ML20203G325
| ML20203G325 | |
| Person / Time | |
|---|---|
| Issue date: | 07/25/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2406, NUDOCS 8608010077 | |
| Download: ML20203G325 (331) | |
Text
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y ! TABLE OF CONTENTS .1 & MINUTES OF THE 311TH ACRS MEETING WASHINGTON, D.C. Sg-gji[6 [f[
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I. Chairman's Report......................................... 1 II. Adequacy of the Seismic Design of the Perry Nuclear 1 Power Plant, Unit 1....................................... 2 l III. Integri ty of Primary Cool ant Systems . . . . . . . . . . . . . . . . . . . . . . 6~ IV. NRC Severe Accident Policy................................ 11 V. Advanced Reactor Designs.................................. 14 I VI. Proposed Implementation of NRC Quantitative Safety Goals .................................................... 16 VII. Report on Visit to Three Mile Island , Unit 2. . . . . . . . . . . . . . 19 VIII. Executive Sessions........................................ 20 A. Sub comi ttee As s i gnme nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
- 1. Revised ACRS Subcommittee Assignments........... 20
- 2. Safety Considerations for Future Nuclear Power Plants.................................... 20 B. Reports, Letters, and Memoranda...................... 20
- 1. ACRS Comments on Proposed Safety Goal Policy.... 20
- 2. ACRS Report on the Perry Nuclear Power Plant, Unit 1 .................................. 20
- 3. ACRS Comments on Proposed Broad Scope Rule Revision to General Design Criterion 4.......... 21
- 4. ACRS Coments on the Implementation Plan for the Severe Accident Policy Statement and Regulatory Use of New Source-Tem Infomation... 21
- 5. PRA Quantification of Public Health Risk........ 21
- 6. Responses to Recommendations of Panel on ACRS Effectiveness.............................. 21
- 7. Pressurized Thennal Shock Letter to the Acting Executive Director for Operations ....... 21
- 8. ACRS Comments on Draft Technical Report on Guidelines for BUR Coolant Pressure Boundary Piping................................. 22
- 9. ACRS Review of NRC Actions Resulting from San Onofre Nuclear Generating Station, Unit 1, November 21, 1985....................... 22
- 10. Move of ACRS to Bethesda, Maryland ............. 22 i
DDIGNATED ORIGINAL 1 e608010077 86072D Certified Ey [ I PDR ACRS PDR 2406
i 11 C. Future Agenda ....................................... 22
- 1. Fu tu re Age n da . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
- 2. Futu re Subcommi ttee Meetings . . . . . . . . . . . . . . . . . . . 22 D. Change to Bylaws .................................... 22 E. The 1986 Ohio Earthquake and Its Impact on the Perry Nuclear Power Plant, Unit 1 . . . . . . . . . . . . . . . . . . . 23 F. Realignment of ACRS Manpower and Fiscal Resources for FY 1987 Consistent with Reductions in ACRS Resources ........................................... 23 G. Conduct of Employees - Conflict of Interest ......... 23 H. Appointment of New Member............................ 23 I. Report of ACRS Management Group ..................... 23 J. National Research Council Study of the Long Term Needs of the NRC Safety Research Program ............ 24 4
- , - --n. . - - . . , .
iii TABLE OF CONTENTS APPENDICES TO MINUTES OF DIE 31131 ACRS MEETING MARCH 13-15, 1985 Appendix I - NRC Attendees Appendix II - Future Agenda Appendix III - ACRS Subcomittee Meetings Appendix IV - NRR Presentation on Perry Earthquake l Appendix V - Applicant Presentations on Perry Earthquake Issue Appendix VI - Earthquake IIazard and Design Research at EPRI Appendix VII - GDC-4 Rulemaking Appendix VIII - NUREG-0313, Rev. 2 Appendix IX - Implementation Plan for the Severe Accident Policy Statemnt and the Regulatory Use of New Source-Tem Infomation Appendix X - NRC Advanced Reactor Program Appendix XI - DOE Presentation on Modular HIGR Appendix XII - DOE Contractor Presentation on Modular HTGR Appendix XIII - NRC Staff Presentation on Safety Goal Policy Appendix XIV - Report on Visit to 'Ihree Mile Island Unit 2 Appendix XV - Revised Subcomittee Assignments Appendix XVI - Additional Documents Provided for ACRS' Use 1 l l l I j 1
I ! 7864 Fcdir:1 Regist:r i Vol. 51, No. 44 / Thursday, March 6,1986' / Natices . /l(,, present evidence and cross-examine and (o Michael W. Mauphin. Esq., implementation of the recommerdations of e witnesses. . Hunton. Williams. Gay and Gibson P.O. the Panel on ACRS Effectiveness. If a hearing is requested, the Box 1535. Richmond, Virginia 23212. . ' 22 AAt-d:#RAL:PerryNuclearPower , Commission will make a final attorney for the licensee. ,. , Plant. Unit J (Open)-ne members will hear determination on the issue of no . Nontimely filings of petitions for leave. and discuss the report ofits. subcommittee . ., , significant hazards consideration.The nsadng the impact oMmcent Mquab, to intervene, amended petitions, "' " final determination will serve to decide w en the hearingis eld. . supplemental petitions and/or requests ", ,Ni , fg*,'h RCS ff and thei for hearing will not be entertained ucensee will make presenta'tions and If the final determination is that the absent a determination by the
- participate in the discussion to the degree '
cmendment request involves no Commission, the presiding officer or the considered appropriate. significant hazards consideration, the Atomic Safety and Licensing Board 4:2 AAt-sco pat: Future ACRS Activities Commission may issue the amendments designated to rule on the petition and/or (Open)-ne members will discuss cnd make them effective, request, that the petitioner has made a anticipated ACRS subcommittee activities. notwithstanding the request for a substantial showing of good cause for Topics proposed for consideration by t!)e full hearing. Any heanng held would take the granting of a late petition and/or committee will also be discussed. place after issuance of the amendments. request.That determination will be sto AAf.do0 AM: Safety Considemtions If a final determination is,that the based upon a balancing of the factors in future Reactors (Open)-The members of cmendments mvolve a sigmficant specified in 10 CFR 2.714(a)(1)(i)-(v) and the Committee will discuss proposed ACRS l hazards consideration, any hearing held - c mments and recommendations regardirs 2.714(d)
- would take place before the issuance of cny amendment. y gg g;g, g spd M M safety considerations in the design of future reactors.
Normally, the Commission will not act. ion, see the application for aa) rat-daoEAt: Discuss Topicsfor I; amendments dated Febraury 6,1988, lasue the amendments until the Afecting with NRC Commissioners (Open)- Expiration of the 30-day notice period. which is available for public inspection The members will discuss topics schedules However, should circumstances change at the Commission's Public Document for discussion with the NRC Commissioners. during the notice period such that failure Room.1717 H Street, NW., Wa hington, including the scope and procedures for DC, and at the Board of Supervisors conduct of ACRS activities; requirements for 13 act in in example. a timely way derating would result, or shutoown of thefor - Cffice Louisa County Courthouse, the CESSAR 11 standardized nuclear plant , f1cility, the Commission may issue the Louisa. Virginia 23093 and the Alderman and safety considerations in future nuclear j Library, hianuscripts Department, plants; and comments regarding the need/ license amendments before the - University of Virginia, Charlottesville, desirability for a federal facility to train Expiration of the 30-day notice period. Virginia 22901. nuclear power plant operators. provided that its final determination is that the amendments involve no Dated at Bethesda Maryland this 27th day Friday, March 4. teas significant hazards consideration.The of February,1988. gy A.A7 9d5A.M Quantitative Safety final determination will consider all laster S. Rubenstein, Cools (Open)-The members will continue public and State comments received. discussion of the proposed ACRS report to Should the Commission take this action, $$,l,#'[f)$#h#[! it will publish a notice of issuance and
, 3 '[*"#'
the NRC regarding esatuation and implementation of quantitative safety goals (FR Doc. 4926 Filed 3-5.e6; 8.45 am) provide for opportunity for hearing after in the regulation of nuclear facihties. issuance.The Commission expects that **G C00E n**-*
- 1000 A M-st:x A.M: Meeting with NRC the need to take this action will occur Commissioners (Open)-The members of the very infrequently. Committee wdl meet with members of the .
A request for a hearing or a petition Advisory Committee on Reactor Commission to discuss items noted above. for leave to intervene must be filed with Safeguards; Meeting Agenda si 45 A M. :x AM. and2xRM-Joo
the Secretary of the Commission. U.S' PM: Integrity of Primary Coolant Systems I i huclear Regulatcry Commission. seci 29 an 1 .o e or c r por s to i s b ommitt e, -
ashington. DC 20555. Attention: Energy Act (42 U.S C. 2039. 2232b). the representatises of the NRC Staff, and the Docketing and Service Branch, or may Advisory Commit'ee on Reactor nuclear industry. as appropriate, regarding be delivered to the Commission s Pubhc Safeguards will hcid a meeting on pmposed changes in NRC regulations on Document Room.1717 H Street. NW., primary s> stem design and inspection. and on
* .htarch 13-15.1986 in Room 1o86.1717 H
%ashington, DC by the abov,e date. Street, NW, Washington. DC. Notice of pressunted thermal shock of reactor pressure Where petitions are-filed denng the last vessets.
this meeting was published in the tin (10) days of tne notice penod. it is Federal Register on February 25.1988 JJ5 EM-515 PM:Serere Accident Policy requested that the petitioner promptly so (Open)--The members wdl discuss the inform the Commission by a toll. free Thursday. March 13,1ses proposed NRC implementation plan for the telephone call to Western Union at (800) Ex A.M-a45 A.M: Reports of ACRS NRC severe accident pahey regarding nuclear I 325-6000 (in Missouri (800) 342-0700). Chairman (Open)-The ACRS Chairman will power plants. Members of the NRC and The Western Union operator should be report briefly regarding items of current nuclear industry will parti,cipate. as given Datagram Identification Number interest to the Committee. appropriate. 3737 and the following message S 45 A M-11.D0 A.M; Quantitative Safety 5:!$ PM-7D0 PM: Advanced Reactor cddressed to Lester S. Rubenstein: Coc/s (Open)-The members will hear and Designs (Open/ Closed)-The members will < (petitioner's name and telephone discuss the report ofits subcommittee hear and discuss the reports of its regarding the evaluatlon of the two. years subcommittee, representatives of the NRC number) (date petition was mailed) trial period for use of quantitative safety Staff. and the Department of Energy. (plant name), and (publication da , page number of this Federal Reg. goals te as a regulatory requirement, and regarding proposed advanced reactor des,igna istar . Representatives of the NRC Staff and the for both gas cooled and liquid metal cooled notice). A copy of the petition should nuclear industry wi!! make presentations and nuclear plants. also be sent to the Executive Legal participate in the discussion as appropriate. Portions of this session will be closed as Director U.SEuclearRegulatory it 000 A.M-J D0 EM: A CRS Effectiveness required to discuss Proprietary Information i Commission. Washington, DC 20555, (Open)-The members will discuss proposed applicable to this matter. b o l e w O
",J" . . ,
; Feders! Regist:r / Vol. 51, No. 44 / Thursday, March 6, i986' f NitJces * . i. .e7865 Saturday, March 15,1ses Proprietary Inforrnation [5 U.S.C. , Fersons interested in submitting a . .
e a30 AN.-22.30 P.M.: Preparation of ACRS 552b(c)(4)] applicable to the matters written statement for the record, or in Reports fopen/ Closed)-.The members will being discussed and information the presenting a five minute oral statement .-
; discuss proposed reports to the NRC release of which would represent a s as well as written testimen/, should .1 regarding items discussed during this and . clearly unwarranted invasion of promptly write the Office of the . . -
previous ACRS meetin s. including proposed personal pris acy (5 U.S.C. 552b(c)(6)]. . Secretary, Postal Rate Commission,1333 Committee reports on se evaluation and Further information regardin8 topics H Street, NW., Suite 300, Washington, 1 to be discussed, whether the meeting DC 20268, or call (202) 789 4640, for . ; repo d chang R has been cancelled or rescheduled, the a7ety goals'refulations regarding reactor policies and pressure vesse pressurized thermal shock, Chairman's ruling on requests for the further details.The Commisalon will attempt to secommodate, on a Arst come and safety considerations in future reactors. opportunity to present oral statements first served basis, those wishing to . Portions of this session will be closed as and the time allotted can be obtained by testify at a specific time,if requests are necessary to discuss Proprietary Information a prepaid telephone call to the ACRS received at least seven days before the applicable to the facilities being discussed. F.xecutive Director, Mr. Raymond F. . bearing in which they would like to a J:30 P.M-J 00 PR.: A CRS Subcommittee Fraley (telephone 202/634-3265), participate. Written statements will be between 8:15 A.M. and 5.00 P.M. accepted through April 21,1986. - . ea di cuss r port ce t
. anticipated ACRS subcommittee activities in Dated: March 3.1986. Cyrill. Pittack, _.
. ' i
& A88/88488 S*CreforY. '
designated areas. including the nomination of John C. Hoyle, candidates proposed for appointment to the (m Doc. 36-4835 Filed 3-6-46,8'45 am) {.. Adviso#7CommitteeManagementOfficer.
*.- ACRS. licensing requirements for the storage - cooe trise-as i' of spent reactor fuelin independent storage (m Doc. 86-4930 Filed 3+46,8 45 am) . , ' ,
i facilities. and resolution of unresolved BILLoso cooe reco-st-as n generic issue A-46. Seismic Qualification of Equipment in Operating Nuclear Power SECURITIES AND EXCHANGE
.. Plants. POSTAL. RATE COMMISSION COMMISSION 7 Portions of this session will be closed as ..c i ,' _ .
t'.Suired to disquss information of a personal nature, the disclosure oTQiich would [ Docket No. $S86-1) ~.. . .. -
, . , IRetesseJ6c,IC-14962; -
O12-4204) -- i Preferred Rate Study; Schedule of
,,; . mpresent a clearly unwarranty invasion of Allied Capital Corp, et al.; Application personal privacy. Hearings on Nonprofit liall for Order Authorizing Joint
*g Procedures forthe conduct of and March 3,19s6.
' Transaction 8 '
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7;- participation in ACRS meetings were published in the Federal Register on Notice is hereby given that following ~ February 2r, tees. - an opening hearing on March 12,1988, at Notice is hereby given that Allied g October 2,1985 (50 FR 191). In ac.codance o wa, h these procedures, oral the Commission's Washington, D.C., .
. Capital Corporation,1825 Eye Street.
4 _ or written statements may be presented headquarters,1333 H Street, N. W., to NW., Washington, DC 20006, a - g - by members of the public, recording' obtain views and comments on federally subsidized mail, pursuant to Notice --- -- registered closed.and ==a=yment
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investment company (" Allied Capital',); k will be permitted only during those
' portions of the meeting when a dated January 24,1986 (51 FR 3867-3969)* hearings will be held in.
Allied Investment Corporation and . - transcript is being kept, and questions Allied Financial Corporation (together
$ may be asked.only by members of the ~
Committee,itsw Aa and Staff. - Atlanta, Geogia: March 1S,1986. 9:00 W "'
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- with Allied Capital," Allied"),both W e= ate 7*'a, az Rich'"a rd"B. "R'us.~"ll se Federal -
~ Persons desiring 76 mHT6tal _. .'"- registertd stment wyeda and w subaWadaclosed.end .)
0 p - statements .%.~M . f, the AGRS ~mV.env fora o i . iUO --w -- -yhich areJiggnseil as smafibuiilness_ of.W &n _-
= p.= _ Executive Director +e hr in edvance 4e- - -
practicable so that appropriate a.m.,DenverMuseum of Naturst-
; j ~ TWest Auditorium,CityPark gmstmenmmpa.nin unMMM ---
J. - _.- @innalcyestment Act olip y; -- 7'- . - . -,*.ge=rt:trran be7nade to allow the - - -. 2001HiMobloFado' Boulevard; , necessary time during the meeting for I,os Angeles, California: March 24,1986, . Bogether,KSubsidianes" 1
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such statements U,se of still, motion _;.a I picture and telev2sms F.mcras during : . -- 9 00 a.m.,-Veterans Administration .-.. Adviso n.-- - this meeting 9 ,hr !Ln'iid lo selected _Beilding --' Medical Ce'nter,W.T12s 'ArigeTes'.---ewned 211rBeenis'JcMheetE ..-AlliedMana ementPaeers. subaad;arpiUuneo - H:_
- - portions of the meeting es determined - 'Wilshire & Sawtelle Boulevards: ; ~ W or y s p e_
Pamardapc eneral pam,er Q( Management ); an,d 4 _by the Cheirten Information regardinS Dallas, Texas Merch 2 Mesa,ema.ms .
.c' ___. the time tobe GT aside f#this perpse W'W"la*MP T r"y.bs 4t_aind hqp' aid telephor.e . m.n.. 7entiallibraiy,'
p,.blic Ubeerv-1615 GroEsiTFloBF Youna Central .. _. f a_ limited partnership _ y-- . W' j call to the ACRS Executive Director. r .the - u ., Street: - -- - - -
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. 5
* : I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o W ASHINGTON, D. C. 20555 Revised: March 11, 1986 SCHEDULE AND OUTLINE FOR DISCUSSION 311TH ACRS MEETING MARCH 13-15, 1986 WASHINGTON, D. C.
Thursday, March 13, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.
- 1) 8:30 A.M. - 8:45 A.M. Report of ACPS Chairman (0 pen) 3 1.1) OpeningStatement(DAW) 1.2) Itemsofcurrentinterest(DAW /RFF)
- 2) 8:45 A.M. - 11:00 A.M. Quantitative Safety Goals (0 pen)
(BREAK - 10:00-10:15) 2.1) Discuss proposed ACRS report to TAB 2------------ NRC regarding proposed implementation of NRC Quantitative Safety Goals (D0/RPS)
- 3) 11:00 A.M. - 1:00 P.M. ACRS Procedures and Practices (0 pen)
INSERT HAND 0UT-TAB 3-------- 3.1) Discuss proposed report on imple-mentation of the recommendations of the Panel on ACRS Effectiveness (DAW /RFF) 1:00 P.M. - 2:00 P.M. LUNCH
- 4) 2:00 P.M. - 4:30 P.M. Perry Nuclear Pcwer Plant, Unit 1 (0 pen) i
.; TAB 4 ------------- 4.1) Report of ACRS Subcommittee regard-ing adequacy of the seismic design ofthisstation(D0/RPS) 4.2) Meeting with representatives of the NRC Staff and the Licensee 4:30 P.M. - 4:45 P.M. BREAK
- 5) 4:45 P.M. - 5:15 P.M. Future Activities (0 pen) l TAB --------------- 5.1) Anticipated Subcomittee A:tivity l (MWL)
TAB --------------- 5.2) ProposedACRSActivities(RFF)
- 6) 5:15 P.M. - 6:15 P.M. Safety Considerations in Nuclear Reactors (0 pen)
TAB 6-------------- 6.1) Discuss proposed ACRS report regarding safety considerations in futureplants(CPS /D0/RKM)
o- 311th ACRS Meeting Agenda .
- 15) 6:15 P.M. - 6:45 P.M. ACRSSubcommitteeActivities(Closed) 15.1) Report of ACRS panel regarding nomination of new ACRS member -
Status report by ACRS screening panel regarding nomination of new ACRS member (HWL/ALN) (Note: Portions of this session will be closed as required to discuss information of a personal nature, the disclosure of which would represent a clearly unwar-
. ranted invasion of personal privacy.)
l l l
.. 311th ACRS Meeting Agenda Friday, March 14, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.
- 8) 8:30 A.M. - 10:00 A.M. QuantitativeSafetyGoals(0 pen) 8.1) Discuss proposed ACRS report to the NRC regarding proposed implementa-tion of NRC Quantitative Safety Goals (00/RPS)
- 15) 10:00 A.M. - 11:30 A.M. Activities of ACRS Subconsnittees (0 pen)
INSERT HAND 0UT-TAB------- 15.2) 10:00 A.M.-10:30 A.M.: Report of Management Group Meeting on (BREAK-10:30-10:45) 3/12/86 - Practices and Pro-cedures in addition to the recom-mendations of the Panel on ACRS Effectiveness (DAW /RFF) 15.3) 10:45 A.M.-11:00 A.M.: Report on proposed revision of 10 CFR Part
- 72. Licensing Requirements for the Storage of Spent Fuel in Independent Fuel Storage Facili-ties (CPS /HA)
TAB------------------ 15.4) 11:00 A.M.-11:30 A.M.: Resolu-tion of A-46, " Seismic Qualifica-tion of Equipment in Operating NuclearPlants"(CJW/AJC) 15.5) 11:30 A.M.-11:45 A.M.: Require-ments for Standardized Nuclear Power Plant Designs (CJW/HA) ,: 10) 11:45 A.M. -12:30 P.M. Integrity of Primary Coolant Systems (0 pen) TAB 10-------------- 10.1) Report of ACRS Subcommittee (PGS/EGI) regarding: 10.1-1) Proposed modification of General Design Criterion-4,
" Requirements for Protec-tion Against Dynamic Effects of Postulated Pipe Ruptures" 10.1-2) t:UREG-0313, Pipe Crack Study Group Report, Revision 2 10.1-3) Proposed NRC regulatory ,
guide on PTS implementa- ' tion 10.2) Meetings with representatives of the NRC Staff, as appropriate i i
. o 311th ACRS Meeting Agenda !
12:30 P.M. - 1:30 P.M. LUNCH
- 11) 1:30 P.M. - 3:00 P.M. Integrity of Primary Coolant Systens (0 pen) 11.1) Complete consideration of items noted above
- 12) 3:15 P.M. - 5:15 P.M. NRCSevereAccidentPolicy(0 pen)
(BREAK - 4:00 P.M.- 12.1) Report of ACRS Subcommittee on pro-4:15 P.M.) posed implementation plan for TAB 12--------- NRC severe accident policy i (WK/DH) 12.2) Meeting with representatives of the NRC Staff and the nuclear industry as appropriate
- 13) 5:15 P.M. - 6:30 P.M. AdvancedReactorDesigns(0 pen) 13.1) Report of ACRS Subcommittee regard-TAB 13----------- ing proposed advanced reactor design for a gas cooled reactor (MWC/MME) 13.2) Meeting with representatives of the NRC Staff and the DOE i
1
/
. .__n 311th ACRS Meeting Agenda Saturday, March 15, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.
- 14) 8:30 A.M. - 12:30 P.M. PreparationofACRSReports(0 pen / Closed) 14.1) Quantitative Safety Goals (D0/RPS) 14.2) Implementation Plan regarding PTS of reactor pressure vessels (PGS/EGI) 14.3) Perry Nuclear Power Plant, Unit 1 (D0/RPS) 14.4) Report of Pipe Crack Study Group (PGS/EGI) 14.5) Severe Accident Policy Implemen-tation Plan (WK/DH) 14.6) General Design Criterion-4 (PGS/EGI) 14.7) Safety Considerations in future nuclear plants (CPS /D0/RKM)
TAB --------------- 14.8) PRA Quantification of Public Health Risk (HWL/RPS) (Note: Portions of this session will be closed as necessary to discuss Proprie-tary Information or detailed safeguards infonnation related to the project being discussed.) 12:30 P.M. - 1:30 P.M. LUNCH
- 15) 1:30 P.M - 3:00 P.M. Complete Preparation of ACRS Reports (0 pen / Closed)
Complete ACRS reports regarding items considered during this meeting
. ~ _ .
l l MINUTES OF THE l 311TH ACRS MEETING March 13-15, 1986 The 311th meeting of the Advisory Comittee on Reactor Safeguards, held , at 1717 H Street, N. W., Washington, D. C., was convened by Chaiman l D. A. Ward at 8:30 a.m., Thursday, March 13, 1986. l [ Note: For a list of attendees, see Appendix I.] 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 Conmittee Act and the Government in the Sunshine Act, Public Laws 92-463 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 I 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 [ Note: R. F. Fraley was the Designated Federal Official for this portionofthemeeting.] Chairman Ward indicated that he, D. Okrent and C. P. Siess have been invited to testify before the National Research Council on April 4,1986 to provide their views and those of the Comittee as appropriate on the National Research Council Study of the Long-Term Needs of the NRC Safety Research Program. Other members (e.g., H. W. Lewis)havebeensentquestionstoanswer. Chairnan Ward indicated that the Subcomittee on Energy and the Environment of the Comittee on Interior and Insular Affairs, U.S. House of Representatives will conduct a hearing on April 8,1986 concerning NRC actions in response to the January 31, 1986 earthquake which occurred in northeastern Ohio. The ACRS has been asked to testify regarding the implication of the 1986 Ohio earthquake for the Perry Nuclear Pcwer Plant, Unit 1. D. A. Ward, C. P. Siess, and D. Okrent were designated to attend the hearing and give testimony (as requested by telecon from Dr. Henry Myers). Chairman Ward announced a special meeting of the Advisory Comittee with the Comission on March 28, 1986 at the request of the Comission to discuss the issue of quantitative safety goals. He also announced a March 19-20, 1986 Human Factors Subcomittee meeting. On the first day of that meeting, the Subcomittee will explore the potential for automating more of the monitoring and control functions of nuclear power plants. The Human Factors Program Plan will be reviewed on the second day and the Subcomittee will be briefed on the status of Emergency Operating
. Minutes 311th ACRS Meeting 2 June 23, 1986 Procedures implementation. Chainnan Ward thought that a number of non-subcommittee members might be particularly interested in the first day of the meeting. A proposed schedule for the meeting was handed out and D. A. Ward indicated that additional meeting status reports would be available for any member who would like to attend.
II. Adequacy of the Seismic Design of the Perry Nuclear Power Plant, Unit 1 (0 pen) [ Note: R. P. Savio was the Designated Federal Official for this portion of the meeting.] D. A. Ward indicated that the Subcommittee on Extreme External Phenomena held a meeting on March 12 to discuss the adequacy of the seismic design of the Perry Nuclear Power Plant. In connection with this meeting, expert advice was solicited from ACRS consultants, P. Pomeroy and M. Trifunac. Dr. Craig Smith from ANCO Engineers, Inc., together with one of his associates, reviewed effects of the earthquake on equipment. He explained that both the Staff and the utility believe that although the response spectra at some frequencies (15 to 25 hertz) exceeded the Regulatory Guide 1.60 spectrum, there was relatively little energy in the earthquake. They assert that, though there were minor displacements, there was no significant visible damage and no functional damage of significance. He mentioned a comment by P. Pomeroy regarding the possible connection to chemical waste injection wells in the vicinity of the Perry Plant. That such deep well disposal has the potential for causing earthquakes; this is based upon experience from the Rocky Mountain Arsenal. P. Pomeroy also raises the point that there was some solution mining between Perry and the site of this earthquake. It involves a high pressure process which conceivably could lead to activation of some fault that already existed relatively deep in the rock, not at the surface. Also mentioned was the fact that the safe shutdown i earthquake was exceeded above 14 hzs. J. C. Mark noted that the l excitation from the earthquake was a little peculiar because it l contained no low frequeny components. He asked if this had l occurred in the case of the Rocky Mountain Arsenal situation. D. l Okrent indicated that this is not the first earthquake in the eastern United States that has shown a high frequency component. It has characteristics of the smaller and more shallow earthquakes I in the eastern United States. He also noted that Dr. Pomeroy i suggested follow-up monitoring regarding the Ohio earthquake. M. Trifunac indicated that some lessons can be learned from this l case regarding whether the procedures used as structural response ' analysis are adequate. He noted that high accelerations of 0.2g at l modified Mercali Intensity V were seen during this earthquake. The Perry Plant is designed for modified Mercali VII. He indicated that this brings up the question of the adequacy of the Regulatory Guide 1.60 spectrum for this type of analysis. It was suggested that too much attention is being paid to forcing a plant to comply with the regulations. It would be more useful to reexamine the placement of instrumentation at this particular plant as i l 1 - - _ __
- Minutes 311th ACRS Meeting 3 June 23, 1986 4
J i l appropriately as possible so that one can infer the types of motions of the structure and foundation. He noted that the high frequency motions and high amplitudes for very small earthquakes appear to be larger than what has been seen in the literature. While there seems to be no difficulty with structures, there may be important implications for equipment. This is due to the fact that j damping values which have been put in the analyses of structures
!; are too large for these types of earthquakes. Equipment may have seen much larger excitations than are calculated. This will be i much more important for equipment because equipment is more responsive to high frequency excitation than structures.
Craig Smith, while speculating that this earthquake could have bothered some of the equipment at the Perry Plant, indicated that he found no major problems with the Plant's structures or equipment. He was satisfied that the safety-related equipment could perform as designed and intended. J. Stefano, NRC Project Manager for Perry, indicated that it is the Staff's intent as a result of the deliberations at this meeting to
- obtain ACRS commitment and agreement with the Staff's action to
, issue a low power operating license which will enable the plant to t
load fuel and to operate at power levels up to 5 percent of thermal , i rated power. The Staff also would entertain any additional ! coninents regarding further follow-on generic work that the Staff < should pursue. He also indicated that petitions have been received ), from Weston Alliance Associates and the Ohio Citizens for Responsible Energy, both interveners in the Perry proceedings. They have expressed concerns regarding the earthquake's impact on the plant design. The Staff is currently preparing responses to be issued shortly. In addition the Ohio Citizens for Responsible Energy, has issued a motion to the Licensing Board to reopen
! hearings on the earthquake event as a new contingent issue. They allege that the design basis of the plant is inadequate. The Staff i responded to that action on March 5,1986 and is awaiting the
^
Board's decision in that regard. D. Okrent asked regarding seismic
! margins at Perry. K. Anderson, NRC, described the Seismic Margins Program as a potential candidate for application to the question at Perry.
P. Sobel, NRC seismologist, sumarized the characteristics of the January 31, 1986 Ohio earthquake and its aftershocks (see Appendix IV). She explained that the earthquake triggered Perry in-plant seismic instruments whose recordings were of short duration and high frequency. Most of the seismic Category I structures for this , site were found on Devonian shale bedrock. There were no capable i faults in the site area. During Plant site excavation some minor faults and folds were discovered. These were investigated by the NRC Staff, the Applicant, the USGS, and the U.S. Amy Corps of i Engineers and found to be due to glacial folding in the area. These structures are non-tectonic and noncapable. The USGS is examining the possibility that earthquakes may be related to injection of chemical wastes in two wells which are about 7 miles from the earthquake epicenter. Past experience with induced
Minutes 311th ACRS Meeting 4 June 23, 1986 seismicity has shown seismicity beginning near the well head and spreading cutward. However, no seismicity has been detected prior to this event near the wells. For these reasons, the Staff considered it unlikely that this seismic event involved injection by these wells. D. Okrent asked what research was planned in connection with geologic structures. P. Sobell indicated that the effort is primarily being conducted by the Applicant. He will be looking at geological, seismological, and geophysical data accumulated in the last 5 years since the FSAR was written. D. Okrent expressed concern regarding what an adequate scope of an examination of this issue would be. R. L. Wesson, USGS, described some of the objectives of the Geological Survey's preliminary study. He indicated that the epicentral locations of the main shocks and aftershocks are fairly well determined. The depth of the earthquake is still an important question which has not yet been resolved. He mentioned P. 3 Pomeroy's suggestion that some form of seismic monitoring with sensitive instruments be conducted for a few years. R. L. Wesson 4 agreed that this was a good recomendation from the point of view of determining whether this earthquake had some relation to the wells. R. Hernon, NRC, indicated that plant walkdowns by the Staff following the earthquake revealed no significant structural damage. Measured structural response analyses and the response calculated using the recorded foundation motion in a fixed base reactor building dynamic model showed that the frequency responses compared well and the amplifications were similar. This confinns a lack of rocking response and insignificant soil-structure interaction. From the quantitative assessments done to date, all of the comparisons for equipment qualified by testing show that there are large margins. The Staff's conclusions are that there was not a significant safety impact on the equipment and structures that have
; been identified. The Staff intends to perform an additional l quantitative assessment of the seismic qualification of a broader sample of equipment types located in different buildings on various elevations. The Staff also intends to perform a generic evaluation of a high frequency short duration earthquake with regard to its energy content and potential safety significance for equipment and structures at Perry. Using the results obtained the Staff will assess the seismic capability of the Perry Plant if another ;
earthquake of similar characteristics but with higher magnitude and a longer duration should occur near the site.
)
M. Edelman, Cleveland Electric Illuminating Company, reviewed the i plant status prior to the seismic event as well as inspections and findings that were made subsequent to the event (see Appendix V). C. Chin, Civil Structural Engineer for Gilbert Commonwealth, discussed a comparison of a recorded 1986 Ohio Earthquake event with Perry design values. He sumarized the nature of the 1986 earthquake as one having high frequencies, short duration, low velocity and low energy displacement. He indicated that the plant design earthquake has broad band frequencies, long duration, high
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l Minutes 311th ACRS Meeting 5 June 23, 1986 l velocity and high energy. In spite of the insignificance of the energy content of this earthquake, Gilbert Comonwealth did a study to reconfirm and quantify the margins of the active components of the equipment. C. Chin discussed the criteria used in selecting the equipment for quantification of margins. He noted that the group of equipment selected was that which would be most sensitive to seismic frequencies above 14 hz, that is, equipment which would be most sensitive to this type of earthquake. The conclusion of the evaluation was that structures and equipment design had ) substantial margins of safety relative to load and stress induced by this earthquake. J. C. Mark pointed out that Gilbert Commonwealth could have made an energy absorption comparison between the design spectrum and the 1986 earthquake by comparing areas under the relative velocity spectra. C. Chin acknowledged that this was not done but was another way of measuring the earthquake's energy. t R. Holt, Western Geophysical, discussed the licensing of the Perry Plant and the seismology of the site. He gave a brief description of the January 31 earthquake and cited specific response spectra. He indicated that the Perry plant was licensed under the concept of the tectonic province which assumes that the largest earthquake in that province would occur at the site. He indicated that the earthquake's epicenter occurred in the town of Leroy approximately 10.5 miles from the plant. The focal depth was approximately five kilometers at the magnitude of 4.9 on the Richter Scale. P. G. Shewmon asked how close the focal depth was determined. R. Holt indicated that it was determined from a series of aftershocks to be somewhere between the surface and 10 kilometers, approximately 5 kilometers deep. T. Leblanc, Western Geophysical, explained that in order to generate a magnitude of five, there is a minimum focal depth required. That was how the five kilometer focal depth was determined. He concluded that the tectonic province approach used was still valid. There was nothing to indicate capable faults nor
- tectonic structures in the epicentral area.
P. Talwani, University of South Carolina, addressed the question of the possibility of injection wells inducing the 1986 earthquake. He indicated that the likelihood of involvement of the wells was very low since there was no report of earthquakes at those wells of magnitude 3 or above before the 1986 earthquake. He noted the lack of any information or any evidence of earthquakes in the vicinity of the wells before this earthquake or after. He did note that there was a general lack of monitoring instrumentation in this area before this particular earthquake so that it is conceivable that one might miss earthquakes of magnitude 2 to 2.5. He suggested that it is the characteristic of injection wells to trigger small numbers of micro-earthquakes which would be picked up by the available instrumentation. Injection wells are characterized by a large number or swarms of these very small earthquakes. The seismic pattern, one main shock and a very few aftershocks dying out in a few days, is very typical of the type of earthquake that has been seen in the eastern United States in Kentucky, New Hampshire, and several other locations. The matter arpes for a f
. Minutes 311th ACRS Meeting 6 June 23, 1986 regular tectonic earthquake rather than one induced by a relation- l ship to the injection wells.
I. B. Wall, Electric Power Research Institute (EPRI), briefly outlined EPRI research addressing some generic issues brought to focus by the 1986 northern Ohio earthquake (see Appendix VI).
; EPRI's seismic research is in four principal areas:
. Seismic hazard in eastern United States
. Seismic margins for eastern United States nuclear plants
. Database for soil-structure interactions
. Database for piping, damping and ultimate capacity I. B. Wall briefly described EPRI's seismic margins project which consists of walk-down procedures and quantification procedures for critical components. He estimated that results of the work would q be available by February 1987. J. C. Ebersole expressed interest in the walk-down procedures to identify seismically critical systems and components. I. B. Wall agreed to provide a draft 1 report in a couple of months. Mention was made of an on-going project to quantify the effectiveness of -northern Ohio type earthquakes which involve short duration, high frequency content.
D. Okrent suggested that it would be useful to determine how much
, one can exceed the design spectra in the 12-20 hz region before adverse effects on equipment. I. B. Wall agreed to.look into that matter. An EPRI project was mentioned involving the deployment of a dense array of accelerometers at a site near Parkfield,
, California to measure the coherency of high frequency seismic l energy such as the tau effect postulated by N. Newmark in the late 1970's. J. C. Stepp, EPRI Seismicity Owners Group, described the l Owners Group development of a probabilistic model to compute earthquake hazards at any point at any geographic location. The motivation for this project is the Charleston earthquake issue. This probabilistic model which focuses on estimating hazard from 4 moderate and large earthquakes would deal with all hypotheses of earthquake causation in the literature to interpret source zones and ultimately the uncertainty about the hazard estimations. The project is divided into two phases: (1) methodology development and (2) methodology review, testing, and enhancement. Six . independent team interpretations or hypotheses are being considered in this study. D. Okrent in'dicated that he did not see any evidence for concern that the earthquake which ' occurred damaged the operation of the plant. He agreed with the Staff conclusion that a 5 percent power letter be issued, and.also thought that the USGS monitoring effort
- for 2 to 3 years should proceed, f III. Integrity of Primary Coolant Systems (0 pen)
[ Note': E. G. Igne was the Designated Federal Official for this portionofthemeeting.]
, Minutes 311th ACRS Meeting 7 June 23, 1986 P. G. Shewmon explained the history behind the NRC's promulgation of a double-ended guillotine break. He explained that massive structural constraints were needed to cope with the massive forces associated with the double-ended rupture of high energy lines in nuclear plants. Pipe restraints and snubbers were needed to handle arbitrary breaks in the primary system. Design changes were put in place to cope with asymmetric blowdown loads. Pipe whip restraints and snubbers were used in the balance of plant. ECCS systems were put in place to handle the consequences of large sudden breaks. In looking for a more rational and safe approach, research on i elastic-plastic fracture mechanics presented a means of justifying '
a more reasonable alternative, the fact that a large through-wali crack in a high energy line will be stable or
" leak-before-break" instead of leading to sudden failura.
