Rule and, in fact, has withdrawn the former proposed rule on fitness for duty (August 1986).

F. J. Remick explained that the policy statement puts the Commission on record as endorsing a concept of a drug- and alcohol-free workplace at operating nuclear power plants, and defers any rulemaking in this areh for 18 months as long as the industry programs produce desired results.

The Edison Electric Institute guidelines (optional--not mandatory) are endorsed by the policy statement. The NRC retains authority and respon-sibility to follow operational events that might be related to fitness for duty, as well as evaluate licensees' efforts to verify effectiveness of the industry programs. The Staff can take enforcement action when regulatory requirements are not met and can utilize 10 CFR 50.54(f) letters to obtain information about fitness-for-duty programs if they desire. However, the NRC Staff plans to exercise discretion in enforce-ment during this time period and will undertake actions only with concur-rence of the Comissioners. The industry will be expected to keep the Commission informed of the status of the program and a method for direct-ly reporting to the Commission is discussed in the Policy Statement.

F. J. Remick indicated that the NRC Staff is charged to evaluate the effectiveness of the program through different methods:

Review the INP0 program status and INP0 evaluation and assistance reports Make periodic observations of INP0 evaluations Direct inspections against the Edison Electric Institute Guide F. J. Remick indicated that the Policy Statenent indicates that the sale, use, or possession of alcoholic beverages or illegal drugs within pro-tected areas is unacceptable. The Policy Statement indicates that violations involving illegal drugs will result in imediate revocation of access to vital areas and discharge from nuclear plant activities.

Violations involving the use of alcohol or abuse of legal drugs inside protected areas will result in imediate revocation of access to vital areas and possible discharge from future plant activities. Any other sale, use, or possession of illegal drugs will result in immediate revocation of access to vital areas, mandatory rehabilitation prior to reinstatement, and possible discharge from nuclear plant activities. He noted that the Policy Statement calls for effective monitoring and testing procedures to provide reasonable assurance that personnel within protected areas shall not be under the influence of any substance, legal or illegal, which adversely affects ability in any way related to safety.

The Edison Electric Institute Guide covers ten program elements which constitute good practices. The Guide speaks to such issues as the r.eed for written policy to buttress the need for utility-wide understanding.

Top management support is essential as is effective policy comunication to generate rank-and-file support. Behavioral cbservation training is

316TH ACRS tiEETING MINUTES 15 important for supervisors and, in some cases, even for operators and security guards. The Guide addresses a need for policy-implementation training for supervisors regarding their need to handle different sit-uations. A union briefing is necessary to foster support for effective implementation. Contractor notification of this requirement, close law-enforcement liaison with law-enforcement bodies, as well as chemical testing of body fluids and employee assistance programs are also covered in the Edison Electric Institute Guide. While the Policy Statement applies only to operating nuclear power plants, the Staff is studying the consequences of personnel not being fit for duty'for all other licensing sectors, such as fuel cycle facilities, material licenses, and non-power

reactors. The NUMARC people indicate that the chief executive officers of all 55 nuclear utilities have committed to establish fitness-for-duty programs. He expressed his belief that the Policy applies to all indi-viduals with unescorted access to vital areas except NRC employees, and the NRC Staff is working, on its own initiative, on developing its own fitness-for-duty program. He noted that the ACRS addressed in its August

, 1983 letter the fact that the exemption of NRC employees from the Policy Statement on Fitness for Duty is inappropriate.

F. J. Remick mentioned a presentation by J. Griffin, Vice President at Arkansas Nuclear-1 and a NUMARC representative, and J. Colvin of INP0, regarding the fact that INP0 is conducting a survey of the status of fitness-for-duty programs that now exist, and has found that all util-ities have some form of fitness-for-duty program. All utilities are I

committed to chemical testing of body fluids for preemployment screening and for cause. Only some utilities plan random sampling. The unions in some cases have challenged the random sampling concept, and while the NRC Staff would prefer random sampling it accepts the current si tuation.

G. A. Reed indicated that he supports random sampling. The Comittee discussed possible endorsement by the ACRS of random drug testing.

F. J. Remick volunteered to draft a letter which he thought reflected the Committee's consensus view on random testing.

F. J. Remick explained that the NRC Staff issued a staffing rule which became effective on January 1,1984. One provision of that rule would require that when a nuclear power unit is in an operational mode or cold shutdown or refueling each licensee shall have a person holding an SR0 license in the control room at all times. As a result, there is a proposed revision to Regulatory Guide 1.114, " Guidance to Operators at the Controls and to Senior Operators in the Control Room of a Nuclear Power Unit," to bring the existing Regulatory Guide in accordance with the rule just mentioned. The proposed revision would clarify the inter-pretation regarding an operator at the controls and a senior operator in the control room. A senior operator is to be inside of or within audible range of the operator at the controls or in audible range of control room annuncation. Administrative procedures are required to

! define and outline, preferably with sketches, a specific area within the control roon where the senior operator should normally remain while performing his designated duties. It also provides some guidance on

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316TH ACRS MEETING MINUTES 16 shift turnover and a conforming change of the Standard Review Plan. He indicated that the Subcommittee was unanimous in recommending a short, favorable letter by the Comittee that would state that the Comittee reviewed the Regulatory Guide change and endorses it.

F. J. Remick explained that the Comission published in the Federal Register an Advance Notice of Proposed Rulemaking on Degree Requirements for Senior Operators. It is claimed that this would be an extension of the Engineering Expertise on Shift Policy Statement issued in October 1985. The Engineering Expertise on Shift Policy Statement established two options for the STA: a combined STA and SRO position and maintenance 1 of a separate STA. This Advance Notice of Proposed Rulemaking is differ-

ent from a proposed rulemaking in that no actual rule is proposed; general comments are being provided to solicit comment from the public before a rule is proposed. The coment period _ is now set to end on September 29, 1986. The Advance Notice of Proposed Rulemaking would be effective on January 1,1991, and would state that SR0 applicants must hold a baccalaureate degree in engineering, engineering technology, or the physical sciences from an accredited insitution. This educational requirement would apply only to SR0s and not R0s. Other baccalaureate degrees from accredited institutions may be accepted on a case-by-case basis if the utility can certify that the applicant has demonstrated high potential for the SR0 position. Degree equivalency will no longer be accepted. SR0s licensed prior to January 1,1991 would be grandfathered and there would be only one reexamination for SR0 applicants who applied just prior to January 1,1991. He noted that the current requirement for SR0s is that they have two years of operating experience. The Advanced Notice of Proposed Rulemaking says that one of the two years must be at greater than 20 percent power operation.

l F. J. Remick indicated that the Comission also intends to prepare a concurrent policy statement to encourage utilities to implement personnel policies that emphasize the opportunities for licensed operators to i assume positions of increased management responsibility (career paths for operational people). Utilities would also be encouraged to develop i programs that would enable currently licensed SR0s and reactor operators to obtain college degrees if desired, and for utilities to try to obtain college credit for appropriate nuclear power plant training and work experience through arrangements with the academic comunity. The purpose i

of the Rule and the Policy Statement would be to enhance engineering and accident management expertise on shift by combining both engineering expertise and operating experience in the SR0 function. This should enhance public safety and provide career paths for SR0s (similarly for R0s if degreed). He noted that Comissioners Roberts and Asselstine provided additional coments to the Advance Notice.

F. J. Remick indicated that the Subcomittee recomends that the Commit-tee not provide any report at this time on the Advance Notice of Proposed 4

Rulemaking since substantial public comments are anticipated from such groups as KMS, the ANS 3.1 Comittee, and NUMARC, as well as individual 1

I 316TH ACRS MEETING HINUTES 17 utility responses. P. G. Shewmon asked about the difference between the current requirements for the separate STA and the STA/SR0 now and after this Advance Notice of Rulemaking is in effect--would the SR0 with a degree qualify as an STA? F. J. Remick indicated that this is the approach that the Comission is trying to take. They are trying to encourage option 1, the joint STA/SR0 option, because it is more strict in its degree requirements than the separate STA option. P. G. Shewmon asked if there is a policy on the part of the NRC to discourage degreed individuals, not necessarily in the control room, from taking the SR0 exam and qualifying to serve as STA. B. Boger, NRC, indicated that, on the contrary, the Staff encourages individuals with degrees to get an SR0 license, but these individuals should meet the operating experience requi rement. G. A. Reed was against the proposed degree requirement for senior operators. because it encouraged " fast trackers." It would make SR0 and shift supervisor jobs short-term career stops. He speculated on a mountain of objecting comments from professionals all the way to utility executives. He thought the Rule a waste of effort on the part of the Regulatory Staff. F. J. Remick suggested that there is a side benefit in that the formalized training and operating experience these individuals get, if they are going to maintain their license, including continuous requalifications, is a benefit for those staff people who need to interact with the operating staff when carrying out their duties.

Another aspect is getting instructors into utility training programs, since people with degrees and specialists are also licensed. D. A. Ward questioned whether the large fraction of the time out of the year these engineers spend preparing for these examinations is an efficient use of their time. M. W. Carbon suggested that G. A. Reed was saying that he thought it not desirable to have SR0s with engineering degrees.

G. A. Reed indicated that he preferred the option of STAS with degrees.

He thought the experience paths much better for developing shift managers without necessarily having a degree. He suggested that the apparent objective of this degree requirement is to create career paths into which to move good shift supervisors after serving two years of operating tine.

This approach will short circuit the career path for a sound high school graduate who has been carefully selected and trained for many years--even though this individual is grandfathered at present. In the future the utility would not have to hire that individual. W. Kerr indicated that he agreed with G. A. Reed but for somewhat different reasons. He did not see anything gained by making the degree a requirement. He thought there ought to be an alternate route that doesn't require a degree. G. A. Reed suggested that the NRC should be cutting back rather than increasing the degree requirements and leaving the option open in the career path to the STA job. The aging shift supervisor, on leaving that position, will have no place to go except out, and this is not in the interest of operational safety. F. J. Remick indicated that the Subcommittee does not propose any action at this time. The Comittee agreed.

F. J. Remick indicated that the Policy Statement on Training and Quali-fication issued in March of 1985 was as a result of a NUMARC approach to the Commission, suggesting a policy statement which would allow for an

m 316TH ACRS MEETING MINUTES 18 industry ' initiative rather than a rule. It endorsed a systematic ap-proach to performance-based training and endorsed the INP0 accreditation program as meeting that approach. The Commission deferred further i rulemaking in this area for at least two years and directed the Staff to i independently evaluate the industry program. The first annual evaluation report by the Staff is SECY-86-118. The Staff monitored INP0 National Nuclear Accrediting Board activities and observed accreditation team visits to plant sites. The Staff found that the accreditation team visits were thorough, constructive, professional, and resource intensive.

They found when observing Accrediting Board meetings that they were professional, high-quality, and independent reviews, and forced senior ,

utility managensent involvement. The Staff has also been reviewing  ;

self-evaluation reports prepared by the utilities and accrediting team j reports. These were found to be consistent with INP0 guidance and thorough. The Staff has also made independent evaluations of licensees.

The Staff points out a high degree of management attention to perfor-mance-based documentation but notices some weakness in administrative areas, such as content of the training programs.

The Staff concludes that progress is being made in improving training, and this progress has required a substantial commitment of resources.

Some deficiencies have been identified and are being addressed by INP0 and the Staff. The current Appendix A to 10 CFR Part 55 constitutes a major problem in this area because utility requalification programs for licensed personnel are dictated by the current Appendix A requirements and it is not systematic or performance based. Therefore, utilities are hesitant to change or improve their requalification training for licensed personnel, because they are concerned that they may go counter to commit-ments already accepted by the Staff. The NRC Staff has sought to solve this problem by proposing a new Appendix A, by updating requalification training for licensed persons to the normal new performance-based or systematic approach, as utilities work their way into existing approved programs. The Staff indicated they believe further rulemaking should continue to be deferred at this time based on that first-year report.

l VI. Seismic Oualification of Equipment (0 pen) l l [ Note: R. K. Major was the Designated Federal Official for this portion i of the meeting.]

l C. J. Wylie indicated that Subcommittee on Reliability Assurance met on August 5, 1986 to review the final resolution of USI A-46, " Seismic Qualification of Equipment in Operating Plants." He mentioned a sumary presentation by the Seismic Qualification Utilities Group (SQUG) of total activities to date toward resolution of A-46. Presented was a progress report of the documentation of the seismic adequacy of the equipment needed for safe shutdown in the case of an earthquake, a generic applica-l tion plan, and a report on the seismic experience gathered to date. An l EPRI presentation concerned equipment anchorage guidelines and relay l evaluation procedures being developed. The NRC Staff is approaching a l

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316TH ACRS MEETING MINUTES 19 final resolution and intends to meet with the CRGR in September 1986 and issue the resolution in a generic letter in October 1986. He noted that there were a number of questions raised at the Subcomittee meeting by members of the Subcomittee. Quite a bit of the documentation has been unavailable to the Subcommittee including the implementation plan, the guidelines, and some of the EPRI reports. The ACRS was briefed on resolution of USI A-46 in July 1983, May 1984, August 1985, and January 1986. He implied that all Committee concerns are being addressed with the exception of the review of the guidelines, reports, and procedures that the ACRS intended to review at this time.

C. Michelson indicated that there were a few items that he would like to see clarified during the current presentation:

Will failure mode and effects analysis (FMEA) be done to determine whether the minimum set of equipment is appropriately protected against a seismic event?

Will FMEA involve simultaneous failure of all equipment which is deemed to have a failure potential during the earthquake, or will failures be taken one element at a time?

How does one handle systems interactions effects when the inter-actions are not direct physical or not direct electrical coupling interactions?

Why are there no moderute-energy pipe breaks either inside or outside of containment being considered?

Why are there no seismically-induced fires considered even in terms of protecting the minimum set of equipment required for safe shut-down?

Why does not the work on handling relay chatter take account of the fact that the devices comonly thought of as relays also include instrument and electrical contacts?

Has it been considered that focusing on the very minimum set of equipment to achieve safe shutdown and not considering failures to any extent in that minimum set or to alternate back-up devices are triming safety margins to a very minimal value and creating large uncertainties regarding the effect of the magnitude of earthquakes on the minimum set of equipment.

