ML20203N347
| ML20203N347 | |
| Person / Time | |
|---|---|
| Issue date: | 09/19/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2415, NUDOCS 8609230333 | |
| Download: ML20203N347 (300) | |
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TABLE OF CONTENTS pa w; mr g*I pO ( j[ l MINUTES OF THE fi i h I' 3 N 312TH ACRS MEETING $M 0 Il [LJ ll W I W APRIL 10-12, 1986 V/ d WASHINGTON, D.C. W h WM[ tf-lf' <?h I. Chairman's Report (0 pen)................................... 1 II. McGuire Nuclear Power Station Proposed Removal of Upper HeadInjectionSystem(0 pen)............................... 1 III. ReactorOperations(0 pen)................................. 7
- 1. Implementation of TMI Action Item I.D.2.-
Safety Parameter Display System at Operating Reactors ............................................ 7
- 2. Failure of Standby Liquid System .................... 8
- 3. Class 1E Station Battery Problems ................... 9
- 4. Component Cooling Water System Problems ............. 10
- 5. Follow-up Discussion to Operating Events ............ 11 IV. Advanced Reactor Designs (0 pen) . . . . . . . . . . . . . . . . . . . . . . . . . . 12 V. Subcomittee Activities Related to Consideration of Human Factors in Nuclear Power Plants (0 pen) .......... 18 VI. Meeting with the NRC Comissioners (0 pen) . . . . . . . . . . . . . . . . 22 A. Anticipated ACRS Activities and Resources in Light of Budget Reduction (0 pen) . . . . . . . . . . . . . . . . . . . . 22 B. Scope of ACRS Activities Regarding TVA .............. 23 C. Responses to Recommendations of Panel on ACRS Effectiveness ....................................... 24 D. Moynihan Proposal for National Training Academy ..... 25 t
E. ACRS Review of GESSAR II ............................ 26 VII. Quantification of Health Effects in PRAs (0 pen) .......... 26 VIII. Quanti tative Safety Goal s (0 pen) . . . . . . . . . . . . . . . . . . . . . . . . . 28 IX. Reliability of Nuclear Components (0 pen) ................. 29 g j X. Report of Congressional Hearing on 1986 Ohio Earthquake (0 pen)......................................... 30 l f ( LDIGUATE] ORIGINAL i 8609230333 PDR ACRS 860919 2415 Certified By - - - 1 PDR-
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312th ACRS MEETlNG June lic 1986 XI. . Auxiliary Feedwater Systems and Resolution of - US I A-45 ( 0 pe n ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 XII. LWR Standard Plant Design (0 pen) . . . . . . . . . . . . . . . . . . . . . . . . . 31 XIII. Fort St. Vrain (0 pen) .................................... 32 XIV. Executive Sessions (0 pen) ................................ 33 A. Subcommittee Assignment.............................. 33
- 1. Status of Emergency Operating Procedures........ 33
- 2. Assignment of the Instrumentation and Control Systems Subcommittee.................... 34
- 3. Fort St. Vrain.................................. 34 B. Reports, Letters, and Memoranda...................... 34
- 1. ACRS Comments on Quantitative Safety Goals...... 34
- 2. ACRS Comments on Quantification of Public Heal th Ri s k s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34
- 3. Proposal by Duke Power Company to Operate the McGuire Plant Without the Upper Head Injection System................................ 35 4
- 4. NRC Review of Advanced Reactor Designs ......... 35 3
- 5. Move to Bethesda................................ 35
- 6. P roposed Change i n ACRS Byl aws . . . . . . . . . . . . . . . . . . 35 C. Future Agenda........................................ 35 I-l 1. Future Agenda................................... 35 t
- 2. Future Subcommittee Meetings.................... 36 l D. Nuclear Reactor Safety Information Exchange Meeting.............................................. 36 E. Appointment of New Members........................... 36 F. PRA Quantification of Public Health Risk............. 36 G. Safety Parameter Display System and Emergency Operating Procedures................................. 37 i
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- ts Endowment for the Arts. Washington. Faendation on the Arts and the WUCtKAR NEGIAATORY'~ i" bC --
DC 20506, or call (202) 682-5433. Humanities Act d 1965, as aumended.' COMMtSSION x
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irr.ludmg ascussion ofinfamention .' John H. Clark, given in cadidence to the Agency by Advisory Cc. ' t on Reactor ' i Director. Office of Cou lendPonel Sefeguarde;RevisedMeetime Agenda a grant applicants. In accordance with the -< Opemnons.Norional wavntforthe Ar s. . s detemination of the Chairman in accordance with the purpo'ses of March 2s.tsee. pubhshed in he Federal Register of sections 29 and 182b.of the Atonde ' ~
, [FR Doc. 86-7352 Filed 4-2-es. e 45 am) ' Febreary 13,1980. these sessions wiH be EnerEr Act (42 U.S.C acus. 22seb), the i m coeg ,,, a closed ta the public persuant to - . Advisory Committee en Reactor ~. . .