P. G. Shewmon explained that the acceptance by the NRC of the a concept of leak-before-break has lead to a series of changes. The first change and application came after resolution of USI A-2, Asymmetric Blowdown Loads. It was determined that the primary piping of the plant would leak well before it broke and one did not need to rebuild the primary system to resist asymmetric blowdown loads. The next application was to the pipe restraints in the primary system embodied in a limited scope revision of general design criterion 4 (GDC 4). The current topic under review is the Broad Scope Revision of GDC 4 which would allow the application of leak-before-break to pipe runs in the balance of plant if they met several criteria such as a low failure probability and are expected to exhibit an easily determined leak well before fast failure. P. G. Shewmon indicated that the Subcommittee on Metal Components met on February 27-28 to hear presentations by the NRC Staff, representatives of Beaver Valley Power Station, Unit 2 (LeadPlant in applying the leak-before-break concept to the entire plant), EPRI regarding its research program on leak-before-break,
- Westinghouse Electric as an interested vendor and a representative of the Atomic Industrial Forum. D. Okrent indicated that he could understand the approach practiced in the Federal Republic of Germany which involves forward-fitting and demands high quality in fabrication of pipe. He had sone reservations, however, regarding approval of the general idea of leak-before-break to all existing U.S. LWRs without carefully examining the specifics of the proposal. P. G. Shewmon noted that there are some tangible benefits to takir.g out restraints such as ease of inspection and making pipe runs more flexible for temperature changes. There would probably be an overall improvement in safety if pipe restraints were removed. D. Okrent suggested that approval for this approach may be premature until the issue of waterhammers large enough to rupture pipes is thoroughly investigated. C.
Michelson expressed interest in the design of sub-compartments and ' compartments relative to the pressure produced by previously i postulated breaks. His interest extended to the design basis fcr sub-compartments. I l 1
, Minutes 311th ACRS Meeting 8 June 23, 1986 J. A. O'Brien, MRC, defined the coverage of the Broad Scope GDC-4 Rule. He explained that pipe whip restraints in general have a negative safety impact. This can be gleaned from a large base of experience data on large pipe breaks in the U.S., as well as elsewhere in the world. He indicated that he could not recall a case where pipe whip restraints would have had a positive impact at nuclear plants. He mentioned several negative facts about pipe whip restraints, including the fact that they will complicate the seismic performance of piping because of the possibility of impact loads between the pipe and the pipe whip restraints during an earthquake. He noted the obvious diminished effectiveness of inservice inspections and increased worker radiation exposures resulting from limited accessibility (see Appendix VII).
J. A. O'Brien explained that the Broad Scope Rule is limited only , to dynamic effects. The double-ended guillotine pipe rupture is still postulated for equipment qualification, containment design, and ECCS performance. In answer to inquiry by C. Michelson, J. O'Brien indicated that pressurization in compartments, sub-compartments, and cavities from pipe rupture has been eliminated except for volumes related to the containment function. C. Michelson was of the opinion that the elimination of the ; pressurization of -sub-compartments was degrading the level of ' safety that is already built in. J. A. O'Brien explained the results of an international leak-before-break seminar. The United Kingdom intends to reject leak-before-break because of what he considered were unrealistic concerns about stress-corrosion-cracking and nondestructive examination. Those in France are inclined to reject leak-before-break but are pursuing the issue. The West Germans are very advanced 'in their application of leak-before-break and the Italians are not far behind the Germans. The Japanese are very assertive in going forward with application of the
; leak-before-break concept and the Canadians are also applying the a concept at their Darlington site.
D. Okrent broached the subject of the waterhammer question and asked whether a steam line can experience a waterhamer event. J. O'Brien indicated that waterhammer problems are addressed in two ways: licensees either document that specific lines have a good history of not having waterhammer or else take measures to mitigate or prevent waterhammer. G. A. Reed suggested that if the Staff
, admits that a steam generator overfill transient . Is possible in a PWR then the Staff must admit that waterhamer is possible. J.
O'Brien Indicated that Staff criteria postulate an extremely low , probability of occurrence for waterhansner and corrosion creep. Licensees must satisfy these criteria before they can apply the leak-before-break concept. C. Michelson again suggested that leak-before-break will mean a decrease in defense-in-depth. W. Kerr indicated that he could not see any possible improvement in safety from this approach. i
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. Minutes 311th ACRS Meeting 9 June 23, 1986 P. G. Shewmon noted that the Staff has not said anything about approving leak-before-break for GESSAR plants. He wondered if these BWRs would be considered under this rule if they could be convinced to reduce or eliminate the probability of stress-corrosion-cracking through either hydrogen treatment or replacement of lines with a stress-corrosion-cracking resistant material. J. A. O'Brien indicated that BWRs are excluded from the Broad Scope Rule because they do not satisfy acceptance criteria.
The Committee discussed the Broad Scope GDC-4 Rule Acceptance Criteria. D. Okrent thought they were not as stringent as the Staff contends and could be interpreted to be stringent or non-stringent as written. He noted a strong emphasis on operational events that lef t him feeling uneasy. He did not see the types of compensating actions proposed that would make one more confident as in the Federal Republic of Germany. P. G. Shewmon
- suggested that it was his concern that the Staff would have so many questions f. ..n licensees that it might turn out to be less expensive not to remove the restraints. C. Michelson contended he knew of some rather large ruptures of relatively high pressure water lines that have dumped a considerable amount of water. He agreed that the double-ended rupture is extremely unlikely but the pressurization of compartments could be an important factor in the case of the rupture of high pressure water lines. He expressed concern that the design basis includes enough drain lines and blow-out panels to take account of large ruptures of high pressure water lines.
P. G. Shewmon discussed another by-product of the NRC Piping Study Group regarding stress-corrosion-cracking and how the Comission intends to cope with this issue (see Appendix VIII). He mentioned a list of recommendations by the Piping Review Committee and a list of four options that can be used to alleviate the stress-corrosion-cracking problem. The Staff wishes to encourage licensees to put
- in ' material resistant to cracking and implement processes for residual stress improvement as well as improved water chemistry.
The Staff intends to give licensees credit for the use of these options, as well as for any repairs made. The Subcomittee thought that Revision 2 was a good package and that the full Committee should approve issuance for public comment. P. G. Shewmon explained that the Subcommittee discussed ACRS , concerns regarding pressurized thermal shock implementation. R. ) Woods, NRR, presented the NRC Staff's responses to certain ACRS ' questions. The Committee discussed the first ACRS question which asked whether some classes of nuclear plants or particular designs are subject to a significantly higher frequency of severe vessel ! overcooling transients and should not be treated generically in the 1 PTS Rule. D. A. Ward explained that while the challenge frequency might be higher in B&W plants, the backflow through the reactor vessel vent valves tends in most areas to provide more of a mixed fluid to the vessel walls. R. Woods explained that a PRA analysis of a generic B&W Plant basically found that although the challenge rate involving the frequency of overcooling events might be
. Minutes 311th ACRS Meeting 10 Jun@ 23, 1986 somewhat higher than with other plants, the vent valves do provide significant mixing and thereby prevent the temperatures from going as low as they might otherwise go. The Staff concluded that the B&W plants should have the same screening criteria applied to them as is applied to other plants. G. A. Reed mentioned a recent SMUD event. D. Okrent mentioned recent events at Rancho Seco, Oconee and Crystal River - all with loss of non-nuclear instrument power.
R. Woods indicated that those events were fully taken into account in the development of the Rule. D. Okrent recalled seeing an INEL report for a class of accidents involving steam generator overfill. This report showed a higher risk number for B&W plants than for Westinghouse plants. R. Woods acknowledged the potential pressurized thermal shock consequences associated with overfill of a steam generator. D. Okrent suggested that the ACRS Staff research this combined INEL Pacific Northwest Laboratory report and pass the information on to the NRC Staff. H. Etherington reminded the Comittee about certain nonconservatisms (regarding the PTS rule) that he mentioned at the January ACRS meeting (309th). He indicated that the crack distribution was probably the biggest uncertainty in the vessel integrity simulation analysis code which uses the OCTAVIA code for flaw size distribution. He noted that one of the reports mentions other crack distributions including one that allows for crack growth. He indicated it is his belief that there is no crack growth allowed in the PTS Rule. He recomended that the Committee review the next request for a waiver of the PTS Rule. P. G. Shewmon indicated that ACRS consultant, I. Catton, questions whether high pressure injection recovery following partial core uncovery is addressed adequately under the PTS Rule. D. Okrent pointed out that the issue appears to have been given up to those studying severe accidents but not looking at pressurized thermal shock transients. He suggested that the Staff has not adequately disposed of the matter probabilistically. P. G. Shewmon acknowledged the Staff admission that there are steam generator overfill scenarios which are considered significant for pressurized thermal shock. D. Okrent expressed an interest in this matter and indicated that he would look into the matter further. The Committee briefly discussed uncertainties involved for some nuclear plants regarding the data on chemical composition of critical welds. Discussion centered on the confidence level on the copper and nickel content that could be expected. P. G. Shewmon noted that there are statistics on variations through a heated weld wire material. D. Okrent wasn't sure whether the uncertainty in the amount'of copper was very small or whether it was significant. P. G. Shewmon indicated that the uncertainty along the wires is not zero but is narrowly bounded. There are standard deviations on the kinds of variations that occur in a weld or a given heated wire and flux. P. G. Shewmon discussed the question of the *xpected consequence of a through wall crack on the likelihocM of core melt, late containment failure, or early containmers failure. He suggested that this is a severe accident scenerio question and not a
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itinutes 311th ACRS Meeting 11 June 23, 1986 pressurized thennal shock question. D. Okrent posed the questio whether the scenario had an estimated frequency of less than 5x10-g l of causing a through wall crack. R. Woods indicated that that i frequency is meant to apply to the sum total composite of all of ' the transients that might occur. He mentioned a paper presented at the Water Reactor Safety Meeting recently held in Gaithersburg j which should have been made available to the Subcomittee members. ' He indicated that that paper was based largely on a contract that NRR had at Pacific Northwest Laboratories. D. Okrent questioned : the peer review done on the document but said he would read the l report. G. A. Reed requested that the NRC Staff review the reasons behind the recent U.K. Central Electricity Generation Board ; decision to provide an automated and integrated control system to depressurize by primary blowdown in order to deal with pressurized ; thermal shock transients. R. Woods indicated that the Staff has l considered that subject for some years and has for good reasons l 3 decided not to pursue the subject in that way. W. Minners, NRC, l indicated that that subject had been explored but that the Staff recognized the difficulty of making such a decision with its attendant political pressures. D. Okrent indicated that he would write a letter to England to cyclore the logic behind that Central Electricity Generation Board decision if the Staff is not interested in pursuing the matter. IV. NRC Severe Accident Policy (0 pen) [ Note: D. Houston was the Designated Federal Official for this portion of the meeting.] W. Kerr indicated that the ACRS Subcomittee on Class 9 Accidents met on February 24, 1986 to review the draft Implementation Plan for the Severe Accident Policy Statement and the regulatory use of new source information. The plan is divided into three general areas: systematic evaluation of individual plants, procedures for PRAs to be used on new plants along with some non-probabilistic 4 criteria, and possible char.ges in rules or regulations that are based on new infonnation developed as a result of the source tem research. !!uch of the meeting was devoted to the systematic evaluation of individual plants being developed by NRR and IDCOR. Z. Posztoczy, NRR, discussed action items associated with the Severe Accident Policy Statement (see Appendix IX). He explained that the Policy Statement is organized into two parts: one for new applicatiors and the other for existing - plants. Under the new application part, there are two action items, one of which is to issue guidanca on the role of PRAs and the other, to consider perfor:nn e criteria for containment systems. Relative to existing plaats, the Commission emphasizes that a systematic approach for tt'a examination of individual plants is needed. Any plant-dependent weaknesses or vulnerabilities would be corrected througt. modification using the backfit policy. The Staff's Implerrantation Plan consists of three elements: , l l 4 l
. Minutes 31*_th ACRS Meeting 12 June 23, 1986
. existing plant evaluation '
. development of guidance on the role of PRAs for new plant ,
applications
- _. changes in rules and regulatory practices .
l D. A. Ward asked whether the examination of plants is really an evaluation of actual nuclear plants or an examination of plant , designs. Z. Rosztoczy indicated that the examination is expected r to involve not only plant design, but some examination of the '
. as-built plant. He noted that the IDCOR methodology is based on a
' plant walkdown as part of the design review . D. A. Ward wondered ,
whether the examination would deal with personnel performance ; (actual crews that man the plant) or be more of a hypothetical ' evaluation. Z. Rosztoczy spoke of a review that would deal with ! equipment under adverse environmental conditions but did not speak of review of the performance of actual plant personnel. Operating 1 procedures are being factored into the review. ;
- 2. Rosztoczy stated that the Staff intends to provide additional infomation for the designers of future plants in order to make
- them more forgiving of r.evere accidents. D. A. Ward asked if the -
Staff will learn anything about operator training as a possible ! sericus cause of a plant being an outlier. Z. Rosztoczy indicated that one of the major goals will be to identify the operator : training necessary so that operators will be ready to handle severe accidents. The purpose of the program is to identify outliers or ; weak plants through the identification of design vulnerabilities. i M. W. Carbon noted that IDCOR has done four or five PRAs already. , He asked if the Staff has confidence that the IDCOR methodology i will find outliers. Z. Rosztoczy indicated that IDCOR has done detailed analysis of plants before but not really PRAs. The results of the analysis of those four plants has been the development of a methodology which may be applied by the owners of , other plants to search for vulnerabilities. IDCOR is rather l confident that its methodology will pick out most of the outliers
' that the PRA would pick up. D. Okrent expressed interest in the reports on the IDCOR methodology which are due in April 1986. J.
C. Ebersole suggested the development of a manual for walkdown procedures. Z. Rosztoczy indicated that he expected that the IDCOR methodology would have some procedures for the walkdown. Z. . ; Rosztoczy briefly mentioned the Staff's development of guidelines and criteria for plant examinations in parallel with the review of ! the IDCOR individual plant examination methodology. D. A. Ward i noted the use of uncertainty ranges in the resolution of technical ! issues. This subject was briefly discussed regarding operating . experience at the plants under study regarding uncertainty ranges ! for operator actions. ; Z. Rosztaczy discussed source term-related rule changes as part of l the program element to make changes in rules and regulatory j practices. He noted that the Staff is exploring the possibility of 4 the development of containment performance criteria. D. A. Ward I asked for a few examples of such criteria. Z. Rosztoczy spoke of an evaluation of the utilization of containment sprays and a requirement on containment leak rate. He spoke of grouping source
Minutes 311th ACRS Meeting 13 June 23, 1986 changes into near-tenn, intermediate and long term revisions (see Appendix IX). The Staff proposes a possible revised approach which is to reexamine the usa of fission product calculations in licensing. The Staff would basically follow the same approach 1 taken in TID-14844 and Regulatory Guide 1.3/1.4 but update the l procedures to today's knowledge and make the process more realistic than it has been in the past. C. Michelson wondered if the Staff has considered the effect of the changes to GDC-4 regarding the leak-before-break concept with the removal of pipe whip restraints. Will the probability of postulating the double-ended guillotine rupture unrestrained cause a much higher probability of containment failure. C. Michelson indicated that the present design base is to restrain pipes from puncturing the containment. The next design bases will not be to restrain the pipes. Z. Rosztoczy indicated that the program relies
- basically on evaluating six reference plants as they stand today.
The contemplated changes to GDC-4 are not included in the current approach. Z. Rosztoczy discussed the relationship of the Severe Accident Implementation Program and the NRC Research Program. He mentioned NUREG-0956 " Reassessment of the Technical Bases for Estimating Source Terms" and NUREG-1150, " Nuclear Power Plant Risks and Regulatory Applications." D. A. Ward asked if the Severe Accident Research Program has a component which seeks to differentiate among different levels of core melts. J. Mitchell, NRC, and 2. Rosztoczy both indicated that they did not know whether the Research Program is doing work on discriminating between classes of core melt. D. A. Ward indicated that he expects an answer from the Staff on this matter. Z. Rosztoczy explained that the IDCOR work up to now and NRC work in support of NUREG-1150 has been limited to internal events in i
- general. This does not mean, however, that there is no ongoing l work on external events. External events for those plants which '
are especially exposed to particular external events have been considered in the licensing process. In addition, there has been a considerable seismic research effort ongoing for a number of years. He indicated that based upon the work with PRAs on external events and NRC's work toward developing simplified methodologies for seismic events, a schedule for handling of external events in the Implementation Program has been developed by the NRC Staff and is under management review. Z. Rosztoczy indicated that the Staff would like to hear the Comittee's views on the Implementation Program. Of particular interest would be the Committee's coments on areas that the Staff might not have addressed in the program. The Comittee deliberated and later in the session produced a report to the Commissioners on i the Implementation Plan for the Severe Accident Policy Statement and Regulatory Use of New Source Term Infonnation.
Minutes 311th ACRS Meeting 14 June 23, 1986 I V. AdvancedReactorDesigns(0 pen) i ! [ Note: M. M. El-Zeftawy was the Designated Federal Official for thisportionofthemeeting.] M. Carbon indicated that DOE and its contractors have been carrying out design studies on an Advanced HTGR and two liquid metal reactors. NRR is interacting with DOE in an early stage in the design effort and the ACRS has been asked to become involved in the review effort, also at an early stage. He indicated that the present session is to be a briefing on the High Temperature Gas Cooled Reactor Design (GCR). He suggested that members pay particularly attention to several specific safety topics. These include: the temperature at which fission products are released from the fuel, the peak clad temperature in worst case accidents, the type of ultimate decay heat removal systems proposed, questions about graphite fires, the need for containment and the need for an emergency preparedness plan. He noted that several major utility
- presidents in the United States are seriously considering offering seed money to produce a module demonstration gas-cooled reactor plant.
T. L. King, NRC, indicated that the work on the GCR design and the two liquid metal reactors (LMR) and NRC's interaction with this ! j program is under the Advanced Reactor Policy Statement which was
! issued for public comment about a year ago. The NRC plans a review process over the next two years to involve conceptual designs for j one GCR and two LMRs. The Staff plans a number of interactions on i key issues and the issuance of a preliminary safety information document, plus an SER on the conceptual design. Due to the impact of the Gram-Rudman Law, Staff resources which were 5 to 6 Staff 4
years per year and approximately $1.25 million of technical assistance per year have been reduced to two Staff years per year and no technical assistance. The impact of this reduction on
- planned reviews is currently being assessed and will probably limit future interactions to a few key issues (see Appendix X). He gave examples of the few key issues: (1) containment vs. confinement; (2) use of interface criteria only. for the balance of plant; (3) use of a single passive highly reliable decay heat removal system; (4) treatment of severe accidents and emergency planning. G. A.
Reed claimed that these key issues could also be applied to current l
. LWRs as a definition of a safety envelope which could be closely i f
regulated releasing the rest of the plant from the burden of heavy regulation. F. J. Remick wondered why there was need to discuss emergency planning for the GCR. T. L. King indicated that the GCR l approach is basically to try and limit the off-site doses at the l site boundary to less than the Protective Action Guidelines which would obviate the need for an Emergency Evacuation Plan. There is a need to consider what accidents one would have to look at to 4 evaluate the design to make sure that it does not exceed the ! Protective Action Guidelines. l F. Gavigan, DOE, indicated that the objective of the Advanced { Reactor Program managed by DOE is the development of a low cost i passively safe, electrical generating option competitive with j contemporary alternatives ~ (see Appendix XI). The characteristics i _ , _ . - . . . -~ . _ . . _ _ . . - -_ ,__ _ _ _ -
Minutes 311th ACRS Meeting 15 June 23, 1986 of these designs are that they tend to be small and intermediate in ' size and modular in the sense that they can be added to a site as a ! function of a load growth demand curve. DOE is emphasizing passive safety and looks for certified and standardized plants. DOE is
- emphasizing shop fabrication to reduce on-site fabrication. The l plants are to be cost effective as a result of very short ,
) construction times and high plant capacity factors. G. A. Reed expressed concern regarding modular construction because of potential difficulties regarding maintenance. F. Gavigan indicated that the preliminary cost estimates are that all three of the plants being considered are competitive with LWRs regarding capital costs and all are less expensive than coal on operating costs. DOE has identified research and development needs for both the GCR and the LMR concepts and resources are being mobilized to answer any questions that may arise. U A. C. Millunzi, DOE, explained that DOE is designing the modular L GCR to meet both user requirements and top level regulatory
, requirements (see Appendix XII). He briefly discussed the design paiemeters and licensing approach. He mentioned the use of an integrated approach to develop requirements such as top level regulatory criteria and evaluate the design selected to meet these requirements. A. J. Neylan, Division Director, G. A. Technologies, Inc., briefly described the key characteristics of the modular GCR.
A key to the design is the use of ceramic-coated fuel particles
- imbedded in the fuel matrix in a graphite block as at Ft. St.
Vrain. It has resulted in very low radioactive releases to the operators. J. C. Ebersole pointed out that the one factor that
- makes the Ft. St. Vrain reactor an operational disaster is the use of water-buffered seals that presistently pump water into the hot environment. A. J. Neylan indicated that the helium gas coolant which is an intrt single phase is indeed a key difference in the GCR design. Although Ft. St. Vrain has been plagued by water 3 ingress and has suffered a terrible availability record because of j it, G. A. Technologies intends to cure that by changing from a steam-driven circulator to an electric driven motor eliminating the o
j' source of water to the bearings. The present design will have d magnetic bearings. !! A. J. Neylan presented an artist's concept of a four unit modular ll GCR plant noting that the modules are located side-by-side below
- grade. He made repeated comparisions to the Ft. St. Vrain design including the fact that the control rod structure is similar. He a focused on the fuel particles which had been successfully developed and demonstrated initially at Peach Bottom, Cores 1 and 2. The fuel to be used in this advanced HTGR is the same triple coat "triso" particle used at Ft. St. Vrain, Core No. 1, which has
- l performed well in terms of limiting releases.- It consists of
! buffer zone and interporphyritic carbon layer, a silicone carbide
- . zone, and a porphyritic carbon outercoat. The oxicarbide fissile
.' particle is 19.9 percent enriched versus the high enriched fuel used at Ft. St. Vrain.
Minutes 311th ACRS Meeting 16 June 23, 1986 F. A. Silady, G.A. Technologies, described modular GCR safety design objectives. He indicated that G. A. Technologies intends to provide a fuel quality and specify normal operating conditions to limit radionuclide inventories outside of fuel to meet 10 CFR 100 doses at the plant boundary. Another objective is to provide a reliable passive design selection to retain the radionuclides s within the fuel to meet the 10 CFR 100 doses at the plant boundary. Additional design selections, largely passive in nature, will be to , i retain radionuclides within the plant to meet Protective Action l Guideline doses at the plant boundary. In ansher to a question by 1 , D. W. Moeller, he indicated that the basic p16nt design is to meet l l the Protective Action Guidelines over a wide spectrum of events 1 l which encompass those for which 10 CFR 100 doses are met. The ' i emphasis will be to show that the fuel alone can meet the requirements of 10 CFR 100. Retention mechanisms such as deposition, plateout. or holdup in the vessel will help meet the I Protective Action Guidelines. He explained that GA technologies has found in this design that radionuclides can be retained in the fuel particles by removal of core heat, control of chemical attack and control of heat generation. l J. A. Neylan discussed design selections to remove core heat which included a passive reactor cavity cooling system, design selections to control chemical attack by water included limiting the sources (magnetic bearings), use of high
- reliable detection in isolation (quality single fuel (lowHe loop). hydrolysis) indicated and that GA is very sensitive to graphite oxidation both from the point of view of water and air and intends to limit air-graphite reaction to retain radionuclides in the core. He stressed, however, that the key to control of chemical attack by water is the fuel quality.
Design selections for reactivity control or control of heat , generation are the fuel negative temperature coefficient, the control rod system and the use of a reserve shutdown system. He concluded that the inherent characteristics and passive features of
- the modular GCR design assure radionuclide retention in the fuel sufficient to obviate the need for off-site sheltering, evacuation, or emergency drills. The program utilizes a systematic licensing approach consistent with the NRC's Advanced Reactor Policy.
VI. ProposedImplementationofNRCQuantitativeSafetyGoals(0 pen) i [ Note: R. P. Savio was the Designated Federal Official for this portion of the meeting.] l D. Okrent indicated that the ACRS Subcomittees on Safety i Philosophy, Technology and Criteria, and Extreme External Phenomena i met on March 12, 1986 to discuss the status of the ED0's review of j the two year trial use of the 1983 proposed Safety Goal Policy and f recommendations for the future use of the Policy. The EDO is still i preparing coments for the Comission. D. Okrent mentioned that J Comissioner Bernthal has asked a few questions and has proposed
! that the qualitative safety goal state that nuclear power is safer i than other viable competing energy generation technologies.
Comissioner Bernthal thought that the general public would not be a i /
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i
. Minutes 311th ACRS Meeting 17 Juns 23, 1986 impressed if nuclear were cast as being as safe as the coal technology. Nuclear should be much better than coal.
Comissioner Asselstine questioned whether the nuclear industry would lose credibility if it were willing to accept a core melt every 10 years. It was suggested that the general public which is now generally negative would be completely turned off by such a statement. D. Okrent pointed out that in the July 17, 1985 Comittee letter that the ACRS suggested that greater attention be placed on wyking toward an objective of a mean core melt frequency of 10~ per reactor year. He thought the Comittee should reiterate that particular objective. He mentioned the coment by Lester Lave, ACRS consultant, that the NRC should include all costs in cost-i benefit analysis even though the Staff may not be able to quantify '
- all of them. It should be expected that not all cost factors will p provide perfectly usable input. Some input might even be slightly j flawed. F. J. Remick thought that it is not the responsibility of l the Commission to protect the nuclear industry. He supported the view that nuclear generation of electricity should be comparable to or have a lower risk than the generation of electricity by competing technologies. Any additional risks must be quantified.
l J. Sniezek, ED0 Staff, indicated that there is not unanimous
- agreement on how the Staff should proceed to implement the safety goals. M. Taylor, NRC, mentioned the issuance of a February 14, i
1986 Staff paper addressing several Comission questions. He indicated that the Directors of NRR, RES and IE support the content ! of the February 14th paper (see Appendix XIII). He explained that i the Staff proposal of February 14 was to assist the Commission in i its deliberations regarding the best course of agency action. He i asserted that the February 14 paper accommodates ACRS coments and i all Staff coments in the Steering Group Report. It recommends issuance of qualitative statements in final form as the Comission
. Safety Goal Policy. It recommends authorizing the Staff to use the / integrated matrix on a trial basis as a quantitative measure of the i qualitative safety goals. He indicated that the Staff concludes j that the Comission should issue.in final fom as its Safety Goal Policy two qualitative statements regarding individual and societal
- risk. It should combine quantitative objectives (e.g., core melt
- frequency, individual mortality risk) and the benefit-cost
{ guidelines into an independent matrix which the Staff can use as a l quantitative measure. These guidelines' should be reviewed as internal objectives not individually discreet statements of expectation. He discussed the key features of the Safety Goal
- Policy Statement and stressed again that the qualitative Safety l l Goals should be incorporated into the regulatory decision making '
process through the Integrated Safety Goal Matrix. He added that the Policy Statement indicates that safety goals are not a
- substitute for existing regulations. The Policy Statement f as an objective a mean core melt frequency of less than 10'gvors per
. reactor year for current plants in accord with ACRS recomendations. Future plants would be expected to achieve a core melt frequency lower than that for current plants. 1
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. Minutes 311th ACRS Meeting 18 June 23, 1986 l 1
M. Taylor indicated that the Policy Statement recognizes that uncertainties must be taken into account in the regulator decision making process. H. W. Lewis expressed concern that uncertainties may not often be correlated with risks. M. Taylor indicated that the Commission is looking for a balanced design. The Integrated Matrix establishes a risk state that would be viewed as a de - minimis level for accident risks. D. Okrent noted that NRC's dii minimis level for latent cancers is about 200 to 300 times lowiir than EPA standards for lifetime risks from pesticides. He suggested that the general public may quarrel with the NRC level being posed as a d_e minimis risk. M. Taylor indicated that the Policy Statement favors use of the Integrated Safety Goal Matrix for review of generic safety regulations in conjunction with deterministic decision making. The Integrated Matrix will be only one factor used in the decision J process. The Comittee discussed parameters in the Integrated Matrix. M. Taylor noted that prompt cancer risk will be contro111nq5 over latent cancer risks. D. W. Moeller noted the target of 10 per reactor year for a large scale core melt frequency. J. Sniezek admitted that PRAs at the current state of the art cannot differentiate between a large core melt and the onset'of core melt. He thought there was a definite impetus to improve the technology in the area of differentiation of various levels of core melts. D. W. Moeller asked if there are standard models for calculations that need to be made to detennine whether a licensee meets the targets in the Integrated Matrix. M. Taylor indicated there were none but NUREG-3568 existed as a value impact analysis guide. He acknowledged however that the Staff needs more in-depth analysis than provided in NUREG-3568. D. Okrent asked why these particular factors were chosen for use in the matrix. M. Taylor indicated . that it was an arbitrary decision that the Staff chose to concentrate on health effects. There was some discussion of the fact that there might be a wide band of uncertainty associated with decision matrix calculations. Mention was made of the sensitivity of mean values and at what point the Staff would take action against a licensee if it didn't meet the matrix limits. J. Sniezek indicated that the Staff has not specifically defined yet how it intendt to implement the matrix. The Staff is asking permission to use the matrix to assess whether individual plants meet the spirit of the two qualitative safety goals that would be issued as the Comission Policy Statement. M. Griesmeyer, ACRS consultant, presented the views of Lester Lave, who was against the matrix formulation. Lester lave thought that the Comission should do the best computations possible but not use the $1000 person-rem as a proxy for detennination of cost-benefit. Lester Lave thought that the Comission should proceed with some sort of Safety Goal Policy although it may be flawed. It has taken too long to develop the Safety Goal. D. McClain, ACRS consultant, thought that the Staff should use some form of cost-benefit. He 1
Minutes 311th ACRS M:eting 19 June 23, 1986 thought that the societal goal should not be: in the Policy Statement. There ought to be some statement regarding a national societal goal and no de minimis cutoff on mandated costs. M. Griesmeyer thought that the matrix from his point of view was a positive step although flawed since it does not describe how to deal with uncertainties. He was not against the matrix fonnulation because he thought the Commission did not intend that it be applied explicitly. It intended to take account of other factors in the decision making process. However, he thought that other factors should be incorporated into the matrix directly to make it more complex. D. McClain thought that a national societal goal would not be unlike what has been suggested by the questions of Commissioners Bernthal and Asselstine. VII. Report on Visit to Three Mile Island, Unit 2 (0 pen) [ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.] D. W. Moeller explained that he spent Friday, March 7, 1986 visiting with NRC Staff and officials of General Public Utilities, Inc. (GPU) at the Three Mile Island, Unit 2 Plant in Middletown, PA. (see Appendix XIV). The principal topic covered during the visit was the clean-up effort at THI, Unit 2, including removal of the fuel debris from the primary system. He noted that GPU has developed some interesting tools for cutting, picking up, and removing the debris. One of the main problems with removing the debris has been the lack of clarity of the water in the reactor pressure vessel. This problem is due to a heavy growth of microbial organisms. These nitrogen-fixing bacteria or algae are being studied. Specialists are trying various methods for control such as various types of filters. While filtration and the use of a nitrogen atmosphere have not proven completely successful, morale is high and studies to detemine a method for controlling the organisns are continuing. 1 He briefly discussed decontamination operations within containment which appear to be progressing well. He mentioned the use of scabbler machines which were particularly interesting in that they can chip up to 1 inch from a concrete surface in one pass. He noted that 22 tons of fuel have been removed from the core and removal operations are continuing essentially around the clock. He mentioned the fact that the ACRS had recommended the hiring of a cesium chemist to assist in their operhtions. D. W. Moeller indicated that when tt uring the facility he was impressed by the size and obvious cost of the plant which is essentially in new condition. He thought it significant that surfaces within the plant that were painted prior to the accident appeared to be far more easily decontaminated than those that were not. He indicated that he was impressed by the on-site NRC Staff who appear to be tough and finn but reasonable in dealings with the licensee.
- . - Minutes 311th ACRS Meeting 20 June 23, 1986 l
. t l* VIII.ExecutiveSessions(0 pen) [ Note: R. F. Fraley was the Designated Federal Official for this i portionoftheMeeting.] ! A. Subcomittee Assignments 1
- 1. Revised ACRS Subcomittee Assignments A revised list of Generic Subcomittee Assignments i
consistent with the one discussed at the meeting of the i ACRS Management Group on March 12,1986 (see memorandum D. Ward to ACRS members, dated March 14, 1986 entitled Revised Subcomittee Assignments) was introduced during the 311th ACRS meeting. Chairman Ward solicited member j comments regarding specific assignments and allocation 3 of tasks within 2 weeks so the assignments in the March
- 14, 1986 memorandum can be made final by April 1, 1986
) (seeAppendixXV). I 2. Safety Considerations for Future Nuclear Power Plants } The Comittee discussed a proposed ACRS report regarding f safety considerations for future light water reactors but 2 was unable to reach a collegial position. The report was 1 tabled indefinitely. A recomendation that each ACRS ' 4 Chairman prepare such a report at the end of his term was ! criticized by several members as a loss of collegiality ! in Comittee activities. Action was deferred. It was i suggested that the newly established Subcomittee on . Advanced Water Reactors could pick this up as a specific
- task.
B. Reports, Letters, and Memoranda [
- 1. ACRS Comments on Proposed Safety Goal Policy l The Comittee prepared a report to the Commissioners i of its continuing review of the proposed j implementation of Safety Goal Policy. A memorandum
! from V. Stello to the Comission, dated February 14, 1986 on Safety Goal Policy served as a focus for this part of the Comittee's review. Additional i coments by W. Kerr and J. C. Mark, and by H. W. Lewis were appended.
- 2. ACRS Report on the Perry Nuclear Power Plant. Unit 1 J
i
! The Comittee prepared a report to the Comissioners l ! of its discussion of the 1986 Ohio earthquake and its review of the implications with respect to the , Perry Nuclear Power Plant, Unit 1.
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- Minutes 311th ACRS Mesting 21 Jun2 23, 1986
- 3. ACRS Coments on Proposed Broad Scope Rule Revision to General Design Criterion 4 Environmental and Missile Design Bases The C' onaittee prepared a report to the Commissioners of its review of the proposed changes to General Design Criterion 4 regarding deletion of the double-ended guillotine pipe break as the DBA for aspects.of the plant such as design of piping restraints, combined blowdown loads, etc.
Additional comments by David Okrent and Glenn A. Reed, and by Paul G. Shewmon and David A. Ward were appended.
- 4. ACRS Coments on the Implementation Plan for the Severe Accident Policy Statement and Regulatory Use
; of New Source-Term Information The Comittee prepared a letter to the Acting Executive Director for Operations regarding its discussion with the NRC Staff of a draft Implementation Plan which is being prepared in accordance with the Comission's Severe Accident Policy Statement.
- 5. PRA Quantification of Public Health Risk The Committee prepared a report to the Comissioners regarding the need for guidance in characterizing dose categories from calculations done in
, probabilistic risk assessments for nuclear power stations and the use of realistic estimates in PRAs as the measure of impact on the public health and ( safety. This report was later withdrawn for further j reconsideration during the 312th ACRS meeting.
- 6. Responses to Recommendations of Panel on ACRS Effectiveness The Committee prepared a report to the Comissioners of its responses to specific recomendations made by the Panel on ACRS Effectiveness in its review of the Comittee's effectiveness as an advisory body to the l Commission.
- 7. Pressurized Thermal Shock Letter to the Acting Executive Director for Operations The Comittee prepared a letter to the Acting Executive Director for Operations indicating that the ACRS has some questions concerning the basis for approval of continued operation of reactor vessels that exceed the screening criteria of 10 CFR 50.61.
The ACRS noted that, if a request for a waiver is
- Minutes 311th ACRS Meeting 22 June 23, 1986 ~
received by the NRC Staff, the ACRS wishes to review l it with the Staff.
- 8. ACRS Coments on Draft Technical Report on Guidelines for BWR Coolant Pressure Boundary Piping The Comittee prepared a letter to the Acting Executive Director for Operations regarding the draft report NUREG-0313, Rev. 2, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" (manuscript completed September 1985).
- 9. ACRS Review of NRC Actions Resulting From the San Onofre Nuclear Generating Station, Unit 1. Novent)er 21, 1985 Event The Comittee prepared a memorandum from the ACRS Executive Director to the Acting Executive Director for Operations regarding its preliminary discussion of the water hamer/ loss of AC power event which occurred at the San Onofre Nuclear Generating Station, Unit 1(SONGS-1)onNovember 21, 1985. The ACRS requested that the NRC Staff meet with the Subcomittee on Westinghouse Water Reactors about two weeks after the completion by the Staff of its Safety evaluation on actions necessary for the restart of SONGS-1. The ACRS will also want to meet with the NRC Staff at some time in the near future to discuss the generic actions resulting from this incident which the NRC Staff will be recomending.
i 10. Move of ACRS to Bethesda, Maryland The Committee prepared a letter calling attention to the adverse administrative impact of separating the ACRS and the Comissioners in the proposed NRC building consolidation plan for the agency. C. Future Agenda
- 1. Future Agenda The Comittee agreed on tentative agenda items for the 312th ACRS meeting, April 10-12, 1986 (see AppendixII).
- 2. Future Subcomittee Meetings A schedule of future subcomittee activities was )
distributed to members (see Appendix III). ' D. Change to Bylaws A proposed change in ACRS Bylaws to provide for discussions by individual ACRS members with individual
- Minutes 311th ACRS Meeting 23 Jun2 23, 1986 4
J i Comissioners was introduced. Chaiman Ward requested that312th the the members studyApril ACRS Meeting (the 1986). text for consideration during I E. The 1986 Ohio Earthquake and Its Impact on the Perry 1 Nuclear Power Plant Unit 1 The Subcomittee on Energy and the Environment of the i Comittee on Interior and Insular Affairs, U. S. House of ! Representatives will conduct a hearing on April 8,1986 l concerning NRC actions in response to the January 31, i 1986 earthquake which occurred in northeastern Ohio. The l ACRS has been asked to testify regarding the implication of the 1986 Ohio earthquake for the Perry Nuclear Power ). Plant, Unit 1. D. A. Ward, C. P. Siess, and D. Okrent were designated to attend the hearing and give testimony as requested by telecon from Dr. Henry Myers. (Note: D. ! A. Ward and C. P. Siess will participate in this hearing. 1 D. Okrent will not.) I F. Realignment of ACRS Manpower and Fiscal Resources for FY j 1987 Consistent with Reductions in ACRS Resources The members discussed and endorsed the breakdown of (ACRS activities into general categories) and priorities regarding the anticipated ACRS workload for FY 1987. A more detailed breakdown will be presented during the April (312th) ireeting. G. Conduct of Employees - Conflict of Interest The Comittee decided to employ an outside legal counsel
; to assist in making legal determinations regarding j potential conflicts of interest of ACRS members.