D. Okrent expressed concern that the resolution of A-46 meet the Safety Goal Policy which states that there should be a reascnable assurance that there is not a core relt for the life of the existing plants. He noted that the Staff, when preparing their implementation plan, should be aware that the Comission will approve the goal of 10~ per reactor year for a major release of radioactivity. He suggested that this should be

316TH ACRS MEETING MINUTES 20 examined, evaluated, and presumably included (or at least at a minimum included for study if not for actual adoption). He also wondered how the Staff plans to factor in earthquakes larger than the SSE. He noted that piping, tanks and heat exchangers are not considered in A-46 except when those tanks and heat exchangers are directly required to achieve and maintain safe shutdown. He thought it very conceivable that other kinds of tanks besides the ones directly involved in achieving safe shutdown  ;

could get one in trouble from the point of view of losing their contents and flooding out vital equipment.

T. Y. Chang, NRC, indicated that Task A-46 was designated as an unre-

. solved safety issue in December 1980 because of the recognition of the concern for assuring the seismic adequacy of cquipment in operating plants (see Appendix VI). Of five tasks identified to address concerns regarding assurance that a plant can be brought to a safe shutdown, use of seismic experience data proved to be the most reasonable alternative to address the current requirement. It is also cost effective. The Staff recognized that during the study of seismic experience data there would be need to collect test data as well to supplement the seismic experience data. The feasibility of using seismic experience data to qualify equipment has been established by a Lawrence Livermore National Laboratory study and also by the pilot program of SQUG. The pilot program studied eight classes of equipment from a number of plants which underwent the San Fernando earthquake and Imperial Valley earthquake in 1971 and 1979, respectively. Since that time data from other earthquakes have been added to the data base. Among the earthquakes were the Coalinga, Morgan Hill, Chile, and Mexico earthquakes. In June 1983, a Senior Seismic Review and Advisory Panel (SSRAP) was formed jointly by SQUG and the NRC to act as an independent panel to advise SQUG on how to use the seismic experience data to qualify equipment. The NRC Staff has participcted in all meetings with SQUG and SSRAP and has closely moni-tored their work. The proposed Staff resolution is based on both the seismic experience and test experience approach. T. Y. Chang indicated that the study of seismic experience data pointed out three major con-cerns. These were equipment anchorages, relay functional capability, and outliers. It appears that equipment, in general, is fairly rugged as far as earthquake loadings are concerned so long as one has adequate anchorages. While the functionality of relays was established after an earthquake, it cannot be established during an earthquake and this has become a concern. The third concern for cutliers involves miscellaneous equipment that is not covered by either the seismic experience or the test data experience. D. Okrent asked what level of seismic safety the Staff is attempting to achieve by the proposed resolution and how does this compare with the safety goal policy. N. R. Anderson, NRC, indicated that the scope and purpose of USI A-46 was to be an alternative to meeting current requirements for seismic qualification of mechanical and electrical equipment at the design level. There is no intent to measure the provisions in USI A-46 against those in the safety goal policy. The seismic risk level the Staff is considering in A-46 is the SSE level for most plants. Most seismic PRAs show that the risk at that level is very

316TH ACRS MEETING MINUTES 21 low. D. Okrent asked if the Staff is considering a separate A-46-type program to look at seismic margins with respect to the Safety Goal Policy and the Severe Accident Policy. N. R. Anderson indicated that the Staff's seismic margins program and also the EPRI s'eismic program will look for plant vulnerability specifically, but it does not quantitatively say what the risk is or whether it meets a quantitative safety goal.

D. Okrent applauded the seismic margin program but indicated that it would not provide enough information for resolution of USI A-46 with respect to seismic levels above the SSE.

C. Michelson noted that in the resolution of A-46 the Staff has iden-tified only one train of equipment to protect. He asked to what extent the Staff will look at the failure of non-seismically qualified equipment whose failure can cause failure of this minimum set, and what kind of losses the Staff will study. N. Anderson indicated that A-46 involves examination of direct physical contact. The flooding review that was dcne in 1972 included large tanks. The flooding issue, including seismic failure from flooding of those tanks, will be covered in the USI A-17 flooding review. C. Michelson pointed out that, in most analyses of this kind, one starts with two trains and, after a failure in one train, there is a remaining train to provide the function. In this case, one is already down to one set of equipment. A second problem exists with respect to fires, where A-46 starts with only one train aveilable to begin with by design. N. Anderson indicated that the Staff has, and intends, to make sure that there are no gaps between fire protection, systems interactions, and other situations. He indicated that he was very reluctant, however, to expand the scope of A-46 at this time.

H. Etherington mentioned a particular problem with high-strength bolts.

N. Anderson indicated that there is another generic issue where the Staff is working on bolting materials. H. Etherington asked when the Staff expects the bolt problem to be resolved. R. Bosnak, NRC, indicated that it should be resolved in 1987.

l l C. Michelson noted the term " alternate component." He asked whether this means an alternate train or another component right beside the one that i might fail. In the case where an alternate component can do the same I function, it can also be flooded out by the same event. Such redundant components are subject to the same failure that failed the original component which is in the minimum set of equipment. N. Anderson conceded that if one postulated that kind of event the alternate component would not assure continued function.

T. Y. Chang discussed the specifics of the proposed resolution of A-46 l for operating plants. He indicated that the Staff intends to perform walk-through inspections to verify anchorages, review seismic systems interactions, and identify and address deficiencies in outliers.

C. Michelson pointed out that only the minimum set of equipment and alternative components are involved in the walkdown. T. Y. Chang in-dicated that the Staff verifies the functional capability of equipment such as relays and contacts. C. Michelson asked if the Staff is looking i

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316TH ACRS MEETING MINUTES 22 4

at tanks, pipes, and other pieces of equipment. T. Y. Chang indicated only up to the extent that they are required for hot shutdown.

, C. Michelson brought out the fact that the Staff is looking only for a limited set of interactions that they have defined which specifically affects the minimum set of equipment that has to be protected. R. Bosnak explained that the Staff looks at the systems required for safe shutdown and any other things that could impact that system or any components, including pipes and duct work, whose loss would cause a loss of function of the system. T. Y. Chang indicated that the proposed resolution package was issued for public comment on September 13, 1985. Since that

time the Staff has received and resolved all coments, and incorporated them in the final proposed resolution package. All NRR Divisions have concurred. The Staff is planning to meet with the CRGR in September 1986 and expects final resolution and issuance of a generic letter on the 1

issue in October 1986.

In answer to a question by D. Okrent, R. Bosnak explained how the resolu-tion of A-46 will not impose any significant additional risk to the health and safety of the public and is in accord with the Commissicn's safety goals. D. Okrent expressed concern that the core melt frequency due to earthquakes is such that the Staff will have difficulty providing reasonable assurance of no core melt in the life of the existing plants, as stated in the safety goal. F. J. Remick objected to the no core melt

. reference as not being the Commission's safety goal. He suggested that F the Commission is not saying that there will never be a core melt; it is saying that the Connission intends to pursue a regulatory objective in

, view of uncertainties that no serious core damage accident shall occur in  ;

[ a U.S. nuclear plant.

R. Bosnak indicated that the Staff seeks a letter from the ACRS of i

' continued endorsement of the resolution of USI A-46. The Staff is now ,

entering the implementation phase and wishes to keep the Committee ,

totally involved in the guidance documents when they become available.

l C. Michelson asked for a clarification regarding the fact that the Staff assumes that no pipes in the plant break during an earthquake irrespec-tive of size or pressure rating. N. Anderson indicated that the assump-tion is no major piping that would either affect the piping system or function of the equipment. The Staff is limiting the situation to piping that is involved in performing the functions or involved in the integrity of the primary coolant system boundary. C. Michelson also asked if the

! Staff is considering a seismically-induced fire. R. Bosnak indicated that fire in the seismic event is not considered in A-46.

VII. Standardized Nuclear Plants (0 pen)

[ Note: H. Alderman was the Designated Federal Official for this portion ofthemeeting.]

C. J. Wylie indicated that the ACRS Subcommittee on Improved LWR Design met on August 5,1986 to discuss the Standardization Policy Statement.

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316TH ACRS MEETING MINUTES 23 The present version of the Standardization Policy Statement was issued by the Commission in 1978. In 1985 the Commission issued its Severe Acci-dent Policy Statement which set forth licensing requirements for new plant designs, both standard and custom. In addition, the Comission has proposed to Congress its draft " Nuclear Power Plant Licensing and Stan-dardization Act of 1985," which would provide for the issuance of com-bined construction and operating licenses in a one-step licensing pro-cess, early site approvals and standard design approvals. Based upon this proposal to Congress, the Comission requested that the Staff prepare a draft revision to the 1978 Standardization Policy Statement.

This was issued as SECY-85-382, " Standardization Policy Statement,"

December 4, 1984. This revised policy statement was reviewed by the Cormissioners and issued for public coment on April 10, 1986. The Staff responded to the April 10, 1986 request for comments with their own version of the Policy Statement on May 14, 1986. The Comission's intent was for the Policy Statement to be simple with the technical implementa-tion to follow in an accompanying NUREG document. He mentioned the July 10, 1986 memorandum from J. C. McKinley, Chief of the ACRS Project Review Branch No.1, which discussed the differences between the April 10 and May 14 versions of the Policy Statement (see Appendix VII). D. Okrent indicated he would vote against the proposed standardization policy because it is standardization without accompanying improvements in safety. G. A. Reed thought that the Commission was in danger of licens-ing systems that are conceptually immature--one should talk more about standardizing some plants. C. Michelson was satisfied with the evolutionary process of standardization and thought it should proceed.

P. G. Shewmon scored the lack of standardization of the balance of plant, which may be taken over more by vendors. C. P. Siess noted the existence of Appendix 0 to 10 CFR Part 50, and noted that this policy adds little in the area of standardization. C. Michelson suggested that the ACRS concentrate on alternate sentences or alternate paragraphs for the Policy Statement as a way to comment; otherwise, the Committee's comments will be very complicated. D. A. Ward wondered whether construction practices in the industry have introduced variability and diversity. C. J. Wylie thought it unfair of the Staff to claim that they are really standardiz-ing a design. H. Berkow, NRC, indicated that the Policy Statement does not preclude other foms of standardization. Both C. Michelson and F. J. Remick thought that the policy statement should focus on design certification. C. Michelson thought it best that the NUREG which has technical guidance be issued at the same time as the policy statement as an enclosure, so that it will be part of the package on which the public would comment.

VIII. Improved Light Water Reactors (0 pen)

[ Note: R. K. Major was the Designated Federal Official for this portion of the meeting.]

W. Kerr discussed his thoughts on accident prevention measures ,

appropriate for consideration in standard plants. His first thought was )

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l 316TH ACRS MEETING MINUTES 24 elimination of concerns about ATWS by designing standard plants so that the likelihood of an ATWS with unacceptable consequences is very low.

He thought this could be accomplished for PWRs by providing a reasonable amount of pressure-relieving capability for the primary system and by keeping the moderator coefficient of reactivity sufficiently negative throughout the fuel cycle. For BWRs, elimination of concerns about ATWS could be accomplished by providing for 100 percent by-pass capacity to the condenser. G. A. Reed thought that the 100 percent by-pass capacity idea was not a viable solution for BWRs because one could not guarantee the availability of the condenser. The Committee discussed the resolution of USI A-45, Decay Heat Removal Systems. C. Michelson comented that W. Kerr's ideas are mostly patches and not conceptual designs. W. Kerr suggested that the ACRS consider informing the Comission that it needs better decay heat removal, and cite the characteristics of what such a decay heat removal process should be.

D. Okrent suggested that the use of a primary blowdown system ought to be considered for advance LWRs. Other desirable features for such a plant might be a modest recirculation capability, makeup capability, as well as resistance to internal and external intrusion (sabotage prevention or mitigation). A new standardized plant should have redundancy and should be single-failure tolerant.

W. Kerr suggested the elimination of the artificial division of plant systems into safety- and non-safety-grade systems which has led to control systems and other balance-of-plant systems being almost ignored as accident initiators. G. A. Reed thought that the balance-of-plant ought to be made up of commercial-grade equipment. W. Kerr continued that it is clear from operating experience that many non-safety systems should have a reliability specification if the industry is to achieve the level of plant reliability sought. D. Okrent agreed that there has not been a test of all electrical systems. W. Kerr also expressed his concern that not enough attention has been paid to the identification of and study of systems interactions. G. A. Reed suggested that nuclear safety systems should be designed to take challenges. D. Okrent thought balance-of-plant designs should be required to meet a given a range of performance.

D. Okrent suggested that members prepare position paragraphs for an ACRS letter, and that these ought to be collected over the next few months.

C. P. Siess discussed the topic of accident mitigation. He suggested that containment systems ought to be designed to keep fission products inside; the containment ought to be a mitigating feature regarding the release of radioactivity. He contrasted the pool scrubbing in the BWP containments with sprays and fan coolers in the containments of PWRs. He recorrended that criteria for containments and containment systems ought to be either performance oriented or prescriptive or design oriented. He thought that a containment could be asked to provide zero release or be designed to minimize leakage, but he thought that a containment could not

. be designed to be adequate to take steam explosions since the loadings l are too high. He thought the guidance in the Comission's safety coal

316TH ACRS MEETING MINUTES 25 precluding core melts confusing and unreasonable. He thought that the terminology " reasonable assurance" woul'd be better. He suggested that the concept should be to postulate a feature for accident prevention, taking account of defense-in-depth. Containment perfomance can be measured by the amount and composition of the fission products released.

Of course, where, how, and when fission - products are released is an important consideration. The Committee discussed how the ACRS should develop recommendations on containments. C. P. Siess noted that the Committee has comented frequently about the need to develop a contain-ment perfonnance design objective. He noted that the Staff is working on one with a deadline of late September 1986. Various approaches by the Staff during a recent workshop were risk oriented and not prescriptive.

D. Okrent suggested straw-man approaches on improved containments for both BWRs as well as PWRs. C. P. Siess suggested that expressing containment performance in terms of the probability of containment failure given a severe accident is not an easy task. For a containment failure due to slow overpressure, whether PWR or BWR, anything that happens a week after the accident starts should show low consequences.

That kind of failure is one that could be~ either prevented or further delayed by increasing the containment capacity. Early containment failure, which might be a failure to isolate a preexisting leak or a steam explosion, would be characterized by high radioactivity concentration but is a low probability scenario. Even filtered venting would be difficult because it does not provide sufficient relief.

C. P. Siess briefly discussed Commissioner Asselstine's suggestion for a containment design objective of 10' per reactor year. D. Okrent noted that the large dry and the Mark III containments are contenders for the best plant containments at present. C. P. Siess suggested that what should be done is to take a look at PRAs that have been done and see what they say about containment performance or containment effectiveness for the spectrum of accident scenarios. D. Okrent made a few suggestions.