subsections (c) (4). (6) and f9](B) of Safeguards wat held a ammetingen Apra , section 552b of Title 5 ited States 38-12.19ss,in Room 10es 17t7 H Street 1 Museum Advisory Pacel;Meettng l Code NW,Washaston DC.Notceof this at to sec6en aM2W the Furtherinformar n with reference to meeting was pehiashed la the Fanieral . Federal Advisory Ce tiee Act Tub. this meeting can obtained from Mr. Ragster on Masch 21. m Portions of
- t. is hereby isory Committe's tWe atinghm been.d ? U to
;. L M ).asa - Joh H'Clari. ;
giren that a the Museem Management icer, National lenge Sechoni to Endowmen orthe Arts. Washington. Advisory Panel ( pa&capants a emesmg the Netsnal r% il on the Arts will be DC 2me. r cam 204682-5C3. held on April tses from 200 a.m. to h as, seas. Thr.sday. Apaf re,2ses Desse - 5:30 p.m Rco 714 of the Nancy Hanks g.h w M h W gACRS sylvania Avenue. John aa. Center.1100 a S. NW, Wash' ston. DC 20506. Di .crfmeofr% - r r @ mel oans.Netrammlsoebwmentpars6e Arse. c3,,raca (Open)-He ACR walreport brieDy regardingitems of This me iag la for the purpose of Doc. awass Pued 44-ee s.1s ml cwreat interest to time Coastuittee. Panel rev' w. discussion. evaluation. 8""8'** M5a.rn.-7t!5e.as McCareNuclear t PbwerSistion (Open/ Closed}-The k: ' a nc as s ce unde t e members wdl hwand dasens the -
" Foun ation on the Arts and the Meanc Advisory Panel;Mee Hu anities Act of1965, as amended. propaed sesmal of tw apper hand D in uding discussion ofinformation injection system. Represamtasyss of the Pumn=n* to section 10( ofthe -
en in confidence to the Agency by IEtC Staff and she 1.ioensee will make v Federal Advisory Commi e Act (Pub. Presentations and parsaspetela the I ant applicants.In accordance with the L 92-.453). as smaa laA - eis hereby discussion, as approprute. determination of the Chairman given that a maatlag Music pub!!shed in the Federal Register of osers Persons ef this session wlR be closed Advisory Penal ( i February 13.1980, these sessions will be Prescreentag) to 'arianal Oundi as as mgwel to disens Propdetary closed to the public ourscant to on April 24-25, Information applicable to this matter. the Arts wiu be Idxo.m.-72Np.m Adranced
' subsectioris (c) [4). (6) and (9)(B) of 19e6 from 9:00 to 5:30 pm, Room 4 section 552b of Title 5. United Sta 714 of the N y Hanks Center.1100 RecetorDesigns (Open/ Closed)-Ths Code. Pennsylv ~ Avenue.NW. members will hear and dScuss features FurtherInformation with rence to of advanced twector desagns being Washinst DC 20 sos.
this meeting can be obtain from Mr. developed by DOE. Representatives of mittee This ingis forthe purpose of
}ohn H. Clark. Advisory the NRC Staff and of DOE wiR make Management Officer. tional Panel view. discussion. evaluation.
. Washington, and commendation on applicatioe6 far presentations and Participate in the Endowment for the I discussion. as appropriate.