H. Appointment of New Member 1 The Comittee discussed background infornation on a i prospective list of applicants ( seven out of a total of
- more than 50) to fill the vacancy on the ACRS that arose
- when R. C. Axtmann left the Comittee in 1985. Members were not entirely satisfied that they had exhausted the field of potential applicants and decided to consider
- additional persons. Further discussion on this matter is scheduled for the 312th ACRS meeting (April 10-12,1986).
i j I. Report of ACRS Management Group l
- D. Ward reported the results of the Group's meeting on Wednesday, March 12, 1986 as follows
l
- i
. Minutes 311th ACRS Meeting 24 June 23, 1986
. Legal advice regarding potential conflict of interest situations - See Item G ,
4
. Implementation of the Recorsnendations of the Panel on ACRS Effectiveness - See Item B.6
. Realignment of ACRS Manpower and Fiscal Resources for FY-87 consistent with reduction in ACRS resources - See Item F
, Revised ACRS Subcommittee Assignments - See Item A.1 i . Proposed change in ACRS Bylaws to provide for individual ACRS members and discussion among individual Comissioners - See Item D
. Periodic Update of ACRS Letters, Books - It was agreed that the books of .iCRS reports published in connection with the 300th ACRS meeting should be
, updated annually
. ACRS International Meeting on Reactor Safety at Wingspread, Wisconsin - It was agreed that the possibility of inviting U.S.S.R. delegates should be explored
. CRGR Activities - It was agreed that an ACRS Project Engineer or Fellow should routinely attend CRGR
, meetings and report the results to the Committee J. National Research Council Study of the Long Term Needs of the NRC Safety Research Program D. Ward, D. Okrent and C. P. Siess have been invited to testify before the National Research Council on April 4 . 1986 to provide their views and those of the Comittee as appropriate. D. Okrent will not testify during the April 4,1986 meeting but he and other members (e.g., H. W. Lewis)havebeensentquestionstoanswer. } The 311th ACRS Meeting was adjourned at 3:00 p.m., Saturday, March 15, j 1986. j i l r
O APPENDIXES TO MINUTES OF THE 311TH ACRS NEETING MARCH 13-15,1986 xts -M66 O
APPENDIX I NRC ATTENDEES 311TH ACRS MTG. O . Thursday, March 13, 1986 0FFICE-OF NUCLEAR REACTOR REGULATION R. Hernan - J. Stefano
- L. Reiter P. Sobel R. Hermann A. Lee R. Bernero H. Culman W. R. Butler S. M. Stern 0FFICE OF RESOURCE MANAGEMENT V. Zeoli EXECUTIVE DIRECTOR FOR OPERATIONS J. Sniezek O
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i NRC ATTENDEES - r 312TH ACRS MTG. , i Friday, March 14, 1986 , OFFICE OF NUCLEAR REACTOR REGULATION M. Spangler R. W. Hernan R. Wood J. T. Chen j L. Soffer
- F. Coffman !
Z. Rosetoczy . N. Anderson ' 0FFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS ' M. A. Taylor I 0FFICE OF NUCLEAR REGULATORY RESEARCH W. R. Pearson O J. O'brien J. Richardson M. Mayseeld J. Mitchell i 0FFICE OF NUCLEAR MATERIAL SAFETY '
& SAFEGUARDS A. T. Clark I i
l i i 1 O
+ x .
i m___________.._.._.__.__.______
PUBLIC ATTENDEES 311TH ACRS MEETING Thursday, March 13, 1986 T. Brazaitis, Cleveland Plain Dealer T. Zogimann, Duquesne LightCo. - A. Noe, Chesebrough Pond's R. E. Schaffsta11, KMC, Inc. C. L. Tully, AIF H. Glasspiegel, Shaw, Pittman S. H. Gibbon, PEco C. Bly, IEAL j M. Beaumont, Westinghouse C. J. Pindzia, Serch Licensing, Bechtel r J. Fuoto, IEAL J. Nurmi, Qatel . R. Borsum, Babcock & Wilcox ' L. Conner, DSA F. T. Stetson, KMC R. B. Leachman, Dept of Anny H. Specter, NYPA T. Ottin, Yankee Atomic R. Wynn, Ohio / Washington News Service J. Hannah, Associated Press N. Wagy, Storer Comm. J. W. Gorman, Storer Comm. O R. Bryek, Detroit Edison L. Benuska, Kinemetrics J. Silberg, Shaw Pittman
)
i i 70 . g3 .; i
INVITED ATTENDEES 311TH ACRS MEETING Thursday, March 13, 1986 J. M. Griesmeyer, Sandia National Labs J. C. Stepp, EPRI T. Stead, Cleveland Electric Illuminating Company (CEI) M. Hayner, CEI . P. Tahwain, Un. South Carolina P. Turner, Westin Geophy K. Pech, C.E.T. E. Levine, Weston Geophysical I. Wall, EPRI J. J. Johnson, NTS/SMA
- 0. K. Herrie, General Electric Company W. H. Fleming, General Electric C. B. Si tt, ANCO R. M. Ginn, CEI J. Smircina, CEI :
A. Uhet, CEI R. Holt, Weston Geophysical G. Leblanc, Weston Geophys K. Campbell, U. S. Geological Survey R. L. Wesson, U. S. Geological Survey G. Sliter, EPRI O M. Edelman, CEI C. Chen, Gilbert Commonweal th J. D. Stevenson, S&A > G. C. K11mkiewicz, Weston Geophysical Corp. L. Bailey, CEI E. Buzzelli, CEI I i I ?
+Y l
i i i PUBLIC ATTENDEES i
- 311TH ACRS MEETING l
Friday, March 14, 1986 3 R. Huston, AIF - i R. H. Buchholz, NUTECH
! E. Fotopoulos, Bechtel } M. Beaumont, Westinghouse ! J. Nurmi, Qatel i i C. Lewe, NUS ,
R. Borsum, B&W - J. Berga, EPRI l P. F. Riehm, KMC L. N. Rib, LNR Associates
- R. E. Schaffsta11, KMC, Inc.
I T. M. Sweeney, Bechtel P. R. Kasten, ORNL , lA L. Walker, SWEC
; J. Recknagel, PSE&G !
W. C. Craign, SWEC ' i S. A. Caspersson, Combustion Engr. J. M. Kendall, YAEC 1 J. F. Davis, NYPA l i f O t . i l ! l I i
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l INVI'E0 ATTENDEES 311TH ACRS MEETING I Friday, March 14, 1986 ' i J. Berga, EPRI l A. Neylan, GA l F. A. Silady, GA A. Kelley, Jr. GCRA l i l i I O l 1 l O gc 5
i APPENCIX II hfh FUTURE AGENDA Af FU1 APRIL ACRS MEETING Advanced Reactors -- Review of conceptual design for advanced 2 hrs LMRs being proposed by DOE Auxiliary Feedwater Systems -- Status report regarding i hr proposed NRC Action Plan to improve auxiliary feedwater system reliability Meeting with Director IE -- Briefing regarding current I hr activities of IE Reliability of Nuclear Plant Components -- Report of ACRS 3/4 hr Subcomittee regarding a generic review of improvements 1 in motor-operated valves, and check valve aging and wear in plant systems McGuire Nuclear Station, Units 1 and 2 -- ACRS coments 2 hrs regarding proposed deletion of the upper head injection system New ACRS Member -- Report of ACRS Selection Comittee 3/4 hr i regarding nomination of candidates for appointment to the ACRS TVA Reorganization -- ACRS subcomittee report regarding I hr technical significance of deficiencies and proposed cor-rective action Davis-Besse Nuclear Power Station, Unit 1 -- ACRS coments 2 hrs regarding proposed restart of the Davis-Besse Nuclear Power Station, Unit 1 following the loss of feedwater incident which occurred on June 9,1985 Reactor Operations -- Report of Subcommittee and represen- 2 hrs tatives of the NRC. Staff regarding recent operating events
, and incidents at nuclear power plants \
Decay Heat Removal Systems -- Subcomittee report regarding i hr status of NRR resolution effort for USI A-45 Safety Goal Policy -- Discussion of plans to implement 4 hrs the proposed NRC Safety Goal Policy I O DE::.: . .
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. . . a wi, Meeting with NRC Comissioners -- to discuss' the following 2 hrs Q- topics:
(a) ACRS resources / workload (Impact of resource limita-tions) (b) Federal Training Academy for Reactor Operators (Senator Moynihan proposal) (c) GESSAR II and safety consideration for future reactors Decay Heat Removal -- Briefing by G. A. Reed of a feed and i hr ' bleed system proposed for the advanced Westinghouse pressurized water reactor Report of the Subcomittee on Human Factors regarding the i hr potential for automating more of the monitoring and control functions of nuclear power plants. The Subcomittee will also review the Human Factors Program Plan and be briefed on the status of Emergency Operating Procedures
- implementation l
i Change in Bylaws regarding interaction and comunication i hr between individual ACRS members and individual Comissioners San Onofre Nuclear Generating Station, Unit 1 -- Comittee defer review of "get-well" plan for restart of San Onofre, to June Unit 1 4 Spent Fuel Storace -- ACRS coments regarding proposed defer changes to 10 CFT Part 72, Licensing Requirements for the storage of spent fuel in an Independent Fuel Storage Facility, including the MRS facility B&W Water Reactors -- Briefing by NRC Staff and B&W Owners defer i Kroup representatives regarding plans for evaluation of to long-term safety of B&W reactors including the implication May of operating experience Systems Interactions -- ACRS coments regarding proposed defer ! NRC generic letter to June 1 NRC Long-Range Plan -- Discuss proposed ACRS outline for defer ' a long range plan to May Containment Venting -- Report of the Subcomittee on defer Regulatory Philosophy, Technology, and Criteria on venting to June of containment during emergency situations O .
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Millstone Nuclear Power Station -- Mil 1 stone 3 PRA defer
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South Texas Project, Units 1 and '2.-- OL review defer to May Safety Characteristics of Future LWRs -- Briefing by defer John J. Taylor, EPRI, regarding EPRI_ study of safety to considerations for future LWRs ' June / July Seabrook Nuclear Plant -- OL review defer indefinitely May ACRS Meetinc - It was agreed that this ineeting will start on Wednescay (5/7) at approximately 1:30 to accommodate a session with representatives of the FRG on Friday (5/9/86) to discuss topics regarding radwaste management and disposal i 1 j h _
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APPENDIX III ACRS SUBCOMMITTEE MEETINGS W6 O Decay Heat Removal Systems, March 18, 1986, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M., Room 1046. The Subcomittee will continue its review of NRR resolution position for USI A-45, " Shutdown Decay Heat Removal Requirements." Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of March 17: i Mr. Ward NONE Mr. Reed DAYS INN Mr. Ebersole CARLYLE Dr. Catton DUPONT PLAZA Mr. Michelson DAYS INN Mr. Davis NONE Human Factors, March 19 and 20, 1986, 1717 H Street, NW, Washington, DC (Schiffgens), 8:30 A.M., Room 1046. The Subcommittee will meet to: (1) examine the potential for automating more of the monitoring and control functions in nuclear power plants to relieve the work burden on plant operators, (2) review 1985 progress on the Human Factors Program Plan, and (3) be briefed on the status of E0P implementation. Attendance by the following is anticipated, and reservations have been made at the hotels ; indicated for the nights of March 18 and 19: Mr. Ward NONE Mr. Remick NONE i Mr. Michelson DAYS INN Mr. Wylie DAYS INN g Mr. Reed DAYS INN Mr. Gimy NONE Joint Emergency Core Cooling Systems / Decay Heat Removal Systems, March 26, 1986, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M., Room 1046. The Subcom-mittee will: (1) continue review of the Duke Power Company's request to delete /- i disable the ECCS-UHI system at McGuire and (2) discuss the proposed NRR resolu-l tion position for Generic Issue 124, " Auxiliary Feedwater System Reliability." Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of March 25: Mr. Ward NONE Dr. Schrock ANTHONY Mr. Ebersole CARLYLE Dr. Sullivan NONE Mr. Reed DAYS INN Dr. The~ofanous ANTHONY Dr. Cattor DUPONT PLAZA Dr. Tien ANTHONY Mr. Davis NONE Ad Hoc Subcommittee on TVA, March 27, 1986, 1717 H Street, NW, Washington, DC (Cappucci), 8:30 A.M., Room 1046. The Subcommittee will discuss TVA Feorganization and related technical and management issues. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of March 26: Mr. Wylie DAYS INN Mr. Reed DAYS INN Dr. Carbon (tent.) CARLYLE Dr. Remick NONE Mr. Ebersole CARLYLE Mr. Ward NONE Mr. Michelson DAYS INN Dr. Hagedorn NONE I O
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I REylSED g R 15 W O Meeting with Comissioners on Quantitative Safety Goals, March 28, 1986, 1717 H Street, NW, Washington, DC (Savio), 10:00 A.M. - 11:30 A.M., Room 1130. The ACRS members listed below will meet with the Commissioners (per T. McCreless' memo of 2/27/86) to discuss' quantitative safety goals. The ACRS attendees are tenta-
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i l tively scheduled to meet in Room 1046 between 8:30 A.M. and 9:45 A.M. to prepare for the meeting with the Comissioners. Attendance by the following is antic-ipated, and reservations have been made at the hotels indicated for the night of March 27: Dr. Carbon CARLYLE Dr. Shewmon NONE Mr. Ebersole CARLYLE Dr. Siess ANTHONY Dr. Kerr LOMBARDY Mr. Ward NONE Dr. Lewis HYATT Mr. Wylie DAYS INN + j Dr. Okrent ANTHONY j Reliability Assurance, April 1, 1986, 1717 H Street, NW, Washington, DC (Major), 8:00 A.M. (Note: Early starting time), Room 1046. The purpose of the meeting will be to discuss the results of EPRI NP-4254, " Improvements in Motor-Operated Valves," dated November 1985. Also to be considered will 1 be NUREG/CR-4380, " Evaluation of the Motor-Operated Valve Analysis and Test System (MOVATS) to Detect Degradation, Incorrect Adjustments, and Other i Abnormalities in Motor Operated Valves," and NUREG/CR-4302, " Aging and I Service Wear of Check Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," both reports were prepared by the Oak Ridge National l Laboratory for the Office of Nuclear Regulatory Research. The Subcomittee will also plan the format for a report which will encompass its considerations a over the past year. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of March 31: Mr. Michelson DAYS INN Mr. Wylie DAYS INN Mr. Ebersole CARLYLE Dr. Lipinski NONE 4' Mr. Reed DAYS INN Dr. Brooks ANTHONY Fort St. Vrain, April 2 8 3, 1986, 16805 WCR 191 (at the Visitors Center of the Fort St. Vrain Power Plant), Platteville, CO (McKinley), 8:30 A.M. The Subcom-i mittee will tour the facility, explore technical problems addressed during the
- recent extended outage, and discuss management changes made as a result of the licensee's independent assessment of management controls. Attendance by the following is anticipated, and reservatinos have been made at the Best Western 3 Centennial Inn (303/776-8700), 3815 Hwy 119 9 I-25, Exit 240, Longmont, C0 for d
the nights of April 1 and 2: Dr. Siess Mr. Ebersole Mr. Ward b i O
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- REVISED Reactor Operations, April 9, 1986, 1717 H Street, NW, Washington, DC (Aldeman),
8:30 A.M., Room 1046. The Subcommittee will review recent significant events. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of April 8: Mr. Ebersole CARLYLE Mr. Reed DAYS INN Dr. Kerr LOMBARDY Dr. Remick NONE Dr. Mark LOMBARDY Mr. Ward NONE Dr. Moeller CARLYLE Mr. Wylie DAYS INN Davis-Besse (Restart), April 9, 1986, 1717 H Street, NW, Washington, DC i (Alderman), 1:30 P.M., Room 1046. The Subcomittee will continue its review of the Davis-Besse restart. Attendance by the following is anticipated, and reser-vations have been mate at the hotels indicated for the night of April 8: Dr. Remick NONE Dr. Siess ANTHONY Mr. Ebersole CARLYLE Mr. Reed DAYS INN 312th ACRS Meeting, April 10-12, 1986, Washington, DC, Room 1046. Waste Management, April 24 and 25, 1986, 1717 H Street, NW, Washington, DC (Merrill), 8:30 A.M., Room 1046. The Subcomittee will review various topics C in the High-level and Low-Level Radioactive Waste Programs. Topics currently identified for review at the April meeting are: (1) Modeling Strategy for HLW performance assessment, (2) Quality Assurance (addressing safety issues of geologic repositories), (3) the NRC LLW program, (4) several research efforts, including international programs and cooperative agreements, results of modeling workshop, setting priorities for HLW research, and LLW shallow land burial (SLB) alternatives; and (5) the Salvaging of Contaminated Smelted Alloys. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the nights of April 23 and 24: Dr. Moeller CARLYLE Dr. Carter NONE Dr. Carbon CARLYLE Dr. Foster NONE Dr. Kerr LOMBARDY Dr. Krauskopf NONE Dr. Mark LOMBARDY Dr. Parker NONE Dr. Remick NONE Dr. Steindler NONE Dr. Shewmon MILLERS Emergency Core Cooling Systems, April 29 and 30, 1986, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M., Room 1046. The Subcomittee will: (1) continue its review of the NRC's proposal to revise 10 CFR 50.46 and Appendix K and (2) continue discussions on defining the themal hydraulic safety issues of most importance that need to be addressed in the future. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the nights of April 28 and 29: Mr. Ward NONE Dr. Catton DUPONT PLAZA On Mr. Ebersole CARLYLE Mr. Schrock ANTHONY V Mr. Michelson DAYS INN Dr. Sullivan NONE Mr. Reed DAYS INN Dr. Theofanous ANTHONY Dr. Tien ANTHONY
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Babcock and Wilcox (B&W) Water Reactors, May 1, 1986, 1717 H Street, NW, Washington, DC (Major), 8:30 A.M. Room 1046. The Subcommittee will consider the B&W Owners Group plans to reassess the long-tenn safety of B&W reactors, including the implications of operating experience on the adequacy of B&W plant designs. The Subcomittee will also be briefed on the NRC Staff's Incident Investigation Team's (IIT) findings related to the 12/26/85 loss of integrated control system power and overcooling transient at the Rancho Seco nuclear power , plant. Lodging will be announced later. Attendance by the following is anticipated: ' Mr. Wylie Mr. Reed Mr. Michelson Mr. Ward Mr. Ebersole Class 9 (Severe) Accidents, May 1 and 2, 1986, AMFAC Hotel, Albuquerque, NM, , (Houston),8:30 A.M. The Subcomittee will review rebaselining studies for five i reference plants; part of NUREG-1150 study; also tour containment model for severe accident performance studies. Attendance by the following is antici- i pated, and reservations have been made at the AMFAC Hotel for the nights of June 30 and May 1: , i Dr. Kerr O Dr. Carbon Dr. Moeller MmWest$ Mr. Bender Dr. Catton ! Dr. Okrent Dr. Corradini , Dr. Shewmon Mr. Davis ! Dr. Siess ; Scram Systems Reliability, May 6, 1986, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M., Room 1046. The Subcomittee will discuss the ATWS ' Rule implementation effort. Attendance by the following is anticipated, j and reservations have been made at the hotels indicated for the night of , May 5: i Dr. Kerr LOMBARDY Mr. Word NONE Mr. Ebersole CARLYLE Mr. Wylie DAYS INN Dr. Lewis HYATT Mr. Davis NONE ' Dr. Okrent ANTHONY Dr. Lipinski NONE i i Long Range Plan for the NRC, May 6, 1986, 1717 H Street, NW, Washington, DC i (Major), 1:00 P.M., Room 1167. The Subcomittee will review the proposed NRC ! Five Year Plan and prepare to address Comittee coments to the Comission. ; Attendance by the following is anticipated, and reservations have been made l at the hotels indicated for the night of May 5 i Dr. Carbon LOMBARDY Dr. Remick NONE Dr. Lewis HYATT Mr. Wylie DAYS INN Dr. Moeller CARLYLE ry 0
4 MAR 151986
. REYlSED Safety Research Program, May 7, 1986, 1717 H Street, NW, Washington, DC ~
(Duraiswamy), 4:30 A.M., Room 1046. The Subcommittee will discuss the proposed NRC Safety Research Program and Budget for FY 1988 and 1989, and gather information for use by the ACRS in its preparation of the annual report to the Comission on the related matter. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of May 6: Dr. Siess AN1FONY Dr. Moeller CARLYLE Dr. Carbo, CARLYLE Dr. Okrent ANTHONY Dr. Kerr LOMBARDY Dr.Remick(tent.) NONE Dr. Mark LOMBARDY Dr.Shewmon(P.M.) MILLERS (May 7) Mr. Michelson DAYS INN Mr. Ward NONE-313th ACRS Meeting, May 3-10, 1986, Washington, DC, Room 1046. Westinghouse Reactor Plants, May 20, 1986 (tentative), 1717 H Street, NW, Washingtoa, DC (Cappucci), 8:30 A.M., Room 1046. The Subcomittee will continue discussions and coment on NRC Staff actions taken with respect to the SONGS-1 waterhamer/ loss of AC power event phich occurred on November 21, 1985. This is a follow-up meeting to the February 12, 1986 Subcomittee meeting on the same subject. Lodging will be announced later. Attendance by the following is anticipated: Mr. Ebersole Dr. Shewmon Mr. Etherington Dr. Siess Dr. Kerr (tent.) Mr. Ward (tent.) Mr. Michelson Mr. Wylie Mr. Reed (tent.) Dr. Catton Safety Research Program, June 4, 1986, 1717 H Street, NW, Washington, DC (Duraiswamy), 8:30 A.M., Room 1046. The Subcomittee will continue its dis-cussion on the proposed NRC Safety Research Program and Budget for FY 1988 and 1989. It will discuss also a Draft ACRS report to the Comission on this matter. Lodging will be announced later. Attendance by the following is anticipated: Dr. Siess Dr. Moeller Dr. Carbon Dr. Okrent Dr. Kerr Dr. Remick Dr. Mark Dr. Shewmon 1 Mr. Michelson Mr. Ward 314th ACRS Meeting, June 5-7, 1986, Washington, DC, Room 1046. ; Wingspread International Conference, October 19-23, 1986, Racine, WI f ' (McCreless). Representatives from the ACRS, RSK, GPR, and Japan will ! exchange information on nuclear reactor safety. O r !Y l
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II % O AEVED V South Texas Units 1 and 2, Date to be detennined (April), Bay City, TX (El-Zeftawy). The Subcommittee will review Houston Lighting and Power Company's application for an operating license. Attendance by the following is anticipated: Dr. Mark Dr. Lewis Dr. Carbon (tent.) Mr. Michelson Mr. Ebersole Dr. Siess Metal Components, Date to be detennined (April /May, tentative), Pittsburch, PA or Charlotte, NC (Igne). The Subcomittee will review the status of FDE of cast stainless steel. Attendance by the following is anticipated: Dr. Shewmon Dr. Okrent Mr. Etherington Mr. Ward Dr. Lewis Dr. Thompson Mr. Michelson Mr. Shack Ad Hoc Subcomittee on TVA, Date to be determined (April /May), Washington, DC (Cappucci). The Subcomittee will review the Staff evaluation of the TVA Corporate Nuclear Performance Plan and issues related to the restart of Sequoyah. Attendance by the following is anticipated: Mr. Wylie fir. Reed Dr. Carbon Dr. Remick Mr. Ebersole Mr. Ward (tent.) Mr. Michelson Dr. Hagedorn Decay Heat Removal Systems, Date to be detennined (May/ June), Washington, DC (Boehnert). The Subcomittee will review NRR's Action Plan to address con-cerns with the reliability of certain plants' AFW systems. Attendance by the following is anticipated: Mr. Ward Mr. Reed Mr. Ebersole Dr. Catton Mr. Etherington Mr. Davis Mr. Michelson Regulatory Policies and Practice 3, Date to be determined (May/ June), Washington,
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DC (Cappucci). The Subcomittee will review backfitting policy implementation. Special focus on industry initiated modifications. Attendance by the following is anticipated: Dr. Lewis Dr. Okrent Mr. Michelson Dr. Remick Dr. Moeller Mr. Wylie O p - /S
MAR 15 W / REVISED Emergency Core Cooling Systems, Date to be detemined (June / July), Washington, DC (Boehnert). The Subcommittee will review General Electric's application Tor use of the SAFER /COREC00L ECCS Code on BWR nonjet pump plants. Attendance by the following is anticipated: Mr. Ward Dr. Catton Mr. Ebersole Mr. Schrock Mr. Etherington Dr. Sullivan Mr. Michelson Dr. Theofanous Mr.- Reed Dr. Tien Metal Components, Date to be determined (June / July), Richland, WA (Igne). The Subcommittee will visit and review steam generator, degraded piping, and NDE facilities and programs. Attendance by the following is anticipated: Dr. Shewmon Mr. Bender Mr. Etherington Mr. Dillon Dr. Lewis Mr. Kassner Mr. Michelson Mr. Rodabaugh Dr. Okrent Mr. Thompson Mr. Ward h) Structural Engineering, Date to be determined (June / July), Albuquerque, NM (Igne). The Subcommittee will visit and review containment integrity and Catego-ry I structures facilities and programs. Attendance by the following is antic-ipated: Dr. Siess Dr. Shewmon
- Mr. Ebersole Mr. Bender Dr. Kerr Dr. Pickel Dr. Okrent Transportation of Radioactive Materials, Date to be detemined (June / July),
Washington, DC (Duraiswa ny). The Subcommittee will discuss the following related to the Transportation System being developed by the Department of Energy (DOE) as required by the Nuclear Waste Policy Act of 1982: (1) Current status and schedule for the development of the Transportation System; (2) Roles of, and the coordination between, DOE and NRC; (3) Safety issues that are foreseen by the NRC Staff; and (4) Criteria to be used by the NRC Staff in the review and certification of the Transportation System. Attendance by the following is anticipated: Dr. Siess Dr. Moeller Mr. Ebersole Dr. Shewmon Dr. Mark g- /9 , w --- e -w. -- w w r,
3AR 151986 REVISED Seabrook Units 1 & 2, Date to be determined (late sumer/early fall), Washington, DC (Major). The Subcomitt.:e will review the application for a full power oilerating license for Seabrook 1 & 2 . Attendance by the following is anticipated: Dr. Kerr Dr. Moeller Dr. Lewis Mr. Michelson Westinghouse Reactor Plants, Date to be detennined, Washington, DC (Cappucci). The Subcommittee will begin the PDA review of the Westinghouse Advanced Pres-surized Water Reactor (RESAR SP/90). Attendance by the following is anticipated: Mr. Ebersole Dr. Shewmon Mr. Etherington Mr. Ward
!4r. Michelson Mr. Wylie Dr. Siess Reliability and Probabilistic Assessment, Date and location to be determined (Savio). The Subcomittee will review the probabilistic risk assessment for Millstone 3. Attendance by the following is anticipated:
Dr. Okrent Mr. Michelson Dr. Kerr Dr. Siess Mr. Ebersole Mr. Ward Dr. Lewis Mr. Wylie Dr. Mark l l l I I i O A?-/7 e 5
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MARCH 18, 1986 DECAY HEAT REMOVAL SYSTEMS (BOEHNERT) Ward, Ebersole, Michelson, Reed Cons.: Catton, Davis PURPOSE: To continue the review of NRR resolution position for USI A-45, " Shutdown Decay Heat Removal Requirements." LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Status of review to be provided to ACRS; at April meeting. What will be done at this meeting? Review at least 2 or more of the Plant Analysis Reports that will form basis of NRR resolution position. What would be the consequence of postponing this meeting? Loss of timely information on A-45 resolution status. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Draft Plant Analysis Reports for the Cooper and Turkey Point plants (will be provided when received from NRR)
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SCHELULE OF ACRS SUBCOMMITTEE MEETING N DATE SUBCOMMITTEE MEETING STAFF ENGR & MEMBERS MARCH 19 & 20, 1986 HUMAN FACTORS (SCHIFFGENS) Ward, 4 Michelson, Reed, Remick, Wylie Cons.: Gimy PURPOSE: The Subcommittee will meet to:- (1) examine the potential for automating more of theoperators, plant monitoring2) (and review control functionson 1985 progress in the NPPs to relieve Human theProgram Factors work burden Plan, on
.and (3) be briefed on the status of the E0P implementation.
LOCATION: WASHINGTON, DC BACKGROUND: , What action is requested; by what date is it needed? N/A What will be done at this meeting? See Purpose. What would be the consequence of postponing this meeting? None PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: To be provided on a timely basis. l l l O g -/7 _
l 1 . l SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MARCH 26, 1986 JOINT ECCS AND (BOEHNERT) Ward, DECAY HEAT REMOVAL SYSTEMS Ebersole, Reed Cons.: Catton, Davis, Schrock, Sullivan, Theofanous, Tien PURPOSE: The Subcommittee will: (1) review Duke Power Compan 's request to delete / disable the ECCS-UHI system at McGuire and 2(y) discuss the proposed resolution position for Generic Issues 124. " Auxiliary Feedwater System Rel iabili ty. " LOCATION: WASHINGTON, DC BACKGROUND: What action is reauested; by what date is it needed?- O Review of licensing action on Duke Power's request; ASAP after 3/14/86. Review of Generic Issues to support submittal to CRGR in April 1986. What will be done at this meeting? See Purpose. What would be the consequence of postponing this meeting? Loss of timely review of deletion / disabling request. Loss of timely review of Generic Issue resolution proposal. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. SER supporting deletion / disabling request.
- 2. Westinghouse's and Duke Power's submittals with bases for request.
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I l SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MARCH 27, 1986 AD H0C SUBCOMMITTEE ON TVA (CAPPUCCI) Wylie, , (tentative) Carbon (tent.), Ebersole, Michelson, Reed, Remick (tent.), Ward Cons.: Hagedorn PURPOSE: The Subcommittee will discuss the TVA reorganization and related technical matters. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? None What will be done at this meeting? Review TVA management issues. What would be the consecuence of postponing this meeting? Delay of ACRS review and comments to the Connission. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Memo dated 1/22/86, fm A. Cappucci to C. Wylie,
Subject:
Project Review Plan for TVA Management Issue.
- 2. Memo dated 2/4/86 fm A. Cappucci to C. Wylie,
Subject:
SECY-86-1A, Status of Staff Actions Regarding TVA.
- 3. Memo dated 2/5/86 fm A. Cappucci to C. Wylie,
Subject:
TVA Suggested Schedule of Topics for the March 11, 1986 Subcommittee Meeting.
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING i l l DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS APRIL 1, 1986 RELIABILITY ASSURANCE (MAJOR) Michelson, (8:00 A.M.) (VALVES) Ebersole, Reed, Wylie Cons.: W. Lipinski Invited Presenter: Boyd Brooks l PURPOSE: To discuss results of EPRI NP-4254, " Improvements in Motor Operated Valves," I dated November 1985. Update of Oak Ridge work on MOV as well as presentation on check valves is expected. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Subcommittee will prepare a report on its investigation into motor operated valves; C will attempt to produce results in the near future. What will be done at this meeting? See Purpose. What would be the consequence of postponino this meeting? Postponement of final report. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: The following have been distributed:
- 1. EPRI NP-4254, " Improvements in Motor Operated Valves," dated November 1985 (now available).
- 2. NUREG/CR-4302, " Aging and Service Wear of Check Valves Used in Engineered Safety-Features Systems of Nuclear Power Plants," December 1985.
- 3. NUREG/CR-4380, " Evaluation of the Motor-Operated Valve Analysis and Test System (M0 VATS) to Detect Degradation, Incorrect Adjustments, and Other Abnormalities in Motor-Operated Valves," January 1986.
- 4. Schedule issued March 3, 1986.
- 5. Status Report issued March 7, 1986.
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/~'\ SCHEDULE OF ACRS SUBCOMMITTEE MEETING V
DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS APRIL 2 & 3, 1986 FORT ST VRAIN (McKINLEY)Siess, Ebersole, Ward PURPOSE: To review helium circulatory bolting failures, PCRV tendon corrosion, control rod drive failures, control rod cable replacement, reserve shutdown material replacement, PSC management improvement actions, equipment qualification, and recent operating experience. LOCATION: PLATTEVILLE, C0 BACKGROUND: What action is requested; by what date is it needed? No action requested of ACRS,; ACRS would be performing its oversight function of monitoring major maintenance, operation, QA, and management functions. What will be done at this meeting? (see above) What would be the consequence of postponing this meeting? Appearance of ACRS disinterest in major management problems at a unique operating (?) power reactor. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Memo from J. C. McKinley to C. P. Siess dated March 27, 1985,
Subject:
" Management at Fort St. Vrain."
- 2. Tentative Schedule issued 3/15/86.
- 3. Status Report to be issued 3/19/86.
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS APRIL 9, 1986 REACTOR OPERATIONS (ALDERMAN) Ebersole, Kerr, Mark, Moeller, Reed, Remick, Ward, Wylie PURPOSE: To review recent significant events. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Review significant events and report to full Comittee. What will be done at this meeting? Review significant events and select a limited number for full Comittee discussion. What would be the consequence of postponing this meeting? Failure of Comittee to be aware of significant events. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Status report and agenda will be prepared prior to meeting.
1 I l g>Y
1 l SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS APRIL 9, 1986 DAVIS-BESSE (ALDERMAN) Remick, (1:30 P.M.) (RESTART) Ebersole, Reed, Siess PURPOSk: The Subcommittee will continue its review of the Davis-Besse restart. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Possible full Committee letter. Not in critical path. What will be done at this meeting? Continue and possibly conclude review of Davis-Besse restart. What would be the consequence of postponing this meeting? NRC Staff would like ACRS comments. Would not affect restart. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Updated SER portions expected about March.
3
)
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS APRIL 24 & 25, 1986 . WASTE MANAGEMENT (MERRILL) Moeller, Carbon, Kerr, Mark, Remick, Shewmon Cons.: Carter, Foster, Krauskopf, Parker, Steindler PURPOSE: To review several High-level and Low-level Radioactive Waste topics: (1) Modeling Strategy for HLW performance assessments; (2) Quality Assurance, addressing safety issues of geologic repositories; (3) the NRC LLW program; (4) several research efforts, including international programs and cooperative agreements, results of modeling workshop, and setting priorities for HLW research, and LLW shallow land burial (SLB) alternatives; and (5) the Salvaging of Contaminated Smelted Alloys. LOCATION: WASHINGTON, DC BACKGROUND: What action is reouested; by what date is it needed? NMSS Division of Waste Management and RES Waste Management Branch have requested ACRS oversight review on the above named topics. What will be done at this meeting? See Purpose. What would be the consecuence of postponing this meeting? None, except for Item (5) above, which Comissioner Bernthal requested the ACRS to review (see Item 3 below). PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Modeling Strategy, Quality Assurance and LLW documents to be provided by NMSS by 3/24/86.
- 2. Documents on research programs to be provided by RES by 3/24/86.
- 3. Staff Requirements, memo, S. Chilk to V. Stello, dtd.1/30/86,
Subject:
SECY-85-373, " Denial of DOE Request for Exemption to Pemit Salvaging Contaminated Smelted Alloys," dtd 11/25/85. 4 Memo for Comissioners fm J. Zerbe, OPE,
Subject:
OPE Coments on SECY-85-373.
- 5. NUREG-0518, Draft Environmental Statement, dtd. October 1980.
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS APRIL 29 & 30, 1986 ECCS (B0EHNERT) Ward, Ebersole, Michelson, Reed Cons.: Catton, Schrock, Sullivan, Theofanous, Tien PURPOSE: To: (1) continue the review of the NRC's proposal to revise 10 CFR 50.46 and Appendix K and (2) continue discussions on defining the thermal hydraulic safety issues of most importance that need to be addressed in the future. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Review of proposed revision to Appendix K in order to support its submittal to CRGR in May 1986. What will be done at this meeting? See Purpose above. What would be the consequence of postponing this meeting? See action requested. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Revised 10 CFR 50.46 and Appendix K.
- 2. Associated Regulatory Guide.
- 3. Technical support paper.
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MAY 1, 1986 B&W WATER REACTORS (MAJOR) Wylie, Ebersole, Michelson,
. Reed, Ward PURPOSE: Consideration of the B&W Owners Group plans to reassess the long-tenn safety of B&W reactors, including the implications of operating experience on the adequacy of B&W plant designs. The Subcomittee will also be briefed on the NRC Staff's Incident Investigation Team's (IIT) findings related to the 12/26/85 loss of integrated control system power and overcooling transient at the Rancho Seco nuclear power plant.
LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? This will be an initial briefing on the B&W Owners Group reassessment of the long-term Os safety of their plants. Comittee coments on the course the review is taking might be appropriate. What will be done at this meeting? See Purpose. What would be the consequence of postponing this meeting?
)
Loss of timeliness. Comittee would become out of phase with NRC Staff and B&W Owners l Group activities. ! PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. NUREG-1195, Loss of Integrated Control System Power Overcooling Transient at Rancho !
Rancho Seco on December 26,1985 (dist. to Comittee on 2/26/86).
- 2. Awaiting a course of action regarding the reassessment of B&W plants by the Owners Group.
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O SCHEDULE OF ACRS SUBCOMMITTEE MEETING V DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MAY 1 & 2, 1986 CLASS 9 ACCIDENTS . (HOUSTON) Kerr, Carbon, (SEVERE ACCIDENT PHENOMENA) Moeller, Okrent, Shewnon, Siess, Ward Cons.: Bender, Catton, Corradini, Davis PURPOSE: To review risk rebaselining studies for five reference plants; part of NUREG-1150 study; also four containment model for severe accident performance studies. LOCATION: ALBUQUERQUE, NM MAY 1 - AMFAC Hotel MAY 2 - Bus tour at Sandia (A.M. only) BACKGROUND: What action is requested; by what date is it needed? Revi w baseline risk calcualtions for five refence plants (Surry, Peach Bottom, Sequoyah, Zion, and Grand Gulf) four containment model. NRR is working on an What will be done at this meeting? Determine the adequacy of the risk calculations. What would be the consequence of postponing this meeting? Could delay the issuance of NUREG-1150 if Subcommittee coments require extensive revision. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: RES will provide some preliminary documents. O
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING O DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MAY 6, 1986 SCRAM SYSTEMS (80EHNERT) Kerr, RELIABILITY Ebersole, Lewis, Okrent, Ward, Wylie Cons.: Davis, Lipinski PURPOSE: To discuss the status of ATWS Rule implementation effort. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? N/A What will be done at this meeting? See Purpose What would be the consequence of postponing this meeting? No significant consequences. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: To be provided on a timely basis, g .3d 1
i SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MAY 6, 1986 LONG RANGE PLAN (MAJOR) Carbon, Lewis, (1:00 P.M.) Moeller, Remick, Wylie PURPOSE: To review the proposed NRC Five Year Plan and prepare to address Committee comments to the Commission. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Estimate is June full Connittee; when asked by Commission. What will be done at this meeting? Exchange information with Staff, consider ACRS comments on Staff's version of a Long Range Plan. What would be the consequence of postponing this meeting? {} Loss of timeliness. 1 PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Five Year Plan has been distributed. -
- 2. Dr. Carbon's Guidelines for a Long Range Plan.
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MAY 7, 1986 SAFETY RESEARCH PROGRAM (DURAISWAMY) Siess, Carbon, Kerr, Mark, Michelson, Moeller, Okrent,Remick(tent.), Shewmon (P.M.), Ward PURPOSE: To discuss the proposed NRC Safety Research Program and Budget for FY 1988 and 1989, and gather infonnation for use by the ACRS in its preparation of the annual report to the Comission on the related matter. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? The ACRS needs to provide its comments to the Comission, by June 1986, on the proposed NRC Safety Research Program and Budget for FY 1988 and 1989. What will be done at this meeting? 2 See Purpose. What would be the consequence of postponing this meeting? Schedule for completion of the report might be delayed. I PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: i
- 1. Proposed Safety Research Program and Budget for FY 1988 and 1989 (expected to be made available to the ACRS during April 1986).