He thought it useful to know what the Swiss are doing on containments. A second point of interest might be to see how the West Germans decide what volume and design pressure to choose for a PUR or BWR. He also noted that the Sandia National Laboratories have a containment event tree computer code which has in it all the chains of events that they could think of. It might be useful to see whether they have developed ideas for a PWR containment or a Mark III containment. He suggested that a contaicment performance criterion might stipulate that scenarios that are expected to lead to early containment failure and core melt have to be well scrutinized and their overall probability less than sone particular number. C. P. Siess noted that the overall probability of release is the prime concern.

C. P. Siess suggested that a good starting figure for the pgobability of large releases, on a scenario-dependent basis, might be 10 per reactor year. This figure might incorporate core melt, core performance, and defense-in-depth.

I 316TH ACRS MEETING 11INUTES 26 J. C. Mark, regarding the subject of sabotage, suggested that for new plants something ought to be done about decay heat removal to make it persistent and protected. J. C. Ebersole suggested the possibility of a shutdown performed automatically from a security door lock-up system, barriers that no one could defeat, to intercept the proper function of that system. P. G. Shewmon suggested that this appears like an opera-tor's nightmare, because of the threat of spurious initiation. D. Okrent asked if the Germans or Swiss, who use systems like this, have the same number of security doors. D. A. Ward, indicated that the Swiss, Germans, and Dutch have bunkered decay heat removal systems that automatically lock up, but they lock up on a signal from the control room operator and can't be entered for a certain number of hours. C. J. Wylie indicated that, from his visits to Germany, he found that the plants did not have as many security doors as plants in the U.S. They have certain key doors that they can lock up, but they are normally open. They just monitor people's activity throughout the plant. He did think that bunkered decay heat renoval systems have merit, and not just from the standpoint of sabotage and security, but from the potential for relaxing many of the restrictions in these plants. This can lead to smoother operations. He noted that with the proper cable type specified, and it is available today, and with a dedicated multitrained bunkered decay heat removal system, one could forget about fire protection in these nuclear plants. There would not be fires in cable tray systems. G. A. Reed spoke in favor of a dedicated decay heat removal system because it cures a multitude of problems for nuclear plants. He suggested that the containment must be a part of the dedicated decay heat removal system itself. J. C. Mark thought that one my to assure security would be to rely on as few people as possible in the plant, have a minimum number of people whose background is very stringently researched associated with fundamental backup decay heat removal and criticality control.

l C. flichelson suggested that a bunkered decay heat removal system could be l easily compromised by the vulnerability to sabotage of a few systems outside of containment. The Committee requested that C. Michelson identify such systems which are located outside of containment.

l C. P. Siess said that sabotage is just another accident scenario that conceivably could be mitigated by the same mitigation features that are used for other accident scenarios. D. Okrent said "not always." C. P.

Siess suggested that perhaps sabotage could be considered as a special accident scenario because the sabotage could be directed to the mitigat-l ing feature at the same time as to accident initiation.

D. W. Moeller noted that the Commission has issued a safety goal policy statement as well as an advanced nuclear power plants policy statement.

The latter was issued on July 8, 1986 and he indicated that he had many problems with it. D. Okrent agreed that you might argue that the two policy statements are not completely compatible. He suggested that the Committee should develop its case, get agreecent around the table, and inform the Commission why. D. W. Moeller noted the change in some key

316TH ACRS MEETING MINUTES 27 words in the policy statement on advanced nuclear power plants from operation of plants "without undue risk" to should be able to be operated with "no undue risk" to the health and safety of the public. He noted that there is a significant difference from this latter statement than from the traditional former statement. He also noted that both policy statements ought to speak to the protection of the health and safety of the workers at the plant as well as to the public. He noted that he found the policy statements inadequate in terms of the Soals for better protection of workers. J. C. Mark noted that the policy statement about new reactors should absolutely have some comment about sabotage mitiga-tion in it. The Cormittee scheduled three additional hours for dis-cussion of this topic at the 317th ACRS meeting.

IX. NRC Regulatory Process (0 pen)

[ Note: G. R. Quittschreiber was the Designated Federal Official for this portion of the meeting.]

W. Kerr presented some thoughts on the regulatory process in response to the request by the Cenmittee that he draft some individual comments on the subject. He suggested a letter for Committee approval which would inform Chairman Zech that the ACRS intends to undertake an indepth study of the regulatory process (see Appendix X). C. P. Siess objected to the tone of the letter which indicated that the ACRS intends to do a better job on this issue. W. Kerr countered that it implies that the ACRS intends to study the situation before making comments and recommendations to the Commission. C. P. Siess mentioned a recent letter from the NRC Chairman about the duties of the ACRS. He suggested that what W. Kerr is proposing is not inconsistent with the charge from Chairman Zech.

D. A. Ward noted a menorandum from G. R. Quittschreiber which suggested that the NRC undertake a study. W. Kerr agreed that Mr. Quittschreiber's letter did differ from his remarks which involved the ACRS studying the issue and eventually making some recommendations for changes. He thought that the ACRS ought to do some preparatory work before counseling the Commission. C. P. Siess thought that the Committee ought not to send the Commission a laundry list such as that included in G. R. Quittschreiber's letter (see Appendix IX). W. Kerr thought G. R. Quittschreiber's list quite appropriate in that it puts on record the fact that the ACRS knows that studies have been made and notes the ones that have been recognized.

G. A. Reed supported W. Kerr's approach rather than the one by G. R.

Quittschreiber. D. W. Moeller thought another approach would be to take G. R. Quittschreiber's memo and amend it to indicate that the ACRS is t setting up a group that wculd interact with a designated section within the NRC on this issue.

l W. Kerr cited the Davis-Besse and TVA situations as examples of failures of the regulators. The Staff knew of the problems but did not enforce the regulations properly. C. J. Wylie endorsed the urdertaking of an ACRS study to decide what the Committee wishes to ao about this issue.

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o 316TH ACRS MEETING MINUTES 28 C. J. Wylie pointed to the Browns Ferry fire incident as an example of poor judgment on the part of TVA in the use of cables--no one else in the industry used that kind of cable. D. A. Ward suggested that if it was important to use that kind of cable there should have been a regulation to provide assurance of its use. Both C. P. Siess and W. Kerr argred that the industry may be overregulated. C. Michelson noted that the experience is on the bad side where the industry is not regulated.

W. Kerr admitted that that might be true for some utilities, but not for all of them. C. P. Siess expressed the concern that there are cases where there is not a clear-cut relationship between regulation and safety; there is a clear-cut assumption that if regulations are satisfied it is safe. He indicated that he had the same concerns about QA and safety. He also noted that, since the ACRS has contributed as much to regulatory complications as anyone else, perhaps an independent outside body ought to be the one to investigate the relationship between safety and regulation. C. J. Wylie noted that some of the major problems are not the fault of the NRC Staff but stem from the failings of utility management. W. Kerr thought that the NRC cught to have either an explic-it or implicit responsibility for developing safe nuclear power. The artificial separation of the responsibility for promoting, as opposed to regulating, nuclear power has led to an adversarial relationship which could be considered very counterproductive to both safety and reliabil-ity. W. Kerr also decried the enforcement policy through the use of fines and levies against organizations. D. W. Moeller suggested that this subject is heavily intertwined with the Committee's attempt to make comments on what the agency should have in a long-range plan. He sug-gested the long-range plan issue ought to be discussed before any de-cisions are made regarding recommendations to the Commission on the regulatory process. F. J. Remick stated that he thought there ought to be a comprehensive look at the regulatory process, but he thought this might be a full-time job for this Committee. He agreed that there are opportunities for improvement, especially with regard to overregulation, but he thought the Committee ought to take account of its limited avail-ability when assessing the job that needs to be done.

D. A. Ward made a motion regarding whether the ACRS should get involved in a study of the regulatory process. The members unanimously agreed to proceed. He then asked whether the Committee knows enough about the issue to develop a set of recomendations for the Commission. W. Kerr suggested that the Committee ask the Subcommittee on Regulatory Policies and Practices to formulate a set of recommendations. D. A. Ward asked how many subccmittee meetings the members thought would be necessary.

C. Michelson thought three or four. C. J. Wylie thought four subcommit-tee meetings would be appropriate. G. A. Reed recomended that the Subcomittee on the State of Nuclear Power Safety handle the sut, ject.

D. A. Ward took it as a consensus of the Committee that the subject be assigned to H. W. Lewis's Subcommittee on Regulatory Policies and Prac-tices.

316TH ACRS MEETING MINUTES 29 X. Long Range Plan (0 pen)

[ Note: R. K. Major was the Designated Federal Official for this portion of the meeting.]

M. W. Carbon reviewed the history of ACRS' involvement in NRC long-range planning which began about one and three-quarter years ago. He mentioned the Comission's conclusion that it needed a long range plan and its directions to OPE and ACRS to work together to develop one. He mentioned the fact that the joint effort of OPE and ACRS encountered administrative problems, and ended when OPE was abolished. He noted that it was his belief that Comissioner Zech is a strong proponent of long-range plan-ning.

R. F. Fraley, ACRS Executive Director, discussed a strategic planning meeting that was held August 1-2, 1986 to deal with a five-year strategic plan for the NRC. All office directors and regional office administra-tors were present. The Comission hired an organization knowledgeable in the mechanisms of strategic planning called the Strategic Decision Group, and it was members of this professional consultant group that led the meeting. Participants at the sessions were encouraged to identify areas of major concern about the Agency's performance, the safety of nuclear power, and the regulatory process. About 200 items were reduced by the group to 7 issues thought to be the most significant for the future of the Agency and the nuclear regulatory program. These items included translating the quantitative safety goals into a regulatory program, as well as deriving a better definition of the responsibilities and the role of various units within the regulatory process. These lists of items associated with the 7 issues will now be integrated into a plan of implementation including resources, time estimates, objectives, and milestones. It would then be implemented by the Commission as its five-year plan. He noted that this appears to be a mechanism that the l Commission may use in development of a long range plan that is truly i strategic in nature, rather than just an extrapolation of existing programs.

M. W. Carbon continued his background history of the ACRS' involvement in NRC long-range planning. From interviews with a dozen or so individuals during meetings of the ACRS Subcomittee on the Long Range Plan and the ACRS' cwn internal thoughts, it was decided that the ACRS should not and could not develop a long range plan, that only the Comissioners them-l selves should and could prepare such a plan. The Subcommittee directed l its efforts toward developing guidelines for an NRC Long Range Plan with the intention that these guidelines be forwarded to the Commissicners.

He suggested that, at this time, the Comittee should decide if it is satisfied with the general direction that the Subcomittee is taking. He presented a list of proposed guidelines which he suggested the full Comittee examine. The ACRS should decide whether it wants to forward them as a full Committee document or as Subcomittee comnents only, or terminate the effort entirely. He indicated his desire for the first

316TH ACRS MEETING MINUTES 30 alternative, with the objective of finalizing the letter to the Commis-sioners by the October meeting. D. A. Ward suggested that the ACRS Executive Director inform J. Rowe, Deputy EDO, that the ACRS is working on a report on an NRC Long Range Plan and expects to issue the report by its October meeting.

The Committee discussed the fonn and substance of a report to the Commis-sioners on an NRC Long Range Plan. M. W. Carbon suggested that the Committee discuss a section entitled "What Could Be Done Differently?"

which was a list of suggestions for the Commission, and a section enti-tied " Establish Goals and Objectives" which suggests directions for the Comissioners to take. He indicated that he thought that there should not be a set of recommendations. W. Kerr thought there wouldn't be any other reason to write the letter besides recommending something.

M. W. Carbon noted that the ACRS has made recommendations in the past -

that they develop a long range, or strategic, plan and they took the Committee's advice and held an initial meeting on August 1-2, 1986.

Before the meeting was adjourned, the Committee decided to discuss this draft letter at the next ACRS meeting.

XI. Executive Sessions (0 pen)

[ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.]

A. Subcommittee Assignments

1. Conduct of ACRS Meetings The members endorsed a practice whereby transcripts will be taken of those ACRS executive session discussions which are held among the members and an extended discussion of members' positions are expected. This record will be useful to the subcommittee chairman in formulating a proposed report, etc.

B. Reports, Letters, and Mecoranda

1. ACRS Comments on the Standardization Policy Statement -

The Committee prepared a report to the Comissioners of its re-view of the Proposed Standardization Policy Statenent as re-quested by the Chairman's memorandum of June 18, 1986. Addi-tional comments by D. Okrent and G. A. Reed were appended.

2. ACRS Report on the Tennessee Valley Authority's (TVA) Manage-ment Reorganization and Shutdown of TVA's Nuclear Pcwer Plants The Committee prepared a report to the Commissioners of its review of the reorganization of TVA's management structure and

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316TH ACRS MEETING MINUTES 31 the issues related to the shutdown of TVA's nuclear power plants. Additional comments by G. A. Reed and D. A. Ward were appended.

3. ACRS Comments on the NRC Policy Statenent on Fitness for Duty of Nuclear Power Plant Personnel The Comittee prepared a report to the Comissioners of its review of the NRC Policy Statement on Fitness for Duty of Nuclear Power Plant Personnel. The ACRS wishes to be kept informed of the progress of the Staff initiative to develop a fitness for duty program which would apply to now-exempt NRC employees who have unescorted access to vital areas in nuclear power plants.

4. ACRS Actions on Regulatory Guide 1.114, Revision 2, " Guidance to Operators at the Controls and to Senior Operators in the Control Room of a Nuclear Power Unit" The Comittee approved a memorandun to the ED0 in which it concurs in the Regulatory Position of Regulatory Guide 1.114, Revision 2, dated June 2, 1986, and in the conforming revisions to the Standard Review Plan Section 13.1.2.

5. ACRS Comments on Various NMSS and RES Waste Management Topics The Comittee prepared a letter to the ED0 regarding the following topics which were discussed during a report of the July 21-23,1986 meeting of the Subcommittee on Waste Manage-ment:

a) NMSS radioactive waste management program, including the Division of Haste Management's five-year plan, the pro-posed Federally Funded Research and Development Center (FFRDC), proposed use of rulemaking to bring key preli-censing issues to closure, and alternative methods to shallow land burial.

b) Programs of the Waste Management Branch, Office of Nuclear Regulatory Research (RES), including the development of field data on the movement of radionuclides within the environment and the associated impact of heat-water-rock interactions, and the predicted performance of repository systems under realistic field conditions.

c) Generic technical positions en " Determination of Radionuclide Sorption" and "Detemination of Radionuclide Solubility" for high-level nuclear waste repositories.

o e

316TH ACRS MEETING MINUTES 32 d) Development of residual radiation limits for the disposition of land, buildings, equipment, and metals resulting from the decontamination and decommissioning of nuclear power plants and fuel facilities.

e) Salvaging of contaminated smelted alloys containing technetium-99 and/or low-enriched uranium as residual radioactive contamination.

f) NRC Staff policy statement and implementation of NRC policy on radioactive wastes "below regulatory concern."