DC 20506, or call
' I assistance under the '
)682-5473. Portions of this session wi!! be closed ion on the Arts and t as necessary to discuss applicahfe Dated: March 1986. umanities Act of 1965, as ded,
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n H.Cla incldng dis os of un Proprietary Information. Dancsar.O aceofCmacilendPanel given ta tenfidence to Agency bY 1:M p.m. 2,*15p.m Quantificationof
. qparati Nouar alEudowmenfarthe Arts. srent applicantsJa with the Health 27/ects in Paobabliistic Aisk , ill sN354 Filed +-2.ae; aAs amj hairman Assessmeals (Open)-The mesibers w (FE determination of ca.g ,,, . published in the Registerof ' discuss the quantifications in PRA of e February 13.1 these ensaiome wdl be health affeets andpublic risk. -
< blic persuant to 2:15p.m.-J:45p.ma Reoctor closed to
' Music AdvisoryPanet; Meeting Operations (Open/ Closed)-The ection b 5. United tes members will hear and discuss a report i
Pwsuost to section 1olaM4 of the Federal AdvisoryComnattee Act(Pub.- Cod L 9e-4a3), as amended. notice is hereby infanmauan with reference to h$.Its ating g- meeting com he aimmined fan Mr. auclear power plants. given that a meeting of the Music
,e Advisory Penel(Composets Fellowships H.Qari % rn=dttee ' ' " " ' ' ***** *I Prescreening) to the National Council m Manageenent Officer. National **"*U ""** 'U the Arts will be held on May 4-6,1966 Endowment for the Arts. Washington. '"" *"#"I frern st00 a.m.to sc30 pm Room 714 oI DC 20506, or cali(202) se2-6433. luformation partmining to die incilities t}'e Nancy Hanks Conser.1100 Dated: Men:h as.ises.
being db-4 A 4:00p.m.-deptma Qocat:%styre
- Pennsylessia Avenue.NW
- John H. Clark, Wasirington DC 3e50tk Sofety Goals (Open}-ne members will
$ Director. Office o/Councilandranet discuss psopened methods of r - This rueeting is far the purpose of 07'rvem Maann/Kademmntfor the Ans. lamplem =Mme the NRC's quantitative
- Paned review, discussion, evehsetion.
. and r===rmiendation on appucations for [FR Dec. as-T336 Fuad 44-se: ass ami
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11494 - Feder'al Register / Vol. 51, No. 64 / Thursday, Xpril 3,'1985'/ Noticks amp.m.4:45p.m.: Future ACRS : Nuclear Station. and Quantification of ' Draft Regulatory Guide;lasuance, Activities (Open)-The members will Health Effects in PRAs. Availability discuss anticipated ACRS Subcommittee Portions of this session will be closed s as necessary to discuss Proprietary The Nuclear Regulatory Commission ictiviues. Topics proposed for has issued for public comment a draft of consideration by the full Committee will Information., . a new guide planned for its Regulatory ; also be discussed. - W p.m.-J:Ja p.m.:ACRS By/aws
- s .
(Open)-The members will discuss a Guide Series together with a draft of the < fridaymApril12 Isse proposed change in the ACRS Bylaws. associated value/ impact statement. This series has been developed to describe
&Joa.m.4Wa.m. Reliabilityof Procedures for the conduct of und and make available to the public Nuclece Components (Open)--The participation in ACRS meetings were publishedin the Federal Register on methods acceptable to the NRC staff of members will hear a report by the Subcommittee on Reliability Assurance. October 2.1985 (50 FR 191). In ' implementing specific parts of the accordance with these procedures, oral Commission's regulations and in some rwa.m.-Rd5a.m. Preparationfor . cases, to delinea te techniques used by
. Meeting with NRC Commissioners - . or written statements may be presented by members of the public, recordings the staff in evaluating specific problems (Open)-The members will discuss the. or postulated accidents and to provide
, subjects of the meeting to be held with willbe permitted only during those the NRC Commissioners, . .. Portions of the meeting when a guidance to applicants concerning ,
transcript is being kept and questions certain of the information needed by the fas a.m. -71:30 a.m. Meeting with ' staffin its review of applications for NRC Commissioners (Open)--ne may be asked only by members of the members of the ACRS will meet with the Committee, its consultants, and Staff. permits and licenses. Persons desiring to make oral The draft guide, temporarily identified NRC Commissioners to discuss the . statements should notify the ACRS by its task number, SI 501-4 (which scope and priorities of AERS activity, the design of the GESSAR 11 and safety. Executive Director as far in advance as should be mentioned in all . practicable so that appropriate correspondence concerning this draft considerations for future plants, and a arrangements can be made-to allow the guide),is entitled " Station Blackout" proposed Federal academy for training necessary time during the meeting for and is intended for Division 1. " power nuclear power plant personnel. such statements. Use of still, motion . Reactors."It is being developed to J1:45 a.m.