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MAY 20, 1986 WESTINGHOUSE REACTOR PLANTS (CAPPUCCI) Ebersole. (Tentative) Kerr(tent), Etherington, Michelson, Okrent, Reed (tent), Shewmon, Siess, Ward (tent),Wylie Cons.: Catton PURPOSE: To continue discussions and coment on NRC Staff actions taken with respect to the SONGS-1 waterhamer/ loss of AC power event. Follow-up Subcommittee meeting to February 12, 1986 meeting on same subject. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? None requested. Subcommittee Chairman's action. What will be done at this meeting? Complete review of the SONG-1 event and prepare coments for full Comittee action. What would be the consequence of postponing this meeting? Input to NRC Staff restart decisions will be untimely. ! PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. NRC Staff restart SER (to be published on May 20,1986).
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS JUNE 4, 1986 SAFETY RESEARCH PROGRAM (DURAISWAMY)Siess, Carbon, Kerr, Mark, Michelson, Moeller, Okrent, Remick, Shewmon, Ward PURPOSE: To continue discussion on the proposed NRC Safety Research Program and Budget for FY 1988 and 1989. Also, to discuss a Draft ACRS report to the Commission on this matter. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? The ACRS needs to provide its comments to the Comission, by June 1986, on the proposed NRC Safety Research Program and Budget for FY 1988 and 1989. What will be done at this meeting? See Purpose. What would be the consequence of postponing this meeting? Schedule for completion of the report might be delayed. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Proposed Safety Research Program and Budget for FY 1988 and 1989 (expected to be made available to the ACRS during April 1986).
- 2. Budget Review Group mark (if available).
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/' SCHEDULE OF ACRS SUBCOMMITTEE MEETING V DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS OCTOBER 19-23, 1986 WINGSPREAD INTERNATIONAL (MCCRELESS)ACRS CONFERENCE Members PURPOSE: To exchange nuclear reactor safety information with the RSK, GPR, and representatives from Japan. LOCATION: RACINE, WI BACKGROUND: l What action is requested; by what date is it needed? N/A What will be done at this meeting? l See Purpose. What would be the consequence of postponing this meeting? Wingspread is available only on certain days. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: Exchange papers will be prepared after agenda planning is completed in April. O l W
1 SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED SOUTH TEXAS 1 & 2 (EL-ZEFTAWY) Mark, (APRIL) Carbon (tent.), Ebersole, Lewis, Michelson, Siess PURPOSE: To review Houston Lighting and Power Company's application for an OL. LOCATION: BAY CITY, TX BACKGROUND: What action is requested; by what date is it needed? Issue an ACRS " full power" OL letter; at the May 1986 ACRS meeting. What will be done at this meeting? Subcomittee OL review in time for Comittee consideration at the May 1986 ACRS meeting. What would be the consequence of postponing this meeting? Possible delay of South Texas full power operation. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: To be provided later. The NRC Staff anticipates publishing the SER on, or about, end of March. O y 3 c- ,
SCHEDULE OF ACRS SUBCOMMITTEE MEETING r DATE SUBCOMMITTEE MEETING' STAFF ENGR. & MEMBERS TO BE DETERMINED METAL COMPONENTS (IGNE) Shewmon, (APRIL /MAY) - Etherington, Lewis, (Tentative) , Michelson, 0krent, Ward Cons.: B. Thompson, W. Shack PURPOSE: To review with the NRC Staff and industry'the status of NDE of cast stainless steel. ~ LOCATION: Pittsburgh, PA (Westinghouse) or Charlotte, NC (EPRI-NDE Center) A Subcommittee meeting at either of these' locations is necessary because demonstrations of actual NDE equipment and procedures on large specimens are planned. BACKGROUND: What action is requested; by what date is it r.eeded?
/
The Subconnittee requested that it keep abreast' of the progress of NDE of cast L stainless steel materials. Early spring,1986 -- before GDC-4 broad scope rule is promulgated. What will be done at this meeting? Review the reliability of NDE to detect and size flaws in cast stainless steel materials. , What would be the consequence of postponing this meeting? Nothing. Will not be able to intelligently comment on GDC-4 broad scope rule regarding the detection and sizing of flaws in cast stainless steel. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Status Report on NDE of cast stainless steel, WOG Materials Subcommittee (Tentative Schedule, end March /early April). ,
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED AD HOC TVA (CAPPUCCI) Wylie, (APRIL /MAY) Carbon, Ebersole. Michelson, Reed, Remick, Ward (tent) Cons.: Hagedorn PURPOSE: To review the TVA Corporate Nuclear Plan and the NRC Staff evaluation of the Plan. Also issues' related to the restart of Sequoyah. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? None requested by NRC. Full Comittee assignment. What will be done at this meeting? Review TVA "get will" plans and NRC Staff evaluation. Prepare ACRS coments to Comission. What would be the consequence of postponing this meeting? No coments to the Comission by May/ June full Comittee meeting. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. TVA Nuclear Performance Plan, Volume 1, Corporated dated 3/10/86.
- 2. Staff SER (if available).
l J
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l l l l SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED DECAY HEAT REMOVAL SYSTEMS (BOEHNERT) Ward, (MAY/ JUNE) Ebersole, Etherington, Michelson, Reed Cons.: Catton, Davis PURPOSE: The Subcommittee will review NRR's Action Plan to address concerns with the reliability of certain plants' AFW systems. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Meeting needs to be scheduled in coordination with submittal of Action Plan to CRGR. What will be done at this meeting? See Purpose. What would be the consequence of postponing this meeting? See above. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Action Plan package to be provided when available, i
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A SCHEDULE OF ACRS SUBCOMMITTEE MEETING U DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED REGULATORY POLICIES AND (CAPPUCCI) Lewis, (MAY/ JUNE) PRACTICES Michelson, Moeller, Okrent, Remick, Wylie PURPOSE: To discuss backfitting policy implementation with specific focus on utility initiated modifications. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? None requested by NRC. Full Comittee assignment. What will be done at this meeting? Review MC-0514 on Plant Specific Backfitting, use of 50.59 analyses with respect to backfitting policy, and prepare coments for possible full Comittee actions. What would be the consecuence of postponing this meeting? None PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Memo, Cappucci to Lewis, 3/5/86, Implementation of Manual Chapter 0514, Management of Plant-Specific Backfitting of Operating Power Reactors.
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(3 SCHEDULE OF ACRS SUBCOMMITTEE MEETING V DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED ECCS (80EHNERT) Ward, (JUNE / JULY) Ebersole, Etherington, Michelson, Reed Cons.: Catton, Schrock, Sullivan, Theofanous, Tien PURPOSE: To review GE's application for use of the SAFER /COREC00L ECCS code on BWR NJP plants. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Support NRC review of SER on SAFER /COREC00L; SER requested by September 1986. What will be done at this meeting? See Purpose. What would be the consequence of postponing this meeting? Possibly lose coordination with Staff review. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. GE Licensing Topical Report on SAFER /COREC00L (will be provided when issued - early June ?).
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED METAL COMPONENTS (IGNE)Shewmon, (JUNE / JULY) Etherington, Lewis, Michelson, Okrent, Ward Cons.: Bender, Dillon, Kassner, Rodabaugh, Thompson PURPOSE: To visit and review steam generator, degraded piping, and NDE facilities and programs. LOCATION: RICHLAND, WA (PNL) BACKGROUND: What action is requested; by what date is it needed? RES requested that we visit and review the above programs during sumer of 1986. What will be done at this meeting? See Purpose. What would be the consequence of postponing this meeting? None PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: To be provided on a timely basis. O d- P
SCHEDULE OF ACRS SUBCOMMITTEE MEETING V DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED STRUCTURAL ENGINEERING (IGNE) Siess, Ebersole, (JUNE / JULY) Kerr, Okrent, Shewmon Cons.: Bender, Pickel
- PURPOSE
- To visit and review containment integrity and Category I structures facilities and programs.
LOCATION: ALBUQUERQUE,NM BACKGROUND: What action is requested; by what date is it needed? RES requested that we visit and review the above programs during summer of 1986. What will be done at this meeting? See Purpose. What would be the consequence of postponing this meeting? None , PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: To be provided on a timely basis. U
SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED TRANSPORTATION OF (DURAISWAMY) Siess, (JUNE / JULY) RADI0 ACTIVE MATERIALS Ebersole, Mark, Moeller, Shewmon PURPOSE: To discuss the following related to the Transporatation System being developed by DOE as required by the Nuclear Waste Policy Act of 1982:
- Current status and schedule for the development of the Transportation System.
- Roles of, and the coordination between, DOE and NRC.
- Safety issues that are foreseen by the NRC Staff.
Criteria to be used by the NRC Staff in its review and certification of the Transportation System. LOCATION: WASHINGTON, DC BACKGROUND: O What action is requested; by what date is it needed? U Comission requested that the ACRS review matters associated with the Nuclear Waste Policy Act. No definite date for action. What will be done at this meeting? See Purpose What would be the consequence of postponing this meeting? None PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: O
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b SCHEDULE OF ACRS SUBCOMMITTEE MEETING V DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED SEABROOK UNITS 1 & 2 (MAJOR) Kerr, Lewis, (late summer /early fall) Moeller, Michelson PURPOSE: Full power approval for the Seabrook plant. Currently ACRS has written a 5% power letter (4/19/85). Outstanding issues include emergency planning and Staff review of a probabilistic safety assessment performed for the Seabrook plant. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Conclusion of ACRS OL review. Prior to operation above 5% power. What will be done at this meeting? Review outstanding issues and consider this plant for a full power ACRS letter. Conclude OL review. What would be the consequence of postponing this meeting? Postponing this meeting could impact plant operations. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. SER on Emergency Planning and review of the PRA expected by late summer /early fall.
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SCHEDULE OF ACRS SUBCOPNITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED WESTINGHOUSE REACTOR (CAPPUCCI) Ebersole, PLANTS Etherington, Michelson, (CLOSED) Shewmon, Siess, Ward, Wylie . Cons.: Davis PURPOSE: To begin PDA review of Westinghouse Advanced PWR (RESAR SP/90). LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? ACRS letter on PDA approval by 11/86. , What will be done at this meeting? , Begin reviewing design modules. What would be the consequence of postponing this meeting? Delay in the completion of ACRS PDA review. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. RESAR SP/90 Standard Plant Design (50-601).
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SCHEDULE OF ACRS SUBCOP9fITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED RELIABILITY AND PROBABILISTIC (SAVIO) Okrent, Kerr, ASSESSMENT Ebersole, Lewis, Mark, Michelson, Siess, Ward, Wylie PURPOSE: To review the PRA for Millstone 3 (not an OL critical path item). LOCATION: To be determined BACKGROUND: What action is requested; by what date is it needed? Review of the Millstone 3 PRA; th.e meeting is to be scheduled after the completion of the NRC Staff's review of the PRA (estimated to be by the end of May 1985). There is no ACRS action date. What will be done at this meeting? Review of the Millstone 3 PRA for information. What would be the consecuence of postponing this meeting? ACRS has stated that this review need not be completed prior to full power operation. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Millstone 3 PRA (d.istributed).
- 2. NRC Staff report on the results of the NRC/LLNL review of the Millstone 3 PRA (expected by the end of May 1985).
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APPENDIX IV NRR PRESENTATION NRR STAFF PRESEB g . ACR ON PERRY EARTHQUAKE
SUBJECT:
PERRY EARTHQUAKE DATE: MARCH 12 - 13, 1986 PRESENTER: J. STEFANO, P. SOBEL, A. LEE 1; PRESENT5R'S TITLE / BRANCH /DIV: DIVISION OF BWR LICENSING - l PRESENTER'S NRC TEL. NO.: J. STEFANO, X29473 SUBCOMMITTEE: EXTERME EXTERNAL PHENOMENA O 6 'l 7
i J. STEFANO - SUMMARIZE EARTHOUAKE EVENT, PLANT RESPONSE, STAFF ACTION / FINDINGS, POST-LICENSE CONFIRMATORY ISSUES P. SOBEL - 3 GE0 LOGY / SEISM 0 LOGY FINDINGS AND OBJECTIVE OF RELATED CONFIRMATORY ITEMS i TO BE PURSUED POST-LICENSE A. LEE - EQUIPMENT QUALIFICATIONS AND RELATED ITEMS STRUCTURAL FINDINGS AND OBJECTIVE OF RE O C NFIRMATORY ITEMS POST-LICENSE J. STEFANO - CONCLUDING REMARKS i l I O p-ss i
.,,______,.,...,.---r----- ~ ~ ~ ' ' ' ' ' ' ' " ' ' ' ' ' " ^ ~ ' ' ' ' '
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STAFF FINDS (SSER N0, 9 - MARCH 5, 1986) NO OBSERVED SIGNIFICANT DAMAGE TO PLANT
. FROM EARTHOUAKE -
DESIGN OF FLANT STRUCTURES AND EQUIPMENT HAVE SUBSTANTIAL MARGINS OF SAFETY RE 1 LOADS / STRESSES INDUCED BY EARTHOUAKE NO BASIS AT THIS TIME TO REVISE SEISMIC DESIGN' BASES FOR PLANT l SEVERAL CONFIRMATORY MATTERS NEED FURTHER ANALYSIS AND REVIEWS BY CEI/ STAFF (LISTED !() BELOW) f i I O
+ Sl
i 1 O 1 _ CONFIRMATORY ITEMS MOST COMPLETED PRIOR'TO FULL POWER LICENSE i (ALL EXCEPT LONG TERM GENERIC ITEMS) FAULT PLANE. SOLUTIONS INJECTION WELLS ! FAULTS AT PLANT SITE 1 ENRICHED'HIGH FREQUENCY I HIGH FREQ, SHORT DURATION EARTHOUAKES i O RELOCATION OF SEISMIC INSTRUMENTS '3 - SEISMIC QUALIFICATION OF EQUIPMENT i I f 1-4 1 J i !O
, I r O ! CONCLUSION PLANT SEISMIC DESIGN ADEQUACY HAS BEEN REAF CONFIRMATORY WORK NOT EXPECTED TO RESULT IN A DESIGN CHANGES ~ 5% LICENSE ON MARCH 14 i O O g s3
F e () i JANUARY 31, 1986 EARTHOUAKE 1
! LOCATION - 41.65'N, 81.16*W .
i o, l ABOUT 10 MILES SOUTH OF PERRY i * ! MAXIMUM INTENSITY VI (MM) t
* 't AFTERSHOCKS - ABOUT TEN
- 1 TO 6 MI DEEP 1
- LARGEST WAS MAGNITUDE 2.4
- FREE-FIELD RECORDINGS I' i
INPLANT RECORDINGS OF MAIN SHOCK SHOW EXCEEDANCES {
! SSE AND OBE AT HIGH FREQUENCIES (ABOVE 15 Hz). 6
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6 ~ STAFF REVIEW (SER-1982) ' MOST SEISMIC CATEGORY I STRUCTURES FOUNDED ON DEVONIAN SHALE BEDR,0CK. NO CAPABLE FAULTS IN THE SITE REGION. FAULTS IN THE INTAKE AND DISCHARGE TUNNELS AND IN THE PLANT SITE EXCAVATIONS ARE NOT CAPABLE, PERRY SITE IN CENTRAL STABLE REGION. SITE SPECIFIC RESPONSE SPECTRA FOR A NEARBY MAGNIT 5.3 EVENT COMPARED TO THE SSE (0.15G, RG 1.60). O O g sf
1 l O CONFIRMATORY ISSUES LOCATION OF MAIN SHOCK AND AFTERSHOCKS. FAULT PLANE SOLUTIONS AND STRESS DIRECTION. SEARCH FOR ASSOCIAT'ED GEOLOGICAL STRUCTURE (S). ASSESSMENT OF FAULTS WHICH WERE BELIEVED TO BE INDUCED BY PLEISTOCENE GLACIATION. POSSIBILITY THAT EARTHQUAKES ARE RELATED TO INJEC OF CHEMICAL WASTES IN TWO WELLS 7 MILES NORTH OF THE EARTHQUAKES.
' ASSESS FREE-FIELD GROUND MOTION RECORDINGS WITH
' RESPECT TO WORLD-WIDE DATA BASE; ASSESS SOURCE OF HIGH FREQUENCIES AND POSSIBLE SITE EFFECT. i i i 1 O 9
- O STRUCTURAL SEISMIC DESIGN i
PLANT WALKDOWNS REVEALED NO SIGNIFICANT STRUCTURE DAMAGE. REGION 111, SQRT, AIT A GOOD CORRELATION OF MEASURED IN-STRUCTURE RESPONSE AND THE RESPONS$ CALCULATED USING THE RECORDED FOUNDATION MOTION AND A FIXED-BASE REACTOR BUILDING 1 DYNAMIC MODEL WAS OBTAINED. THIS CONFIRMS A LACK OF ROCKING RESPONSE; AND HENCE, INSIGNIFICANT SOIL-STRUCTURE INTERACTION. ' RECORDED FOUNDATION MOTIONS ARE SIMILAR TO THE CORRESPONDING FREE-FIELD GROUND MOTIONS IN FhEQUENCY [ CONTENT; ABOUT 20 Hz. WITH A FIXED-BASE 3-D MODEL AND THE RECORDED FOUNDATION ACCELERATION TIME HISTORY AS INPUT, THE CALCULATED RESPONSE
- SPECTRA AT HIGH ELEVATION OF REACTOR BUILDING INDICATE SIMILAR AMPLICATIONS AS THE MEASURED RESPONSES OVER TH I MEASURED FOUNDATION RESPONSE SPECTRA, AT 20 HZ REGION.
f a 1 i I I
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- THE ORIGINAL PERRY REACTOR BUILDING DYNAMIC MODEL IS, THEREFORE, ADEQUATE IN PREDICTING HIGH FREQUENCY RESPONSE.
THE HIGH-FREQUENCY, SHORT-DURATION EARTHQUAKE HAS AN INSIGNIFICANT ENERGY CONTENT. FOR REACTOR BUILDING, THE RECORDED MOTION WOULD NEED TO BE SCALED BY A FACTOR OF HIGHER THAN TWO TO ACHIEVE DEFORMATIONS CORRESPONDING TO THE DESIGN LEVEL OF FORCES. I A QUANTITATIVE ASSESSMENT INDICATES THAT THE DYNAMIC STRESSES IN CONTAINMENT BUILDING INDUCED BY THE RECORDED i FOUNDATION TIME HISTORY IS WELL BELOW THE DESIGN VALUES l 1 O
+ sr ,
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S EQUIPMENT SEISMIC QUALIFICATION PLANT WALKDOWNS REVEALED NO DAMAGE TO EQUIPMENT ITSELF, THE SUPPORTS, AND MOUNTING CONFIGURATIONS. THE HIGH-FREQUENCY, SHORT-DURATION EARTHQUAKE HAS AN INSIGNIFICANT ENERGY CONTENT TO CAUSE DAMAGES TO EQUIPMENT. ALL OF THE 39 SAFETY-RELATED AND 36 NON-SAFETY RELATED SYSTEMS THAT WERE ENERGIZED DURING THE EARTHQUAKE HAD FUNCTIONED AS DESIGN. THE TRIPPING OF THREE NON-SAFETY RELATED SYSTEMS WAS CONSIDERED TO BE EITHER INSIGNIFICANT OR EXPECTED BY DESIGN. A QUANTITATIVE ASSESSMENT OF A SAMPLING OF EQUIPMENT REVIEWED SO FAR INDICATES THAT THE ORIGINAL SEISMIC QUALIFICATION IS ADEQUATE. ( l l O 5
& 57
O CONCLUSIONS NO SIGNIFICANT SAFETY IMPACT OF THE EARTHQUAKE ON
- EQUIPMENT AND STRUCTURES HAS BEEN IDENTIFIED.
THE DESIGN-BASIS EA'RTHQUAKES MAY HAVE BEEN EXCEEDED AT A HIGH, NARROW FREQUENCY REGION OF THE RESPONSE SPECTRA, THE ORIGINAL PLANT SEISMIC DESIGN WAS NOT AFFECTED. THE STAFF CONCLUSIONS AS STATED IN THE PREVIOUS SER AND ITS SUPPLEMENTS REGARDING THE ADEQUACY OF THE l SEISMIC QUALIFICATION PROGRAM REMAIN VALID. O i i i O
& Gd 1
O CONFIRMATORY ACTIONS BY THE APPLICANT PERFORM AN ADDITIONAL QUANTITATIVE ASSESSMENT ON THE SEISMIC QUALIFICATION OF A BROADER SAMPLE OF EQUIPMENT TYPES, LOCATED IN d!FFERENT BUILDINGS ON VARIOUS ELEVATIONS. PERFORM A GENERIC EVALUATION OF A HIGH-FREQUENCY, SHORT-DURATION EARTHQUAKE WITH REGARD TO ITS ENERGY CONTENT AND POTENTIAL SAFETY SIGNIFICANCE OF EQUIPMENT AND STRUCTURES AT PERRY. USING THE RESULTS OBTAINED, ASSESS THE SEISMIC CAPABILITY OF THE PERRY PLANT, IF ANOTHER EARTHQUAKE OF SIMILAR CHARACTERISTICS, BUT WITH HIGHER MAGNITUDE AND/OR LONGER DURATION SHOULD OCCUR NEAR THE SITE. 1 O d 0( J
___ _ _m _. _. . - - - E E APPENDIX V APPLICANT PRESENTATIONS ON PERRY EARTHQUAKE ISSUE AGENDA I. INTRODUCTION AND OVERVIEW M. R. EDELMAN, CEI I SEISMIC DESIGN EVALUATIONS DR. C. CHEN GILBERT COMMONWEALTH
- EARTHQUAKE ANALYSIS AND SEISMICITY R. HOLT
; WESTON GEOPHYSICAL SEISMICITY AND INJECTION WELLS DR. P. TALWANI UNIVERSITY OF SOUTH CAROLINA 1
SUMMARY
AND CONCLUSIONS M. R. EDELMAN, CEI i i i i ,i l l ? 1 i O l
- g. 4 ^
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-,r.
i, OVERVIEW
- SEISMIC EVENT 01/31/86
- PLANT RESPONSE l
- SUBSEQUENT EVALUATIONS j - PHYSICAL PLANT
- DESIGN BASIS .
i. j - EARTHQUAKE ANALYSIS 1 l PRIOR ACRS MEETINGS
- )
J^
- DESIGN CONFIRMED
- PLANT READY TO LOAD FUEL I
O l i 1 1 i i 1 i i ee
- O l 3
g4 l l I 1
t 5 O V l 1 PLANT STATUS PRIOR TO SEISMIC EVENT
- ONGOING TESTING, CALIBRATION, WORK ACTIVITIES PREPARATION FOR DIVISION II DIESEL GENERATOR TESTING STARTUP SOURCES NOT YET MOVED SYSTEMS ENERGIZED (IN OPERATION AND STANDBY MODE) i 39 SAFETY SYSTEMS 36 NON-SAFETY SYSTEMS I
i ) O 4 J l, 4 O d cY i
I INSPECTIONS AND FINDINGS FOLLOWING THE SEISMIC EVENT OPERATOR SURVEY (NO STRUCTURAL DAMAGE) WALKDOWNS BY PLANT MAINTENANCE PERSONNEL (NO STRUCTURAL DAMAGE) SYSTEMATIC PLANT WALKDOWNS BY 65 ENGINEERS AND TECHNICIANS (NO STRUCTURAL / EQUIPMENT DAMAGE) SITE SURVEY PLANT SETTLEMENT SURVEY COOLING TOWER WALKDOWN SEISMIC CLEARANCE INSPECTION ENERGIZED ELECTRICAL EQUIPMENT STUDY ~7006 8 ON GOING SVI's , 1 l SYSTEMS ENERGIZED AND NOT-ENERGIZED DURING EVENT POST EARTHQUAKE WORK REQUEST REVIEW PROCEDURE
) -
ASSURED CEI/NRC REVIEW OF ALL POTENTIAL EARTHQUAKE-RELATED ITEMS TIME FRAME: FEB. 3 THRU 28 TOTAL WR's REVIEWED DURING PERIOD: 2401 EARTHQUAKE RESPONSE PROCEDURE REVISIONS ADDITIONAL GUIDANCE TO OPERATORS ON USE OF INFORMATION FROM AVAILBLE INSTRUMENTATION LONGERTERM EVALUATION OF SETPOINTS AND ACTIONS BASED ON EXPERIENCE AND RESULTS OF ALL ENGINEERING EVALUATIONS O g- p S
- , CHARACTERISTICS OF THE 1986 OHIO EARTHQUAKE
(] e HIGH FREQUENCIES e SHORT DURATION e LOW ENERGY , e LOW VELOCITY e SMALL DISPLACEMENT
~
, CH ARACTERISTICS OF TH'E PERRY DESIGN B ASIS EARTHQUAKE (SSE) e BROAD BAND FREQUENCIES
- e LONG DURATION e HIGH VELOCITY e LARGE DISPLACEMENT e H!GH ENERGY a6 l O -
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I . O - SEISMIC DESIGN BASIS e BROAD BAND FREQUENCY DESIGN RESPONSE SPECTRA i e SMOOTHED,84 PERCENTILE 1 SPECTRA l l e COMPOSITE TIME HISTORIES WITH LONG DURATIONS AND HIGH
=
i O ENERGv S
.. l O ;
,+ 7 6 l \
. ZPA COMPARISON e ZERO PERIOD ACCELERATION ~ .
e
- RECORDED ZPA'S VARY FROM BELOW OBE VALUES TO 74% OF SSE VALUES e EXCEPT AT CONTAINMENTVESSEL ELEVATION 686'
' ~
e BUT RELATIVE DISPLACEMENTS 1 AND STRESSES HE.RE (AS WELL A5 ALL OTHER LOCATIONS) ARE LOW l REASON:
)
e HIGH FREQUENCY AND LOW
~
ENERGY OF THE 1986 EARTHQUAKE eS O g- 7/
O CONTAINMENT STRESSES COMPARISON t gg DYNAMIC FOPCES g4IC STRESS FFOM CAICUIATED STFESSES-NNte 1 SEISMIC CGTOET P Mr Mr P/A + Mr/5 or P/A + M /5 (K) (FtK) (FtK) (K/In2) 688'-6* 1,339 44,220 31,820 0.414 (X/In2)
.398 Bbte 2 644' 6* 1,589 46,970 44,820 0.464 .802 592'3* 1,674 38,000 53,670 0.510 1.320 Bbte 3 NOIE 1:
O THE DESICN OF CONTAINMENT VESSEL IS CChTROLLED BY INIEFNAL PRESSURE IDADS, NOT SEISMIC IDADS. NCTIE 2: STRESSES AT THIS ELEVATICN DOES NOT CCNTROL DESIGN. SIBCE THIS IS A UNIFORM THICKNESS SHELL, DESIGN IS CCNTFOLLED BY MAXIMCM STRESS AT ELEVATICN 592'-3". NOTE 3: THEYIEIDSTRESSISg8K/IN IN COPARISCN WITH THE DYNAMIC STRESS OF 0.510 K/IN . O i
- g. 9 b
- - - y---v-w-.mm
RESPONSE SPECTRA COMPARISON e PERRY DESIGN RESPONSE SPECTRA ARE FAR
. ABOVE THE RECORDED SPECTRA IN THE FREQUENCY REGION BELOW 11 Hz.
o CERTAIN RECORDED RESPONSE SPECTRA EXCEED DESIGN SPECTRA VALUE5 IN THE REGION AROUND 20 Hz. e CORRESPONDING SMALL DISPLACEMENTS O (EXAMPLE: 7/100 INCHES OR BELOW AT FOUNDATION MAT) e RECORDED VELOCITY SPECTRA SHOW MUCH LESS ENERGY THAN THE DESIGN l RESPONSE SPECTRA ! O
~
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._ !., . O E5?.IHC'dArst JAhvARY ai, iSES IIAE001 PERRY NUCLEAR POWER PLAN! C D.9
- S OU T .H SMA35/N :ES-:L
( DA.981NG VALUES ARE 2 l'ERCENT OF CRI T IC AL ( . 10 FREOUENCY - HZ 1 *O 3.. .. . . .. . 3.. .. . . . .
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1 - I 10 - c-0 1d 1
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EQUIPMENT SELECTION CRITERIA ACTIVE SAFETY CLASS EQUIPMENT REQUIRED FOR SAFE SHUTDOWN. - EQUIPMENT LIST COMPILED BY LAWRENCE LIVERMORE NATIONAL LABORATORY, WITH FREQUENCIES HIGHER THAN 14 Hz AND HCLPF VALUES LESS THAN 0.5g. SUPPLIED BY MULTIPLE VENDORS. ACTIVE COMPONENTS QUALIFIED BY ANALYSES. VALVES & MOTOR OPERATORS SUPPPORTED BY PIPING SYSTEMS. ELECTRICAL SWITCHGEAR & INSTRUMENT RACKS. VERTICAL PUMPS. BATTERIES & BATTERY RACKS. CRITERIA RESULT IN MORE COMPREHENSIVE SAMPLES OF EQUIPMENT. APPROXIMATELY 75% OF SELECTED EQUIPMENT TYPES HAVE ALREADY BEEN EVALUATED WITH AMPLE QUALIFICATION MARGINS. BALANCE TO BE COMPLETED BEFORE JUNE 1986. O 1
.--n-.. - - - .,----,,_,,,.c--. - - , - - - , ,,- , - ,, . . , , , , , - , ~ , ,n,, e --
O REPRESENTATIVE EQUIPMENT EVALUATED FOR SEISMIC QUALIFICATION MARGIN Qualification Components Bldg. Elev. Methods Vendor ~ Motors Auxiliary 568' Test / GE Analysis Pumps Auxiliary 568' Analysis Byron Jackson Control Intermediate 620' Test GE Switchgear Blowers Auxiliary 620' Test GE Pressure Auxiliary 568' Test Rosemount Transmitters Power Control 620' Test Brown Boveri Switchgear Flow Auxiliary 568' Test Rosemount Transmitters Valves Reactor 688' Test / Henry Pratt' Analysis Actuators Reactor 688' Test / Dettis Analysis Limitorque Instrument Auxiliary 568' Test GE Racks O 4 g ar
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, Run No. 60, HCA, 1:: (SSE) ( 2 % Page 26 og 63 '
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, COMPARISON OF THE CUST0MIR'S HORIZONTAL SSE RAS TO THE HORIZONTAL !
TRONT-TO-BACK TRS OF THE SEK-350 SWITCH 0ZAR TEST SPECIMDI f.I : GAL MPL ' IR22 S0% , it 22 S007, I R11.-S003 1
O . CONCLUSION THE 1986 OHIO EARTHQUAKE:
- 1 HIGH FREQUENCIES SHORT DURATION LOW ENERGY LOW VELOCITY ,
SMALL DISPLACEMENT STRUCTURAL & EQUIPMENT DESIGN HAS SUBSTANTIAL MARGINS OF SAFETY RELATIVE TO LOADS & STRESSES INDUCED BY THE EARTHQUAKE O e0 0
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M AINSHOCK JANUARY 31,1988
- LOCATION
- TOWN OF LEROY LAKE COUNTY, OHIO DISTANCE TO PLANT : APPROX.10.5 MILES APPRO X.17.0 KM.
ORIGIN TIME : 16hr dem 4 2.3 s U.T. U.S. MODEL JEFFREYS-8ULLEN TABLE O t l LATITUDE : 41.6 5 0 N 41.6 4 9 N LONGITUDE : 81.16 2 W 81.10 5 W l 1 FOCAL DEPTH : 5 KM (CONS 7R AINED) i Me (TELESEISMIC) 4.9 (10 ST ATIONS) l i i SOURCE: MATION AL EARTHOUAKE INFORM ATION SERVICE (USGS) t l l
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O . 5AFE 5HUTDOWN EARTHQUAKE MODIFIED MERCALLIINTENSITY = Vil PEAK GROUND ACCELERATION -0.15 g REGULATORY GUIDE 1.60 SPECTRA SITE SPECIFIC SPECTRA O - e 5.3 MsLG t .5 e 5.5 M ata t .3 o
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1 DNPP - BASIC SSRS Ci4 Components] Dampins = 0.05 2 l 10 , . i iii ii i i i ii l 1 e i 6 ___Ii i g - W M ] N 10 z
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- NO CAPABLE FAULT
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- MODIFIED MERCALLI VII VS. VI (1/31/86)
SITE SPECIFIC RESPONSE SPECTRA FOR SSE
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EXCEEDANCE OF SAFE SHUTDOWN EARTHQUAKE SHORT DURATION (4 0.5 SEC.)
- HIGH FREQUENCY (20 Hz)
CONFIRMATORY PROGRAM OF STUDIES l O i
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LAKE ERIE Paaav
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1 O SEISMICITY AND CALHIO INJECTION WELLS NO KNOWN SEISMICITY NEAR WELLS - BEFORE/AFTER JANUARY 31, 1986 EVENT EARTHQUAKES - 11 KM S OF WELLS. NO SEISMICITY BETWEEN THEM HISTORICAL SEISMIC ACTIVITY IN AREA BEFORE WELLS OBSERVED TIME LAGS BETWEEN INJECTION'& SEISMICITY AT OTHER LOCATIONS IS IN DAYS / WEEKS. HERE 2 4 YEARS NUMEROUS MINOR EVENTS AT OTHER KNOWN LOCATIONS OF INJECTION RELATED SEISMICITY. NOT SEEN HERE LACK OF EYDROLOGICAL CONNECTION BETWEEN WELLS
* /
ANY RELATIONSHIP'BETWEEN INJECTION WELL AND JANUARY 31, > 1986 EVENT IS HIGHLY UNLIKELY O h l i
SUMMARY
AND CONCLUSIONS
- DEMONSTRATED ADEQUATE SEISMIC DESIGN FOR JANUARY 31, 1986
, EARTHQUAKE INTENSITY VI, 4.9 RICHTER
- SAFETY RELATED PLANT STRUCTURES AND SYSTEMS UNAFFECTED BY EARTHQUAKE
- NO CHANGE TO CONCLUSIONS ON GEOLOGY AND SEISMOLOGY
- DESIGN EARTHQUAKE BOUNDS JANUARY 1986 EVENT (MAGNITUDE, ENERGY ETC.)
- DESIGN SPECTRUM IS ADEQUATE EVEN WITH THE INCLUSION OF RECENT EVENT
- PLANT SEISMIC DESIGN ABLE TO ACCOMMODATE JANUARY, 1986 EARTHQUAKE
- SHORT DURATION, HIGH FREQUENCY, LOW ENERGY
- MEASURED RESPONSE EXCEEDANCES RESULT.IN CALCULATED STRESSES WITHIN DESIGN
- REPRESENTATIVE SET OF EQUIPMENT EVALUATIONS CONFIRM SEISMIC QUALIFICATION MARGINS CONFIRMATORY PROGRAMS IN PROGRESS O
,9-99
I SEISMIC DEISGN MARGINS
- ACRS RECOMMENDED SCHEDULE FOR COMPLETION, INCLUDING ANY NEEDED MODIFICATION, PRIOR TO STARTUP FOLLOWING SECOND REFUELING (END OF 1989)
- CURRENTLY MONITORING / PARTICIPATING GENERIC INDUSTRN
& NRC EFFORTS TO DEVELOP METHODOLOGIES AND ACCEPTANCE CRITERIA -
FOLLOWING DEVELOPMENT OF GUIDANCE, AND DATABASE INFORMATION, ESTABLISH SEISMIC MARGIN ASSESSMENT PROGRAM DURING FIRST CYCLE OF OPERATIONS COMPLETION OF EVALUATION AND IMPLEMENTATION OF RESULTS DURING SECOND CYCLE OPERATIONS O O
,p v
APPENDIX Vi l EARTHQUAKE F ""$$5fd$EN!"EESIGN l0 RESEAI 1 I IAN B. WALL J. CARL STEPP H. T. TANG ELECTRIC POWER RESEARCH INSTITUTE PALO ALTO, CALIFORNIA O i l l O er #
4 SOIL STRUCTURE INTERACTION RESEARCH O GENERATE DATABASE TO VALIDATE REALISTIC SSI MODELS . WORK WITH NRC ON REVISION TO SRP 3.7.2 O l O g f0 5
EPRI SEISMIC RESEARCH i O : SEISMIC HAZARD IN EASTERN UNITED STATES ;
)
SEISMIC MARGINS FOR EUS PLANTS DATABASE FOR SOIL-STRUCTURE INTERACTIONS DATABASE FOR PIPING DAMPING ~ AND ULTIMATE CAPACITY + 0 O
'#/st . -, ... , , . _ _ _ _ _ _ . . . . _ _ _ _ _ _ _ _ _ _ _ . . _ .-- ....m_, . . _ _ - , _ -_.. ...~,_- ,~.-.c_..----.--r-.. . , - , - , -
EPRI SEISMIC MARGINS PROJECT C WALKDOWN PROCEDURE
- IDENTIFY CRITICAL COMPONENTS
- AVAILABLE MARCH 1986 QUANTIFICATION PROCEDURE FOR CRITICAL COMPONENTS
- INTEGRATE PRODUCTS FROM OTHER EPRI WORK
- AVAILABLE SEPTElillER 1987 FF82u AR y O
O
- g. tex
O EPRI SEISMICITY OWNERS GROUP SEISMIC HAZARD RESEARCH PROGRAM PURPOSE 4 DEVELOP PROBABILISTIC SEISMIC HAZARD
! METHODOLOGY AND INTERPRETATIONS TO ADDRESS THE
" CHARLESTON EARTHOUAKE ISSUE" "THERE IS A FINITE PROBABILITY THAT LARGE EARTHOUAKES MAY OCCUR AT ANY LOCATION IN THE CENTRAL AND EASTERN UNITED STATES WHERE FAVORABLE GEOLOGIC STRUCTURE EXISTS, INDEPENDENTLY OF PRIOR SEISMICITY"
) ' 2 JCS/LHM/3854ST6B O .
O EPRI SEISMICITY OWNERS GROUP SEISMIC HAZARD RESEARCH PROGRAM PARTICIPATION FORTY-TWO NUCLEAR UTILITIES IN CENTRAL AND EASTERN UNITED STATES. PROGRAM MANAGEMENT O - EPRI NUCLEAR POWER DIVISION SEISMIC CENTER OWNERS GROUP OVERSIGHT COMMITTEES EXECUTIVE COMMITTEE TECHNICAL ADVISORY COMMITTEE LICENSING STEERING COMMITTEE ~ JCS/t.HM/3854ST6s 1
- O
. g. M Y
- - . ~ - ..... . - -
io EPRI SEISMICITY OWNERS GROUP SEISMIC HAZARD RESEARCH PROGRAM i PROGRAM HAS TWO PHASES
- PHASE 1 - METHODOLOGY DEVELOPMENT DECEMBER 1983 TO APRIL 1985 O
- PHASE 2 -METHODOLOGY REVIEW, TESTING AND ENHANCEMENT APRIL 1985 TO JULY 1986 1
i i , i l JCS/LHM/3854ST68 3 i O
. . . ..__. .. . _ ... . 2 . . _.. .. _ ... .