6. ACRS Comments on Loss of Power and Water Hammer Event at San Onofre, Unit 1 The Committee met with representatives of the NRC Staff and the Licensee to discuss the loss of power and the resulting water hammer and loss of feedwater event which occurred at the San Onofre Nuclear Generating Station, Unit 1, on November 21, 1985. The ACRS concluded that an ACRS report was not warranted regarding this matter.

C. Generic Issues

1. ACRS Comments on the Resolution of USI A-46 (Seismic Qualification of Equipment in Operating Plants)

The Committee discussed a draft report to the Commissioners of its review of the proposed resolution of USI A-46 (Seismic Qualification of Equipment in Operating Plants). Time did not permit full consideration of this report, and it will be scheduled for further discussion during the 317th ACRS meeting (September 1986).

2. Improved Light Water Reactors During a discussion of proposed ACRS comments on characteris-tics of improved light water reactors (LWRs), G. A. Reed spoke in favor of a dedicated bunkered shutdown decay beat renoval system for future LWRs as essential to enhanced plant safety and sabotage protection. The containment would be part of such a dedicated decay heat removal system. C. Michelson suggested that such a bunkered system could be easily compromised by the vulnerability to sabotage of a few systems outside of contain-ment. The Committee requested that C. Michelson identify such systems which are located outside of containment.

C 316TH ACRS MEETING MINUTES 33

3. D_gign Performance of Containments D. Okrent requested that information be obtained about the Swiss pressure suppression containment regarding the sup-pression pool arrangements within their dynamic containments.

He also asked that information be provided regarding any changes which the FRG has incorporated in volume and pressure requirements for their BUR contalment to accommodate serious accidents.

D. Future Agenda

1. Future Agenda The Committee agreed on tentative agenda items for the 317th j ACRS Meeting, September 11-13, 1986 (see Appendix II).

2. Future Subcommittee Activities A schedule of future subcomittee activities was distributed to members (see Appendix III).

E. Proposed Power Level Increase for the North Anna Nuclear Station, Units 1 and 2 During the discussion of anticipated ACRS activities for the 317th ACRS meeting September 11-13, 1986 note was taken of a proposed power level increase at the North Anna Station, Units 1 and 2 from 2775 MWt to 2893 MWt. J. C. Ebersole, Chairman of the Reactor Operations Subcomittee, concluded that a review of this power level increase did not appear necessary. The Committee concurred in this recomendation that no action was necessary regarding this power level increase. Several items were discussed as a possible basis for a review of North Anna operations, but the Committee decided not to pursue them at this time. The ACRS Executive Director was instructed to inform the NRC Staff regarding this decision.

F. RSK Report on the Chernobyl Accident D. Okrent called the following quote from the RSK paper entitled

" Intermediate Report of the Reactor Safety Comission about a Preliminary Evaluation of the Accident at the Chernobyl Nuclear

Pcwer Plant as it Affects Nuclear Power Plants in the Federal Republic of Germany," dated June 6,1986, to the attention of the ACRS members:

"The RSK also believes it is appropriate that for pressurized water reactors specific installation recommendations be made, such as i could prevent the failure of the containment after a core meltdown with a slow pressure build-up. The advantages and disadvantages of these measures have to be exanined in detail."

a 316TH ACRS HEETING MINUTES 34 The 316th ACRS meeting was adjourned at 3:05 p.m., Saturday, August 9, 1986.

_)

APPEllDICES

,1 TO 5 THE 316TH ACRS MEETING

[' ' AUGUST 7-9, 1986 hbf$#kY5f e

3 0

e

APPENDIX I ATTENDEES NRC ATTENDEES 316TH ACRS MEETING Thursday, August 7, 1986 0FFICE OF NUCLEAR REACTOR REGULATION g V. Scale Hi

0. D. T. Lynch, Jr.

R. Hernan 6 N. R. Anderson R. Bosnak T. Y. Chang H. Reponen 1

O q .

i k~l i

)

PUBLIC ATTENDEES 3I6TH ACRS MEETING Thursday, August 7, 1986 E. Fotopoulos, SERCH Licensing, Bechtel H. Eckert, NUS R. Smith, IEM .

g G. A. Brown, Stone & Webster Takashi Takedor, JAERI L. Cuoco, Frank Fried.

. M. Paris, New York Power Authority J. C. Carter, III, I. T. Corporation J. E. Thomas, Duke Power D. E. Schaffstall, KMC, Inc.

W. K. Schmidt, MPR Assoc.

T. G. Roberson, Houston Lighting & Power L. Zeidman, JARVIS Corp.

G. G. Young, YES, Inc.

D. H. Williams, Arkansas Power & Light Company R. Smith, IEAL i

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9 NRC ATTENDEES 316TH ACRS MEETING Friday, August 8,1986 0FFICE OF NUCLEAR REACTOR REGULATION o R. Dudley '

A. W. Serkiz R. Hernan

. B. A. Boger P. C. Goodman '

K. G. Ramirez A. Bournia T. Collins OFFICE OF INSPECTION AND ENFORCEMENT

11. Bailey E. Weiss J. Posentha D. Allison e

INVITED ATTENDEES 316TH ACRS MEETING Fridai, August 8, 1986 SOUTHERN CALIFORNIA EDISON J. Rainsberry g M. O. Medford e

en A-f

l 1

I i

PUBLIC ATTENDEES i

316TH ACRS MEETING i

Friday', August 8,1986 L. Cuolo, Fred, Frank

R. Smita, IEAL S. Letourneau, Bechtel g L. Connor, DSA M. Paris, New York Power Authority C. J. Allen, NUS Corp.

. D. L. Lambert, TVA R. S. Powelson, Knoxville News-Sentinel

=

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APPENDIX II g FUTURE AGENDA APPENDIX A FUTURE AGENDA SEPTEMBER ACRS MEETING MeetingwithNRCCommissioners--DiscussACRSreport 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dated August.12, 1986 on NRC Policy Statement on Standard Nuclear Plants Proposed Revision of ECCS Requirements -- Proposed 2-3 hours o revision of 10 CFR 50.46, Acceptance Criteria for ECCS for LWRs and Appendix K, ECCS Evaluation Models

- NRC Regulatory Process and Procedures -- Report of 3/4 hour subcommittee regarding the IIT review of Davis-Besse loss of feedwater incident. Discuss proposed response to Comissioner Asselstine's inquiry of ACRS regarding adequacy of regulatory processes associated with the design, review / approval of, and the NRC inspection and enforcement program regarding Davis-Besse Nuclear Station along with a '

proposal, which was adopted by the Comittee, for con-sideration of an ACRS review of the existing regulatory process with the aim of providing some recomendations for IO-l-%

a more comprehensive program to the Comissioners. This was a spin-off from W. Kerr's proposal for a review of the existing system of NRC regulations and requirements, and the NRC Staff's enforcement of these requirements.

Improved Light Water Reactors -- Continue discussion of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> proposed ACRS coments regarding characteristics of advanced nuclear power plants Severe Accident Policy Implementation Plan / Removal of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Radioactivity f rom Containment Atmosphere -- ACRS comments requested

! B&W Water Reactors -- B&W Owners Group presentation 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> regarding reevaluation of B&W reactors' long-term safety Briefing Regarding Chernobyl Reactor Accident -- Report 1-It hours

~

of IAEA Meeting with USSR representatives Long-Range Plan -- Continuation of discussion of the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> proposed ACRS guide for a long-range plan ACRS Management Subcomittee Meetings of July 9,1986 and September 10, 1986 -- Discuss proposed changes in ACRS procedures and practices, and work priorities Aptitude Testing -- Discuss proposed ACRS comments 1/2 hour regarding use of aptitude testing in selection of nuclear plant personnel k '~b 1 -

. ... =.

316th ACTIONS & AGREEMENTS Seabrook.. Nuclear Plant, Units 1 and 2 -- Briefings deferred of ACR5 regarding proposed change in emergency to October planning zone NRC Staff Briefing on the French Paluel Reactor -- deferred Report regarding 25 safety " improvements" to October n

Briefins recardins the NRC General Council response deferred i to the Pay 27,19E6 ACR5 request for a legal opinion to October

- on Backfit Rule applicability to USI A-17, Systems Interactions Subcommittee Reports on:

Maintenance Practices and Procedures -- Completion of 1/2 hour Phase I of the NRC Maintenance and Surveillance Program Plan Metal Components -- Research related to primary system I hour integrity IE Programs -- Functional Inspection of Safety Systems I hour and Status of the SALP Program Decay Heat Removal -- Generic Issue A-124, Auxiliary 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Feedwater System Reliability (An ACRS report is likely to be proposed)

Scram System Reliability -- Status of the ATkS Rule deferred implementation effort to October 1

Extreme External Phenomena -- Report of meeting to review

, the importance of seismic risk to nuclear power plants Procedure and Administration -- August 9,1986 meeting r

to consider the Effectiveness Panel's recommendations regarding ACRS officers' tenns, and the makeup of the ACRS Management Subcomittee

'

• Containment Performance Requirements -- NRC draft position deferred paper on containment performance design objective as an to October j

addition to the Safety Goal Policy Schedule for ACRS Activities -- Discuss proposed ACRS 1/4 hour meeting dates for CY 1987 A-7 1

AUG 0 81986 7 ...

APPENDIX III ACRS SUBCOMMITTEE MEETINGS

'ACRS SUBCOMMITTEE MEETINGS Maintenance Practices and Procedures, August 13, 1986, 1717 H Street, NW, Washington, DC (Alderman), 1:00 P.M., Room 1046. The Subcommittee will review the report on Phase I of Maintenance Program Plan. Attendance by the following is anticipated, and members will arrive the morning of August 13:

Mr. Reed Mr. Wylie Mr. Michelson I&E Programs, August 14, 1986, 5th Floor Hearing Room, East West Towers -

West Building, 4350 East West Highway, Bethesda, MD (Boehnert), 10:30 A.M.

The Subcommittee will review I&E Programs with focus on the Safety System Function Inspection (SSFI) Program, and the risk-related inspection methodology. A tour of the I&E Incident Response Center is also planned.

Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of August 13:

Mr. Reed DAYS INN Mr. Michelson DAYS INN Mr. Ebersole CARLYLE Mr. Wylie DAYS INN Nuclear Plant Chemistry, August 26, 1986, 1717 H Street, NW, Washington, DC (Alderman), 8:30 A.M., Room 1046. The Subcomittee will discuss various topics relevant to plant chemistry, i.e., Na0H in containment spray, suppression pool scrubbing, and chemical engineering research. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of August 25:

Dr. Moeller CARLYLE Mr. Etherington HOLIDAY INN Mr. Ebersole CARLYLE Mr. Reed DAYS INN Thermal Hydraulic Phenomena, August 27, 1986, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M., Room 1046. The Subcommittee will continue its review of the RES-proposed revision to the ECCS Rule (10 CFR 50.46 and Appendix K). Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of August 26:

Mr. Michelson DAYS INN Dr. Catton DUPONT PLAZA

. Mr. Ebersole CARLYLE Mr. Schrock NONE Mr. Reed DAYS INN Dr. Sullivan NONE Mr. Ward NONE Dr. Tien NONE Decay Heat Removal Systems, September 9, 1986, 1717 H Street, NW, Washington, DC (Boehnert), 1:00 P.M., Room 1046. The Subcommittee will review NRR's7ction Plan to address concerns with the reliability of certain plants' AFW systems. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of September 8:

Mr. Ward NONE Mr. Reed DAYS INN Mr. Ebersole CARLYLE Dr. Catton DUPONT PLAZA Mr. Michelson DAYS INN (9/9) Mr. Davis NONE A-r

Management Subcommittee (CLOSED), September 10, 1986, 1717 H Street, NW, Washington, DC (Fraley), 8:30 A.M., Room 6110 Maryland National Bank Building. Specific topics have not been selected as yet. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of September 9:

Mr. Ebersole CARLYLE Mr. Ward NONE

= Dr. Lewis HYATT 317th ACRS Meeting, September 11-13, 1986, Washington, DC, Room 1046. ,

Containment Requirements, September 23, 1986, 1717 H Street, NW, Washington, DC (Houston), 8:30 A.M., Room 1045. The Subcommittee will review a draft position paper on containment performance design objective as an addition to the Safety Goal Policy, and a draft of a proposed generic letter on Mark I containment requirements for severe accidents.

Lodging will be announced later. Attendance by the following is anticipated:

Dr. Mark Dr. Okrent Mr. Ebersole Dr. Siess Dr. Kerr Mr. Wylie Severe (Class 9) Accidents, September 24, 1986, 1717 H Street, NW, Washington, DC (Houston), 8:30 A.M., Room 1046. The Subcommittee will review the NRR Implementation Plan for Severe Accidents and the IDCOR Methodology for Individual Plant Evaluation. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Kerr Mr. Ward Dr. Carbon Mr. Bender l Dr. Mark Dr. Catton Dr. Okrent Dr. Corradini Dr. Shewmon Dr. Davis Dr. Siess International Operating Experience, September 25, 1986 - CANCELLED Westinghouse Reactor Plants (CLOSED), September 25, 1986, 1717 H Street, NW, Washington, DC (El-Zeftawy), 8:30 A.M., Room 1167. The Subcommittae will begin the PDA review of the Westinghouse Advanced Pressurized Water Reactor (RESAR SP/90). Attendance by the following is anticipated:

Mr. Reed Mr. Wylie Mr. Michelson Mr. Davis Dr. Shewmon AA

Joint Seabrook/0ccupational and Environmental Protection Systems, September 25,1986,1717 H 5treet, NW, Washington, DC (Major /Igne),

1:00 P.M., Room 1046. The Subcommittees will gather and exchange information with the NRC Staff and PSNH. Under consideration will be efforts by the applicant to reduce the size of the EPZ. This effort uses results of the Seabrook Probabilistic Safety Assessment to justify a smaller EPZ. A primary focus will be the credit taken for the strength of the Seabrook containment. A status report on the Seabrook EPZ will also be discussed. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Kerr Dr. Siess Dr. Moeller Decay Heat Removal Systems, September 26, 1986, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M., Room 1046. The Subcommittee will continue its review of NRR's proposed resolution position for USI A-45,

" Shutdown Decay Heat Removal Systems." Lodging will be announced later.