O EPRI SEISMICITY OWNERS GROUP SEISMit HAZARD RESEARCH PROGRAM PHASE 1
- DATA BASE COMPILATION
- METHODOLOGY DEVELOPMENT PROBABILISTIC PROCEDURES INPUT INTERPRETATION PROCEDURES O - INPUT INTERPRETATIONS SElSMIC SOURCE ZONE INTERPRETATIONS SEISMICITY OCCURENCE PARAMETER
~
l INTERPRETATIONS
- TEST COMPUTATIONS JCS/LHM/3854ST68 4 O
g- a +
l v SEISMICITY OWNERS GROUP EARTH SCIENCE TEAMS Team Members Bechtel Group, Inc. Dr. Thomas Buschbach Dr. Robert D. Hatcher, Jr. Dr. Joseph Litehiser* Dr. Rolfe Stanley Dr. Isidore Zietz Dames & Moore Prof. Charles Fairhurst Prof. Robert Herruann Prof. Lyle McGinnis Mr. James McWhorter* Dr. Rene Rodriguez Law Engineering Testing Company Prof. Robert Butler Dr. Martin Chapman O Dr. John Dwyer Prof. Arch Johnston Prof. Timothy Long Mr. Malcolm Schaeffer Mr. William Seay Dr. Robert White
- Rondout Associates, Inc. Ms. Noel Barstow*
Prof. William Hinze Prof. Pradeep Talwani Prof. Barry Voight Weston Geophysical Corporation Mr. Richard Holt Dr. George Klimkiewicz* Dr. Gabriel LeBlanc Prof. Donald Wise Woodward-Clyde Consultants Dr. Terry Engelder Dr. John Kelleher Dr. Richard Quittmeyer Mr. Thomas Statton* l Dr. Thomas Turcotte
- Team Leader g-/s2 ,
O EPRI SEISMICITY OWNERS GROUP SEISMIC HAZARD RESEARCH PROGRAM PHASE 2
- METHODOLOGY ENHANCEMENTS
- COMPARATIVE EVALUATIONS WITH NRC/LLNL METHODOLOGY O - PARAMETRIC ANALYSIS SCIENTIFIC PEER REVIEW i
- SEISMIC HAZARD METHODOLOGY TOPICAL REPORT JCS/LHM/3854ST68 5 0
- p. /sr
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JA06 w GDC-4 RULEMAKING a - E 'i @E SCHEDULE FINAL PROPOSED D(
)
LIMITED SCOPE RULE JULY 85 MARCl1 86 BROAD SCOPE RULE MAY 86 DECEMBER 86 O O O
BROAD SCOPE GDC-4 RULE RFSPONSE TO ACRS METALS SUBCOMMITTFF QUESTIONS
- 1. IS THE BROAD SCOPE Rule LIMITED TO HIGH ENERGY PIPING?
YES, THIS WAS THE INTENT. PIPE WHIP RESTRAINTS AND JET SHIELDS ARE ONLY LOCATED NEAR SUCH PIPING. THE SUPPLEMENTARY INFORMATION IN THE N RULE IS MODIFIED TO STATE THAT ONLY HIGH ENERGY PIPING IS AFFECTED AND A DEFINITION OF HIGH ENERGY PIPING IS PROVIDED (SYSTEM PRESSURE GREATER THAN 275 PSIG OR OPERATING TEMPERATURE GREATER THAN 200 F)
- 2. ARE INDIVIDUAI BREAKS EXCLUDED OR ARE Al1 BREAKS IN A FIUID SYSTEM EXCIUDFD?'
THE RULE SPEAKS OF THE " PROBABILITY OF FLUID SYSTEM PIPE RUPTURE" BEING EXTREMELY LOW. THE RULE DOES HOI DEAL WITH SPECIFIC BREAK LOCATIONS. WHEN LEAK-BEFORE-BREAK IS APPLIED, IT IS APPLIED TO FLUID SYSTEM PIPING OR PORTIONS THEREOF. A PORTION OF A FLUID SYSTEM PIPING MEANS AN ANALYZABLE SEGMENT OF PIPING, USUALLY BETWEEN ANCHORS. THE RULE IS MODIFIED TO CLARIFY THIS. O O O
BROAD SCOPE GDC-4 RULE
- 3. WOUID ADDITIONAL ATTENTION BE GIVEN TO FLANGED JOINTS WHEN IFAK-REFORF-BREAK IS APPLIED TO WELDED JOINTS?
NO. IN THE JUDGMENT OF THE STAFF, CATASTROPIC RUPTURES IN FLANGED JOINTS NEED NOT BE POSTULATED. PAST EVALUATIONS HAVE INDICATED THAT FLANGED JOINTS DO NOT RUPTURE. NO MODIFICATION NEEDED TO RESPOND g TO THIS QUESTION. Q
- 4. WIII THE RuiE REQUIRF AN EVALUATION OF CREEP RUPTURE IN PIPING?
YES. IN LWRS CREEP DOES NOT OCCUR AT THE TEMPERATURE AND FOR THE MATERIALS WHICH EXIST. HOWEVER, SINCE THE RULE APPLIES TO GAS AND METAL COOLED REACTORS, AN EVALUATION OF CREEP MAY BE REQUIRED IN SOME CASES. THE RULE IS MODIFIED TO REFLECT THIS COMMENT.
- 5. IS THE INFLUENCE OF PLUGGING INCLUDED IN THE LEAK RATE ESTIMATIONS?
NO, PLUGGING IS NOT CONSIDERED A PROBLEM FROM THE PRACTICAL POINT OF VIEW BECAUSE OF HIGH PRESSURES AND FLAW GE0METRIES (N0 TIGHT CRACKS); HOWEVER, RESEARCH IS PLANNED AND ON-G0ING THAT ADDRESSES THIS ISSUE. NO MODIFICATION NEEDED TO RESPOND TO THIS COMMENT. O O O
BROAD SCOPE GDC f4 RULE
- 6. IF THERE ARE PUBLIC SAFETY BENEFITS. SHOUlDN'T THE RulF BE MANDATORY RATHER THAN PERMISSIVE?
WHILE THE EFFECTIVENESS OF INSERVICE INSPECTION IS IMPROVED, AND PIPE RUPTURE PROBABILITIES ARE REDUCED (BECAUSE OF ELIMINATION OF INADVERTENT RESTRAINT OF THERMAL GROWTH) THESE BENEFITS x HAVE NOT BEEN QUANTIFIED, AS A CONSEQUENCE, THERE IS NO SECURE BASIS FOR MAKING THE RULE MANDATORY. ON THE OTHER HAND, THE INCREASED PUBLIC RISKS ASSOCIATED WITH REMOVING DEVICES WHICH MITIGATE THE AFFECTS OF POSTULATED PIPE RUPTURES HAS BEEN SHOWN TO BE INSIGNIFICANT FOR REACTOR COOLANT LOOP PIPING. O O O
BROAD SCOPE GDC-4 RULE SAFETY ISSUE APPR0XIMATELY 15,000 PIPE WHIP RESTRAINTS IN SERVICE TODAY OF VARYING SIZES AND DESIGNS (INSTALLED AT A COST OF ROUGHLY $2 BILLION fl985 DOLLARS) IN DIRECT COSTS; IF STRETCHED OUT SCHEDULE AND FINANCE COSTS ARE INCLUDED, FIGURE JUMPS TO $6 BILLION), MORE THAN 100,000 EXPERIENCE YEARS ACCUMULATED WITH PIPE WHIP RESTRAINTS, WITif0VT L) HOWEVER, EVEN ONE INSTANCE WHERE A PIPE WHIP RESTRAINT WAS NEEDED TO ;g INDIAN POINT,-DUANE ARNOLD AND MAINE PERFORM A SAFETY FUNCTION. t YANKEE, WHICH TO DATE REPRESENT THE MOST SEVERE LEAK / BREAK PIPING PROBLEMS IN SERVICE, DID NOT REQUIRE PIPE WHIP RESTRAINTS TO MAINTAIN S: SAFETY. WE BELIEVE FOREIGN EXPERIENCE IS SIMILAR. PIPE WHIP RESTRAINTS CAN DEGRADE SAFETY BY LIMITING THERMAL EXPANSION WHEN CONTACT BETWEEN THE PIPE WHIP RESTRAINT AND PIPE INADVERTENTLY OCCURS DUE T0: A. INTERACTION WITH ANY OF THE ESTIMATED 2000 LOCKED (FAILED) SNUBBERS NOW INSTALLED (APPR0XIMATELY 40,000 SNUBBERS IN SERVICE, WITH 4000 FAILED " FREE" AND 2000 FAILED " LOCKED"). G e
BROAD SCOPE GDC-4 RULE SAFETY ISSUE (CONTINUED) B. INABILITY TO LIMIT, CONTROL OR ESTIMATE THE INFLUENCE ON TilERMAL GROWTH OF PIPE SUPPORT GAPS. 4 C. DIFFICULTIES WITH MAINTAINING TOLERANCES AND ALIGNMENTS IN PIPE s
\
WHIP RESTRAINTS, WHICH IN THE COLD CONDITION MAY BE SEVERAL 8 INCHES AWAY FROM PIPING, WHILE DURING OPERATION ARE TO BE A FRACTION OF AN INCH AWAY IN SOME CASES. I THESE CONTACT PROBLEMS MAY CAUSE CRACKS TO GROW AT LOCATIONS NOT NOW PROTECTED AGAINST PIPE RUPTURE BECAUSE OF MODIFIED STRESSES. PIPE WHIP RESTRAINTS LIMIT ACCESSIBILITY FOR AND DIMINISil EFFECTIVENESS OF INSERVICE INSPECTION WHILE INCREASING WORKER RADIATION EXPOSURES. O 9 e
I l l BROAD SCOPE GDC-14 RULE REASONS FOR LIMITING LBB TO DYNAMIC EFFECTS 0 LEAKS, VALVE MALFUNCTIONS AND OTHER SOURCES OF BLOWDOWN IMPOSE REQUIREMENTS FOR CONTAINMENTS, ECCS AND ENVIRONMENTAL QUALIFICATION WHICH CANNOT BE ELIMINATED BY INVESTIGATING AND DEMONSTRATING PIPING INTEGRITY. O IF LBB IS APPLIED TO THE CONTAINMENT, ECCS AND ENVIRONMENTAL QUALIFICATION DESIGN BASES TO ELIMINATE THE DEGB ACCIDENT, THEN A b REPLACEMENT PIPE RUPTURE ACCIDENT MUST BE DEVELOPED FOR THESE k ASPECTS OF FACILITY DESIGN. NONE NOW EXISTS. O MAJOR LONG TERM RULEMAKING WOULD BE NEEDED TO ADDRESS OTHER THAN DYNAMIC EFFECTS, THEREBY FORESTALLING IMMEDIATE PAYOFF IN ELIMINATING PIPE WHIP RESTRAINTS, JET IMPINGEMENT BARRIERS AND UNDERTAKING OTHER BENEFICIAL FACILITY MODIFICATIONS. O DYNAMIC EFFECTS LEAD TO THE PLACEMENT OF COUNTER PRODUCTIVE HARDWARE WHICH NEGATIVELY AFFECTS PLANT PERFORMANCE IN TERMS OF DEGRADED SAFETY AND INCREASED COSTS. OTHER ASPECTS OF FACILITY DESIGN MAY BE NEGATIVELY AFFECTED BY POSTULATED DEGB'S, BUT NOT TO THE SAME DEGREE THAT DYNAMIC EFFECTS IMPOSE COUNTER PRODUCTIVE REQUIREMENTS.
# 9 e
BROAD SCOPE GDC-4 RULE DEFINITION OF DYNAMIC EFFECTS 0 PIPE WHIP AND OTHER PIPE BREAK REACTION FORCES. g N 0 JET IMPINGEMENT FORCES. O DECOMPRESSION WAVES WITHIN THE RUPTURED PIPE (AFFECTS PIPE COMPONENTS AND THEIR INTERNALS). . O PPESSURIZATION IN CAVITIES, SUBCOMPARTMENTS AND COMPARTMENTS, EXCEPT WHEN THESE VOLUMES ARE PART OF THE CONTAINMENT SYSTEM, 0 PIPE RUPTURE GENERATED MISSILES (INSULATION, PIPE SUPPORT BOLTS, ETC.). E* *
- BROAD SCOPE GDC-4 RULE INTERNATIONAL LBB SEMINAR ON OCT 28-30, 1985)
FOREIGN PRACTICES: (SOURCE: GENERAL NOTES: 1. N0 NATION ACCEPTING OR CONSIDERING LBB AT TilIS TIME WILL MODIFY REQUIREMENTS FOR CONTAINMENTS, ECCS OR ENVIRONMENTAL QUALIFICATION.
- 2. EVERY NATIONAL ACCEPTING OR CONSIDERING LBB AT THIS TIME WILL INCLUDE, OR IS DISPOSED TO INCLUDE, CRITERIA FOR LEAKAGE DETECTION.
UK STRONG INCLINATION TO REJECT LBB FOR SIZEWELL BASED ON > CONCERNS WITH SCC AND NDE (CEGB AT ODDS WITH NII ON THIS). FRANCE UNDECIDED, BUT WEAKLY INCLINED TO REJECT LBB AT THIS TIME PARTLY BECAUSE OF COMMITMENT TO STANDARDIZATION. RESEAR LBB IN PROGRESS. FRG STRONG COMMITMENT TO LBB IN PWR MAIN COOLANT, MAIN FEED AND MAIN STEAM INSIDE CONTAINMENT. ITALY CLOSE TO FRG PRACTICES.
- O e
BROAD SCOPE GDC-11 RULE FOREIGN PRACTICES: (CONTINUED) JAPAN INCLINED TO ACCEPT LBB. HEAVY INVESTMENT IN LBB RESEARCH, d x CANADA INCLINED TO ACCEPT LBB FOR CERTAIN PIPING SYSTEMS AT THE DARLINGTON FACILITY. Q MOST OTHER COUNTRIES SEEM TO BE LESS ACTIVELY INVOLVED IN LBB, OR ARE OBSERVING DEVELOPMENTS IN THE LEADERSHIP COUNTRIES AB0VE.
- O e
Camparisen Setween U.S. exd FRG Proptsed Revisions fyr pipe Rupture Desten Reevirements at nuclear Power Plants FRG U.S.
; Scope of piping Limited to pWR main coolant loop, pressurizer Any piping in any reactor i surge line, main steam lines, feedwater line, type which arets rigorous RHR lines and other high energy lines inside acceptance criteria. piping the containment. SWR piping not included. subject to failure from water Piping outside containment no_Tincluded, hammer, corrosion indirect sources of rupture and high and low cycle fatigue ex-cluded from consideration.
Only pWR primary coolant loops known to satisfy
, rigorous acceptance criter.
1a at this time. Old piping vs. Limited to new piping which has several Age not a factor. piping in new piping quality improvements, such as forged operating plants, plants fittings, double us11ed piping. Iower under construction and i stress levels, better in-service inspection, future designs included. snubbers replaced with bumpers on components and high standards for leak detection.
! Requirements for Not affected, not affected.
omergency core
; cooling systans, environmental qualification and contatsment pressurization I
i Dynamic leads on Retained in modified form; previous stynamic Eliminated. heavy component requirements replaced with a static analysis
- supports from with a safety factor of 2.
I pipe rupture Replacement for los flow area longitudinal or circuerferential None; however, breaks may double-ended break or some other requirement based on still be postulated in pipe rupture for fracture mechanics analyses. connecting branch pipes. stynamic loads pressurization Retained. tiiminated except in volumes in compartments, related to the contairunent subcompartments function. , and cavities from pipe rupture
- Effects of Double-ended pipe rupture eliminated in pri- Eliminated.
l decompresston asry coolant loop; homever double-ended pipe waves on heavy rupture retained in asin steam and feeduster component inter- lines for the steam generators.
- mais i
Leakage detec- Leakage detection depended upon, and described hst be reliable, redundant. tion in detail in "Appendia 1 of Draft of Trans- diverse and sensitive so that , O actions of 173rd Meeting of R$K." a mergin greater than 10 on detection of unidentified leakage from throughwall flaws exists.
; + /# / . - - - - - , _ _ - . , _ - _ - . - , . . , . , , , , . _ . - _n-,,,, _ . , _ _ _ _ , , , , - , , . , -m- .
BROAD SCOPE GDC-4 RULE VALUE/ IMPACTS 0 FOR THE LIMTIED SCOPE RULE, APPLYING LBB TO PRIMARY COOLANT LOOP PIPING IN A FORECASTED POPULATION OF 85 PWRS LED TO THE FOLLOWING BEST ESTIMATE RESULTS (BASED SOLELY ON PIPE WHIP RESTRAINTS AND JET IMPINGEMENT BARRIERS): AVERTED RADIATION EXPOSURES: 34,000 MAN-REM N REDUCED COSTS: $186 MILLION 4 N HOWEVER, HEAVY COMPONENT SUPPORT REDESIGN CAN YIELD EVEN GREATER VALUE/ IMPACTS FLORIDA POWER CORP IS ESTIMATING COST SAV)NGS OF ,
$20 MILLION AND AVERTED RADIATION EXPOSURES OF 2000 MAN-REM DUE y SOLELY TO REDESIGN OF REACTOR COOLANT PUMP SUPPORTS AT CR-3.
O IF LBB COULD BE APPLIED TO BWR RECIRCULATION LOOP PIPING UNDER THE BROAD SCOPE RULE, THE FIGURES ABOVE COULD BE INCREMENTED AS FOLLOWS (BASED ON A FORECASTED 38 BWRS AND CONSIDERING ONLY PIPE WHIP RESTRAINTS AND JET IMPINGEMENT BARRIERS): AVERTED RADIATION EXPOSURES: 8,600 MAN-REM REDUCED COSTS: $30 MILLiON 0 FOR ANY FUTURE PLANT, COST SAVINGS OF APPR0XIMATELY $100 MILLION ARE ESTIMATED PER PLANT THROUGH EXCLUSION OF PIPE BREAKS (DUE TO IMPROVED CONSTRUCTION SCHEDULES, REDUCTIONS IN FINANCING COSTS AND DIRECT DESIGN AND CONSTRUCTION COSTS). SOME OF THESE PIPE BREAKS CAN BE EXCLUDED VIA THE PROPOSED REVISION TO SRP 3.6.2. O O O
BROAD SCOPE GDC-4 RULE NRC RESOURCE DEMANDS 0 GE (FOR GESSAR) AND DUQUESNE LIGHT CO. (FOR BEAVER VALLEY, UNIT
- 2) ARE ALREADY MAKING DEMANDS FOR NRC RESOURCES UNDER Tile BROAD p SCOPE RULE. q N
0 IT IS EXPELTED THAT IF THE BROAD SCOPE RULE IS NOT PUBLISHED THE , COMMISSION WILL BE INUNDATED WITH REQUESTS FOR SYSTEM AND PLANT g UNIQUE EXEMPTIONS TO GDC-4 IN ADDITION TO PETITIONS FOR RULEMAKING. THIS EXPECTATION ARISES BECAUSE THE PIPING REVIEW COMMITTEE PUBLICALLY SUPPORTED A BROAD SCOPE RULE AND THE LIMITIED SCOPE RULE STATED THAT "THE COMMISSION WILL PROPOSE A BROADER AMENDMENT TO GDC-4". TWO-THIRDS OF COMMENTERS ON THE LIMITED SCOPE RULE URGED THAT THE BROADER RULE BE EXPEDITED. O APPR0XIMATELY TEN TO TWENTY NRC MAN YEARS OF EFFORT PLUS ADDITIONAL RESEARCH OVER THE NEXT SEVERAL YEARS ARE ESTIMATED TO RESPOND TO INDUSTRY INITIATIVES TAKEN UNDER THE BROAD SCOPE RULE. NOTE: INDUSTRY HAS ALREADY EXPENDED AND IS CONTINUING TO EXPEND CONSIDERABLE RESOURCES FOR LBB RESEARCH, ANALYTICAL TECilNIQUES AND CRITERIA DEVELOPMENT. 9 O e
BROAD SCOPE GDC-4 RULE RELATED NRC REGULATORY ACTIONS SRP 3.6.2 IS BEING REVISED TO ELIMINATE REQUIREMENTS FOR ARBITRARY INTERMEDIATE PIPE BREAKS. THIS REVISION WILL ALSO ALLOW THE REMOVAL OF PIPE WHIP RESTRAINTS AND JET IMPINGEMENT BARRIERS, BUT IS y INDEPENDENT OF Tills RULEMAKING. HOWEVER, THE SRP REVISION HAS ABOUT THE SAME POTENTIAL MAGNITUDE OF VALUE/ IMPACTS. THE PIPE RUPTURES BEING ELIMINATED VIA THE SRP REVISION ARE: ,
- 1. ARBITRARY IN THE SENSE THAT THEY ARE REQUIRED WITHOUT PHENOMEN0 LOGICAL BASES, THAT IS, EVEN TiiOUGH STRESSES AND USAGE FACTORS (A FATIGUE MEASURE) ARE ACCEPTABLY LOW, THESE BREAKS ARE STILL POSTULATED.
- 2. INTERMEDIATE TO CONTRAST WITH TERMINAL END BREAKS WHICH ARE STILL POSTULATED. TERMINAL END BREAKS ARE POSTULATED WITHOUT SPECIFIED PHENOMEN0 LOGICAL BASES BECAUSE HISTORY TEACHES THAT PIPE RUPTURES ARE MORE LIKELY TO OCCUR AT TERMINAL ENDS, G
G e
BROAD SCOPE GDC-4 RULE RELATED NRC REGULATORY ACTIONS (CONTINUED) THIS RULEMAKING AFFECTS POSTULATED BREAKS WHICH ARE NOT NORMALLY ARBITRARY INTERMEDIATE BREAKS. ABOUT FIFTEEN NUCLEAR POWER UNITS HAVE ALREADY BEEN ALLOWED TO RELAX ARBITRARY INTERMEDIATE BREAK REQUIREMENTS EVEN THOUGH THE REVISION IS STILL IN PROGRESS, 4 N i s 9 O O
APPENDIX VIII NUREG 0313 REV 2 NUREG 03 .O EXPANDS REV. 1 COVERAGE INCLUDES ALL STAINLESS PIPING (cL 1, 2, 3) REV. 1 HAD LIMITED SCOPE REQUIRES F_0RMAL QUALIFICATION OF NDE EXAMINERS AND PROCEDURES
. d 4 tat y p rLL - WC Cs Q, p_ .
REV. 1 JUST RECOMMENDED THAT IMPROVED UT PROCEDURES BE USED () PROVIDES GUIDLINES FOR EVALUATION AND REPAIR OF CRACKED WELDS REV. 1 REQUIRED REPLACEMENT OF CRACKED WELDS O g /1 4
NUREG 0313 REV 2 GENRALLY FOLLOWS RECOMMENDATIONS OF PIPING REVIEW , COMMITTEE NUREG 1061 VOL. 1 RECOMMENDS: USE OF IGSCC RESISTANT MATLS REPLACEMENT OF SUSCEPTIBLE PIPING PROCESSES FOR RESIDUAL STRESS IMPROVEMENT IMPROVED WATER CHEMISTRY O PROVIDES SPECIFIC INSPECTION SCHEDULES CONSIDERING: MATERIAL IGSCC RESISTANCE STRESS IMPROVEMENT PROCESSING WATER CHEMISTRY IMPROVEMENT REPAIRS AND CRACKING CONDITION PROVIDES GUIDELINES FOR CRACK EVALUATION AND REPAIRS UPGRADES LEAKAGE LIMITS AND MONITORING O
- 4. p,
APPENDIX 1X IMPLEMENTATION PLAN FOR THE SEVERE ACCIDENT POLICY STATEMENT AND THE REGULATORY USE OF NEW SOURCE-TERM INFORMATION O NRR STAFF PRESENT ACRS ,
~~
SUBJECT:
IMPLEMENTATION PLAN FOR THE SEVERE ACCIDENT POLICY STATEMENT AND THE REGULATORY USE OF NEW SOURCE-TERM INFORMATION DATE: MARCH 14, 1986 Z0LTAN R. ROSZTOCZY ]RESENTER: PRESENTER'S TITLE / BRANCH /DIV: CHIEF REGULATORY IMPROVEMENTS BRANCH DIVISION OF SAFETY REVIEW 8 OVERSIGHT PRESENTER'S NRC TEL. NO.: 492-8016 G -
,9. p r ng -- - - - - - p g a -
c's r-- -'
SEVERE ACCIDENT POLICY STATEMENT - ACTION ITEMS POLICY STATEMENT NEW APPLICATIONS EXISTING PLANTS
- s. GUIDANCE ON THE ROLE SYSTEMATIC APPROACH OF PRAs FOR THE EXAMINATION 11 0F INDIVIDUAL PLANTS x
R}, *
- PERFORMANCE CRITERIA IMPLEMENT MODIFICATION
'ND FOR CONTAINMENT THROUGH BACKFIT POLICY SYSTEMS CHANGES IN RULES AND REGULATORY PRACTICES, -
AS NEEDED i,
^
IMPLEMENTATION PROGRAM El.EMENTS
- 1. EXISTING PLANT EXAMINATION
- REVIEW OF THE IDCOR INDIVIDUAL PLANT EXAMINATION METHODOLOGY
- DEVELOPMENT OF GUIDELINES AND CRITERIA FOR PLANT EXAMINATIONS
- 2. DEVELOPMENT OF GUIDANCE ON THE ROLE OF PRAs FOR NEW PLANT APPLICATIONS
- DETERMINISTIC REQUIREMENTS
({])
- ACCEPTABLE CONTENT OF PRAs
- CRITERIA FOR THE REGULATORY REVIEW AND INTERPRETATION OF THE PRA RESULTS
- 3. CHANGES IN RULES AND REGULATORY PRACTICE
- SOURCE TERM PELATED CHANGES
- SEVERE ACCIDENT RELATED CHANGES i
O 2 g'/3d
EXPECTED ACCOMPLISHMENTS PLANT SPECIFIC VULNERABILITIES WILL BE IDENTIFIED AND' FIXED (BACKFIT RULE) IF GENERIC VULNERABILITIES ARE IDENTIFIED, APPROPRIATE DESIGN AND/0R OPERATIONAL CHANGES WILL BE REQUIRED (RULEMAj(ING) O- 'esSONS 'enaNeD WI u Ne'e DeveL0eneNT Os IneR0veD DESIGNS WITH SAFETY BENEFITS A NEW, MORE REALISTIC REGULATORY APPR6XCFfON SOURCE TERMS WILL BE PURSUED (SOURCE TERM RELATED CHANGES) O l
,$/J /
SUMMARY
OF EXPECTED ACCOMPLISHMENTS SEVERE ACCIDENT POLICY FMPLEMENTATION COMPLETE THE NRC ANALYSIS OF SIX 6/86 PEFERENCE PLANTS FOR SEVERE ACCIDENTS INCLUDING SOURCE TERM CALCULATIONS , RESOLVE IDCOR/NRC TECHNICAL ISSUES 7/86 COMPLETE THE REFERENCE PLANT SENSITIVITY 7/86 STUDIES (EVALUATION OF UNCERTAINTIES) COMPLETE REVIEW 0F IDCOR METHODOLOGY 10/86 FOR INDIVIDUAL PLANT EXAMINATIONS BRIF.F COMMISSION ON THE FINDINGS AND 12/86 PECOMMENDATIONS FOR THE INDIVIDUAL PLANT EXAMINATIONS ISSUES GUIDANCE FOR PUBLIC COMMENT 2/87 ON THE ROLE OF PRAs FOR NEW PLANT APPLICATIONS ISSUE FOR PUBLIC COMMENT RULE CHANGES 4/87-NECESSARY TO RESOLVE GENERIC SEVERE ACCIDENT RELATED VULNERABILITIES O g y/3 A
O
~
O O p
~s p-----
/ Reference Plant g
/
Regulatory I Analyses l . . nnciple
\ 6/86 / \ j
~._
s _J % _l / Preparation of Strawman [ Guidelines lf 9/86 p----- Technical Issue Evaluation of Final / \ Guidelines Research Resolution + Reference Plants Es Criteria l Update I 7/86 8/86 \ 10/86 / 10/86 g
- -_ j Development # -
of Proposed J L Criteria 9/86
/ -- % 5 mk l
/ Sensitivity Analyses
\
I Commi,ssion Briefing
\ 7/86 / 12/86 W % _________ /
s lf Standards for Review of Acceptable IDCOR ' Methodology Methodology 3/86 10/86 LEGEND Program Activities I~~') Input from other Programs C ~~~~'N
'---- / IDCOR \
l Methodology I
\ ()$/8G
/
}
Figure 3.1 Program Eleinent 1. Development of Guidance for Individual Plant Examinations
O TABLE 3.2 LISTING OF MILESTONES 3.1 PEVIEW 0F THE IDCOR INDIVIDUAL PLANT EXAMINAITON METHODOLOGY STANDARDS FOR AN ACCEPTABLE METHODOLOGY 3/86 SUB91TTAL OF IDCOR PEP 0 PTS FOR TWO BWRS 3/86 AND TWO PWRS SUBMITTAL OF PEFAINING THREE IDCC'R PEPORTS 7/86 NRC COW ENTS TO IDCOR ON METHODOLOGY 7/86 EVALUATION OF THE APPLICATION OF THE IDCOR - 10/86 METHODOLCGY TO SEVEN PLANTS 3.2 CEVELOPMENT OF GUIDLINES AND CRITERIA FOR PLANT EXAMINATIONS 3.2.1 TECHNICAL ISSUE RESOLUTION
- DEFINE UNCERTAINTY PANGES FOR SURRY, -
2/86 PEACH BOTTOM 8 SEGUOYAH
- DEFINE UNCERTAINTY PANGES FOR ZION, -
3/86 GRAND GULF & LASALLE DRAH NRC/IDCOR ISSUE PAPERS - 5/86 O - et"^' "ac'tocca issue e^ mas - 7/88
,9. / ) V 6
TABLE 3.2 (CONTINUED) d ! 3.2.2 EVALUATION OF REFEPENCE PLANTS j - COPPLETION OF IDCOR SEVERE ACCIDENT EVALUATIONS (WITH UNCERTAINTY ANALYSIS) - 7/86 COMPLETION Cf SARRP RISK EVALUATIONS:
- SURRY -
4/86 PEACH BOTTOM - 5/86 ZION - 5/86 1 SEQUOYAH - 6/86 GRAND GULF - 6/86 i l CCPPLETE PEFERENCE PLANT RISK PROFILE - 8/86 EVALUATION OF REFERENCE PLANTS - 8/86 1 3.2.3 PPEPARATION OF STRAWAN GUIDELINES 3.2.4 DEVELOPENT OF PROPOSED CRITERIA STRAWAN GUIDELINES & PROPOSED CRIIERIA PEACH E0110M - 6/86 OTliER PLANTS - 9/86 3.2.5 DEVELOPENT OF FINAL GUIDELINES & CRITERIA q - FINAL GUIDELINES & CRITERIA - 10/86 V
- g. /3S 7 !
l
% m l
/ \
( Resolution of a USis and GSls
\
3
/ ~\ \_ ./ Evaluation of Deterministic Deterministic
/ Assessment of \1 - Requirements Requirements 12/86
{ Reference GESSAR & IP Plants - - 10/86 8/86 /
~ - . _ 2
/ PSA h IREP \
Procedures k Guides Procedures for
\ 8## Core Damage j.-~~ g /
%__--__- / Frequency
/ \ 10/86
/ IDCoR l Methodology G *" ' "
Commission Commission ( 10/86
/
n6 mum Content of PRAs
- Paper - Approval k _ .,/ 12/86 I#8I IIII
/a.- - - - Procedures for OO / Guidelines and Containment b g Criteria for IPE Consequence ,
\ 10/86 / Analyses Y % ____ sl %
[ [ Reference Plant Analysis i 6/86 I Use of Safety
\ ~
/ Goats With
[ Safety Goal 6 g "# *rt nties gf Containment l Performance Guidance on
\ Objective / Criteria for 6/86 N%---s/ Regulatory use 12/86 LEGEND f "~ ~ N Criteria for PSA Manual. \ incremental Risk l Program Activities
/ 9/86 l e PRA lnsights I Reports j (,,,,,,)
Input from other Pro 0 rems
\ $/85 j
%.- -_ /
Figure 4.1 Program Element 2 - Development of Guidance on the Role of PRAS 4 l i i t
O O O Source Term Calculation \
\
for Reference Plants l
\- _ _6/86
___/ g------
/ Development of \ Capability for Development of I Source Term Codes I '
Source Term Calculation ' New Source Terms - Source Term
\ 12/85 / 3/86 12/86 r Related Changes
\ -- /
Selection of f Regulatory Principle Research Update g 10/86 4/86
\ /
s -- - x p~ Containment g Development of e ,/ Performance - ' Containment 1 Design objective ' " Performance Criteria
\ \ 6/86 ,/ 10/86 Q , - - - - - - -
R eronco lan s Containment Related Changes g 6/86 j Identification of Generic Vulnerabilities 9/86 Other Severe Accident Related Changes p---- % p---- _ LEGEND [ Hesolution of / Examination of
) I
( USis and GSis j k Individual Plants l Program Activities
~
( _ ,,,_ _ _ _ ,,/ \_______/ ~ { } Input from Other Programs Figure 5.1 Program Element 3 - Changes in Rules and Regulatory Practice
.r. !
.? I
_. . =. . .. . - - . -- . _ . -_ O P0lENTIAL SOURCE TERM CHANGES i NEAR-TERM INTERMEDIATE LONG-TERM REVISED TREATMENT EMERGENCY PLANNING SITING 0F ACCIDENTS IN EIS 4 REMOVAL OF SPRAY CONTAINMENT LEAK RATES ACCIDENT O ADDITIVES <ewa) 4ND 1NTEGaITY MONIr0aINe SUPPRESSION POOL ENV, QUALIFICATION
- CPEDIT (BWR) 0F EQUIPMENT O
- c-SAFETY ISSUE EVALUATION l
1 CONTROL 900M HABITABILITY j AND AIR FILTRATION SYSTEMS . O
- g. /35 10
-- .- . = - .-. - - - -_ . - - - - . . - . - . . - . - - - - . . . . . . .
Q V POSSIBLE REVISED APPROACH WISH TO RE-EXAMINE THE USE OF FISSION PRODUCT CALCULATIONS IN LICENSING (TID-14844 AND REG. GUIDES 1.3/1.4) . POSSIBLE APPROACH CONSIDER A NUMBEP OF ACCIDENT SEQUENCES LEADING TO COPE DEGRADATION, MELT AND RELEASE INTO CONTAINMENT EVALUATE ACTIONS OF NATURAL REMOVAL PROCESSES AS /] WELL AS FISSION PRODUCT CLEANUP SYSTEMS TO EVALUATE PROGRESSION OF ACCIDENT AND CALCULATE RELEASE OF FISSION PRODUCTS TO CONTAINMENT G: . c-FROM ABOVE, DETERMINE FISSION PRODUCT TYPES, AMOUNTS AND TIME-DEPENDENT CONCENTRATIONS IN CONTAINMENT THAT GENERALLY ENVELOPE SEQUENCES ESTIMATE TIME-DEPENDENT CONTAINMENT LEAK RATE AND CALCULATE FISSION PRODUCT LEAKAGE FROM CONTAINMENT CALCULATE DOSES TO HYPOTHETICAL INDIVIDUALS AT EAB AND LPZ AND COMPARE TO PART 100 (MAY HAVE TO ADD OTHER ORGAN DOSE CRITERIA) s- /37 11
(~\
SUMMARY
OF EXPECTED ACCOMPLISHMENTS SOURCE TERM RELATED CHANGES ISSUE FOR COMMENT PEVISED SRP SECTION 6.5.2 , 9/86 SPECIFYING THE NEED FOR SPPAY ADDITIVES ~IN PWRs ISSUE FOR COMMENT REGULATORY GUIDE 1.3 AND THE , 9/86 APPROPPIATE SECTION OF THE SRP ON FISSION PRODUCT SCRUBBING IN SUPPRESSION P0OLS (BWRs) ISSUE FOR COMMENT PROPOSED CHANGES TO 10 CFR 50.47 2/87 AND 10 CFR 50, APPENDIX E ON EMERGENCY PLANNING REVISE NRR OFFICE LETTER 16 WITH RESPECT TO THE 2/87 USE OF SOURCE TERMS IN SAFETY ISSUE EVALUATION O ISSUE FOR COMMENT CHANGES IN CONTAINMENT LEAK 3/87 RATE REQUIREMENTS, INCLUDING POTENTIAL CHANGES IN 10 CFR 50 APPENDIX J G: c-REVISE 10 CFR 50.49 AND REGULATORY GUIDE 1.89 6/87 WITH RESPECT TO THE RADIATION ENVIRONMENT FOR EQUIPMENT QUALIFICATION, FOR COMMENT BY ISSUE FOR COMMENT REVISIONS OF SITING CRITERIA 10/87 (10 CFR 100) BASED ON NEW SOURCE TERM INFORMATION ISSUE FOR COMMENT REVISED REGULATORY GUIDE 1.97 12/87 ON ACCIDENT MONITORING AND MANAGEMENT O g& 12
O
's_,) PELATIONSHIPS WITH OTHER PROGRAMS RES PROGRAMS ,
- NUREG-0900 SUPPLEMENT, RESEARCH PLAN FOR ,
SEVERE ACCIDENTS ._
- NUREG-0956, REASSESSMENT OF THE TECHNI. CAL BASES FOR ESTIMATING SOURCE TERMS ,
- NUREG-1150, REFEPENCE PLANT ASSESSMENT
- UPDATE ON SEVERE ACCIDENT RESEARCH NRR PROGRAMS
- SAFETY GOALS: FINAL VERSION, CONTAINMENT _
PERFORMANCE DESIGN OBJECTIVE
- UNPESOLVED AND GENERIC SAFETY ISSUES:
(~) STATION BLACKOUT, SHUTDOWN DECAY HEAT REMOVAL
- PPA REVIEWS AND INSIGHTS PEPORTS: INDIAN C3R POIET., ZION, LIMERICK, AND GESSAR; PPA INSIGHTS REPORTS, PROCEDURES GUIDE AND REVIEW MANUAL .
INDUSTRY PROGRAMS
- IDCOR: REFERENCE PLANT ANALYSES, TECHNICAL ISSUES, INDIVIDUAL PLANT EXAMINATION METHODOLOGY
- AIF: SOURCE TERM ISSUES, PRA ISSUES (3
s_/ g- / Y/ 13
Q CONS!CEPATION OF EXTERNAL EVENTS IN THE V IPPLFitNTATION PROGRAM IDCOR WORK, l!P TO NOW, AND NRC WORK IN SUPPORT OF NUREG-1150 WEPE LIMITED TO INTERNAL EVENTS,'IN GENERAL.
*~
SEISMIC EVENTS WERE CONSIDERED UNDER SEPARATE PPOGPAMS, NPC HAS A SEPARATE SEISMIC PROGRAM, INDilSTRIES EFFORTS ARE C00PDINATED BY THE SEISMIC QUALIFICATION USERS GROUP (SQUG), SENIOR SEISMIC PEVIEW ADVISOPY PANEL, EPRI, AIF ETC. THE GOAL OF THE POLICY STATEFtNT AND ITS IMPLEIENTATION IS , TO BRING STABILITY TO LICENSING AND PEGLILATION WITH RESPECT TO ALL SEVEPE ACCIDENT ISSUES. THIS CANNOT BE ACCCFPLISHED WITHOUT CONSIDEPATION OF EXTERNAL EVENTS. THE AVAILABILITY OF ANALYTICAL E THODS FOR EXTERNAL EVENTS IS COMPARABLE TO THAT OF INTEPNAL EVENTS. BOTH NRC AND THE l INDUSTRY HAS DEVELOPED METHODOLOGY FOR SEISMIC RISK ANALYSIS. APPROXIMATELY 20 SEISMIC PRAs HAVE BEEN COMPLETED OR ARE CLOSE O V
-l f)14
T C MPLETICN. SIMPLIFIED MEll!CDS APPROPRIATE FOR IDEhTI-O FYING WLNEPABILITIES ARE BEING DEVELOPED BY BOTH NRC AND EPRI. SAMPLE APPLICATIONS OF THE METliODS APE PLANNED, AN APPROACH TO THE HANDLING OF EXTERNAL EVE RS IN THE IMPLE- , PENTATION PRCGRAM HAS BEEN DEVELOPED BY THE NRC STAFF AND , IT IS UNDER MANAGEPENT REVIEW, O i l k' g 15
+,--m-y --
--y a %y __.--
, . _ _ . , _ , . _ _ . . . . _ _ . . ,___y _ . , m. __-, - , . . . ._.w__77.-._.,_,m. ,
-.em - - - ,r,-.----.-.
REcutENDED APPROACH TO EXTEPNAL EVENTS EXTERNAL EVENTS NEED TO BE CONSIDERED: HOWEVER, THE EXIENT OF THE REVIEW AND THE SCHEDULE FOR EXTEPNAL EVENTS DOES NOT HAVE TO BE THE SAME AS FOR INTEPNAL EVENTS. THE EFFORT SHOULD CONCENTRATE ON FLANI WLNEPABILITIES DUE TO EXTERNAL EVENTS. RESOLUTION OF THE QUESTION WHAT IS THE CON-TRIBUTION OF EXTERNAL EVENTS TO OVERALL RISK IS NOT A NECESSITY, THE FIRST PHASE OF THE PROGRAM SHOULD CONCENTPATE ON: A.) ESTABLISHING THE EXTENT TO WHILH PLANTS HAVE BEEN PEVIEWED FOR EXTERNAL WLNEPABILITIES IN THE PASI OR WILL BE REVIEWED l] UNDER ONGOING PROGRAMS, LIKE A4I6. B.) ESTIMATING THE MARGIN THESE PEVIEWS WILL ASSUPE PELATIVE T0 LVENTS BEYOND THE DESIGN BASIS, FOR EXAMPLE EARIHOUAKES l BEYOND SSE. s C.) IDENTIFYING EXTEPNAL tvENTS THAT NEED TO BE INCLUDED IN THE WLNERABILITY SEARCH. D.) IDENTIFYING THOSE AREAS WHERE EXAMINATION FOR WLNERABILITIES IS NEEDED, AND ARE NOT COVERED UNDER ANY OF IHE EXISTING PROGRAMS, O l v . J 1 8- / V! 16
= , - - - - - - . - - . -. . - - - .. . . __
i l RECC^19DED APPROACH TO EXTEPNAL EVENTS - (CONTINUED) O " NRC SHOULD SOLICITY INDUSTRY PAPTICIPATION IN THE FIRST PHASE OF THE PROGRAM WHEN THE FIRST PHASE IS CTPLETE, THE STAFF WILL ADDRESS THE FOLLOWING CUESTIONS: A.) IS THEPE A NEED TO CONTINUE THE PROGRAM? B.) WHAT FORM SHOULD THE SECOND PHASE OF THE PROGRAM TAKE? lo i 1 l, a 17
APPENDIX X NRC ADVANCED REACTOR PROGRAM j e NRR STAFF PRESEN7 ACRS
SUBJECT:
NRC AD\ANCED REACTOR PROGRAM DATE: i%RCH 14,1986 PRESENTER: TH0f%S L. KING PRESENTER'S TITLE / BRANCH /DIV: SECTION LEADER SAFETY PROGRAM EVAll'ATION BRANCH DI\'ISION OF SAFETY REVIEW & OERSIO!T PRESENTER'S NRC TEL. NO.: 49 2-7014 SUBCOMMITTEE: FULL C0ft11TTEE O g.199
1 l PROGRAM O I) ORIGINAL PLANS: - 1
- NRC REVIEW OVER THE NEXT TWO YEARS CONCEPTUAL DESIGNS FOR ONE HTGR AND TWO LMRs.
- INTERACTIONS AMONG THE STAFF /ACRS/D0E/ REACTOR DESIGNERS ON KEY ISSUES FOLLOWED BY NRC REVIEW 0F A PRELIMINARY SAFETY INFORMATION DOCUMENT (PSID) AND PRA ON EACH CONCEPT.-
- REQUIRED STAFF RESOURCES WERE 5-6 STAFF YEARS PER YEAR AND APPROXIMATELY $1.25M TECHNICAL ASSISTANCE PER YEAR.
- ALL RESOURCES WERE TO BE FROM NRR.
~
II) CURRENT DIRECTION: (~'
- REDUCE EFFORT TO 2 STAFF YEARS PER YEAR AND NO TECHNICAL ASSISTANCE.
- IMPACT ON PLANNED REVIEWS IS CURRENTLY BEING ASSESSED.
WILL PROBABLY LIMIT FUTURE INTERACTIONS TO A FEW KEY ISSUES. I a e 9 O g in
. _ _ _ . . - _ . . _ _ _ _ . _ ~ _ .__. _ _ ___ _ . . . . _ _ . . - _ _ . - _ ._
7 . i f l i ~ KEY ISSUES-EXAMPLES i O j 1) CONTAINMENT / CONFINEMENT ! 2) USE OF INTERFACE CRITERIA ONLY FOR THE BALANCE OF PLANT, j 3) USE OF A SINGLE, PASSIVE, HIGHLY RELIABLE DECAY HEAT ! REMOVAL SYSTEM. l
- 4) TREATMENT OF SEVERE ACCIDENTS AND EMERGENCY PLANNING, j' <
i i - 1 4 i i l0 l ,9-ivt i : I
. _ - . , _ , . _ . . , _ , . - - . _ _ . - . _ _ _ _ _ _ _ - . . .._ _ .. _ _ ,,- _._ _ . _ .. _ _ _.. _ _ _..__.__=_ . - _ ,... ,, _ _ _ ,, _ . _ _ . -- -
MAJOR ACCOMPLISHMENTS TO DATE [ I) MEETINGS HELD ON KEY ISSUES:
- HTGR:
- TOP LEVEL DESIGN CRITERIA
- APPROACH FOR EMERGENCY PLANNING
- DESIGN APPROACH
- LMR:
i
- SHUTDOWN HEAT REMOVAL
- LICENSE BY TEST APPROACH
- METAL FUEL
!
- APPLICATION OF LWR-GSIs
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- SAFEGUARDS AND SECURITY II) REVIEW :
- DOCUMENTS CURRENTLY UNDER REVIEW ON:
- HTGR TOP LEVEL CRITERIA
- LMR SHUTDOWN HEAT REMOVAL
- LMR SAFEGUARDS AND SECURITY 9
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1 ND Nk E V I e-T I T E P M O C N O I T P O G N I T A R E N E G M L A A R C G I O R R T P C E R L O E T C A E R E F A O S . D S E Y E C L V N E I A V T V I A D S N A S R A E P T L A T S Y O R C A R R WO G O P T L M H E A T D N N P O A O C E L S V E H I V T I 'R T E M C DW E L E P J O B C O - S - o o
PROPOSED PLANT CONFIGURATIONS (RATINGS ARE APPROXIMATE) SAFR PRISM HTGR 1,400 1,250 560 PLANT ELECTRICAL OUTPUT (MWE) s' 4 3 2 TG UNITS PER PLANT REACTORS PER PLANT 4 9 4 4 h
\
1 3 2 REACTORS PER TG UNIT J 1 1 1 CONTROL ROOMS PER PLANT 925 430 350 REACTOR THERMAL RATING (MWT) 350 420 280 TG ELECTRICAL RATING (MWE) 350 140 140 EQUIVALENT ELECTRICAL RATING PER REACTOR (MWE)
UTILITY PROBLEMS / ISSUES _ o PUBLIC ACCEPTANCE o LICENSING PROCESS AND UNCERTAINTY o SHORTER AND MORE PREDICTABLE CONSTRUCTION TIME COST AND FINANCIAL RISK UNCERTAINTIES q o N o IMPROVED PLANT CAPACITY FACTORS o FINANCING / INSTITUTIONAL ALTERNATIVES k o SliARED OWNERSHIP RISKS 'l 9 9 e
ADVANCED REACTOR CHARACTERISTICS o SMALL/ MODULAR o PASSIVELY SAFE CERTIFIED AND STANDARDIZED PLANT o W o SHOP FABRICATION h N i o PRE-ASSEMBLED - BARGE OR RAIL TRANSPORTABLE I o COST COMPETITIVE o SHORT CONSTRUCTION TIME o HIGH PLANT CAPACITY FACTORS O O e
NRC ADVANCED REACTOR POLICY STATEMENT REQUIREMENTS RtLIABLE, SIMPLE HEAT REMOVAL o EASY MAINTAINABILITY o SYSTEMS o FEWER SUPPLEMENT.;L SAFETY o REDUCED PERSONNEL EXPOSURE vs FEATURES o LONGER TIME CONSTANTS o MULTIPLE BARRIERS y o INCREASED STANDARDIZATION AND o SIMPLIFIED SAFETY SYSTEMS FABRICATION UTILIZE INilERENCY, RELIABILITY, o EXPERIMENTALLY VERIFIABLE o REDUNDANCY, DIVERSITY, SAFETY FEATURES INDEPENDENCE o RELIABLE BOP EQUIPMENT l
ADVANCED REACTORS ACCOMPLISHMENTS TO DATE o LMR's/HTGR - MIDWAY THROUGH CONCEPTUAL DESIGN . s t o NRC INTERACTION SCHEDULES ESTABLISilED AND UNDERWAY - POSITIVE EX o COSTS - PRELIMINARY ESTIMATES COMPETITIVE WITH LWR's AND C0AL x o SAFETY - ENHANCED PASSIVE CORE SilVTDOWN AND HEAT REMOVAL PROCES o R&D NEEDS EVOLVE FROM CONCEPTS - FOCUSES THE SUPPORTING R&D PROGRAM o UTILITY INTERACTION - ACTIVELY PURSUED - POSITIVE TO DATE 9-e- - - -
f
- >~ s ;
r m I i i ACRS BRIEFING ON MHTGR )% AGENDA N e PROGRAM STATUS h l v e DESIGN, SAFETY, AND LICENSING APPROACH i e DESIGN OVERVIEW m,e . 8=== l e SAFETY CHARACTERISTICS Ejj!
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! EE33 l == l VAXD:[WAUZEY]10 13 7-MA I
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! F 7 4 MHTGR PROGRAM STATUS i DESIGN, SAFETY, AND LICENSING APPROACH 4.. l l Q '% PRESENTED TO THE ACRS MARCH 14,1986 l l - l A. C. MILLUNZI i DEPARTMENT OF ENERGY } l N Y VAXD:[WAUZEY]10 7 11-WAR-86
I Q . O - O l l F 3 i i 1 PROGRAM OBJECTIVE ! 4 k DEVELOP HTGR'S FOR BROAD RANGE OF l Q APPLICATIONS IN SUPPORT OF ! COMMERCIAL / USER INTERESTS IN SAFETY } AND HIGHER TEMPERATURE I CHARACTERISTICS OF THESE PLANTS. l k ) j - 4 _ m ,3 , , , . ..
i O O . O NRC INTERACTIONS I I FY 1985 l FY 1986 l FY 1987 LICENSING l PLAN ' PROCEDURAL e . APPROACH I
- I I AGREEMENT -
AGRE MENT ON ON TO 2-LEVEL LICEN ING BASES k TECHNICAL APPROACH ! l i N I I Q- l l PSID
- DESIGN / TECHNOLOGY l Q
I FAMILIARIZATION ' i I I l l LICENSABILITY l l STATEMENT PSER DESIGN / TECHNOLOGY l
. e 7
e i REVIEW ' l I I i
- .l l V1
O - O - O r , l l . PROCEDURAL APPROACH INTERACTION LICENSING PLANT DRAFT COMPLETE g NRC ADVANCED REACTOR ! { POLICY ISSUED FOR CONNENT COMPLETE l x h LICENSING PLAN FORMAL SUBMITTAL COMPLETE INDUSTRY CONMENTS ON POLICY COMPLETE NRC ACCEPTANCE OF LICENSING PLAN COMPLETE l N VA l13 2 Il-WAR-86
i f 3 i i i i ) l TECHNICAL APPROACH INTERACTION BRIEFING SUBMITTAL TOP-LEVEL CR1TERIA COMPLETE COMPLETE g BRIDGING METHODS COMPLETE 2/86 0 g ACCIDENT SELECTION METHOD COMPLETE 2/86 ! SAFETY CLASS SELECTION COMPLETE 2/86 i ) PRINClPAL DESIGN CRITERlA COMPLETE 2/86 l ACRS BRIEFING 1/86 NA i l VA l13 3 u-uax-as 4
f 3 i 1 DESIGN / TECHNOLOGY FAMILIARIZATION j ISSUE BRIEFING l MODULAR HTGR DESIGN 12/85 I 4 FUEL 3/86 l . DECAY HEAT REMOVAL 5/86 h REACTIVITY CONTROL 5/86
! u CORE SUPPORT STRUCTURE 5/86 ISI 7/86
- WATER / AIR INGRESS 7/86
! CONTAINMENT / CONFINEMENT 8/86 i
BOP CLASSIFICATION 8/86 l MULTIPLE MODULE CONTROL 8/86 { STANDARD PLANT ISSUES 8/86 ACRS BRIEFING 9/86 ] i k j VA l13 4 11-MAR-86 l
i i O . O . O i I I 3 i DESIGN / TECHNOLOGY FAMlUARIZATION (cont) DOCUMENT DATE l g' TECHNOLOGY PLAN 9/86 ! PRA 9/86 { h PSID 9/86
'l - OUTLINE COMPLETE
- FULL SUBMITTAL 9/86 DESIGN AND TECHNOLOGY REVIEW PSER 6/87 j LICENSABILITY STATEMENT 9/87
%A 11 3 5 u-uAn-us
l i O O - O i HTGR DESIGN & LICENSING APPROACH i I USER TOP-LEVEL J REGULATORY REQUIREMENTS f CRITERIA , ! ; LICENSING BASIS
, r I
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6 INTEGR ATED APPRO ACH 4 l BRIDGE (-) LICENSING BASIS EVENTS l l i hm
, , EQUIPMENT CLASS
! _\ ENGINEERING PRODUCT OTHER BASES I PLANT DESIGN, ETC. 8
l 0 . o . o , f 3 l I I i I 1 MODULAR HTGR ) MEETS UTILITY USER REQUIREMENTS
- % e NOMINAL PLANT SIZE 500 MW(e) i x e EQUIVALENT AVAILABlUTY >80%
l * -5
- PROBABlUTY OF PLANT LOSS <10 / MODULE-YR I
q e MEET EXISTING SAFETY AND UCENSING CRITERIA WITH NO . l -7 PUBUC SHELTERING FOR EVENT FREQUENCIES >5x10/YR l e 10% POWER COST ADVANTAGE OVER COAL VAXD:[WAUZEY]IO 14 7-MAR-86 i e/ -
o o . .o r , PROPOSED BASIS FOR TOP-LEVEL CRITERIA SELECTION 4
- 1. CRITERIA MUST BE DIRECT STATEMENTS k OF ACCEPTABLE CONSEQUENCES OR RISKS TO THE PUBLIC OR THE ENVIRONMENT
- 2. CRITERIA MUST BE INDEPENDENT OF PLANT DESIGN
- 3. CRITERIA MUST BE QUANTIFIABLE J
VA 11 3 6 11-WAR-85
F m i i 1 1 i PROPOSED SOURCES AND CANDIDATES FOR TOP-LEVEL REGULATORY CRITERIA j e NUREG-0880 4 - INDMDUAL AND SOCIETAL MORTAUTY RISKS x e 10CFR50 APPENDIX l ! $ - NUMERICAL DOSE GUIDEUNES l e 10CFR100 l - NUMERICAL DOSE GUIDEUNES i t e EPA-520 j - PAG DOSES i ( ) VAXD:[MAUZEY]10 15 7-MAR-86 \ _ _ _ _ _ _ _ - - -
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O O - O F 3 ) l l i PURPOSE OF THE INTEGRATED APPROACH 4 THE INTEGRATED APPROACH IS USED TO: i 1 i h w o DEVELOP REQUIREMENTS ! 0 EVALUATE DESIGNS SELECTED TO MEET REQUIREMENTS ! o COMMUNICATE i t ; VAXD:[WAUKY]IO 7 6-MAR-86 i
O O O r m l SAVINGS FROM THE INTEGRATED APPROACH l l THE SAVINGS ENVISIONED FROM THE USE OF THE INTEGRATED APPROACH ARE DUE TO: j g' 9 A CLEAR UNDERSTANDING BY THE DESIGNERS, CONTRACTORS, AND OPERATORS OF WHAT THEIR ROLES i
] AND RESPONSIBlUTIES ARE.
% 9 AN EARLY IDENTIFICATION OF INTERFACES WHICH REDUCES THE RISK OF LATER MORE COSTLY
! REVISIONS. i 9 VISIBILITY OF THE BASIS FOR DESIGN REQUIREMENTS.
- 9 EUMINATION OF UNJUSTIFIABLE RETROFITS.
l 9 JUSTIFICATION FOR, OR DELETION OF, DEVELOPMENT ! PROGRAMS. L J 13 7 11-WAR-86
O O O
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, SAFETY PHILOSOPHY l e PROVIDE DEFENSE-IN-DEPTH THROUGH PURSUIT OF l FOUR GOALS:
1 1 - MAINTAIN SAFE PLANT OPERATION j 2 - MAINTAIN PLANT PROTECTION
- j. 3 - MAINTAIN CONTROL OF RADIONUCUDE RELEASE l
g( 4 - MAINTAIN EMERGENCY PREPAREDNESS i d e GOAL 1 TO BE ACHIEVED BY HIGHLY REUABLE OPERATION AND WITH WELL TRAINED PERSONNEL 4 l e GOALS 2 AND 3 TO BE ACHIEVED THROUGH l UTIUZATION OF INHERENT CHARACTERISTICS AND ! PASSIVE SAFETY FEATURES j e GOALS 1 - 3 TO BE ACHIEVED SO WELL THAT l MINIMAL RELIANCE NEED TO BE PLACED ON GOAL 4 i L J VA [13 8 11-WAR-8f
O O O i 3 i l i LICENSING METHODOLOGY
SUMMARY
i e USE PRA TO IDENTIFY UNEUHOOD OF EVENTS AND 4 - CLASSIFY EVENTS INTO THREE REGIONS FOR COMPARISON AGAINST TOP LEVEL REGULATORY CRITERIA i x ! v e EXAMINE EVENTS TO IDENTIFY REQUIRED FUNCTIONS TO ! I MEET THE TOP LEVEL REGULATORY CRITERIA [ i e CHOOSE DESIGN SELECTIONS TO ACCOMPUSH REQUIRED l . FUNCTIONS TO MEET THE TOP LEVEL REGULATORY CRITERIA i j j J mes.m p,
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i; MHTGR SAFETY AND LICENSING APPROACH CONSISTENT WITH g ADVANCED REACTOR POLICY
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- EARLY INTERACTION WITH NRC l W e CRITERIA SPECIFIC TO ADVANCED REACTORS
- e EMPHASIS ON INHERENT, PASSIVE SAFETY l
L J VAXD:[WAUZEY]t0 16 7-WAR-86
l O O O ' r 3 i MODULAR HTGR . t DESIGN OVERVIEW ) l S x PRESENIt-D TO THE ACRS a MARCH 14,1986 A. J. NEYLAN, DIVISION DIRECTOR 1 s GA TECHNOLOGIES INC. N A VAXD:[PUFFENBUR]46 1 11-WAR-86
l MODULAR HTGR KEY CHARACTERISTICS
- BASIC HTGR FEATURES l
4
- HEllUM GAS COOLANT -INERTISINGLE PHASE l - GRAPHITE MODERATOR - LONG RESPONSE TIMES
{ - GRAPHITE CORE STRUCTURE - HIGH TEMPERATURE STABILITY I Q
- CERAMIC COATED FUEL EMBEDDED IN GRAPHITE MATRIX - LOW RELEASESILOW WORKER EXPOSURE x
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- SPECIAL MODULAR FEATURES h -
CONFIGURATION SELECTED USING PASSIVE FEATURES ASSURES J PUBLIC SAFETY STANDARDIZED, PRE-LICENSED, FACTORY-FABRICATED ASSEMBLY
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O O O m GATechnologies M SIMPLIFIED FLOW DIAGRAM FOR MODULAR HTGR ii i j CONTROL ROD DAlVt/ REFUEtiMG PEMTRATIONS lF~ MACTM 49f** O O 1"ji ijj;i f MACien e ' _ mAls TUASWE M MRAIM CIRCHATOR f ' l SilAIE y M8ERAI88 gr -
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O O O h THE MODULAR HTGR HAS HIGH THERMAL EFFICIENCY i
- NUCLEAR SYSTEM PARAMETERS THERMAL POWER, [4 x 350 MW(t)] -
1400 HEllUM PRESSURE, PSIA 925 HEllUM TEMPERATURE, DF 497/1268 l 3' POWER SYSTEM PARAMETERS I ' POWER CYCLE NON-REHEAT
- D TURBINE INLET PRESSURE / TEMPERATURE, PSIA /cF 2415/1000 ll SYSTEM PARAMETERS NET ELECTRICAL DUTPUT, MW(e) 558 NET THERMAL EFFICIENCY, % 39.9 I , .. .-__-_--______-_-__I
O O . O M GATechnologies M SHUTDOWN COOLING SYSTEM ii ii P O O lljl;
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-J FISSILE AND FERTILE FUEL PARTICLES m GATechnologies M OUTER ISOTROPIC PYROLITIC CARBON p SILICON CARBIDE INNER ISOTROPIC PYROLITIC CARBON
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O O O THE MODULAR HTGR ASSURES PUBLIC SAFETY 4 WITHOUT OPERATOR ACTIONS h
- WITHOUT OPERATION OF ANY POWERED SAFETY SYSTEMS n
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'o WITHOUT THE NEED OF PUBLIC SHELTERING OR EVACUATION l
i l H-293(2) 7-11-8 5 PASSIVE SAFETY i i } _ _ _ _ _ ._ _ _ -- _ _ _ _
O O O r m MHTGR SAFETY CHARACTERISTICS
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PRESENTED TO THE ACRS N 1 MARCH 14,1986 l F. A. SILADY, SAFETY AND RELIABILITY MANAGER j GA TECHNOLOGIES INC. k J l VAXD:{WAUZEY]10 1 11-W A9-86
O O O r T I MHTGR SAFETY DESIGN OBJECTIVES
- 1. PROVIDE FUEL QUALITY AND SPECIFY NORMAL OPERATING CONDITIONS TO LIMIT RADIONUCLIDE INVENTORIES OUTSIDE OF FUEL TO MEET 10CFR100 DOSES AT PLANT g BOUNDARY
! 2. PROVIDE REUABLE PASSIVE DESIGN SELECTIONS 4 TO RETAIN RADIONUCLIDES WITHIN FUEL TO MEET 10CFR100 DOSES AT PLANT BOUNDARY ! 3. PROVIDE ADDITIONAL DESIGN SELECTIONS TO RETAIN j RADIONUCLIDES WITHIN THE PLANT TO MEET PAG DOSES I AT PLANT BOUNDARY L J vAxo:[uAuztyJio s' H-WAR-86 i
O . O O MHTGR MEETS 10CFR100 BY _ , _ RADIONUCLIDE RETENTION IN FUEL 3. D iiiUIiNY$fkIbNNE[E!
- 1 2 1 ENGTES.EE1 1.c,,,,, FUNCTIONS NEEDED
;;;,,,,0NucuoE RELEASES I
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!fddNI$$ONANEshI!! CONTROL TRANSPORT CONTROL TRANSPORT CONTROL TRANSPORT
! FROM CCRE IN PRIMARY CIRCulT FROM REACTOR BLDG FROM SITE i -- l 3.1.1.2.1.1 l 3.1.1.2.1.2 I
!!ii5YNiN'NN6idNUELION5!! RETAIN RADIONUCLlDES I ' . .I.N. . . .F.U. .E. .L. .P. .A. .R.IN.T. .I.C.GRAPHITE CORE .L.E. .S. . . ' H-558(1) 3-7-86 l
O O . O r m
.. .. .ma COMPARISON OF l-131 ACTIVITY REDUCTION REQUIREMENTS FOR THE MHTGR REQUIREMENT 10CFR100 (150 ren) PAG (5 rem)
A) ALLONABLE HUMAN UPTAKE 10-# Ci 3x10-8 Ci 7 7 B) CORE INVENTORY 10 Ci , 10 Ci Q i C) PERMITTED UPTAKE FRACTION (A/B) 10- 3x10-'3 'N L> ATTENUATION FACTORS D) DISTANCE 4x10-7 4x10-7 (425 M EAB) (425 M EPZ)
- E) RESIDUAL RELEASE 3x10-8 1x10-8 FRACTION PERMITTED (C / D)
(i.e., considering fuel retention, vessel retention, plateout,settIing,etc.) VAXD:[WAUZEYJ10 10 10-W Ak-86
O O O r m FUEL QUALITY LIMITS PRIMARY CIRCUlT ACTIVITY RELEASE FUEL FRACTION LIMITED RELEASE FROM FUEL PARTICLES DURING NORMAL OPERATION
- PLATE OUT ACTIVITY 1x10-5
- CIRCULATING ACTIVITY (10-8)
SUBTOTAL 1x10-* (meets 10CFR100)
$ OTHER LIFTOFF LIMITED BY LEAK AREA SIZE (41n Iine) .01 HELIUM BOUYANCY CAUSES ELEVATED RELEASE .03 TOTAL 3x10-8 (meets PAG)
L J VAXD:(WAUZEY]10 11 11-WAR-86
O O .O REQUIRED FUNCTIONS TO RETAIN _ _ _ _ RADIONUCLIDES IN FUEL PARTICLES b i RETAIN RADIONUCLIDES IN FUEL PARTICLES ,
-Q l
4 x k CONTROL REMOVE CONTROL CHEMICAL ATTACK CORE HEAT HEAT GENERATION l U"8? i
O O O r m a 4 DESIGN SELECTIONS ! TO REMOVE CORE HEAT gs
- ADDITIONAL TO ACTIVE HEAT REMOVAL SYSTEMS x (HTS and SCS)
(( e PASSIVE REACTOR CAVITY COOLING SYSTEM
- (RCCS) j - PRESSURIZED
- DEPRESSURIZED i
( ) VAXD:[WAUZEY]Io 2 8-MAR-86
O O .O r m PASSIVE CORE HEAT REMOVAL RESULTS i FROM MATERIAL CHARACTERISTICS l AND DESIGN FEATURES i l
- SMALL THERMAL RATING l - UMITS AMOUNT OF AF IENHEAT l g e CORE GEOMETRY l
1
- PROVIDES FOR REMOVAL OF AF IERHEAT BY PASSIVE l Q CONDUCTION AND RADIATION l
"
- PASSIVE HEAT SINK
! - UTILIZES REACTOR CAVITY C00UNG SYSTEM (RCCS) e SLOW HEAT UP OF MASSIVE GRAPHITE CORE I e HIGH TEMPERATURE STABILITY OF REACTOR CORE AND FUEL
- FISSION PRODUCTS RETAINED IN COATED PARTICLES L J l
VAXD:[WAUZEY]l0 12 8-WAR-86
O O O PRESSURIZED DECAY HEAT REMOVAL BY
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NATURAL CONVECTION AND RADIATION TO PASSIVE REACTOR CAVITY COOLING SYSTEM
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@ C0ATED FUEL PARTICLES MAINTAIN INTEGRITY TO VERY HIGH TEMPERATURES l
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! s 0.2 - PEAK FUEL g t OPERATING 0 I TEMPERATURE a e i i i M 88g 0 i 1000 1200 1400 1600 1800 2000 2200 2400 2600 ! TEMPERATURE (8C) i N-150(9) 9-4-84 l l _ _ _ _ - _ _ _ _
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O O l f 3 PASSIVE RETENTION DURING TEMPERATURE TRANSIENTS i RELEASE FUEL FRACTION LIMITED HEATUP RELEASE FROM 2x10-8 i FUEL PARTICLE COATINGS g' DELAYED RELEASE OF ACTIVITY (METEOROLOGY) .3 SUBTOTAL 6x10-6
'O (meets 10CFR100)
% OTHER
- Y I VESSEL HOLDUP AND RETENTION ONING .1 TO INHALATION AND DEPOSITION i
j HOLDUP AND RETENTION WITMiN .01 1 REACTOR BUILDING TOTAL 6x10-8 l (meets PAG)
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VAXD:[WAUZEY]t0 13 10-WAR-86 j
_ _ _ _ _ , , _ _y, _ ,,__ . ,,_ ,,... __._, ,,,. , . y__._% - % _ _. _ ,, _. O O O r m M SA Techne 4egfee M DESIGN SELECTIONS TO CONTROL CHEMICAL ATTACK BY WATER ( e LIMITED SOURCES (MAGNETIC BEARINGS) e HIGH QUALITY FUEL (LOW HYDROLYSIS) l e RELIABLE DETECTION AND ISOLATION (SINGE LOOP) VAXD:[WAUZEY]t0 3 11-WAR-86
O O O r m
. . ima RETENTION DURING WATER INGRESS 1
RELEASE FUEL FRACTIdN NO HYDROLISIS OF FUEL PARTlCLES %HICH DO NOT ALREADY HAVE FAILED COATINGS 2x10-# i I LIMITED RELEASE FROM PARTICLES WITH h' FAILED COATINGS .015 SUBTOTAL 3x10-6 D (meets 10CFR100) O OTHER
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HYDROLiSlS OF FUEL LIMITED BY AMOUNT l OF WATER %HICH ENTERS CORE .2 l RETENTION IN VESSEL AND AUXILIARY j BUILDINGS .1 1 TOTAL 6x10-a (meeisPAG) \ L J VAXD:{WAUZEY]10 4 II-WAR-86
O O . D h LIMITED AIR-GRAPHITE REACTION RETAINS RADIONUCLIDES IN CORE i 1 l 10-3 - S 5 ) t3 ~ 5 ec s [ 10 E _ i3 SQ. IN. LEAK I /' I b EE - g == / . g ,
/ 0.1 SQ. IN. LEAK _,
g 58 10-5
' ''l !' ' ' ' il i e i i'i' 1 2 3 10" 10 10 10 TIME (HRS) PAST i
I l ) PRIMARY COOLANT LEAK LEADING TO CONDUCTION C00LDOWN 1 j H-558(11) 3-7-86
O O . - O r 3 PASSIVE RETENTION DURING AIR INGRESS RELEASE 4 FUEL FRACTION OXIDATION RELEASE LIMITED TO ACTIVITY 2x10-5 HELD UP IN GRAPHITE AFTER CORE HEATUP e o DELAYED RELEASE OF ACTIVITY (METEOROLOGY) .3 h SUBTOTAL 6x10-6 (meeis 10CFR100) OTHER OXIDATION LIMITED BY PRIMARY COOLANT 1x10-4 i BOUNDARY BREACH AND REACTOR BUILDING TOTAL 6x10-'
- (meets PAG)
- k J VAXD
- [WAUZEY]10 21 10-W AR-86 l
O O - O r m
~
DESIGN SELECTIONS g TO CONTROL HEAT GENERATION
% e FUEL NEGATIVE TEMPERATURE COEFFICIENT ~ '
- CONTROL ROD SYST5M ~
e RESERVE SHUTDOWN SYSTEM
?
( ) i VAXD:[WAUZEY]10 6 11-WAR-86
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Z U . A - [ A W D X Y V A _ B K C S NIA S F OlA I I C T I T CL R A EA P O LC EI T L E SM E N A L U F L E NH O O S S L C I GC I C E E S M A S L G V E C S N E O, Y R E E N V C _ DR I T T I L U C A O D D T A U S E E N E D D D R Q D D O N H E l E B B E C O G I M i M M N H I L E E O e e e e e T F e O 4' b x4 1 4 l .iii! .!!!llll !
l O O O 1 r , l l CONCLUSIONS e HIGH QUAUTY FUEL UMITS NORMAL OPERATION RADIONUCUDE INVENTORIES IN PRIMARY SYSTEM TO LEVELS WITHIN 10CFR100 g,
@ 10CFR100 UMITS CAN BE MET FOR ALL EVENTS RELYING ON PASSIVE DESIGN SELECTIONS TO RETAIN RADIONUCUDES-p WITHIN FUEL l *
- PAG UMITS CAN BE MET WITH ADDITIONAL REUANCE ON i PASSIVE MECHANISMS EXTERNAL TO THE FUEL WITHOUT A CONVENTIONAL LEAK TIGHT CONTAINMENT l
j k J VAXD:[WAUZEY]to 14 ll-WAR-86 i
O O . O F 3
SUMMARY
e INHERENT CHARACTERISTICS AND PASSIVE FEATURES ASSURE RADIONUCLIDE RETENTION IN FUEL SUFFICIENT TO OBVIATE l % OFFSITE SHELTERING OR EVACUATION
) e PROGRAM UTluZES SYSTEMATIC, TRACEABLE UCENSING APPROACH SPECIFIC TO MODULAR HTGR CONSISTENT WITH l ADVANCED REACTOR POUCY i
l c > f VAXD:[WAUZEYJ10 22 11-W AR-86
9 & l 1 APPENDIX XIII l-NRC STAFF PRESEilTATION ON SAFETY G0AL POLICY !O i i i I I
- I l
l l 1 PRESENTATIONS TO THE ACRS l I - i J !
~
I ON l l i l SAFETY G0AL POLICY i e d i MARCH 13, 1986 t i-
?
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- g. > t 3
O l HISTORY OF PRA AND SAFETY GOALS e REACTOR SAFETY STUDY (WASH-1400) e- KEMENY COMMISSION AND NRC RESPONSE 4 e PUBLIC WORKSHOPS 1981 e PROPOSED SAFETY GOAL POLICY STATEMENT FOR PUBLIC COMMENT FEB. 1982 e FOUR PUBLIC MEETINGS 1982
,() .
e REVISED POLICY STATEMENT MARCH 1983 i e 2-YEAR EVALUATION PERIOD . 1 e STEEPING GROUP REPORT e ACRS AND STAFF COMMENTS / MEETINGS e FEBRUARY 14, 1986 STAFF PAPER O (7 p / 't i
STAFF PROPOSAL OF FEBRUARY 14, 1986 e ~ PURPOSE - ASSIST COMMISSION IN ITS DELIBERATIONS REGARDING THE BEST COURSE OF AGENCY ACTION e BELIEVE IT ACCOMMODATES ACRS COMMENTS AND ALL STAFF COMMENTS ON STEEPING GROUP REPORT o RECOMMEND ISSUANCE OF QUALITATIVE STATEMENTS IN FINAL FORM AS COMMISSION SAFETY GOAL POLICY e RECOMMEND AUTHORIZING STAFF TO USE INTEGRATED MATRIX ON A TRIAL BASIS AS A QUANTITATIVE MEASURE OF QUALITATIVE SAFETY , GOALS e ALL PROGRAM 0FFICE DIRECTORS (NRR, RES, IE, NMSS) SUPPORT THE FEBRUARY 14 PROPOSAL
~
i e STAFF PREPARED TO IMPLEMENT COMMISSION DECISION O p is 1 l
, - , , y- - - , v.,-.,. -.- , , , . , , , , , ,n ,---n. v. , ,n.,,-,.y,-. y. w.ev,n,,--n,e m --n g
O STAFF CONCLUSIONS e ISSUE IN FINAL FORM, AS THE COMMISSION'S SAFETY GOAL POLICY, TWO QUALITATIVE STATEMENTS REGARDING INDIVIDUAL AND SOCIET RISK. e COMBINE QUANTITATIVE OBJECTIVES (E.G., CORE-MELT, FREQUENCY, INDIVIDUAL MORTALITY RISKS) AND THE BENEFIT-COST GUIDELINES INTO AN INTEGRATED MATRIX WHICH THE STAFF CAN USE AS A QUANTITATIVE MEASUREMENT OF THE COMMISSION'S QUALITATIVE SAFETY GOALS. THESE SHOULD BE VIEWED AS INTEGRAL OBJECTIVES, , NOT INDIVIDUALLY DISCRETE (OR ISOLATED) STATEMENTS OF EXPE i A O
- g. ,1/f'
; O QUALITATIVE SAFETY GOALS 9 INDIVIDUAL RISK GOAL - INDIVIDUAL MEMBERS OF THE PUBLIC SHOULD BE PROVIDED A LEVEL OF PROTECTION FROM THE CONSEQUENCES OF NUCLEAR POWER PLANT OPERATION SUCH THAT INDIVIDUALS BEAR NO SIGNIFICANT ADDITIONAL RISK TO LIFE AND HEALTH 0 SOCIETAL RISK G0AL - SOCIETAL RISKS TO LIFE AND HEALTH FROM NUCLEAR POWER PLANT OPERATION SHOULD BE COMPARABLE TO OR LESS THAN THE RISKS OF GENERATING ELECTRICITY BY VIABLE COMPETING TECHNOLOGIES AND SHOULD NOT BE A SIGNIFICANT ADDITION TO OTHER SOCIETAL RISKS O
,9- 7 t ?