Attendance by the following is anticipated:

Mr. Ward Dr. Catton Mr. Ebersole Mr. Davis Mr. Michelson Reactor Operations, October 8, 1986, 1717 H Street, NW, Washington, DC, (Alderman), 8:30 A.M., Room 1046. The Subcommittee will review recent operating events at nuclear reactors. Lodging will be announced later.

Attendance by the following is anticipated:

Mr. Ebersole Mr. Reed Mr. Michelson Dr. Remick Dr. Moeller Mr. Wylie 318th ACRS Meeting, October 9-11, 1986, Washington, DC, Room 1046.

i Wingspread International Conference (CLOSED), October 19-23, 1986, Racine, WI (McCreless). Representatives from the ACRS, RSK, GPR, and Japan will

. Echange information on nuclear reactor safety.

Waste Management, October 30-31, 1986, 1717 H Street, NW, Washington, DC (Merrill), 8:30 A.M., Room 1046. The Subcommittee will review pertinent radioactive waste management topics as to be determined during the agenda-planning meeting with the Division of Waste Management in late August, 1986. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Moeller Dr. Shewmon i

Dr. Carbon Dr. Carter Dr. Kerr Dr. Foster Dr. Mark Dr. Orth Mr. Reed Dr. Steindler Dr. Remick g

319th ACRS Meeting, November 6-8, 1986, Washington, DC, Room 1046.

Metal Components, November 12 (tour) and 13, 1986, near Pittsburgh, PA (on 12th)andWashington,DC(on13th?(Igne). The Subcomittee will visit Beaver Valley 2 on the 12th and (:.) hear a status report of the Whipjet program (application of broad scope GDC-4 criteria) as applied to lead

, plant BVPS-2; and (2) review public comments on NUREG-0313, Rev. 2 (long range fix for BWR-IGSCC problems) per ACRS letter. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Shewman Dr. Okrent Mr. Etherington Mr. Ward Dr. Lewis Mr. Bender Mr. Michelson Mr. Rodabaugh Extreme External Phenomena, November 20, 1986, 1717 H Street, NW, Washington, DC (Savio), 8:30 A.M., Room 1046. The Subcomittee will continue its review of the Diablo Canyon long-term siemsic program.

Lodging will be announced later. Attendance by the following is anticipated:

Dr. Siess Dr. Lewis Dr. Carbon Dr. Moeller Mr. Etherington Mr. Wylie S)ent Fuel Storage, November 21, 1986, 1717 H Street, NW, Washington, DC (ierrill), 8:30 A.M. , Room 1046. The Subcomittee will continue its review of 10 CFR Part 72 and Monitored Retrievable Storage (MRS). Attendance by the following is anticipated:

Dr. Siess Dr. Remick Dr. Kerr Dr. Shewmon

! Dr. Moeller Safety Research Program (CLOSED), December 10, 1986, 1717 H Street, NW, Washington, DC (Duraiswamy), 8:30 A.M., Room 2046. The Subcomittee will discuss the following and gather information for use by the ACRS in its

. preparation of the annual report to the Congress on the NRC Safety Research Program and budget for FY 1988: (1) proposed NRC Safety Research Progran and and budget for FY 1988, (2) preliminary OMB Mark and the impact of the

! . OMB-proposed reductions on the continuing and proposed research, and (3)

RES responses to ACRS recommendations contained in its June 11, 1986 report to the Comission. Lodging will be announced later. Attendance by the j following is anticipated:

1 Dr. Siess Dr. Okrent

! Dr. Carbon Dr. Remick l

Dr. Kerr Dr. Shewmon Mr. Michelson Mr. Ward Dr. Moeller Mr. Wylie A-//

l 320th ACRS Meeting, December 11-13, 1986, Washington, DC, Room 1046.

Regional bperations, Date to be determined (September / October), Chicago, IL (Boehnert). The Subcommittee will begin its review of the activities of the NRC Regional Offices. This meeting will focus on the activitie, of the Region III Office. Attendance by the following is anticipated:

Dr. Remick Mr. Reed Dr. Carbon Mr. Wylie Mr. Michelson AC/DC Power Systems Reliability, Date to be determined (October / November),

Washington, DC (El-Zeftawy). The Subcommittee will review the proposed Station Blackout rule (SECY-85-163). Attendance by the following is anticipated:

Dr. Kerr Mr. Reed Mr. Ebersole Mr. Wylie Dr. Lewis Seabrook Units 1 and 2, Date to be determined (fall / winter), Washington, DC TMajor). The Subcommittee will review the application for a full power operating license for Seabrook land 2. Attendance by the following is anticipated:

Dr. Kerr Dr. Moeller Dr. Lewis Mr. Michelson Metal Components, Date to be determined (November / December), Oak Ridge, TN, (Igne). The Subcommittee will review the HSST program, including dosimetry program by HEDL. Attendance by the following is anticipated:

Dr. Shewmon Dr. Okrent Mr. Etherington Mr. Ward Dr. Lewis Mr. Bender Mr. Michelson Structural Engineering, Date to be determined (early 1987), Albuc uerque, NM (Igne). The Subcommittee will review containnent integrity anc Category T structures programs and test facilities. Attendance by the following is anticipated:

Dr. Siess Dr. Okrent Dr. Carbon Dr. Shewmon Mr. Ebersole Mr. Bender Dr. Kerr Dr. Pickel A -/.o

NRR STAFF PRESENTATION TO THE ACRS APPENDIX IV NRR PRESENTATION ON SONGS-1 I WATER HAMMER EVENT

SUBJECT:

SAi1 Of10FRE 1 -

NCVEFSER 21, 1985 WATER HAlliiER EVEi;T DATE: AUGUST 8, 1986 PRESENTER: RICHARD DUDLEY PRESENTER'S TITLE / BRANCH /DIV: Sail ON0FRE 1 PROJECT MAi1AGER

- PAD #1: DPLA PRESENTER'S NRC TEL. NO.: 49-27218 l

\

SUBCOMMITTEE: WESTillGHOUSE WATER REACTORS

\ A -/3 i

i f

_. NOVEMBER 21, 1985 SAN ON0FRE UNIT 1 SEQUENCE OF EVENTS -

SUMMARY

INITIAL CONDITIONS

. Reactor Power 60%

S.G. continuous blowdown in progress (100 gpm)

Troubleshooting ground indication on 4ky (abnormal electrical alignment)

. Fox 3 computer system disabled Five check valves in feedwater/AFW have undetected seat failures at this time TRANSIENT INITIATOR 0451 Aux Transformer C phase to phase fault 4kv bus 2C automatically deenergized SYSTEM RESPONSE 0451 Loss of 4ky bus 2C deenergized:

- East Feedwater Pump

- 120 VAC Vital bus 4 ENS phone rings spuriously on loss of power Diesel Generator 2 starts automatically but does not load per design

_ East feedwater pump discharge check valve fails to seat as pump coasts down Running west feedwater pump pressurizes the east condensate train.

East flash evaporator ruptures 0451 Reactor manually tripped per procedure Turbine trip / generator trip deenergizes remaining 4ky buses (1C, lA, IB)

All inplant power lost except for 120 VAC vital buses (battery / inverter)

Diesel Generator 1 starts automatically but does not load per design i

Electric and steam driven AFW pumps receive automatic initiation signals .

- Steam pump begins 31 minute warm up period

- Electric pump has no motive power As west feedwater pump stops, all steam generators begin to empty their feedwater lines to the ruptured flash evaporator Spurious annunciation of safety injection j/_./q(

Operators begin attempts to restore offsite power

. 0455

• Steam driven AFW pump begins to deliver 130 gpm flow. Flow is lost to the ruptured flash evaporator Operators determine that the Loss of Voltage Sequencer did not complete sequence. Attempt closing breakers from control room with numerous difficulties, succeeding on fourth try.

• Station power restored 0455 Electric AFW pump starts Operators shut feedwater isolation valves per procedure, unknowingly stopping further voiding of steam generators AFW flow starts filling feedwater lines RCS pressurizer pressure and level noted low 0458 South charging pump started 0501 Started RCP B to provide pressurizer spray

. 0502 Tenninated AFW flow to minimize RCS cooldown. Resumed AFW flow at 25 gpm.

0506 OcenseedeclaredanUnusualEventonsite 0507 A loud bang was heard in turbine building mezzanine. The water hammer had occurred in the feedwater line. Steam / water leak occurs in a check valve bonnet 0527 Wide range level indication dropped offscale low in all three steam generators 0528 Increased AFW to S.G. A and C to 70 gpm Blowdown from S.G.'s secured 0530

• Wide range level returned on scale on S.G.'s A and C 0630 Started emergency boration for cold shutdown

. 0835 Secured steam driven AFW pump due to low steam pressure 0910 Attempted to open RHR suction valves; but pressure interlock did not clear although pressure was below 400 psig 0918 RHR interlock overriden, RHR valves opened .

0935 RHR pumps started 0940 Unusual Event terminated A -tf t

1045 Feedwater leak through check valve manually isolated 1508 The plant entered Mode 5 (cold shutdown)

November 22, 1985 0100 Operators entered containment sphere and identified damaged pipe, supports, and insulation on the "B" S.G. feedwater line O

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t ACRS PRESEflTATI0fl Of' SOf!GS-1 RESTART 4 .

AUGUST 8, 1986 ELLIS W. MERSCH0FF i VPB/IE

- 492-9045 l

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IE REVIEW 0F GENERIC ASPECTS

1. APPROACH _

1, INSPECT VEflDORS MCC PACIFIC ATWOOD a MORRILL VELAN CRANE ANCHOR DARLING ,

2. VISIT LICEf! SEES

~

SONGS-1 Il0RTH AtlNA MILLST0t!E HOPE CREEK tIMERICK PEACH BOTTOM

3. REVIEW FAILURE DATA NPRDS LER 50 55(E)

PAAT21 Pfl0 VE!! DOR RECORDS LICEllSEE RECORDS

4. REVIEW IN SERVICE TESTIt!G COMMITMENTS - SAMPLED 5 SYSTEMS AT 16 PLAtlTS IMPLEMEllTAT10tl - LOOKED AT TESTING PROCEDURES II, RESULTS

1. IDErlTIFI'ED POTE!!TI ALLY GEllERIC ISSUES LICEllSEE/VEllDOR lt!TERFACE WITH CHECK VALVE SUPPLIERS

- LARGELY INEFFECTIVE CHECK VALVES CHALLENGED BY N0f!-0PTIMAL:

l 0

• SIZit'G

!

• LOCATION

• DESIGil APPLICATI0fl i -

IST IflEFFECTIVE AT DETECING DEGRADED C0flDITIONS .

l

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l III. RESOLUTI0tl

1. MEET WITH OWilERS GROUPS E IrlP0 1 i

APRIL 7  !

MAY 21 I

2. ~IflDUSTRY PLAtl

- ISSUE SOER MAltlTEllAt!CE TESTIllG ISSUE DESIGfl AllD APPLICATI0fl GUIDE SIZIflG

- LOCATI0flS APPLICATI0f!S

3. flRC STAFF WILL C0llTittuE TO WORK WITH IflDUSTRY TO RESOLVE ISSUES e

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LICENSEE MODIFICATIONS T0 PREVENT FEEDLINE WATER HAMMER O PRIMARY PROTECTION - PREVENT FEEDLINE VOIDING BY FIXING CHECK VALVES -

- NEW DESIGN, IMPROVED IN-SERVICE TESTING, PERIODIC INSPECTION O FURTHER MODIFICATIONS TO PREVENT WATER HAMMER:

ADDITION OF CHECK VALVES INSIDE CONTAINMENT

~

' PREVENT VOIDING

- AUTO-CLOSURE OF FW REG. VALVES (LOMFW, T. TRIP, AFW INIT.)

REPLACE "B" FEEDLINE WITH PIPING SLOPED AWAY FROM STEAM GENERATdR i% -;t9

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WATER HAMMER REVIEW' 9

1 l -

AW. SERKIZ NRR/DSRO 8-8-86 -

ACRS SONGS-1 MEETING A -32

WATER HAMMERS 1969 to 1980 -

. 14 8 OCCURREXCES REPORTED

. FREQUEXCY = 0.29 / RX-YR

. UNDERLYIXG CAUSES WERE DESIGX & OPERATIOXS RELATED

. REVIEW OFDAMAGE IXCURRED

& SAFETY SIGXIFICAXCE WAS .

BASIS FOR A-1 RESOLUTIOX.

. NO BACKFIT REQUIREMEXTS 1

WATER HAMMERS 1981 to 1985 e

. . 40 OCCURRENCES REPORTED

. FREQUEXCY = 0.13 / RX-YR

. SIMILAR UXDERLYIXG CAUSES A-33 l

l

WATER HAMMER HISTO RY ANNUAL OCCURENCE' 20 LEGEND 18-- - EM BWRs 16- -

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WATER HAMMER HISTORY FREQUENCY OF .OCCURENCE

.co 2 LEGEND

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1 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 CALENDAR YEAR _ ._

3

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N UC LE AR POWER PLANT OPERATING HISTO RY C U M M U _ATIVE YEAR S

~

600  : LEGEND

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200 [ l O l g

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IU' F -' - -

0 - -

68 70 72 74 76 78 80 82 84 69 71 73 75 77 79 81 83 85 CALENDAR YEAR

Table 1: Reported Hamer Occurrences Reported (CY81 to CY85)

Water Hammer LER Date Plant Me No. System 03/28/81 Brunswick 1 BWR 325/81-034 HPCI 04/14/81 Brunswick 1 BWR 324/81-046 RHR 07/04/81 Nine Mile Point 1 BWR 220/81-031 RCIC 09/03/81 Dresden 2 BWR 237/81-057 HPCI

. 10/05/81 Dresden 2 BWR 237/82-061 CCSW 10/28/81 Hatch 2 BWR 366/81-104 HPCI 01/21/82 Arnold BWR 331/82-008 LPCI 02/19/82 Arnold BWR 331/82-014 RHR 05/09/83 Cooper .