O KEY FEATURES OF SAFETY G0AL POLICY STATEMENT 8 TWO QUALITATIVE SAFETY G0ALS REMAIN UNCHANGED 8 THE QUANTITATIVE DESIGN OBJECTIVES AND BEN
- GUIDELINES TREATED INTEGRALLY NOT AS INDIVID DISCRETE STATEMENTS OF EXPECTATION i e OVALITATIVE SAFETY G0ALS INCORPORATED INTO TH REGULATORY DECISIONMAKING PROCESS THROUGH AN INTEGRATED SAFETY GOAL MATRIX l ,
! 9 MATRIX COMPRISED OF QUANTITATIVE PARAMETERS F ~
FREQUENCY OF CORE-MELT ACCIDENTS, PROMPT AND LATENT f , l I CANCER FATALITY RISKS AND A BENEFIT-COST GU i , i I I I iO
+ xP j i
__ ,._____,_,__..,_.,_...r _........_,__..,_,__-.m.__.__,., ..
KEY FEATURES OF SAFETY G0AL POLICY t POLICY STATEMENT INDICATES SAFETY G0ALS ARE NOT A SUBSTITUTE FOR EXISTING REGULATIONS l
- 0 POLICY STATEMENT EXHORTS ACHIEVING A MEAN CORE-MELT FREQUENCY OF LESS THAN 1 IN 10,000 PER REACTOR YEAR FOR CURRENT PLANTS IN ACCORD WITH 7/85 ACRS RECOMMENDATIO S THE POLICY STATEMENT INDICATES THAT FUTURE PLANTS WO BE EXPECTED TO ACHIEVE A CORE-MELT FREQUENCY LOWER T CURRENT PLANTS 7
i e POLICY STATEMENT INDICATES THAT THE STAFF SHOULD
- ATTEMPT TO ENSURE THAT CORE-MELT FREQUENCY AND i
MORTALITY RISKS ARE NOT DRIVEN BY A SINGLE ACCIDENT SEQUENCE (MANAGEMENT OF UNCERTAINTIES) i lO M-?/f
1 KEY FEATURES OF SAFETY GOAL POLICY 0 POLICY STATEMENT, THROUGH THE INTEGRATED MATRIX, ESTABLISHES A RISK STATE THAT WOULD BE VIEWED AS THE DE MINIMIS LEVEL FOR ACCIDENT RISKS AND ALSO A RISK STATE ABOVE WHICH A FACILITY SHOULD HAVE'A SENSE OF URGENCY FOR SAFETY IMPROVEMENTS OR SHUTDOWN. O POLICY STATEMENT, THROUGH THE INTEGRATED MATRIX,
- INTRODUCES AN ALARA APPROACH TO AID STAFF SAFETY DECISIONS O BENEFIT-COST GUIDELINE EMBODIED IN MATRIX EMPHASIZES THAT ACHIEVEMENT OF THE MORTALITY RISK OBJECTIVES IS THE PRIORITY CONSIDERATION IN THE SAFETY G0AL CONCEPT (PUBLIC HEALTH AND SAFETY EMPHASIS)
O BENEFIT-COST GUIDELINE EMPHASIZES A SLIDING EXPENDITURE SCALE TIED TO CORE-MELT FREQUENCY AND DRIVES TOWARDS A PEDUCTION IN CORE-MELT FREQUENCY (ACCIDENT' PREVENTION EMPHASIS) 0 POLICY STATEMENT RECOGNIZES THAT MAINTENANCE, OPERATIONS AND MANAGEMENT MAY HAVE A SIGNIFICANT IMPACT i ON THE STATE OF PLANT RISK (RISK MANAGEMENT EMPHASIS) O
,9. a s
KEY FEATURES OF SAFETY G0AL POLICY t POLICY STATEMENT INDICATES THAT WE ARE ABOUT READY SAFETY G0ALS (MATRIX) IN THE REVIEW 0F GENERIC SAFETY j REQUIREMENTS - STILL NEED TO DEVELOP METHODOLOGY FOR - ACCOUNTING FOR INTEGRATED SAFETY IMPROVEMENT IMPACT CORE-MELT FREQUENCY AND DEVELOP AN AGGREGATE COST SYSTEM, OTHER AREAS WILL BE IMPLEMENTED GRADUALLY
- 0 POLICY STATEMENT INDICATES THAT USE OF THE SAFETY G0AL , INTEGRATED MATRIX WILL BE USED IN CONJUNCTION WITH DE l MINISTIC DECISIONMAKING AND WILL BE ONE FACTOR USED I DECISION PROCESS 0 POLICY STATEMENT RECOGNIZES THAT UNCERTAINTIES MUS j
1, INTO ACCOUNT IN THE REGULATORY DECISIONMAKING PROCESS k - t POLICY STATEMENT, THROUGH THE INTEGRATED MATRIX, EMPHASIZES [ i THE NEED TO STRIVE FOR SAFETY IMPROVEMENTS WHICH ARE EFFECTIVE IN TERMS OF SAFETY AND COSTS t POLICY STATEMENT EXPRESSES FUTURE INTENT TO ESTABLISH
)
QUANTITATIVE OBJECTIVES TO GUIDE CONTAINMENT DESIGN AN PERFORMANCE, STAFF SHOULD PROPOSE THEM BY EARLY FY '87 4 INCLUDES INTERNAL AND EXTERNAL INITIATORS f N/
O PARAMETERS REFLECTED IN MATRIX e HEALTH EFFECTS PROMPT FATALITY RISK - THE RISK TO AN AVERAGE INDIVIDUAL IN THE VICINITY OF A NUCLEAR POWER PLANT OF PROMPT FATALITIES THAT MIGHT RESULT FROM REACTOR ACCIDENTS DOES NOT EXCEED ONE-TENTH OF ONE PERCENT (0.1%) 0F THE SUM OF PROMPT FATALITY RISKS RESULTING FROM OTHER ACCIDENTS TO WHICH MEMBERS OF THE U.S. POPULATION ARE GENERALLY EXPOSED. LATENT CANCER FATALITY RISK - THE RISK TO THE POPULATION IN THE AREA NEAR A NUCLEAR POWER PLANT OF CANCER FATALITIES THAT MIGHT RESULT FROM NUCLEAR POWER PLANT O OPERATION DOES NOT EXCEED ONE-TENTH OF ONE PEPCENT (0.1%) 0F THE SUM OF CANCER FATALITY RISKS RESULTING l FROM ALL OTHER CAUSES. O LARGE-SCALE CORE-MELT FREQUENCY - THE LIKELIHOOD OF A NUCLEAR j REACTOR ACCIDENT THAT RESULTS IN A LARGE-SCALE CORE-MELT IS , j ALLOWED TO VARY. THE IMPETUS TO IMPROVE THE CORE-MELT FREQUENCY
- IS DEPENDENT UPON HOW MUCH THE FREQUENCY IS GREATER THAN 10-5/RY AND THE STATUS OF THE HEALTH EFFECTS.
O BENEFIT-COST GUIDELINE - THE BENEFIT OF AN INCREMENTAL REDUCTION OF MORTALITY RISKS AND IN THE FREQUENCY OF LARGE-SCALE CORE-MELT ACCIDENTS IS COMPARED WITH THE ASSOCIATED i COSTS ON THE BASIS OF $1,000 PER PERSON-REM AVERTED PLUS A i PERCENTAGE OF THE ONSITE RADIOLOGICAL, INCLUDING ECONOMIC, i COSTS AVERTED. i O
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l O . l COST CALCULATIONS USING SAFETY GOAL INTEGRATED MATRIX (Sensitivity of Percentage of Onsite Cost , Versus Core Melt Frequency) l Palisades Site (1,100 MWs Reactor) 108 . , l l r I I Onsite costs calculated using t
)Percentages Proposed multiplied safety goal l Offstte cost plus. by matrix
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/ / q plus offsite cost j
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t J COST CALCULATIONS USING Q SAFETY GOAL INTEGRATED MATRIX ; (Sensitivity To Site Population) Health effect goals are not met Health effect goals are met 10' - a 1 3 b 8 High
$ 30 h
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- Site 8
l l g 10s l l l r 3 x 10-8 3x104 1 x 10 8 PROBABILITY OF CORE MELT (Per RY)
- g. ?23
TREn1reN1 0e uNCeRraiNTies O e UNCERTAINTIES NOT GENERATED BY PRA METHODOLOGY e PEER REVIEW 0F PRA RESULTS e UPPEL BOUNDS AND MID RANGE VALUES CAN BE INFERR EXPERIENCE e USE OF MEAN VALUE INTRODUCES A SKEWNESS TOWARD p CONFIDENCE REGION 4 h t e US.E WITH UNDERSTANDING OF MAGNITUDE OF UNCERTA INVOLVED - LOOK TO ROOT CAUSE CONTRIBUTORS e DISPLAY MEAN, MEDIAN AND CONFIDENCE LEVEL RANGES i
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' APPENDIX XIV REPORT ON VISIT TO THREE MILE ISLAND UN O
REPORT ON VISIT TO THREE MILE ISLAND UNIT 2 March 7, 1986 I. Introduction On Friday, March 7,1986, I spent the day visiting with NRC Staff and GPU officials at the Three Mile Island Unit 2 plant in Middletown, PA. People with whom I had discussions included: Mr. Curtis J. Cowgill, i TMI-2 Project Section Leader U. S. Nuclear Regulatory Commission Mr. Frank Standefer Vice-President / Director, TMI-2 GPU Nuclear Corporation Mr. Michael Slobedein p Radiation Protection Manager ( Three Mile Island Nuclear Station GPU Nuclear Corporation The principal topics covered were the cleanup of TMI Unit 2 including removal of the fuel debris from the primary system; decontamination of the contain-ment and other buildings; and the tests being conducted to determine the extent of the contamination and requirements for cleanup. II. Removal of Core Debris
, The work of removing the fuel and other debris from the reactor pressure vessel appears to be well underway. As of March 7,1986, some 22 canisters had been filled and removed, representing some 22,000 pounds of material, and removal operations are continuing essentially around the clock. Included in the tour was a visit to the Defueling Test Assembly, used for training the plant staff in the defueling operations, and to the pool for storing the filled canisters prior to preparing them for shipment offsite.
Since the damaged reactor core includes objects having a variety of sizes and shapes, a number of tools have been developed for cutting, picking up and removing the debris (See Figures 1 through 17). In addition, special facil-ities have been built above the reactor pressure vessel to help guide the l l
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s RPT/TMI-1 VISIT /3/7/86 2 operators in handling various fuel debris removal tools and in loading the material into the canisters (See Figures A through D). One of the main problems in removing the debris has been the lack of clarity of the water in the reactor pressure vessel. The problem is due to a heavy growth of microbial organisms. These microbes, which are being studied by specialists, include both anaerobic and aerobic organisms and indications are that it is a " flourishing colony." Although chlorination might be one method for control, the Plant staff wants to avoid the introduction of chlorides into the primary coolant system. Attempt to control the organisms by the introduction of ~a nitrogen atmosphere quickly demonstrated that nitrogen could serve as a food for them. Data also show that the organisms can feed on other materials, such as the hydraulic fluids that leak from machinery used near the core. Studies to determine a method for controlling the organisms are continuing. III. Decontamination Operations Work on the decontamination of critical areas within containment appears to be progressing well. Although exposure rates in the flooded areas of the basement still range in the hundreds of R/hr, exposure rates in most of the working areas are now down to several mR/hr. (See Figures E through G). This has been accomplished by a combination of hydrolazing scabbling, , shielding, and other approaches. Because of the success of these efforts, plant officials now project that they will be able to accomplish the cleanup of TMI Unit 2 with collective occupational doses less than the estimates given in the most recent Environmental Impact Assessment. The scabbling operation proved particularly interesting in that it is done with machines that can pound and grin up to 3/16 inch off of a concrete surface in one pass. The scabbler is air operated with the removed debris being carried via a " vacuum" system into a barrel, and the air being subsequently exhausted through a high efficiency filter. Once a surface is satisfactorily decontaminated, it is sealed (coated) with epoxy resin paint. Surfaces that were exposed to steam and moisture appear not to be contaminated to a depth of more than perhaps a half an inch or so. Surfaces within the basement, however, which were flooded, may be contaminated extensively, versus concrete depending)on the porosity blocks and whether of the material the surface (poured had previously concrete been painted. To assess the extent of the problem in the flooded basement, and the applicability of various decontamination methods, core samples of the concrete are currently being obtained (using remotely operated machines) and O
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RPT/TMI-1 VISIT /3/7/86 3 Oi analyz'e f The GPU Staff promised to provide some of the data from these tests to us within the nexts few weeks. IV. Commentary In the way of commentary, I offer the following observations:
- 1. Morale at the plant (TMI-2) appears good. One example was a sign that has been erected onsite to maintain a running tabulation of the amount of core debris removed, versus the total anticipated to be in need of removal. In addition, it might be noted that the Plant Staff clearly understands that the current goal is to clean up the plant and they see this as within the scope of their capabilities.
- 2. In touring the facility, one cannot help but be impressed by the tremendous size and obvious cost of the plant which (since it had only been in operation about one year) is essentially in "new" condition. The Plant Staff appears to take pride in keeping the undamaged portions in a clean condition, and in the progress they are making in reclaiming portions of the plant that were formerly contaminated.
l
- 3. The Plant Staff (if my impressions were correct) is beginning now A to believe that less of the fuel from the core was transported
() outside the reactcr pressure vessel than originally thought. If this is the case, it may facilitate plant cleanup.
- 4. Although it is a minor point, it is significant that surfaces within TMI-2 that were painted prior to the accident appear to have been far more easily decontaminated than those that were not. This might be kept in mind in building and maintaining such facilities in the future.
- 5. I was particularly impressed by the onsite NRC Staff. They appear to be tough and firm, but reasonable, in their dealings with the Licensee. Although there is some degree of an adversarial rela-tionship, it is obvious that communications between the two groups are good and that they are meeting and resolving problems as they develop, and doing so in a constructive manner.
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- NOTE 1: The given tool length, T, includes one (1) 15 ft. pole section and two (2) l 7 ft. pole sections. If a 7 ft. pole section is removed, 7 ft. shall be added to the required minimum rigging length, R.
NOTE 2: The minimum rigging length (R) is based on maintaining the 4 ft. exclusion zone. This length can be reduced up to 3 ft. to allow for larger size debris loading into fuel canisters. When the shorter rigging length is used. Radiological Controls monitoring must be in progress when the debris enters the 4 ft. exclusion zone. For determining when the debris enters the 4 ft. exclusion zone, the lower 4 ft. of the End Effector Handling Tool (Figure 1) is painted red. (Which roughly corresponds to the 323 ft. i marking on the End Effector Handling Tool aligning with the handrall). 5 1 2a l
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k g##*%, UNITED : e '$ NUCLEAR REGULN APPENDIX XV 2 REVISED SUBCOMMITTEE ASSIGNMENTS Is i ADVlsORY COMMITTEE OP t WASHINGTOP
~h (V %,...,*/ March 14 MEMORANDUM FOR: ACRS Members NY I FROM; D. Ward, Ch irman
SUBJECT:
REVISED SU8 COMMITTEE ASSIGNMENTS i Enclosed is the revised list of Subcomittee Assignments. Please consider your assignments and provide any coments you may have to me or Mort Libarkin as soon as possible. Subcomittee Chairmen should also consider the listing of specific tasks for each Subcomittee and suggest additions or deletions within 2 weeks. I would like to make these assignments final by April 1,1986. Finally, please note that this listing does not include any changes ir consul-tant assignments in support of Subcommittees. Prospective Chairmen are urged to consider these assignments also and modify them as appropriate. O O
Enclosure:
As stated cc: R. Fraley ACRS Staff ACRS Fellows 1 l l U f- )h
03/13/86
, SUBCOMMITTEE ASSIGNMENTS MAX W CARBON CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES:
STAFF ENGINEER BELLEFONTE PLANT UNITS 1 & 2 AJC ADVANCED NON-LWR DESIGNS MME LONG RANGE PLAN FOR THE NRC , RKM MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN COMBUSTION ENGINEERING REACTOR PLANTS WYLIE , DAVIS-BESSE STATION UNIT 1 REMICK FERMI, ENRICO UNIT 2 KERR SEABROOK NUCLEAR PLANT UNITS 1 & 2 KERR WATERFORD STATION UNIT 3 WARD ADVANCED LWR DESIGNS OKRENT EXTREME EXTERNAL PHENOMENA OKRENT REGIONAL OPERATIONS REMICK PROBABILISTIC RISK ASSESSMENT OKRENT SAFEGUARDS AND SECURITY MARK ' SAFETY RESEARCH PROGRAM SIESS WASTE MANAGEMENT M0ELLER SEVERE (CLASS 9) ACCIDENTS KERR , STANDARD PLANT DESIGN WYLIE l l O V if J Y)
03/13/86 SUBCOMMITTEE ASSIGNMENTS /N FOR d JESSE C. EBER50LE CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: STAFF ENGINEER CATAWBA NUCLEAR STATION UNITS 1 & 2 , JOS PALO VERDE UNITS 1-3 MDH V0GTLE, ALVIN W. PLANT UNIls 1 & 2 JOS WATTS BAR UNITS 1 & 2 AJC INSTRUMENTATION AND CONTPOL SYSTEMS MME REACTOR OPERATIONS HA SECONDARY SYSTEMS MME MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN BABC0CK & WILC0X WATER REACTOR PLANTS WYL'IE BEAVER VALLEY POWER STATION UNIT 2 WYLIE DAVIS-BESSE STATION UNIT 1 REMICK DIABLO CANYON UNITS 1 & 2 SIESS GENERAL ELECTRIC REACTOR PLANTS OKRENT GRAND GULF NUCLEAR STATION UNITS 1 & 2 OKRENT MIDLAND PLANT UNITS 1 & 2 OKRENT NINE MILE POINT UNIT 2 SHEWMON RIVER BEND UNITS 1 & 2 OKRENT p SUSQUEHANNA STATION UNIT 2 REMICK Q WPPSS NO. 2 WESTINGHOUSE REACTOR PLANTS MARK REED ADVANCED LWR DESIGNS OKRENT AUXILIARY SYSTEMS MICHELSON THERMAL HYDRAULIC PHENOMFNA MICHELSON GENERIC ITEMS SIESS HUMAN FACTORS REMICK MAINTENANCE PRACTICES & PROCEDURES REED PROCEDURES & ADMINISTRATION WARD SAFEGUARDS AND SECURITY MARK STRUCTURAL ENGINEERING SIESS AC/DC POWER SYSTEMS RELIABILITY KERR SCRAM SYSTEMS RELIABILITY KERR ! DECAY HEAT REMOVAL SYSTEMS WARD . I RELIABILITY ASSURANCE WYLIE l GAS COOLED REACTOR l'LANTS SIESS SYSTEMATIC ASSESSMENT OF EXPERIENCE LEWIS TRANSPORTATION OF RADI0 ACTIVE MATERIALS SIESS STANDARD PLANT DESIGN WYLIE NUCLEAR PLANT CHEMISTRY M0ELLER CONTAINMENT REQUIREMENTS MARK INSPECTION AND ENFORCEMENT PROGRAMS REED PLANT OPERATING PROCEDURES MICHELSON O l
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1 03/13/86 SUBCOMMITTEE ASSIGNMENTS l FOR HAROLD ETHERINGTON CHAIRMAN OF THE F0LLOWING SUBCOMMITTEES: 1 STAFF ENGINEER , MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN i THREE MILE ISLAND UNITS 1 & 2 M0ELLER EXTREME EXTERNAL PHENOMENA .0KRENT THERMAL HYDRAULIC PHENOMENA MICHELSON METAL COMPONENTS SHEWMON DECAY HEAT REMOVAL SYSTEMS WARD NUCLEAR PLANT CHEMISTRY M0ELLER l l !O ) I i I I i I \ U i 1 g ,Ly9 I
I 03/13/86 SUBCOMMITTEE ASSIGNMENTS 4
'T FOR '
) WILLIAM KERR !
CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: ! STAFF ENGINEER CLINTON STATION UNITS 1 & 2 PAB FERMI, ENRICO UNIT 2 PAB LASALLE COUNTY STATION EGI , LIMERICK UNITS 1 & 2 RPS , NAVAL REACTORS PAB SEABROOK NUCLEAR PLANT UNITS 1 & 2 RKM CORE PERFORMANCE MDH AC/DC POWER SYSTEMS RELIABILITY MME SCRAM SYSTEMS RELIABILITY PAB SEVERE (CLASS 9) ACCIDENTS MDH STATE OF NUCLEAR POWER SAFETY AJC MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN BABC0CK & WILC0X WATER REACTOR PLANTS WYLIE BEAVER VALLEY POWER STATION UNIT 2 WYLIE BELLEFONTE PLANT UNITS 1 & 2 CARBON CATAWBA NUCLEAR STATION UNITS 1 & 2 EBERSOLE GENERAL ELECTRIC REACTOR PLANTS OKRENT
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(,,, HOPE CREEK UNIT 1 MILLSTONE UNIT 3 SIESS SHEWMON PALO VERDE UNITS 1-3 EBERS0LE SUSQUEHANNA STATION UNIT 2 REMICK THREE MILE ISLAND UNITS 1 & 2 M0ELLER WESTINGHOUSE REACTOR PLANTS REED INSTRUMENTATION AND CONTROL SYSTEMS EBERS0LE THERMAL HYDRAULIC PHENOMENA MICHELSON FUEL CYCLE SHEWMON HUMAN FACTORS REMICK FUEL PERFORMANCE SHEWMON PROBABILISTIC RISK ASSESSMENT OKRENT j SAFETY PHILOSOPHY, TECHNOLOGY, AND CRITERIA OKRENT I SAFETY RESEARCH PROGRAM SIESS DECAY HEAT REMOVAL SYSTEMS WARD GAS COOLED REACTOR PLANTS SIESS i SYSTEMATIC ASSESSMENT OF EXPERIENCE LEWIS { SPENT FUEL STORAGE SIESS STANDARD PLANT DESIGN WYLIE CONTAINMENT REQUIREMENTS MARK INSPECTION AND ENFORCEMENT PROGRAMS REED PLANT OPERATING PROCEDURES MICHELSON UNIVERSITY TEST REACTORS MARK l J l
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03/13/86 SUBCOMMITTEE ASSIGNMENTS \ /" FOR i HAROLD W. LEWIS CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: STAFF ENGINEER SAN ONOFRE UNIT 3 MDH . REGULATORY POLICIES AND PRACTICES AJC COMMITTEE ACTIVITIES MWL SYSTEMATIC ASSESSMENT OF EXPERIENCE EKM MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN COMBUSTION ENGINEERING REACTOR PLANTS WYLIE DIABLO CANYON UNITS 1 & 2 SIESS LIMERICK UNITS 1 & 2 KERR NAVAL REACTORS KERR PALO VERDE UNITS 1-3 EBERS0LE SEABROOK NUCLEAR PLANT UNITS 1 & 2 KERR SOUTH TEXAS 1 & 2 MARK INSTRUMENTATION AND CONTROL SYSTEMS EBERSOLE EXTREME EXTERNAL PHENOMENA OKRENT METAL COMP 0NENTS SHEWMON PROCEDURES & ADMINISTRATION WARD REACTOR OPERATIONS EBER50LE PROBABILISTIC RISK ASSESSMENT OKRENT ( SAFETY PHILOSOPHY, TECHNOLOGY, AND CRITERIA OKRENT KERR AC/DC POWER SYSTEMS RELIABILITY SCRAM SYSTEMS RELIABILITY KERR LONG RANGE PLAN FOR THE NRC CARBON STATE OF NUCLEAR POWER SAFETY KERR
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03/13/86 SUBCOMMITTEE ASSIGNMENTS FOR O J. CARSON MARK CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: STAFF ENGINEER BRAIDWOO9 STATIONS EGI BYRON STATIONS EGI SEQUOYAH PLANT UNITS 1 & 2 AJC SOUTH TEXAS 1 & 2 MME WPPSS N0. 2 PAB SAFEGUARDS AND SECURITY JOS CONTAINMENT REQUIREMENTS MDH UNIVERSITY TEST REACTORS EGI MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN LASALLE COUNTY STATION KERR NINE MILE POINT UNIT 2 SHEWMON SUSQUEHANNA STATION UNIT 2 REMICK ADVANCED NON-LWR DESIGNS CARBON CORE PERFORMANCE KERR WASTE MANAGEMENT M0ELLER SEVERE (CLASS 9) ACCIDENTS KERR SYSTEMATIC ASSESSMENT OF EXPERIENCE LEWIS q TRANSPORTATION OF RADI0 ACTIVE MATERIALS SIESS b/ OCCUPATIONAL AND ENVIRONMENTAL PROTECTION SYSTEMS M0ELLER O i l g, J .$ - l 4-.- ,, , _ - , . , . , , e - --n---n ea-- e,- - g -w-- r ---w,-
03/13/86 SUBCOMMITTEE ASSIGNMENTS FOR O' CARLYLE MICHELSON CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: STAFF ENGINEER PERRY PLANT UNITS 1 & 2 PAB AUXILIARY SYSTEMS SD THERMAL HYDRAULIC PHENOMENA PAS PLANT OPERATING PROCEDURES JOS MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN BABC0CL & WILC0X WATER REACTOR PLANTS WYLIE CATAWBA NUCLEAR STATION UNITS 1 & 2 EBERSOLE COMBUSTION ENGINEERING REACTOR PLANTS WYLIE COMANCHE PEAK UNITS 1 & 2 OKRENT GENERAL ELECTRIC REACTOR PLANTS OKRENT LIMERICK UNITS 1 & 2 KERR MIDLAND PLANT UNITS 1 & 2 OKRENT SEABROOK NUCLEAR PLANT UNITS 1 & 2 KERR SHOREHAM STATION SIESS WESTINGHOUSE REACTOR PLANTS REED ADVANCED LWR DESIGNS OKRENT CORE PERFORMANCE KERR INSTRUMENTATION AND CONTROL SYSTEMS EBERS0LE U'O GENERIC ITEMS SIESS HUMAN FACTORS REMICK MAINTENANCE PRACTICES & PROCEDURES REED METAL COMPONENTS SHEWMON QUALITY & QUALITY ASSURANCE IN DESIGN AND CONSTRUCTION REED FUEL PERFORMANCE SHEWMON REACTOR OPERATIONS EBERSOLE REGIONAL OPERATIONS REMICK REGULATORY ACTIVITIES SIESS REGULATORY POLICIES AND PRACTICES LEWIS SAFEGUARDS AND SECURITY MARK SAFETY PHILOSOPHY, TECHNOLOGY, AND CRITERIA OKRENT SECONDARY SYSTEMS EBERS0LE COMMITTEE ACTIVITIES LEWIS DECAY HEAT REMOVAL SYSTEMS WARD RELIABILITY ASSURANCE WYLIE SYSTEMATIC ASSESSMENT OF EXPERIENCE LEWIS STATE OF NUCLEAR POWER SAFETY KERR STANDARD PLANT DESIGN WYLIE INSPECTION AND ENFORCEMENT PROGRAMS REED UNIVERSITY TEST REACTORS MARK l g a s3
03/13/86 SUBCOMM?.TTEE ASSIGNMENTS - n FOR CADE W. MOELLER , CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: STAFF ENGINEER THREE MILE ISLAND UNITS 1 & 2 RKM WASTE MANAGEMENT OSM NUCLEAR PLANT CHEMISTRY HA OCCUPATIONAL AND ENVIRONMENTAL PROTECTION SYSTEMS JOS MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN BRAIDWOOD STATIONS MARK BYRON STATIONS MARK CATAWBA NUCLEAR STATION UNITS 1 & 2 EBERSOLE CLINTON STATION UNITS 1 & 2 KERR MILLSTONE UNIT 3 SHEWMON SEABROOK NUCLEAR PLANT UNITS 1 & 2 KERR EXTREME EXTERNAL PHENOMENA OKRENT FUEL CYCLE SHEWMON MAINTENANCE PRACTICES & PROCEDURES REED REACTOR OPERATIONS EBERS0LE REGULATORY POLICIES AND PRACTICES LEWIS SAFETY RESEARCH PROGRAM SIESS w LONG RANGE PLAN FOR THE NRC CARBON SYSTEMATIC ASSESSMENT OF EXPERIENCE LEWIS SPENT FUEL STORAGE SIESS STATE OF NUCLEAR POWER SAFETY KERR TRANSPORTATION OF RADI0 ACTIVE MATERIALS SIESS INSPECTION AND ENFORCEMENT PROGRAMS REED l \ __
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03/13/86 SUBCOMMITTEE ASSIGNMENTS FOR fO DAVID CKRENT CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: STAFF ENGINEER COMANCHE PEAK UNITS 1 & 2 HA GENERAL ELECTRIC REACTOR PLANTS RKM GRAND GULF NUCLEAR STATION UNITS 1 & 2 HA MIDLAND PLANT UNITS 1 & 2 JOS RIVER BEND UNITS 1 & 2 RPS ADVANCED LWR DESIGNS MME EXTREME EXTERNAL PHENOMENA RPS PROBABILISTIC RISK ASSESSMENT RPS SAFETY PHILOSOPHY, TECHNOLOGY, AND CRITERIA RPS MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN l BABC0CK & WILC0X WATER REACTOR PLANTS WYLIE CARBON l BELLEFONTE PLANT UNITS 1 & 2 BROWNS FERRY UNITS 1-3 WYLIE DAVIS-BESSE STATION UNIT 1 REMICK DIABLO CANYON UNITS 1 & 2 SIESS HOPE CREEK UNIT 1 SIESS LASALLE COUNTY STATION KERR PERRY PLANT UNITS 1 & 2 MICHELSON k\ SAN ONOFRE UNIT 3 LEWIS THREE MILE ISLAND UNITS 1 & 2 M0ELLER V0GTLE, ALVIN W. PLANT UNITS 1 & 2 EBERSOLE WATTS BAR UNITS 1 & 2 EBERSOLE ADVANCED NON-LWR DESIGNS CARBON CORE PERFORMANCE KERR METAL COMPONENTS SHEWMON l REGULATORY POLICIES AND PRACTICES LEWIS SAFETY RESEARCH PROGRAM SIESS l SECONDARY SYSTEMS EBERS0LE STRUCTURAL ENGINEERING SIESS SEVERE (CLASS 9) ACCIDENTS KERR STATE OF NUCLEAR POWER SAFETY KERR CONTAINMENT REQUIREMENTS MARK UNIVERSITY TEST REACTORS MARK l l I 1 O V
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03/13/86 SUBCOMMITTEE ASSIGNMENTS i FOR l' ( GLENN A. REED CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: STAFF ENGINEER WESTINGHOUSE REACTOR PLANTS AJC MAINTENANCE PRACTICES & PROCEDURES HA - QUALITY & QUALITY ASSURANCE IN DESIGN AND CONSTRUCTION EGI INSPECTION AND ENFORCEMENT PROGRAMS AJC MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN BABC0CK & WILCOX WATER REACTOR PLANTS WYLIE BELLEFONTE PLANT UNITS 1 & 2 CARBON COMBUSTION ENGINEERING REACTOR PLANTS WYLIE NAVAL REACTORS KERR V0GTLE, ALVIN W. PLANT UNITS 1 & 2 EBERSOLE ADVANCED LWR DESIGNS OKRENT AUXILIARY SYSTEMS MICHELSON THERMAL HYDRAULIC PHENOMENA MICHELSON FUEL CYCLE SHEWMON GENERIC ITEMS SIESS HUMAN FACTORS REMICK FUEL PERFORMANCE SHEWMON I-) REACTOR OPERATIONS REGIONAL OPERATIONS SAFEGUARDS AND SECURITY EBERS0LE REMICK MARK SECONDARY SYSTEMS EBERS0LE AC/DC POWER SYSTEMS RELIABILITY KERR SCRAM SYSTEMS RELIABILITY KERR DECAY HEAT REMOVAL SYSTEMS WARD STATE OF NUCLEAR POWER SAFETY KERR NUCLEAR PLANT CHEMISTRY M0ELLER OCCUPATIONAL AND ENVIRONMENTAL PROTECTION SYSTEMS M0ELLER PLANT OPERATING PROCEDURES MICHELSON O 4 2s6
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03/13/86 SUBCOMMITTEE ASSIGNMENTS () FORREST CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: REMICK STAFF ENGINEER DAVIS-BESSE STATION UNIT 1 HA SUSQUEHANNA STATION UNIT 2 EGI HUMAN FACTORS JOS REGIONAL OPERATIONS AJC MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN
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BROWNS FERRY UNITS 1-3 WYLIE COMBUSTION ENGINEERING REACTOR PLANTS WYLIE GENERAL ELECTRIC REACTOR PLANTS OKRENT MIDLAND PLANT UNITS 1 & 2 OKRENT NAVAL REACTORS KERR FROCEDURES & ADMINISTRATION WARD QUALITY & QUALITY ASSURANCE IN DESIGN AND CONSTRUCTION REED REACTOR OPERATIONS EBERSOLE REGULATORY ACTIVITIES SIESS PROBABILISTIC RISK ASSESSMENT OKRENT SAFEGUARDS AND SECURITY MARK SAFETY PHILOSOPHY, TECHNOLOGY, AND CRITERIA OKRENT g WASTE MANAGEMENT M0ELLER 5g COMMITTEE ACTIVITIES LONG RANGE PLAN FOR THE NRC LEWIS CARBON SPENT FUEL STORAGE SIESS OCCUPATIONAL AND ENVIRONMENTAL PROTECTION SYSTEMS MOELLER PLANT OPERATING PROCEDURES MICHELSON UNIVERSITY TEST REACTORS MARK i I O g 2n
03/13/86 I SUSCOMMITTEE ASSIGNMENTS O FOR L' PAUL G. SHEWMON CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: STAFF ENGINEER MILLSTONE UNIT 3 JOS NINE MILE POINT UNIT 2 JOS FUEL CYCLE HA METAL COMPONENTS EGI FUEL PERFORMANCE HA MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN BEAVER VALLEY POWER STATION UNIT 2 WYLIE BRAIDWOOD STATIONS MARK BYRON STATIONS MARK FERMI, ENRICO UNIT 2 KERR GENERAL ELECTRIC REACTOR PLANTS OKRENT RIVER BEND UNITS 1 & 2 OKRENT WESTINGHOUSE REACTOR PLANTS REED CORE PERFORMANCE KERR AUXILIARY SYSTEMS MICHELSON SAFETY RESEARCH PROGRAM SIESS STRUCTURAL ENGINEERING SIESS WASTE MANAGEMENT M0ELLER O. SEVERE (CLASS 9) ACCIDENTS GAS COOLED REACTOR PLANTS KERR SIESS SPENT FUEL STORAGE SIESS TRANSPORTATION OF RADI0 ACTIVE MATERIALS SIESS NUCLEAR PLANT CHEMISTRY M0ELLER O g.2sr i
03/13/86 SUBCOMMITTEE ASSIGNMENTS FOR O CHESTER P. SIESS CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: STAFF ENGINEER DIABLO CANYON UNITS 1 & 2 EGI HOPE CREEK UNIT 1 MME SliOREHAM STATION HA GENERIC ITEMS SD REGULATORY ACTIVITIES SD SAFETY RESEARCH PROGRAM SD STRUCTURAL ENGINEERING EGI GAS COOLED REACTOR PLANTS JCM SPENT FUEL STORAGE HA TRANSPORTATION OF RADI0 ACTIVE MATERIALS SD MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN DAVIS-BESSE STATION UNIT 1 REMICK GRAND GULF NUCLEAR STATION UNITS 1 & 2 OKRENT LIMERICK UNITS 1 & 2 KERR MILLSTONE UNIT 3 SHEWMON NINE MILE POINT UNIT 2 SHEWMON SAN ONOFRE UNIT 3 LEWIS SEQUOYAH PLANT UNITS 1 & 2 MARK n" SOUTH TEXAS 1 & 2 WATERFORD STATION UNIT 3 MARK WARD ADVANCED NON-LWR DESIGNS CARBON EXTREME EXTERNAL PHENOMENA OKRENT PROCEDURES & ADMINISTRATION WARD QUALITY & QUALITY ASSURANCE IN DESIGN AND CONSTRUCTION REED REGULATORY POLICIES AND PRACTICES LEWIS PROBABILISTIC RISK ASSESSMENT OKRENT SEVERE (CLASS 9) ACCIDENTS KERR COMMITTEE ACTIVITIES LEWIS RELIABILITY ASSURANCE WYLIE STANDARD PLANT DESIGN WYLIE CONTAINMENT REQUIREMENTS MARK O ga n
03/13/86 SUBCOMMITTEE ASSIGNMENTS O FOR DAVID A. WARD CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES: STAFF ENGINEER WATERFORD STATION UNIT 3 MDH PROCEDURES & ADMINISTRATION RFF DECAY HEAT REMOVAL SYSTEMS PAB MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN BABC0CK & WILC0X WATER REACTOR PLANTS WYLIE GENERAL ELECTRIC REACTOR PLANTS OKRENT LASALLE COUNTY STATION KERR MILLSTONE UNIT 3 SHEWMON NAVAL REACTORS KERR SAN ON0FRE UNIT 3 LEWIS V0GTLE, ALVIN W. PLANT UNITS 1 & 2 EBERSOLE WATTS BAR UNITS 1 & 2 EBERSOLE ADVANCED LWR DESIGNS OKRENT THERMAL HYDRAULIC PHENOMENA MICHELSON HUMAN FACTORS REMICK METAL COMPONENTS SHEWMON REGIONAL OPERATIONS REMICK PROBABILISTIC RISK ASSESSMENT OKRENT O_ SAFETY PHILOSOPHY, TECHNOLOGY, AND CRITERIA OKRENT SIESS SAFETY RESEARCH PROGRAM COMMITTEE ACTIVITIES LEWIS GAS COOLED REACTOR PLANTS SIESS SYSTEMATIC ASSESSMENT OF EXPERIENCE LEWIS LJ .
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03/13/86 SUBCOMMITTEE ASSIGNMENTS FOR (]
'v CHARLES J. WYLIE CHAIRMAN OF THE FOLLOWING SUBCOMMITTEES:
STAFF ENGINEER BABC0CK & WILCOX WATER REACTOR PLANTS RKM BEAVER VALLEY POWER STATION UNIT 2 HA BROWNS FERRY UNITS 1-3 AJC COMBUSTION ENGINEERING REACTOR PLANTS PAB RELIABILITY ASSURANCE RKM STANDARD PLANT DESIGN HA MEMBER OF THE FOLLOWING SUBCOMMITTEES: CHAIRMAN SEQUOYAH PLANT UNITS 1 & 2 MARK V0GTLE, ALVIN W. PLANT UNITS 1 & 2 EBERSOLE WESTINGHOUSE REACTOR PLANTS REED ADVANCED LWR DESIGNS OKRENT AUXILIARY SYSTEMS MICHELSON INSTRUMENTATION AND CONTROL SYSTEMS EBERS0LE GENERIC ITEMS SIESS HUMAN FACTORS REMICK MAINTENANCE PRACTICES & PROCEDURES REED OUALITY & QUALITY ASSURANCE IN DESIGN AND CONSTRUCTION REED O PEACTOR OPERATIONS EBERSOLE V REGULATORY ACTIVITIES SIESS REGULATORY POLICIES AND PRACTICES LEWIS SAFEGUARDS AND SECURITY MARK SAFETY PHILOSOPHY, TECHNOLOGY, AND CRITERIA OKRENT SAFETY RESEARCH PROGRAM SIESS SECONDARY SYSTEMS EBERSOLE AC/DC POWER SYSTEMS RELIABILITY KERR COMMITTEE ACTIVITIES LEWIS DECAY HEAT REMOVAL SYSTEMS WARD LONG RANGE PLAN FOR THE NRC CARB0k STATE OF NUCLEAR POWER SAFETY KERR CONTAINMENT REQUIREMENTS MARK OCCUPATIONAL AND ENVIRONMENTAL PROTECTION SYSTEMS M0ELLER PLANT OPERATING PROCEDURES MICHELSON O
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i l ADVISORY COMMITTEE ON REACTORS SAFEGUARDS s PROJECTS SUBCOMMITTEES REVISED: l l PROJECTS (COGNIZANT ACRS STAFF MEMerR) MEMBERS BABC0CK & WILC0X REACTOR PLANTS (RKM) --------- WYLIE, Ebersole, Kerr, (B-SAR-205 - Docket No. 50-561) Michelson, Okrent, Reed, Ward Specific Task:
- Responsible for review of specific events at B&W-designed plants.