BWR 298/83-006 RHR 06/04/83 Fitipatrick BWR 333/83-053 RHR 04/19/84 Cooper BWR 298/84-007 SBGT 09/06/84 Brunswick I BWR 325/84-023 CSS 09/08/84 Grand Gulf 1 BWR 416/84-042 SSWS 11/27/84 Brunswick 2 BWR 324/84-014 RHR 12/07/84 . Brunswick 1 BWR 325/84-034 MSS 03/31/85 Pilgrim 1 BWR 293/85-008 HPCI 04/27/85 Susquehanna 2 BWR 388/85-016 RHR 05/18/85 Pilgrim 1 BWR 293/85-012 HPCI 12/09/82 San Onofre 2 PWR-CE 361/82-165 SG 088 01/25/83 Maine Yankee PWR-CE- 309/83-002 FW #2 SG 04/21/84 Clavert Cliffs 2 PWR-CE 318/84-720 Main Feedwater 05/08/85 Waterford 3 PWR-CE PN0/V86-21 Emerg FW System 08/13/85 Arkansas Nuclear 2 PWR-CE 368/85-017 SG 10/04/85 San Onofre 2 PWR-CE 361/85-049 SDCS

- 05/04/81 San Onofre 1 PWR-WE 206/81-008 SG B 10/14/81 Indian Point 2 PWR-WE 247/81-023 SG 21 05/29/82 Turkey Point 3 PWR-WE 250/82-007 SG 3C 09/06/82 Robinson 2 PWR-WE 261/82-012 CVCS 10/07/82 Diablo Canyon 1 PWR-WE 275/82-009 ASW SYS 03/27/83 McGuire 1 PWR-WE 369/83-016 RHR 04/07/83 McGuire 1 PWR-WE 369/83-018 - SHR , .

10/06/83 Salem 2 PWR-WE 311/83-057 AFWS 12/05/83 Sumer 1 PWR-WE 395/83-138 FW 04/06/84 Salem 2 PWR-WE 322/84-015 MFW 08/05/84 McGuire 2 PWR-WE. 370/84-017 CYCS/R,HR

. 02/17/85 Diablo Canyon 1 PWR-WE 275/85-011 FW BYPAS 04/17/85 Salem 2 PWR-WE 311/85-005 SG Feed Pump 04/29/85 McGuire 2 PWR-WE 370/85-009 Main Steam 05/18/85 Diablo Canyon 1 PWR-WE 275/85-014 AFW 11/21/85 San Onofre 1 PWR-WE 206/85-107 MFW l

OVERALL WATERHAM MER i:.

.i.

STUDY COMPARISON .i.

~/  :-

i EPRI / ITI NRC / EG&G  :

=c r 2 _=,- - . , - - - .

s m ,:: .&._ _, _ -- . .-:. .=u: = : .,z . = .--- . . .

r = r .-- --- -

! DATES INCLUDED 1981 - 1985 1969 - 1981 Y CALENDAR YEARS S 12 .

BWR EVENTS 30 81  :::

PWR EVENTS 26 40  :::-

SGWH EVENTS 2 27  :

{

TOTAL EVENTS 58 148 ,: <

1 l

l

Reference:

EPRI PROJECT: RP-2856-1, July 1986 Note: This report contains no references to substantiate i number of water hammers cited to have occured in CYs 81-85, nor related inf ormation. )

); 38

1981 to 1985

. . 18 BWR OCCURREXCES

. 22 PWR OCCURREXCES

. UXDERLYIXG CAUSES:

.. VOIDED LINES

.. STEAM -

WATER ENTRAIXMEXT

_ .. SYSTEM SWITCH-OVERS

.. OPERATIOXAL ERRORS 6

A -39

~

CONCLUSIONS-

~

. S0$GS-1 11/21/85 WATER HA M MER WAS THE RESULT OF GROSS _

CHECK VALVE FAILURES

. REVIEW OF CY 81 -

85 EVEXTS HAS XOT REVEALED XEW PHENOMEXA

. FREQUEXCY OF OCCUREXCE HAS HAS DECREASED

~

. LEVELS OF DAMAGE SIMILAR

. NEW PLANTS WERE NOT MAJOR CONTRIBUTORS

. GEXERIC BASIS FOR RE-OPENIXG USI A-1 DOES XOT EXIST

/-46 l

WATER HAMMER ISSUE: OVERVIEW

SUMMARY

THE ISSUE IS WELL TRACKED AND MONITORED.

INSTITUTIONALLY IT IS BEING ADDRESSED.

IT IS-NOT A CATASTROPHIC IMPACT SAFETY PROBLEM. IT TENDS TO BE MORE AN ECONOMIC / AVAILABILITY /0PERATIONAL

. MATTER.

THE ISSUE WILL BE CONTROLLED OR CHALLENGED BY UTILITY .

OPERATIONAL ELEMENTS.

. IT IS VERY IMPORTANT THAT OPERATING EXPERIENCES BE TRACKED AND EVALUATED. THE LESSONS LEARNED SHOULD BE CANDIDLY SHARED AND HEEDED.

f

Reference:

Power,J.W., EPRI/NASC Presentation on Water Hammer j . Issue, Workshop on Water Hammer in Nuclear Power l Plants, Boston, Mass., July 9-10, 1986 l

I A -41 JWP 7/2/86

[

e'

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c UNITED STATES NUCLEAR REGULATORY COMMISSION .

)

E I WASHINGTc N, D. C. 20555 I

(!

\ . . . . j[ July 14, 1986 , '2' I

(

AEOD/E508 MEMOPANDUM FOR: Themis P. Speis, Director .. /

Division of Safety Review and Oversight -

Office of Nuclear Reactor Regulation FROM: Frederick J. Hebdon, Deputy Director .

Office for Analysis and Evaluation of Operational Data -

SUBJECT:

RE-EXAMINATION OF WATER HAMMER OCCURRENCES Enclosed is AE00 engineering evaluation report (AE0D/E608) concerning a review of water hammer events which have occurred since the resolution of Unresolved Safety Issue (USI) A-1, " Water Hamer." The study was initiated by AE00 to support NRR's response to staff action Item No. 2 enclosed with the February 4, 1986 EDO memorandum: " Staff Actions Resulting from the Investigation of the November 21, 1985 San Onofre Nuclear Generating Station, Unit 1 Event (NUREG-1190)."

The study was primarily concerned with identifying and evaluating water hamer i

events caused by check valve leakage or failure. The study found that the underlying causes and general nature of the water hamer events which have occurred Qver the past five years do not appear to indicate any new generic concern not already identified and examined by the staff. Check valves were involved in only 2 of the 40 water hamer events evaluated. Furthermore, check valves were found to have been specifically cited as contributing to only five of the almost 200 water hamer events evaluated since 1969.

Therefore, the study concludes that check valve leakage or failure was not and is not a significant generic cause of water hamer.

The infomation, analysis and evaluation contained in the enclosed report would appear to be useful for supporting your assessment of this issue. In this regard, we understand that your proposed response for action Item No. 2 incorporates elements of the enclosed AEOD report.

If you or your staff have any question concerning the enclosed engineering evaluation report, please contact Eric Leeds of my staff on extension 24445.

. Frederick J. Hendon, Deputy Director Office for Analysis and Evaluation of Operational Data

Enclosure:

As Stated cc w/ enclosure:

A. W. Serkiz, NRR V. Hodge, IE W. Minners, NRR [ f$

D. Humenansky, OCM J. Sniezek, DEDROGR

- - eup s, wm l MEMORANDUM FOR: Victor Stello, Jr.

l' Executive Director for Operations Fc0M: Harold R. Denton, Director Office of Nuclear Reactor Regulation SUEJECT: REVIEW AND ASSESSMENT OF WATER HAMMER _'

OCCURRENCES SINCE CY 1981 This memorandum is our response to staff action Item No. 2 enclosed with your February 4, 1986 memorandum: " Staff Actions Resulting from the Investigation of the November 21, 1905 San Onofre Nuclear Generating Station, Unit 1 Event

. (NUREG-1190)." The principal reason for this reassessment was the Chairman's --

question during the San Onofre IIT's briefing on January 22, 1986 regarding the need to reopen USI A-1.

We reviewed the reported water hammer occurrences and find that the frequency of water hamer o::currence has decreased since last reviewed in 1981 as part of the resolution of USI A-1 and that no new underlying causes have emerged.

!!c de not expect to achieve a total elimination of water hamer occurrences (as noted previously in the r~esolution of USI A-1) nor do we see a stronger safety implication than was previously derived.

e find the San Onofre Unit I water hamer attributable to grossly failed

' check valves in the feedwater system and not_a basis for reopenino the water-hamer safety issue.( Moreover, the resolution of USI A-1 recognized that water namer events, such as that which occurred at SONGS-1, would continue to occur, and that such events would not produce unacceptably high contributions to core melt frequency or public risk. The water hamer event at SONGS-1, in and of itself, did not result in a "near miss".to a core melt accident.

Finally, we reconfim our original conclusion that the imposition of new requirements related to further reducing water hammer events is not supportable by current cost-benefit guidelines.

A brief review and assessment of such water hamers has been prepared by A. W. Serkiz of my staff..With assistance from Eric Leeds (AE0D), and is enclosed. .

original signed il

_ptped -

Harold R. Denton, i)irector

. Office of Nuclear Reactor Regulation

Enclosure:

Review & Assessment of Water Hamer Occurrences - CY 1981 cc: See next page through CY 1985

• See previous concurrences Distribu ion: See next page RSIB:DSRO* RSIB:DSRO* DD:DSR0* D:DSRO* D y D:N ASerkiz:sim WMinners BSheron TSpeis ollmer HDe .on

/}/86 1'q /86 6/26/86 6/26/86 6/30/86 7/02/86

-N -

./

r e

l

\- _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

~

WATER HAMMER

• REFERENCES - ACRS Meeting B-8-86 SONGS-1

. 1. NUREG-0927, REV 1, " Evaluation of Water Hammer Occurrence in Nuclear Power Plants, Technical Findings Relevant to USI A-1", March 1984

2. NUREG-0993, REV 1, " Regulatory Analysis for USI A-1, Water Hammer", March 1984

3. H.R. Denton to V. Stello, Jr. memo dated July 7, 1986, with Encrosure,"R'e view and Assessment of Water Hammer Occurrences Since 1981.

4. F.J. Hebdon to T.P. Speis memo dated July 14, 1986, with Enclosures," Re-Examination of Water Hammer Occurrences" (AEOD/E608)

5. House, R.K., Water Hammer Data Base", EPRI Project

RP-2856-1, July 1986 l

6. Power, J.W. "EPRI/NSAC Presentation Water Hammer Issue:

Overview:" Workshop on Water Hammer in Nuclear Power Plants, Boston, Mass., July 9-10, 1986 W

S

APPENDIX V RECENT SIGNIFICANT EVENTS AGENDA FOR ACRS Meeting on August 8, 1986 3:15 p.m.

Room 1046, H Street -

RECENT SIGNIFICANT EVENTS O

Presenter / Office Cate Plant Event Telephone Paae 6/1/86 LaSalle 2 Feedwater Transient without A. Bournia, NRR 1 Reactor Scram 28698 6/25/86 Inadequate Design of PWR H. Bailey, IE h Safety Injection Minimum Flow 29006 6/14/86 Vermont Yankee Problems with Scram Solenoid E. Weiss, IE /@

Valves 29005 O

4

LA SALLE ? - FEEDWATER TRANSIENT WITHOUT REACTOR SCRAM JUNE ], 1986 (A. BOURNIA, NRR1 PROBLEM FEEDWATER CONTROLS INDUCE TRANSIENT L0 Rx WATER LEVEL SCRAM SWITCHES (3 0F h) DID NOT TRIP

. SIGNIFICAN.CE ALL DP SWITCHES BY THIS MANUFACTURER (STATIC "0" RING) ARE

. SUSPECT GENERIC APPLICABILITY TO SEVERAL BWRS AND PWPS PRECURSOR EVENTS AT OYSTER CREEK (JAN, 1986) AND LA SALLE (FEB 1986, MAY 1986)

DISCUSSION NEW LEVEL SWITCHES (SOR, INC) INSTALLED TO MEET E0 RULE PLANT APPR0XIMATELY 90% AND PERFORMING SURVEILLANCE OF MAIN FEEDWATER PUMPS AT APPROXIMATELY 4 A.M. JUNE 1, 1986 PLANT EXPERIENCES TRANSIENT AND PLANT OPERATORS INITIATION OF MOTOR DRIVEN MFP TO AMELIORATE EXCURSION (INDICATION WATER LEVEL REACHED NEAR TRIP POINT)

LEVEL REC 0VERED IN REACTOR, POWER RESTORED TO APPR0XIMATELY 60% RECIRC, PUMP ON-COMING SHIFT REALIZED Rx LEVEL FELL BELOW NORMAL Rx TRIP TRIP SETPOINT WITHOUT SCRAM: STUDIED STAR-TREC OUTPlIT (NARROW RANGE) SUBSTANTIATING LOW LEVEL PROBLEM ALERT DECLARED PLANT SHUTDOWN AUGMENTED INSPECTION TEAM DISPATCHED R-III ISSUES CONFIRMATORY LETTER LICENSEE RECALIBRATION SWITCHES AND FOUND ERRATIC BEHAVIOR 0F SETPOINT INFORMATION NOTICE (86-47) AND IE BULLETIN (86-02) ISSUED i

UTILIZATION OF THESE SWITCHES IN OTHER APPLICATION; DETERMINED l TO HAVE SAME PROBLEM LICENSEE PERFORMED A DETAIL TEST FOR ALL APPLICATIONS OF SWITCHES

/ - hb FOLLOWUP l NRR ASSESSING STARTUP OF UNIT 2 A

LaSalle Unit 2 Foodwater TrEnstent '

June 1, 1986, Startrec Data i 8 Startrec initiates

! @ +l8 inches 4

[ Narrow Range level j-

[ Af ter 2 feedpump lockup with high flow O Reset -

C "A" TDRFP 6 0 ,

l i R

I

! B 62 Y ,-

NSO Begins increasing I "A" TDRFP F low l g 44 c,-

1.

NORHAL LE' 'EL (36 INC IES) g j Y

. DRF Begin Reducing g N Core Flow I y '

t i

1/2 Scras,'r, RR downshi ft I $ 28 -

! l i i r

, 20 -

F CONF IRHED u

• DATA AT 1815

. p BEli W 11 INCHE i HOUR i ON 6/1/86

. 12 7 tUR 1.25 SEC Start HDRFP-4 to 12 (Sec) 24 36 48 1 :0.0 1:12 1:24 1:38 1:48 vs,.