BEAVER VALLEY POWER STATIONS UNIT 2 (HA) ------ WYLIE, Ebersole, Kerr, Shewmon (Docket No. 50-412) BELLEFONTE PLANT UNITS 1 & 2 (AJC) ------------ CARBON, Kerr, Okrent, Reed (DocketNos. 50-438,-439) BRAIDWOOD STATIONS (EGI) ---------------------- MARK, Moeller, Shewmon (DocketNo. 50-456,457) BROWNS FERRY UNITS 1-3 ( AJC) ------------------ WYLIE, Okrent, Remick (DocketNos. 50-259,-260,-296) BYRON STATIONS (EGI) -------------------------- MARK, Moeller, Shewmon (Docket Nos. 50-454, 455) CATAWBA NUCLEAR STATION UNITS 1 & 2 (JOS) ----- EBERSOLE, Kerr, Michelson, (CocketNos. 50-413,-414) Moeller CLINTON STATION UNITS 1 & 2 (PAB) ------------- KERR, Moeller (DocketNos. 50-461,-462) Cognizant Staff Merabers: HA - Heman Aldeman MWL - Morton W. Libarkin PAB - Paul A. Boehnert RKM - Richard K. Major AJC - Anthony J. Cappucci, Jr. JCM - John C. McKinley SD - Sam Duraiswamy OSM - Owen S. Merrill , MME - Medhat M. El-Zeftawy GRQ - Gary R. Quittschreiber RFF - Raymond F. Fraley RPS - Richard P. Savio MDH - M. Dean Houston JOS - John 0. Schiffgens EGI - Elpidio G. Igne i O
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_. . _ . . . ~ _ _ . . _ . _ p Q REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS COMBUSTION ENGINEERING REACTOR PLANTS (PAB) --- WYLIE, Carbon, Lewis, (CESSAR-System 80/Palo Verde) Michelson, Reed, Remick (Docket No. 50-470) Specific Tasks:
- Provisions for removal of decay heat in CE System 80 (CESSAR) type plants, specifically, extra requirements on the reliability of the auxiliary feedwater system and the integrity of the steam generators and consideration of the potential for adding valves of a size to facilitate rapid depressurization of the primary system to allow more direct methods of decay heat removal (per ACRS report on FDA for System 80 dtd.12/15/81 and discussion at 302nd ACRS mtg.). .
Responsible for review of specific events at CE-designed plants. COMANCHE PEAK UNITS 1 & 2 (HA) ---------------- OKRENT, Michelson (Docket Nos. 50-445,-446) DAVIS-BESSE STATION UNIT 1 (HA) --------------- REMICK, Carbon, Ebersole, m) l (Docket No. 50-346) Okrent, Siess DIABLO CANYON UNITS 1 & 2 (EGI) --------------- SIESS, Ebersole, Lewis, Okrent (DocketNos. 50-275,-323) FERMI, ENRICO, UNIT 2 (PAB) ------------------- KERR, Carbon, Shewmon (DocketNo.50-341) GENERAL ELECTRIC REACTOR PLANTS (RKM) --------- OKRENT, Ebesole, Kerr, (GESSAR II) Michelson, Remick, Shewmon, Ward Specific Task:
- Responsible for review of specific events at GE-designed plants.
GRAND GULF NUCLEAR STATION UNTIS 1 & 2 (HA) --- OKRENT, Ebersole, Siess (DocketNos. 50-416,-417) HOPE CREEK UNIT I (MME) ----------------------- SIESS, Kerr, Okrent (Docket No. 50-354) LASALLE COUNTY STATION (EGI) ------------------ KERR, Mark, Okrent, Ward (DocketNo. 50-373,-374) LIMERICK UNITS 1 & 2 (RPS) -------------------- KLRR, Lewis, Michelson, Siess j (DocketNos. 50-352,-353) f hh
REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) _ MEMBERS MIDLAND PLANT UNITS 1 & 2 (JOS) --------------- OKRENT, Ebersola, Michelson, (Docket Nos. 50-329,-330) Remick MILLSTONE POINT UNIT 3 (JOS) ------------------ SHEWMON, Kerr, Moeller, Siess, (Docket No. 50-423) Ward NINE MILE POINT UNIT 2 (JOS) ------------------ SHEWMON, Ebersole, Mark, Siess (Docket No. 50-410) PALO VERDE UNITS 1-3 (MDH) -------------------- EBERSOLE, Kerr, Lewis (Docket Nos. 50-528,-529,-530) PERRY PLANT UNITS 1 & 2 (PAB) ----------------- MICHELSON, Okrent (DocketNos. 50-440,-441) RIVER BEND UNITS 1 & 2 (RPS) ------------------ OKRENT, Ebersole, Shewmon (DocketNos. 50-458,-459) SAN ON0FRE UNIT 3 (MDH) ----------------------- LEWIS, Okrent, Siess, Ward (Docket No. 50-362) SEABROOK NUCLEAR PLANT UNITS 1 & 2 (RKM) ------ KERR, Carbon, Lewis, (DocketNo. 50-443,-444) HTclielson, Moeller Specific Task:
- Review emergency evacuation plans prior to issuance of full power OL.
SEQUOYAH PLANT UNITS 1 & 2 (AJC) -------------- MARK, Siess, Wylie (DocketNos. 50-327,-328) SHOREHAM STATION (HA) ------------------------- SIESS, Michelson (Docket No. 50-322) SOUTH TEXAS UNITS 1 & 2 (MME) ----------------- MARK, Lewis, Siess (DocketNos. 50-498,-499) SUSQUEHANNA STATION UNIT 2 (EGI) -------------- REMICK, Kerr, Ebersole, Mark (Docket No. 50-388) THREE MILE ISLAND UNITS 1 & 2 (RKM) ----------- M0ELLER, Etherington, Kerr, (DocketNos. 50-289,-320) Okrent V0GTLE, ALVIN W., PLANT, UNITS 1 & 2 (JOS) ---- EBERSOLE, Okrent, Reed, Ward, (Docket Nos. 50-424,-425) Wylie n ( ) U gr. ,2 & V
I REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS WPPSS NO. 2 (PAB) ----------------------------- MARK, Ebersole (Docket No. 50-397) WATERFORD STATION UNIT 3 (MDH) ---------------- WARD, Carbon, Siess (Docket No. 50-382) WATTS BAR UNITS 1 & 2 (AJC) ------------------- EBERSOLE, Okrent, Ward (Docket Nos. 50-390,-391) WESTINGHOUSE REACTOR PLANTS (AJC) ------------- REED, Ebersole, Kerr, i RESAR SP/90 (WAPWR) RIHelson,Shewmon,Wylie l (Docket No. 50-601) (Cons.: Theofanous) 4 Specific Task:
- Review the proposed new Westinghouse advariced standard plant design.
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u.____... l 1 i l l REVISED: l PROJECTS (C0GNIZANT ACRS STAFF MEMBER) MEMBERS AC/DC POWER SYSTEMS RELIABILITY (MME) --------- KERR, Ebersole, Lewis, Reed, WylTe (Cons.: Davis, Lee,
. Lipinski,Mueller)
Specific Tasks: Review reliability of AC/DC power systems in nuclear facilities, including the potential for disruption of offsite power sources. ADVANCED LWR DESIGNS (MME) -------------------- OKRENT, Carbon, Ebersole, Michelson, Reed, Ward, Wylie Specific Task:
- Consider the generic safety implications of advanced LWR designs which are not sponsored by a particular NSSS vendor (e.g., the EPRI program).
ADVANCED NON-LWR DESIGNS (MME) ---------------- CARBON, Mark, Okrent, Siess (Cons.: Avery, Bush, Golden, Lipinski, Plesset, Seals, Siegel) Specific Tasks:
- Consider safety implications of advanced designs of reactor systems which are fundamentally different from current LWRs, e.g., LMRs, Advanced GCRs, PIUS, breeder concepts, etc.; --
- Consider advanced reactor designs (e.g., LMFBR, GCFBR, HTGR, spectral shift, heavy water and light water breeders) including the design bases, criteria, and applicable regulatory guides; Develop proposed GDC and related RSR program needs; Prepare appropriate section of annual RSR reports to NRC/ Congress.
AUXILIARY SYSTEMS (SD) ------------------------ MICHELSON, Ebersole, Reed, Shewmon, Wylie Specific Task: Consider generic safety implications of the performance of systems not covered by other subconinittees, e.g., fire protection, air powered systems, cleanup systems, chilled water systems, etc. 0 g 2.&$'
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PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS COMMITTEE ACTIVITIES (MWL) -------------------- LEWIS, Michelson, Remick, Siess, Ward, Wylie Specific Tasks:
- Consider ACRS procedures for dealing with " cosmic" safety issues (e.g.,
Annual Report to NRC);
- Organize an annual ACRS retreat to discuss / exchange views on safety philosophy; 1 CONTAINMENT REQUIREMENTS (MDH) ---------------- MARK, Ebersole, Kerr, Okrent, l Siess, Wylie Specific Tasks:
Consider nonstructural aspects of containment perfonnance; Consider the development of containment performance criteria as related to the safety goal; O V - Evaluate proposed Mark III containment venting guidelines; Reexamine the proposed NRC Rule on hydrogen control; consider progress thereon, and proposed approaches to its implementation. CORE PERFORMANCE (MDH) ------------------------ KERR, Mark, Michelson, Okrent,
'90smon (Cons.: Lee, Lipinski)
Specific Tasks: Consider core physics; Consider power distribution measurement and control (including neutron flux at PB wall); Consider effect of positive moderator coefficient; Consider reactivity effects (e.g., calculation of rod drop and rod ejection accident). Prepare appropriate section of annual RSR reports to NRC/ Congress. O 43n
REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS DECAY HEAT REMOVAL SYSTEMS (PAB) -------------- WARD, Ebersole, Etherington, Kerr, Michelson, Reed, Wylie (Cons.: Catton, Davis) Specific Tasks: Consider general safety issues related to performance of all shutdown heat removal systems, e.g., RHR, AFW, ECCS;
- Examine the use of alternate and improved systems for removal of decay heat under non-LOCA and/or small LOCA conditions, including use of natural circulation as a mechanism for decay heat removal; evaluate " feed and bleed" procedures and systems as an alternate method of decay heat removal; evaluate dedicated (bunkered) high pressure forced circulation system and shock condensers; Review safety research and full-scale power plant test procedures in these areas; Review USI A-45 resolution position.
EXTREME EXTERNAL PHENOMENA (RPS) -------------- OKRENT, Carbon, Etherington, Lewis,Moeller,Siess(Cons.: Ang, Bush, Maxwell, Page, Philbrick, Pomeroy, G. Thompson, Trifunac) Specific Tasks: Periodic review of the Diablo Canyon Long-Term Seismic Program; 1 Consider criteria for extreme external phenomena (e.g., earthquakes, l tornadoes, tsunamis, seiches, hurricanes, floods, explosions, airplane crashes, release of noxious chemicals, ef.h of LNG fires / explosions or othersimilarevents);
- Review of the SSMRP; Seismic design of auxiliary feedwater systems; l
Consider related research program and prepare appropriate sections of l annual RSR reports to NRC/ Congress- 1 l Reevaluation of the extreme external design basis events for the SEP plants. D d pp. ) CO
O O REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS FUEL CYCLE (HA) ------------------------------- SHEWMON, Kerr, Moeller, Reed (Cons.: Foster, Healy,
. Lawroski, Orth, Parker, Steindler)
Specific Tasks:
- Consider research related to fuel cycle facilities (other than reactors) regarding effluent control and effects of fission product dispersal;
- Consider radiobiological effects and occupational exposures, including decommissioning and decontamination of fuel cycle facilities;
- Consider onsite treatment of radwastes;
- Consider safety associated with aspects of the fuel cycle and assurance of compliance with the ALARA concept for fuel cycle facilities other than those specifically related to nuclear reactors; Prepare appropriate section of annual reports to NRC/ Congress.
FUEL PERFORMANCE (HA) ------------------------- SHEWMON , Kerr, Michel son, Reed (Cons.: Solomon) Specific Tasks: Consider new and modified fuel design, including proof testing; Consider thermal-hydraulic and mechanical fuel performance during normal and abnormal conditions; Consider pellet-cladding interaction; Consider fuel failure propagation;
- Consider high burnup fuel performance; Evaluate replacement fuel designs and qualification testing; Prepare appropriate section of annual RSR reports to NRC/ Congress.
GAS COOLED REACTOR PLANTS (JCM) --------------- SIESS, Ebersole, Kerr, Shewmon, Ward Specific Task: Consider safety issues related to operation of the Fort St. Vrain plant and any proposals for new plants of similar design.
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REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS GENERIC ITEMS (SD) ---------------------------- SIESS, Ebersole, Michelson, Reed, Wylie Specific Tasks: Conduct a preliminary evaluation of proposed generic issues, and with the Committee's concurrence, proceed with further action (e.g., refer to NRC Staff for action, assign to an ACRS topical subcomittee for further evaluation,etc.); Provide oversight and coordination of " established" generic matters, handling those items it is competent to deal with and assigning others to ACRS topic subcommittees for review as considered appropriate. Established generic items include those recognized by the ACRS and/or NRC Staff including Task Action Plan Items, Unresolved Safety Issues, etc.; Periodic meetings with NRC Staff to discuss NRC Generic Issues Tracking System (GITS) quarterly report and quarterly Unresolved Generic Issues report (Aqua Book) (sea Fraley memo to Mark dtd. 7/29/81 re Generic Issues Tracking System); Coordination and review
- of priority ranking of generic issues and NRC task action plans for resolution of safety issues, including referral to cognizant topical or project subcomittees, where appropriate; Generic Items Subcomittee will review safety-related items not assigned to other standing subcomittees.
Review of implementation of resolved generic items, including referral to cognizant topical or project subcomittees where appropriate; include i consideration of backfitting criteria and followup regarding application to specific plants and/or classes of plants; l Review NRC Staff guidelines proposed to evaluate and set priorities for resolution of safety issues; followup regarding ACRS letter to E00 dtd. 9/15/81 regarding the development of a defined methodology to evaluated the impact of nuclear power plant changes required by NRC. O g , d '?d
(7 REVISED: LJ PROJECTS (C0GNIZANT ACRS STAFF MEMBER) MEMBERS HUMAN FACTORS-(JOS) --------------------------- REMICK, Ebersole, Kerr, Michelson, Reed, Ward, Wylie (Cons.: Gimmy) Specific Tasks: Qualification and organization of operating group, including plant review and audit conunittees; qualification and organization of corporate management;
- Operator training, qualification, and requalification for all licensed facilities, including the HLW repository and other facilities required by NWPA;
- Review qualifications, training, etc., of infrastructure (supporting personnel) at nuclear facilities; Man-machine interaction, including design and arrangement of the control room; operator response under stress;
- Criteria for preparation of normal, abnormal, and emergency operating J procedures; Consider evaluation of plant / project management as outlined in the 7/3/85 letter from the EDO to Chairman, TVA. (Assigned during 303rd ACRS mtg.)
Review proposal for a National Academy for training nuclear power plant operators (305th ACRS mtg.) Prepare appropriate section of the annual RSR reports to NRC/ Congress. INSPECTION AND ENFORCEMENT PROGRAMS (AJC) ----- REED, Ebersole, Kerr, HT5elson,Moeller Specific Task: Review of I&E programs, including enforcement, the development of inspection methods used by regional and resident inspectors, etc. INSTRUMENTATION AND CONTROL SYSTEMS (MME) ----- EBERS0LE, Kerr, Lewis, Michelson, Wylie (Con.: Davis, Epler, Lipinski) Specific Tasks: Consider general safety issues related to performance of all I&C systems, electrical, electronic, hydraulic and pneumatic, as appropriate; Consider design design and performance of nonnal and emergency power supplies, plant instrumentation and control systems, safety actuation systems, leak detection, vibration and loose parts monitors;
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REVISED: J PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS INSTRUMENTATIONANDCONTROLSYSTEMS(CONT'D) Specific Tasks (Cont'd) ! - Consider design and performance of plant computers, diagnostic and display equipment, etc., provided to assist the operator in control of the plant during normal and abnormal conditions;
- Followup regarding instrumentation to detect inadequate core cooling (per ACRS ltr. to EDO dtd. 6/9/81, Comanche Peak OL dtd. 12/17/81, St. Lucie 2 OL dtd.11/17/81, etc.);
- Provide coments regarding proposed Regulatory Guide (IC 121-5), Response Time Testing of Protection System Instrument Channels for incorporation in the final guide to be promulgated by NRC;
- Evaluate proposed design and perfomance of reactor control and safety systems, including the digital safety systems in nuclear plants (e.g., CE units such as AN002/ San Onofre 2) and use of N-16 detectors in the overpower trip circuitry of W RESAR-3 type plants (e.g., Comanche Peak);
i - Prepare appropriate section of annual RSR reports to NRC/ Congress. LONG RANGE PLAN FOR THE NRC (RKM) ------------- CARBON, Lewis, Moeller, Remick, Wylie f Specific Tasks: j
- Develop advice for the Comission on a plan to deal with the technical issues related to the regulation of nuclear power plant safety and safety regulation over the next five to ten years, by offering coments on the Long Range Plan being developed by OPE and EDO (per discussion at 301st :
ACRS mtg.); 1
- Consider the present and anticipated state of the industry, i.e., i operational concerns with over 100 power plants and over 50 licensees, ;
and safety research considering the spectrum of possible directions that j j a resurging industry could take or any special needs which should be i ] addressed by the Comission, in the fact of a dwindling industry i J comanding over-decreasing resources. Emphasis should be on the l j technical issues and potential technical issues as opposed to l j administrative and political concerns; ' 1 I
- The advice should be directed to the Comissioners as an aid to their planning, but should also consider as an audience the Congress in its role of creating public policy to support improvements in reactor safety, 1 the NRC Staff, and eventual instruments of new NRC policies, and the various segments of the nuclear power industry; 1
(' ' A REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS LONG RANGE PLAN FOR THE NRC (CONT'D) Specific Tasks (Cont'd): ,
- The advice should be develo* ped through subcommittee meetings and interactions with the full ACRS on such a schedule that the full Comittee can issue a report describing the Plan at its October 1985 meeting. The first report would be based primarily on what the ACRS sees as problem areas from its vantage point as an independent advisory body.
Discussions with representatives of the NRC Staff and other groups involved in the regulatory process (e.g., vendors and licensees) and past members of the NRC and its staff would be appropriate for the first round. This group of people might be expanded as considered appropriate for subsequent reports; MAINTENANCE PRACTICES AND PROCEDURES (HA) ----- REED, Ebersole, Michelson, Ro'dTler, Reed,Wylie Specific Tasks:
- Review NRC criteria / requirements for maintenance of nuclear facilities Os (per 280th ACRS mtg.);
Consider applicants' programs for maintenance of nuclear power plant equipment, including the potential for inducing comon mode failures (per 280th ACRS mtg.);
- Vendor instructions regarding adjustment, periodic / preventive maintenance.
METAL COMPONENTS (EGI) ------------------------ SHEWMON, Etherington , Lewis , Michelson, Okrent, Ward (Cons.: Bender, Bush, Catton, Dillon, Gall, Hutchinson, Kassner, Pense, Pickel, Rodabaugh,B. Thompson) Specific Tasks: l
- Consider design and performance of metal components, including metal containment structures, reactor pressure vessels and other components 1 such as piping, valves, pumps, snubbers, and other high- energy restraints, rod drives, turbines (including probability, size, etc., of missiles resulting from turbine failure);
- Consider material properties (e.g., maximum allowable stresses, response to radiation damage, etc.) and response to environmental conditions
(, (e.g., stress corrosion cracking, reaction to containment spray additives, etc.) including the effect of water conditions / treatment on steam g>nerator tube degradation, physical / chemical effects of deconta,ninaticn pescesses and fluids on plant components and systems; q})
REVISED: PROJECTS (C0GNIZANT ACRS STAFF MEMBER) MEMBERS METAL COMPONENTS (CONT'D) Specific Tasks (Cont'd): Consider environmental qualification of metal components / systems; Consider failure modes of metal components (e.g., " leak-before-break" behavior vs. " instantaneous, double-ended pipe break");
- Prepare appropriate sections of annual RSR reports to NRC/ Congress; 1nsider proposals for review of NRC requirements for piping design, other than seismic design requirements; Consider the adequacy of nondestructive examination techniques in detecting and sizing flaws in metal components and piping systems.
NAVAL REACTORS (PAB) -------------------------- KERR, Lewis, Reed, Remick, Ward Specific Task: s Conduct the review of Naval Reactors' proposal for a Moored Training Facility. r NUCLEAR PLANT CHEMISTRY (HA) ------------------ M0ELLER, Ebersole, Etherington, Reed, Shewmon Specific Tasks: H2 addition to BWRs and concommitant hydrogen embrittlement;
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and steam explosions; i Severe accident chemistry and its relationship to the codes that purport to model that chemistry; Water quality, e.g., in the secondary side of steam generators; 10 CFR 20, occupational dose reduction from electro-polishing or pre-oxidation of BWR replacement recirculation piping; i Microbiological 1y induced corrosion of containment service water systems (IE Notice 85-30); Phenomenology of N 16 in BWR steam; Radiation effects on equipment in accident aerosols; Metal oxide sludges in PWRs.
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b' / REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS OCCUPATIONAL AND ENVIRONMENTAL PROTECTION SYSTEMS (JOS) ------------------------------- MOELLER , Mark , Reed , Remi ck , Tylie Specific Task:
- Consider issues related to reactor radiological effects on plant personnel and the environment, e.g. ,10 CFR 20, perfonnance of plant HVAC as.related to personnel safety, etc PLANT OPERATING PROCEDURES (JOS) -------------- MICHELSON, Ebersole, Kerr, Reed,Remick,Wylie Specific Task:
Review of development and use of plant procedures, particularly E0Ps. PROBABILISTIC RISK ASSESSMENT (RPS) ----------- OKRENT, Carbon, Kerr, Lewis, Remick, Siess, Ward (Cons.: Avery, Bush, Davis, Dietrich, O Epler, Hickman, Lipinski, Mueller, Petersen, Powers, Saunders, Wilson) Specific Tasks: Consider reactor safety studies and probabilistic risk assessments of nuclear plants; Consider reliability assessment o'f systems and components; Consider containment isolation provisions; Consider integrated reliability evaluation programs (IREP, mini-IREP, NREP, and RSSMAP); Consider quantitative risk criteria and safety goals;
- Review of PRAs for Indian Point / Zion / Limerick stations; Prepare appropriate section of annual RSR reports to NRC/ Congress.
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m O %J REVISED: l PROJECTS (C0GNIZANT ACRS STAFF MEMBER) MEMBERS PROCEDURES AND ADMINISTRATION (RFF) ----------- WARD, Ebersole, Lewis, ) Ro'eTler, Remick, Siess j
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Specific Tasks: . Consider ACRS procedures and bylaws; ;
- Policy and oversight regarding Fellowship program; i
- Selection of new ACRS members (e.g., disciplines needed, screening of candidates,etc.);
Proposed improvements in SARs (per discussion during 255th ACRS mtg. regarding M. Bender 1tr. to Fraley dtd. 4/13/81); The current chairman and vice-chairman and the most recent past ACRS chairman will act as a subgroup to work as needed with the ACRS Executive Director to discuss / resolve / implement proposed / required changes in ACRS Staff support and related ACRS activities resulting from proposed Commission actions regarding manpower, budget, facilities, etc.
> QUALITY AND QUALITY ASSURANCE IN DESIGN AND CONSTRUCTION (EGI) -------------------------- REED, Michelson, Remick, Siess, Wylie Specific Tasks:
Consider proposed NRC rules and regulations, Regulatory Guides, etc., dealing with QA during construction of nuclear plant; 1
- ACRS review of NRC " study of existing and alternative programs for i improving QA and QC in the construction of connercial nuclear power plants" (per P.L. 97-415 dtd. January 4, 1983);
- Review proposed definition of plant features important to safety and related criteria for such systems / components, including graded requirements regarding reliability, qualification, etc., of systems / I components with different levels of safety significance; l Prepare apprcpriate section of annual RSR reports to NRC/ Congress.
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i REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS REACTOR OPERATIONS (HA) ----------------------- EBERS0LE, Lewis , Michel son, - Moeller, Reed, Remick, Wylie (Cons.: Catton, Lipinski, Mathis) Specific Tasks: Operational QA; preoperational and startup testing and inspection; inservice inspection and testing; evaluation of recurring incidents / transients which are applicable to a type or class of reactor (e.g., primary system blowdown transients in BWRs, overpressurization incidents in PWRs, overcooling transients in PWRs); Screen proposed power level increases to determine if further action by generic or project subcommittee (s) is warranted;
- Review LERs to determine if there are trends in the performance of systems or components, including the contribution of human error, design deficiencies, need for RSR, etc., which warrant attention; screen 3 incidents and abnormal occurrences to determine if full Connittee action is warranted; detennine if component and system failure rate are reasonable; identify causes of systems interactions; Review and evaluation of NRC procedures and processes for review, evaluation, dissemination, and implementation of action related to operating experience gained at nuclear facilities (e.g., Manual Chapter 0515, Operational Safety Data Review, Periodic Overview of Operational Data Activities);
Evaluate regulatory procedures for factoring operating information into the, licensing process; Review proposed changes / improvements in the LER system (e.g., NUREG/CR-1928, Development of Licensee Event Report Sequence Coding and Search Procedure); Review NRC Enforcement Policy, particularly the use of fines for deficiencies in management, operation, etc.; Meet periodically (i.e., approximately bimonthly) with IE/NRR/AE0D to discuss trends and patterns in operating experience (285th ACRS mtg.). REGIONAL OPERATIONS (AJC) --------------------- REMICK, Carbon, Michelson, Reed, Ward Specific Task:
- Review of the NRC's Regional programs.
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N~ (d REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS REGULATORY ACTIVITIES (SD) -------------------- SIESS, Michelson, Remick, Wylie Specific Tasks:
- Conduct and coordinate review of regulatory guides and regulations as appropriate;
- Ad hoc review of regulatory staff proposals for new approaches, and referral as appropriate;
- Examine CRGR procedures and impact on regulatory activities.
REGULATORY POLICIES AND PRACTICES (AJC) ------- LEWIS, Michelson, Moeller, Okrent, Siess, Wylie Specific Tasks: 0
- Review the recommendations of the NRC Regulatory Reform Task Force regarding proposed changes in the regulatory process;
- Examine specific aspects of the NRC regulatory processes as considered appropriate. Consider changes in emphasis needed in NRC re policies and practices; study the issue of overregulationassigned (gulatory during 296th ACRS mtg.);
- Identify important safety issues needing increased (or less) attention and/or resolution in the NRC regulatory process;
- Review Rule / implementation on plant-specific backfitting;
- Consider alternative proposals for performing NTSB-type investigations:
(1) NRC Staff plan, and (2) use of ASLBP (assigned during 303rd ACRS mtg.);
- Review of SECY-82-256, Enforcement Policy on Vendors (304th ACRS mtg.).
RELIABILITY ASSURANCE (RKM) ------------------- WYLIE, Ebersole, Michelson, Siess (Cons.: Bender) Specific Tasks: 1
- Provide a focal point for coordination / liaison of this program with the NRC Staff;
- Conduct and/or coordinate review of related regulatory criteria, NUREGs, etc.; assign review to cognizant ACRS topical subconnittees as appropriate;
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REVISED: PROJECTS (C0GNIZANT ACRS STAFF MEMBER) MEMBERS RELIABILITY ASSURANCE (CONT'D) Specific Tasks (Cont'd):
- Consider seismic and environmental qualification of mechanical, electrical, and electronic components / systems (including resolution of USI A-46). Consider the NRC program to evaluate plant aging (except for those aspects being considered by the Metal Components Subcomittee, e.g., PV embrittlement, steam generator tube degradation, thermal aging of cast SS piping / components);
Review proposed NRC Reliability Assurance Program Plan;
- Valve operability assurance and reliability;
- Pump operability assurance and reliability; Prepare appropriate section of annual RSR reports to NRC/ Congress.
SAFEGUARDS AND SECURITY (J05) ----------------- MARK, Carbon, Michelson, Reed, y" Remick, Wylie (Cons.: l Lawroski, Woodcock) l I Specific Tasks: 1
- Consider industrial security / access control; Consider design features to preclude or mitigate effects of sabotage; review improved Westinghouse and GE standard plant designs proposed during the 261st ACRS meeting; reconsider the " issue of sabotage" (290th ACRS mtg.);
Consider material accountability for special nuclear material; 1
- Consider measures to inhibit / preclude unauthorized diversion of SNM; Plant arrangements to provide plant security; Prepare appropriate section of annual RSR reports to NRC/ Congress.
Consider safeguards determinations required by NWPA for Spent Fuel Transport, MRS, etc. O
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N REVISED: (D PROJECTS (C0GNIZANT ACRS STAFF MEMBER) MEMBERS SAFETY PHILOSOPHY, TECHNOLOGY, AND CRITERIA (RPS) ------------------------------ OKRENT, Kerr, Lewis, Michelson, Remick, Ward, Wylie Specific Tasks:
- Consider generic implications of reactor accidents / incidents regarding regulatory criteria and policy;
- Review GDC and other criteria (e.g., single failure criteria
- and guides as needed based on operating experience, advances in related technology and research results;
- Systems interactions, including consideration of functional inter-actions (e.g., failure of a common power supply on several safety systems), the adequacy of the single failure criterion, the potential for common mode failures such as compartment flooding, high compart-ment pressure, etc., and nonfunctional interaction of systems such as pipe whip, jet forces, etc., including the impact of nonsafety grade system failures on safety systems, separation criteria, missile protection.
- Post-accident recovery and cleanup equipment (e.g., water and gas treatment systems, emergency power supplies);
Review seismic and nonseismic induced systems interactions resulting from nonfunctional interactions (e.g., pipe whip, environmental effects, etc.); Independent test of the DEDROGR technique for assigning priorities to unresolved safety issues (per 260th ACRS meeting); Consideration of important societal resources in the siting of nuclear plants. SAFETY RESEARCH PROGRAM (SD) ------------------ SIESS, Carbon, Kerr, Moeller, Okrent, Shewmon, Ward, Wylie Specific Tasks: Consider scope and balance of RSR program; Coordination of annual RSR report to Congress; > Coordination of annual RSR report to the Comission. O v q..~>Pd
3 REVISED: J PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS SCRAM SYSTEMS RELIABILITY (PAB) --------------- KERR, Ebersole, Lewis, Reed TCons.: Davis, Lee, Lipinski, Mueller) Specific Tasks: Review issues associated with implementation of the ATWS Rule; Review reliabilities used in reliability calculations for scram systems, including PRAs and other studies such as those reported in NUREG-0460; Review observed reliabilities as detemined from operating histories; Review reliabilities that might be expected if UV trip is used, if shunt trip is used, if the two are used together. SECONDARY SYSTEMS (MME) ----------------------- EBERSOLE, Michelson, Okrent, Reed, Wylie Specific Task: Consider generic safety issues related to the performance of nonnuclear parts of the plant (e.g., steam and electrical generation systems). SEVERE (CLASS 9) ACCIDENTS (MDH) -------------- KERR, Carbon, Mark, Okrent, 3hiiwmon, Siess (Cons.: Bender, Catton, Corradini, Davis, Lee, Powers) Specific Tasks: Determine if low probability accidents with consequences greater than the DBA should be taken into account in the licensing process; Consider process leading from hot fuel to melt-through to detemine if core cooling can be accomplished inside the reactor PV or the containment; consider nature and characteristics of related events such as core coolant interactions (e.g., steam explosions; core containment interactions (e.g., reaction with concrete, heating of structural elements, etc.); consider impact on containment design, including need i for and design characteristics of a core catcher; need for and characteristics of filtered vents and/or other provisions to handle large quantities of gasses which may be generated as the result of a core melt or major degradation; Characteristics and design of molten core retention devices, including C the proposed core ladle for the FNP, Zion, and Indian Point nuclear plants. Provide guidance / assistance to the FNP, Zion, and Indian Point Project Subcomittees as needed in evaluating these features; f ,.).cf f
REVISED: PROJECTS (C0GNIZANT ACRS STAFF MEMBER) MEMBERS SEVERE (CLASS 9) ACCIDENTS (CONT'D) Specific Tasks (Cont'd):
- Evaluate " pre-vent" containment (per 253rd ACRS meeting); - Review proposed NRC rules regarding degraded cores; Prepare appropriate section of annual RSR reports to NRC/ Congress.
SPENT FUEL STORAGE (HA) ----------------------- SIESS, Kerr, Moeller, Remick, Shewmon Specific Tasks: Evaluate the design of spent fuel storage pools;
- Consider provisions to preclude or withstand missile damage (including caskdropaccidents); - Consider seismic effects; - Consider the effects of tornadoes (including missiles);
Consider pool storage capacity; Consider provisions to preclude criticality; Consider provisions to cool the spent fuel under normal and abnormal conditions; Consider resistance to sabotage; Evaluate proposed design criteria for offsite, spent fuel storage facilities; Review of NWPA requirements for development of additional on-site spent fuel storage capacity; licensing of any spent fuel storage technologies; development of monitored retrievable storage. O nm
REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS STANDARD PLANT DESIGN (HA) -------------------- WYLIE, Carbon, Ebersole, Kerr, Michelson, Siess Specific Tasks:
- Evaluate / propose appropriate scope of standard plant designs;
- Consider policy regarding how and when changes will be pemitted to new or previously licensed standard plants [per ACRS report on CE System 80 (CESSAR) FDA dtd. 12/15/81];
- Consider the approach being proposed by EPRI to the establishment of a defined list of Generic Issues to be addressed in the design of a future standardized LWR.
STATE OF NUCLEAR POWER SAFETY (AJC) ----------- KERR, Lewis, Michelson, i MoeTler, Okrent, Reed, Wylie Specific Tasks:
- Develop a comprehensive draf t statement of the ACRS' perspectives on the s present state of safety, in the domestic nuclear power industry, should identify those areas where improvements is needed or desired, and on what scheduled (priority). The draft statement should be prepared on a scheduled that will permit the full Coninittee to issue its report on the subject at the September 1985 ACRS meeting.
STRUCTURAL ENGINEERING (EGI) ------------------ SIESS, Ebersole, Okrent, Shewman (Cons.: Bender, Pickel,Rodabaugh) ! Specific Tasks: l
- Review of NRC Staff proposal to revise seismic design methodology (e.g.,
- increased damping, reduce imposed loads, etc.) for piping systems;
- Consider structural and civil engineering aspects of nuclear power 1 plants, including concrete containment and concrete reactor pressure i vessel design bases and criteria;
- Consider structural aspects of steel containments; 1
- Consider structural aspects of structures such as shield buildings /
walls, turbine buildings, auxiliary buildings, etc.; 1
- Prepare appropriate section of the annual RSR reports to NRC/ Congress.
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i REVISED: PROJECTS (C0GNIZANT ACRS STAFF MEMBER) MEMBERS SYSTEMATIC ASSESSMENT OF EXPERIENCE (RKM) ----- LEWIS, Ebersole, Kerr, Mark, ' Michelson, Moeller, Ward I Specific Tasks:
- Responsible for an integrated review of experience;
- Consider the LER program (as distinguished from individual LERs) and NPRCS; j
- Serve as the ACRS interface with AE0D;
- Review INP0 and NSAC work, particularly SOERs.
THERMAL HYDRAULIC PHENOMENA (PAB) ------------- MICHELSON, Ebersole, Etherington, Kerr, Reed, Ward (Cons.: Catton, Plesset, Schrock, Sullivan, Theofanous, Tien) Specific Tasks: Consider design of current ECCS and improved ECCS designs; Consider thermal-hydraulic performance of primary system during LOCA; Consider related test facilities such as LOFT, Semiscale, etc.;
- Consider BWR containment, fluid dynamics, and heat transfer (e.g.,
Mark I, Mark II, and Mark III programs);
- Consider RPV asymmetric loads; Consider containment subcompartment pressures and dynamic loads during LOCA blowdown; Consider dynamic effects of relief valve operation, etc.;
- Consider waterhammer problems and corrective measures.
Prepare appropriate section of annual ACRS reports to NRC/ Congress. ! O V
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REVISED: PROJECTS (COGNIZANT ACRS STAFF MEMBER) MEMBERS TRANSPORTATION OF RADI0 ACTIVE MATERIALS (SD) -- SIESS, Ebersole, Mark, Ro'dTTer,Shewmon (Cons.: Healy, Langhaar, Orth, Shappart,Steindler) Specific Tasks:
- Conduct such reviews as requested by the Coninission (280th ACRS mtg.);
Cosnider Transportation of Spent Fuel and Radwaste as requested by NWPA. UNIVERSITY AND TEST REACTORS (EGI) ------------ MARK, Kerr, Michelson, Okrent, IIeiiiTek Specific Task: Consider any issues related to the safety of test reactors -- whether at universities, DOE, NBS, etc. WASTE MANAGEMENT (OSM) ------------------------ MOELLER, Carbon, Mark, Remick, Shewman (Cons.: Bair, Carter, O Corradini, Donoghue, First, Foster, Healy, Krauskopf, Lawroski, Orth, Parker, Philbrick,-Steindler, G. Thompson, Wilson) Specific Tasks: Consider radwaste management, including interim storage and long-term disposal of high-level wastes, low-level wastes, and mill tailings; Consider suitable waste classification, waste fanns and methods of waste isolation; Consider NRC's Repository Development and Licensing activities under NWPA; Consider NRC's activities in connection with the Test and Evaluation Facility under NWPA; Consider the study of alternative approaches to management and construction of a HLW repository (as requested by NWPA). Prepare appropriate section of the annual RSR reports to NRC/ Congress. O Ax .2 : s
l l APPENDIX XVI l ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE
- 1. Memorandum, T. G. McCreless, ACRS Assistant Executive Director for Technical Activities, to ACRS Members, Change to ACRS Bylaws, March 13, 1986 1 2. Memorandum, J. O. Schiffgens, ACRS Staff Engineer, to ACRS Members, March 19-20, 1986 Human Factors Subcomittee Meeting, March 12, 1986
- 3. Letter, E. C. Rodabaugh of E. C. Rodabaugh Associates, Inc. to E. G. Igne, ACRS Senior Staff Engineer, Broad Scope GDC-4 Rule, ACRS Metal Components Subcomittee Meeting, February 27-28, 1986, March 3, 1986
- 4. EPRI Report, Ian B. Wall, J. Carl Stepp, et. al, 1986-90 Plan for EPRI Seismic Center, October 1985 1
- 5. Memorandum, V. Stello, Jr. Acting EDO to the Comissioners, Proposed Broad Scope Rule to Modify General Design Criterion 4 to 10 CFR Part 50, March 1986 4 6. Letter, R. F. Walker, President, Public Service Co. of Colorado to Honorable D. Fuqua, Chairman, and Members House Comittee on
, O- Science and Technology, et. al, Joint Participants' Statement on the HTGR Program to Congressional Oversight Committees, February 17, 1986 i
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