O hALLE CoddTY GTATIoM FEET 7kATER TRAQe,tggh "l I 70 FT.

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SPRING RETAINERS i

>$ ,  : o SPRING SEAT , / '

DIAPHRAM RANGE SPRING I g PIS TON (2)

SPRING SEAT PISTON LEVER SPRING LEVER 8 PISTON OPERATING

%. R00 CYLINDER DIS'C .(2)

SPRING HOUSING s BODY PISTON LEVER B EARING BEARING BODY S SWITC H LEVER J_

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Fs Fd b /

ROSS SHAFT F h-FORCE,HI PRESSURE F -FORCE,LO PRESSURE i

FS-FORCE, RANGE SPRING F -FORCE, RESULTANT DIFFERENTIAL

, d (HI-LO)

A -so lo

-. CONCLUSIONS SOR INVESTIGATION SIGNIFICANT SETPOINT SHIFT FROM CAllBRATION PRESSURE TO OPERATING PRESSURE

- SETPOINT SHIFT WITH PRESSURE COMPLETE AFTER 24 HOURS ,

- SETPOINT SHIFT WITH PRESSURE DETERMINED FOR EACH SOR APPLICATION AND INCLUDED IN REVISED CALIBRATION PROCEDURE

- EFFECT OF CYCLING ON VALID "AS-FOUND" AND "AS-LEFT" CALIBRATION

- CALIBRATION PROCEDURES REVISED TO REQUIRE APPROPRIATE CALIBRATION CYCLING TO ACHIEVE VALID "AS-FOUND" AND "AS-LEFT" SETPOINTS-

- THESE SETPOINT AND CALIBRATION METHOD REVISIONS ENSURE THE SWITCHES WILL TRIP WITHIN TECHNICAL SPECIFICATION LIMITS

- THE G.E. ANALYSES SHOW LARGE UNUSED MARGIN BETWEEN THE ANALYZED LIMIT FOR THE PARAMETER AND THE TECHNICAL SPECIFICATION LIMIT

~~

e A-51

?.

INADEQUATE DESIGN OF PWR SAFETY INJECTION PUMP MINIMUM FLOW PAlHS JULY 24, 1985 - JUNE 25, 1986 (HENRY BAILEY.lb)

PROBLEM: POTENTIAL LOSS OF SAFETY INJECTION (SI) PUMP MINIMUM FLOW PATH DUE TO SINGLE FAILURE VULNERABILITY.

SIGNIFICANCE: P0TENTIAL LOSS OF ALL HIGH HEAD SI PUMPS DURING A

- SMALL BREAK LOCA.

DISCUSSION:

JULY 24, 1985, POINT BEACH LICENSEE SUBMITTED 10 CFR 21 REPORT DESCRIBING A DESIGN DEFICIENCY INVOLVING THE MINIMUM FLOW RECIRCULATION VALVES FOR THE SI PUMPS.

DECEMBER 13, 1985, ISSUED IE INFORMATION NOTICE 85-94,

" POTENTIAL FOR LOSS OF MINIMUM FLOW PATHS LEADING TO ECCS PUMP DAMAGE DURING A LOCA." .

FEBRUARY 5, 1986, H..B. ROBINSON LICENSEE SUBMITTED AN LER DESCRIBING THE SAME DESIGN DEFICIENCY AS POINT BEACH, DURING LATE JUNE 1986, LICENSEES FOR GINNA AND TURKEY POINT ALSO REPORTED THESE PLANTS HAD THE POINT BEACH DESIGN DEFICIENCY.

THE SHORT TERM FIX HAS BEEN TO DISABLE THE RECIRCULATION VALVES AND BLOCK THEM OPEN.

THE VALVES ARE TO BE MANUALLY CLOSED BEFORE STARTING ON SUMP RECIRCULATION.

A SIMILAR PROBLEM WAS RECENTLY DISCOVERED ON THE RHR MINIMUM FLOW BYPASS FOR THREE BWR PLANTS, IE COMPLIANCE BULLETIN 86-01, " MINIMUM FLOW PROBLEMS THAT COULD DISABLE RHR PUMPS" WAS ISSUED 0N MAY 23, 1986.

/

FOLLOWUP:

IE HAS DRAFTED A BULLETIN TO ADDRESS THIS MINIMUM FLOW

- DEFICIENCY AND HAS IT UNDER REVIEW.'

e s

= ~ u + . . - - a_ _ .

8 DIAGRAM OF POINT BEACH ECCS , 5' .

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CONT NMENT

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);.53

~

7

VT, YANKEE PROBLEMS WITH SCRAM SOLEN 0ID VALVES (UPDATE)

JUNE 14, 1986 (ERIC WEISS, IE)

PROBLEM: 1 R0D FAILED TO MOVE AND 5 RODS HESITATED 5 SECONDS BEFORE MOVING BECAUSE OF PROBLEMS WITH SCRAM S0LEN0 IDS SIGNIFICANCE: .

LICENSEE WENT CRITICAL WITH INOPERABLE RODS MANUFACURING DEFECTS OF REBUILD KITS MAY AFFECT OTHER

. PLANTS INADEQUATE POST-MAINTENANCE TESTING 0F SCRAM SOLEN 0 IDS CORE DAMAGE POSSIBLE IF TWO STUCK RODS ARE ADJACENT CIRCUMSTANCES:

SCRAM TIME TESTS PERFORMED FOLLOWING A SYSTEM HYDRO AFTER IN-SEQUENCE CRITICAL TEST AND S/D MARGIN TESTS ALL SCRAM SOLEN 0 IDS HAD BEEN REBUILT FOR NORMAL PREVENTATIVE MAINTENANCE (0NCE EVERY 5 YEARS)

R0D CONDITION CAUSE 06-23 NO SCRAM SPRING ON PLUNGER 0U1 0F FLACE 38-23 5 SEC DELAY DIAPHRAGM INSTALLED BACKWARDS

", ',' ',' DIMEySIONSgFCgNEgNPLUyGER jg,23 ,

26-23 " " " " " " " "

22-35 VT, YANKEE HAD 200 0F REBUILD KITS FROM LOT OF 3000 DISTRIBUTED BY GE WHICH HAS ONLY 200 KITS LEFT APPLICABLE TO BWR 2, 3, 4, AND SOME 5s SPRINGS IN REBUILD KITS HAVE OUT-0F-TOLERANCE INNER DIAMETER PLUNGER LAND HAS OUT-0F-TOLERANCE OUTER DIAMETER CONES (CORE NEEDLE) LARGE BECAUSE THEY ARE FOR ANOTHER VALVE SOLEN 0ID BASE SUBASSEMBLIES IS OUT-0F-ROUND -

COULD PREVENT FREE TRAVEL OF PLUNGER SOME TESTS SHOW PROBLEM AFTER A FEW CYCLES SOME TESTS SHOW NO PROBLEM AFTER REASSEMBLY BACKUP SCRAM VALVE WITH CAUSE SCRAM KITS USED TO BE DISTRIBUTED AS " COMMERCIAL GRADE" FOLLOWUP:

IE INFORMATION NOTICE IS UNDER DEVELOPMENT IE VENDOR INSPECTION

/-54 /0

29 30

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1.

ACCUMULATOR CHARGING WATER RISER *

~

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2. DRIVE -WITHORAW RISER 3

3. ISOLATION VALVE - DRIVE WITHDRAW. RISER b)

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4. DRIVE WATER RISER g' p, j4 24

5. ISOLATION VALVE - ORIVE WATER RISER  ;' T

6. ISOLATION VALVE - SCRAM DISCHARGE RISER " 8 k

• 7. SCR AM PILOT VALVE ASSEMBLY '# '

T 9 t

lL OUTLET SCRALi VALVE AND ACTUATOR / ,,

9. TYPICAL E LECTRICAL CONNECTION

10. WIRING THROUGH ASSEMBLY b

MANIFOLO (PART OF PIPING ASSEMBLY) N 11.

DIRECTIONAL CONTROL vat.VES (4 EACH)

• d4AAs* g 12.

3

13. ISOLATION VALVE - ACCUMULATOR CHARGING WATER RISER , ,

14. SCR AM ACCUMULATOR - NITROGEN CYLINDER ,

~

15. ACCUMULATOR G AS PRESSURE INDICATOR 11

16. ACCUMUL ATOR INSTRUMENTATION ASSEMBLY 22 '

17. NEEOLE VALVE - ACCUMULATOR GAS CHARGING

18. FRAME "

- 12

19. SCR AM ACCUMULATOR WATER CYLINDER 2I

20. NEEOld VALVE - ACCUMULATOR WATER CYLINDER DR AIN 21 COOLING CHECK VALVE (IN MANIFOLD) 20 - l 22 SPEED CONTROL VALVES (2 EACH) 13

23. INLET SCR AM VALVE AND ACTUATOR V

24. ISOLATION VALVE - COOLING WATER RISER <

25. ISOLATION VALVE - DRIVE INSERT RISER I 4#

26. ISOLATION VALVE - EXHAUST W ATER RISER  !

27. EXHAUST WATER RISER i j

28. DRIVE INSERT RISER

29. COOLING WATER RISER ig  :

30. SCRAM DISCHARGE RlSER 18 .

; T l

]

9 50 h ~

1s 17 %

% \

'l Wx 16 l

l FIGURE 2.4-3 Hydraulic Control Unit i

1 l

2.h-50

//

m APPENDIX VI-NRC PRESENTATION -

USI A-46 ACRS PRESENTATION USI A - 46 ACRS PRESENTATION AUGUST 7, 198G e

l l

tty 4

i i

A-54

BACKGROUND o TASK A-46 DESIGNATED AS USI IN DECEMBER 1980 o 5 TASKS IDENTIFIED, USE OF SEISMIC EXPERIENCE DATA PROVED TO BE THE MOST REASONABLE ALTERNATIVE o COLLECTION OF TEST EXPERIENCE DATA BY EPRI TO SUPPLEMENT SEISMIC EXPERIENCE DATA 8

e e

A-57

SEISMIC QUALIFICATION OF EQUIPMENT USING EXPERIENCE DATA o STAFFESTABLISHEDFEASIBILITYOFUSINGSEISMICEXPEhlENCE DATA (LLNL STUDY) o SQUG CONDUCTED PILOT PROGRAM TO COLLECT AND EVALUATE SEISMIC EXPERIENCE DATA (SEPTEMBER 82) o ADDITIONAL SEISMIC EXPERIENCE DATA COLLECTED FOR C0ALINGA, MORGAN HILL, CHILE, MEXICO EARTHQUAKES o SSRAP FORMED JUNE 1983, JOINTLY SELECTED BY SQUG AND NRC o EPRI INITIATED TEST sXPERIENCE DATA COLLECTION IN 1984 o NRC STAFF PARTICIPATED IN DATA EVALUATION AND CLOSELY MONITORED SQUG/EPRl/SSRAP EFFORTS o PROPOSED STAFF POSITION BASED ON USE OF SEISMIC EXPERIENCE AND TEST EXPERIENCE e

e

, s-t 4

THREE MAJOR CONCERNS'

1. EQUIPMENT ANCHORAGES

2. RELAY FUNCTIONAL CAPABILITY s 3, OUTLIERS s

\

[

l l k-5f 1

PROPOSED RESOLUTION [

o OPERATING PLANTS

- PERFORM WALK THROUGH INSPECTION VERIFY ANCHORAGES REVIEW SEISMIC SYSTEMS INTERACTION IDENTIFY & ADDRESS DEFICIENCES AND OUTLIERS

- VERIFY FUNCTIONAL CAPABILITY OF E0VIPMENT (RELAYS) o NEW LICENSEES

- NO REQUIREMENTS o IMPLEMENT BY GENERIC LETTER I

t O

i f

ON-GOING S00G & EPRI ACTIVITIES E .

o SEISMIC ADEQUACY OF EQUIPMENT >

o GIP o RELAY SEISMIC EXPERIENCE DATA o CABLE TRAY / CONDUIT EXPERIENCE DATA EPRI o EQUIPMENT ANCHORAGE REVIEW GUIDELINES o GERS o RELAY EVALUATION PROCEDURES O

e e

APPLICABILITYOFBACKFITRULETOA-46IMPLEMENTAT20N

~

o REQUIREMENT TO PERFORM PLANT REVIEW - 50.54F

, o REQUIREMENT TO FIX DEFICIENCIES FOUND DURING REVIEW - 50.109 e

e

USI A Ll6 STATUS O PROPOSEDRESOLUTIONPACKAGEISSUEDFORPUBLICCOMMENTON 9/13/S5 ,

o RESOLUTION OF COMMENTS PRESENTED TO ACRS ON 1/15/86 o RESOLUTION OF COMMENTS INCORPORATED IN PROPOSED RESOLUTION PACKAGE, AND CONCURRED BY NRR DIVISIONS o PLANNED TO MEET WITH CRGR IN SEPTEMBER 1986 o FINAL ISSUANCE OF RE'OLUTION S AND GENERIC LETTER IN OCTOBER 1986 l

O

+

A-U l._. - - - . . . _. . . . _- -_ -. .

pnce:vq 6 ~% UNITED STATES'

[c ' , g j

NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (S WASHINGTON, D. C. 20555

1. ,4 July 10, 1986 APPENDIX VII STANDARDIZATION POLICY MEMORANDUM FOR: A Membe FROM: . . nl Chief Project Review Branch #1

SUBJECT:

STANDARDIZATION POLICY Under Tab 2 of your notebook for the 315th ACRS nteting you were provided a number of documents relating to a proposed policy for standardization of nuclear power plants. On April 10, 1986 SECY (page 41) forwarded to NRR a draft policy statement that reflected the consensus of the Commissioner's assistants and.had been reviewed by the Commissioners. The Staff was re-quested to review the statement and offer comments in the form of a revised statement. In addition, the policy statement was to be supplemented by a detailed NUREG document describing the implementation of the policy. The Commission specifically requested comments on whether to expand the scope of Standardization in the policy statement to include such items as: component and procurement specifications, including acceptance test requirements; installation specifications; emergency operating procedures, training pro-grams; technical specifications; and appropriate empirical information. For entirely new designs, Chairman Palladino, thought that a prototype should be built and operated prior to NRC certification.

NRR submitted a revised Policy Statement on May 14, 1986 (see page 49 in notebook). The two versions of the basic policy statement are very compara-ble with the NRR version being a little more judicious and precise in lan-guage. The major differences that I found are:

1. NRR notes that the proposed statement supersedes the 1978 policy statement.

2. The NRR version includes, in its first paragraph, the concept of construction on pre-acoroved sites.

3. The NRR version adds short term licensing transition options to the implementing NUREG.

4. The NRR version adds cost justification to the backfit provisions.

5. The NRR version adds compliance with (then) current NRC regdlations to certification renewal.

l i

ACRS Members 2 Nothing I saw indicated that NRR addressed the questions on expanding the scope to cover component procurement, acceptance testing, emergency proce-dures, etc.,.or Palladino's thought for a full scale prototype.

NRR did include an outline of the implementing NUREG.

By memorandum dated June 18, 1986, Chairman Palladino requested ISRS comments on the Policy Statement and on the outline for the implementing NUREG.

O f

N_.

l e

- ADVISORYC0h.MITTEEONREACTORSAFEGUARDS b

T APPENDIX VIII DATE: $ June,1986 NOTES BY W. KERR ON IMPROVED T WA M REACTOR MEMO TO: ACRS Members FROM: W. Kerr, ACRS Member

SUBJECT:

Notes on Accident Prevention Measures Appropriate for

- Consideration in Standard Plant Design The following items are suggested for consideration:

1) ATWS. It appears worthwhile to eliminate concerns about ATWS by designing standard plants so that the likelihood of an ATWS with unacceptable consequences is very low, a) For PWRs this can probably be accomplished by providing a reasonable amount of pressure relieving capability for the primary, and by keeping the moderator temperature coefficient of reactivity .sufficiently negative throughout the fuel cycle. The former can be achieved by some appropriate combination of safety and relief valves, although, in order to decrease the likelihood of stuck open I ,

t relief valves, a better quality of valve may be required.

The latter may require the use of burnable coison, at least in some of the fuel loading's. Although this has disadvantages, there some additional advantages in using some burnable poison.

b)ForBWRsthiscouldprobablybe\ccomplishedbyproviding for 100'; bypass capacity in the turbine condenser.

Alternatively a combination of turbine bypass and relief valves might be used, if we don't continue to insist on rapid closure of MSIVs. (Indeed it might be a good idea to eliminate this anyway.) I assume that a source of electric power would be needed to operate the condenser and to close (or open) the MSIVs on demand, but the likelihood of an ATWS

- and the simultanecus less of all AC must be rather low.

2) Since there is general agreement that decay heat removal is of

. major imcortance, we should describe in some appropriate way what we think its characteristics should be. I'm personally not enamored of

" dedicated" decay heat removal systems, mostly because I don't know what that means. Does it mean, for example, that the system is never used except in the case of an accident? If so how does one determine its reliability? But at any rate I do not think the Committee has come to any conclusion on this topic which is perhaps one of the two or three most important in accident prevention.

[

O

3) Is sabotene protection accident prevention? If so has one done as

- much as is feasible if a satisfactory solution is reached to #2 above?

(e.g. does cne have to insure an intact and habitable control room?

Live operators?)

e 4)Reliableldepressurizationoftheprimarysystem. Although this could perhaVs be considered as subsumed by #2, I call it out for special attention because it may also be important in preventing an early challenge to containment failure. Such a system should, e.g.,

make the likelihood of containment heating during a high-pressure core-melt sequence negligibly small.

5) We probably should give some additional attention to the current electrical system design which starts the diesels very rapidly if they are started automatically. Not only do the required tests for quick starting degrade diesel reliability, but the fact that a failure to start quickly is interpreted as a failure to start, coupled with the current Regulatory Guide approach which somehow seeks to demonstrate reliability by a lot more than the -

normally required test starts, makes it less likely that a diesel will be available when needed. (Which is almost never within a few seconds of the beginning of whatever initiating event calls for a diesel to be available eventually).

6) We should insist on the elimination of the artificial division of This division plant systems into safety and non-safety grade systems.

has led to control and other BOP systems being almost ignored as accident initiators. It is clear from operating experience that many

(, so-called non-safety systems need some sort of reliability

'_ specification if we are to achieve the level of plant reliability that we seek. Although USI A-47 was supposed to resolve this issue, most of what the study did was to explore in detail, for a few postulated accident sequen'ces, for a B & W system, the consequences. The general question was left largely unexplored. This issue should be resolved for standard plants by the development of some generally applicable criteria for all plant systems.

7) The question of requiring degrees of licensed operators is still unresolved? How much influence is this decision likely to have on accident prevention? It is not clear to me that a degree should be required, but I suppose I could be con.inced as lcng as the degree is not an MEA.

wk 30-VI-86 A-47

APPENDIX IX REGULATORY PROCESS l

DATE: 4 August, 1986 MEMO TOs Members of ACRS ~

FROM: W. Kerr

SUBJECT:

Regulatory Process After giving some thought to appropriate action by the ACRS in connection with efforts to effect improvements in the regulatory ,

proces's, I have concluded that a Committee study as described in the attached draft letter to the Commission is the approach that is most likely to produce something useful.(Even I therefore though I think the recommend it for your consideration.

process is likely to be slow and painful.)

I also call your attention to a draft letter prepared by Gary Quittschreiber which is perhaps closer than my suggestion tolast the approach that was discussed by the Committee at the end of meeting. This draft has considerable information on previous studies and recommendations, and some worthwhile suggestions, and is an alternative to the approach I am recommending.

' \

9 9

.,g g O

1

/ -48 j /

f dear M'r. Zech:

SUBJECT:

REGULATORY PROCESS

~

Recent experience with licensed operating nuclear power plants suggests that the existing system of Nuclear Regulatory Commission regulations and requirements,' and the NRC Staff's enforcement of these regulations, does not ensure an appropriately high quality of power plant performance.

A number of ACRS reports as well as several studies performed by membe'rs of the Commission staff have suggested possible actions that might result in improvements. Nevertheless the present system does not seem to us to be operating in a manner that results in the quality of plant performance that we believe is desirable.

We have discussed this at some length and have concluded that the ACRS should undertake an in-depth study of this issue. We plan to begin this study within the next month. We expect to report to you periodically as we assemble inf ormation and draw preliminary conclusions. We also plan to discuss this with you at an early meeting with the Commission.

We welcome any comments or suggestions that you may have at this time or as our study progresses. i Sincerely:

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July 22, 1986 William Kerr From: Gary Quittschreiber

Subject:

The Regulatory Process I have written down some at thoughts for your consideration for the draft the August ACRS Meeting on the " regulatory ACRS report to be considered process". Please consider the f ollowing:

1986, the ACRS discussed the During its 316th Meeting, August 7-9, Nuclear Regulatory Commission 's regulatory process with regard It is to the evident that and operation of

,conatruction licensees can build and operate nuclear power plants in nuclear power plants.

nuclear plant the Nuclear Regulatory Commissions's rules and cecordance with all of requirements and still have a high frequency of accidents which pose a threat to the public health and safety.

The regulatory process for licensinginside and regulation of nuclear plants has and outside the NRC, since the bacn examined by many groups, from both Three Mile Island Unit 2 accident on March 28, 1979. ManyA studies have been few specific pcrf ormed which have looked at the regulatory process.

cxcmples include the following:

1. Speci al Inquiry Group Island 's Report to the Nuclear Regulatory Commission on (Rogovin Report), dated January 1980 o2.

Accident A Review at Three of MileLight Water Reactor Regulatory Requirements, NUREG/CR-4330, wated April 1986 A Survey by Senior NRC Managment to Obtain Viewpoints on the Safety 3.

Impact of Regulatory Activities from Representative Utilities Operating and NUREG-0839, dated August 1981 Constructing Nuclear Power Plants,

4. Report of the Independent Ad Hoc Group for the Davis-Besse Incident, NUREG-1201, dated June 1986

5. Regulatory Reform Task Force Lewis, etc.

6. Presidents Committee (Governor Babbett, H.

and many problems have beer Even though many studies have been performed, significant events continue tc identified and recommendations have been made, 1985, happen, such as the loss of feedwater event at Davis Besse on 1, June 9, on November the loss of power and water hammer event at San Onofre, Unit system power and overcooling 25, 1985, the loss of integrated control December 26, 1985. In addition the Tennessee transient at Rancho Seco onother utilities have shut down their nuclear plantt Valley Authority and some due to management related problems.

We have not yet completed These are a few of the studies performed. We still owe our review of the June 9, 1985 event at Davis-Besse. 1985 letter to the ACI Commissioner Asselstine a response to his September 9, responses to some of his questi which asked for views on NRC StaffWe are committed to respond after our review of the Ad regulatory process.

Hoc Report mentioned in item 4 above.

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Tha ACRS has insund many lottcroff cnd th9 Problems have bGsn identificd.

crataining numerous recommendations for consideration by the NRC Sta ission. Some of these recommendations have been but responded have to, w L ars have not. testimony about problems with the regulatory process, congressional cocn little action to correct the problems which have been identified.

~ balanced regulatory process We suggest that the NRC establish a well improvement group to look at the regulatory process as a wholebyand to make the ACRS mako a concerted effort to evaluate the many recommendations made and by others, and to make specific proposals to implement those In addition, yccommendations which are determined to have significant merit.t renolve, such as:

co*atruction, and 1.

Do NRC's requirements with regard to design, .ress and frustration on operation of nuclear plants lead to low morale, high and if so, how can this the part of the nuclear facility owners and operators, a

be improved? nature of the regulatory process a detriment or

2. Is the Do adversarial fines being levied on utilities for violation of NRC's bsnafit?

! rsquirements lead to improved safety?

3. Does the NRC need to improve its ability to evaluate licenseeto shut down plants l

parformance so it can act faster oafety problems are evident? the NRC's safety research l

What should be the nature and scope of 4.

Should industry be performing all or part of this research?

program?Should the NRC be looking at f uture reactor designs or concentrate on 5.

f uting reactors? evaluating plant performance? Are

6. What is the most effective way of more or Doesless reactor trips better?

7.

the present backfit rule improve or detract from safety?

8.

What is the effectiveness of emergency planning with regard to ovacuation vs. sheltering? investigations be conducted? Should equipment be

9. How should accident Independent Safety Board be the accident? Should an frozen at the time ofis the proper balance between experience and inoependence established? What for investigators? for the utilities conducting aptitude

10. Should the NRC show its support plant operators and other workers.

tests for its new nuclear We suggest that the regulatory process improvement function group have soathat direct its

. line of communication with the Commission and followupWe would be happy to work findings can be implemented in a timely manner.the many recommendations and with this group to evaluate the merit of unanswered questions that have been made with regard to the regulatory process.

i Sincerely David A. Ward Chairman s

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.  : E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS "o / WASHINGTON, D. C. 20666 August 7, 1986 NRC STRATEGIC PLANNING MEMORANDUM FOR: David Wa

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FROM: RayFralid

SUBJECT:

NRC STRATEGIC PLANNING Attached are some materials from the meeting I attended on NRC Strategic planning.

The Chairman's Office has suggested that this effort be brought to the attention of the Committee.

P.S. The SDG reps, were Carl Spetzler and Patricia Evans.

cc:

ACRS Members 9

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AGENDA FOR STRATEGIC PLANNING MEETING AUGUST 1-2, 1986 Holiday Inn 8120 Wisconsin Avenue Bethesda, Maryland Friday. August 1, 1986 1:00 - 6:00 p.m.(VersaillesIII) s Keynote Address

- Chairman Zech wil1 introduce the meeting, explain its purpose and describe his expectations for what will be accomplished.

Framework of Strategic Planning Strategic Decisions Group will provide a formal introduction to strategic planning, including a description of the steps in the strategy development process, the deliverables to be produced from each step, and the roles of various participants in the process.

Issue Generation Session Working in two subgroups, participants in the meeting will list potential strategic issues and planning topics they would like considered for inclusion in the strategic planning process. The purpose in this session is to generate the topics, but not to evaluate or solve them at this time.

Saturday. August 2. 1986 .

_ 7:30 a.m. - 1:00 p.m. (Versailles III)

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Consolidation and Categorization of the topics List into Challenges Working together, participants will merge all the issues generated into one list and group issues that are similar or related. Issues will also be categorized as purely strategic, purely operational, or both. Then groups of related issues that are at least partly strategic will be formulated into major challenges. The set of challenges define the major broad concerns to which the strategic alternatives (to be developed later) should respond.

Conclusion ,

Strategic Decisions Group will sumarize the challenges that have been identified and describe how they will be used in the next steps of the planning process.

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\ Kickoff Meeting For NRC l Strategic Planning Effort

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' NRC Senior Management l

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Washington, D.C.

August 1-2,1986 ,

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i l STRATEGIC DECISIONS GROUP 3000 Sami Hill Road

• Menlo Park, Caldornia 91025-7127 * (415) 854-9000 I

We interviewed several commissioners and the EDO, and asked two questions.

What significant issues should be addressed in the strategic planning process?

k How would you judge success or failure of the g process a year from now?

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We found general agreement on the importance of strategic planning and on the keys to success in the effort. (I) ,

Success Creation of a " quality" plan for what the agency should be in 1992 that:

A - Demonstrates significant new initiative and -

4 grapples with serious problems h - Is proact,ive '

- Has broad content with specific appendices

- is easily understood.

Alignment of the organization -- participation, commitment

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We found general agreement on the importance of strategic planning and on the keys to success in the effort. (II) ,

Failure The plan does not get implemented.

- Not tied to resource commitments and operating plan s - Not used on a day-to-day basis h The planning process generates too much controversy.

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- Diverts commission

- Creates internal management conflict rather than resolves issues l

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Our purpose in the strategic planning effort is to help the NRC protect the public health and safety in the conunercial use of nuclear power by: I Achieving a significant improvement in the strategic direction of the NRC

} Developing a five-year plan that includes a clear y statement of direction and an allocation of NRC resources consistent with the direction Aligning and motivating the organization along the chosen direction.

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P FRAMEWORK FOR

STRATEGYDEVELOPMENT 4

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APPENDIX XI ADDITIONAL DOCUMENTS PROVIDED

,f FOR ACRS USE APPENDIX XI ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE

1. Memorandum, R. F. Fraley to D. A. Ward, ACRS Chairman, NRC Strategic Planning, August 7, 1986

2. Memorandum, R. F. Fraley to ACRS members, Appointment of Former k Commissioner Peter J. Bradford as Chairman of New York State Public Service Commission, August 6, 1986 0

3. Memorandum, G. R. Quittschreiber, Chief, Project Review Branch No.

2 (ACRS) to ACRS members, Legal Opinion on Backfit Rule Applicability to USI A-17, August 7, 1986

4. Memorandum, R. F. Fraley to ACRS members, ACRS Activities, August 7, 1986 k

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