-12:15p.m. Hearing on 1988 describe a method acceptable to the picture and television cameras during Ohio Earthquake (Open)-The members this meeting may be limited to selected NRC staff for complying with the will hear a report concerning a Portions of the meeting as determined . proposed regulation that would require Congressional hearing on the 1986 Ohio by the Chairman. Information regarding light-water-cooled nuclear power plants Earthquake and its effects. - the time to be set aside for this purpose to be capable to coping with a station 2:15p.m.-2:30p.m.t Human Factors may be obtained by a prepaid telephone blackout (i.e loss of the offsite electric (Open)-ne members will hear a report callto the ACRS Executive Director R. power system concurrent with reactor of recent activities of the Subcommittee F. Fraley prior to the meeting. In view of trip and unavailability of the onsite ( m Human Factors. the possibility that the schecule for emergency AC electric power system) 2; Jap m.-4:30p.m. Quantitative for a specified duration. Safety Coals (Open)-ne members willChairman ACMS mee, tings may as necessary be adjusted to facilitate the by the This draft guide and the associated i continue the discussion of * .- conduct of the meeting. persons value/ impact statement are being issued implementation of the NRC's ' planning to attend should check with the to involve the public in the early stages quantitative safety goals. ,. ACRS Executive Director if such - of the development of a regulatory
. 4:45 p.m.-400p.m.:S.ebcommittee rescheduling would result in major i position in this area.They have not Activities (Open)-ne members will received complete staff review and do hear reports of ACRS Subcommittee I"C""V'"I'"C
I d' l " "' n t represent an official NRC staff activities concerning auxilia feed- ,, ,n L 92 tha water system reliability, reso ution of
- f th ' P 'I'i "*
Public comments are being solicited j USI A-45, and LWR standard plant ",'C$', " ed ab to d s on both drafts. the guide (including any desfgn. l Proprietary information (5 U.S C. implementation schedule) and the draft l '& p.m.-7 2 p.m.: Decay Neot Removal 552b(c)(4)) detailed security information value/ impact statement. Comments on (Closed}-ne members will discuss a [5 U.S.C. 552b(c)(3)]. and information the the draft value/ impact statement should proposed decay heat removal system. release of which would represent a be accompanied by supporting data. This session will be closed to discuss clearly unwarranted invasion of Written comments may be submitted to Proprietary Information related to this personal privacy [5 U.S.C. 552b(c)(6)]. the Rules and Procedures Branch. topic. 3 Further information regarding topics Division of Rules and Records. Office of Satunlap. Apri/ 22.1988 . to be discussed. whether the meeting Admimstration. U.S. Nuclear Regulatory
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has been cancelled or rescheduled, the Commission. Washington. DC 20555. 8:J0 c.m.-R00 a.m. No m m.ation of- Chairman's ruling on requests for the Comments may also be delivered to New Member (Closed)-ne members- l Room 4000. Maryland National Bank
,will hear a report of the screening panel and the ' time opportunity allotted tobe can present obtainedora statements by Building. 7735 Old Georgetown Road.
l on nomination of a candidate for a prepaid telephone call to the ACRS Bethesda, Maryland from 8:15 a.m. to appointment to the Committee: ' ' Executive Director Mr. Raymond F* 5:00 p.m. Copies of comments received nis session willbe closed to discuss Fraley (telephone 202/634-3265)' may be examined at the NRC Public information the release of which would between 8:15 a.m. and 5:00 p.m. Document Room.1717 H Street. NW represent a clearly unwarranted Washington. DC. Comments will be Invasion of personal privacy. .:, . Dated: March 31.1986. l Joha c. Hoyle.- most helpfullf received by June 19.1986 i Rop a.ma-222 Noon: Preparation of . Advisory Committee Management Officer. Althbugh a time limit is given for t . 4CAS Reports (Open/Closedb-ne < . comments on these drafts, comments
' members will discuss ACRS reports on - [71t Doc. 86-74:9 Filed 4-2-es: 8:43 aml sauna come non-e-w - and suggestions in connection with (1)
Quantitative Safety Goals, the McGuire
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~ ,i, UNITED STATES
! NUCLEAR REGULATORY COMMISSION
$ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o wAsmNGTON, D. C. 20555
**... APRIL 1, 1986 SCHEDULE AND OUTLINE FOR DISCUSSION 312TH ACRS MEETING APRIL 10-12, 1986 WASHINGTON, D. C.~
Thursday, April 10, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.
- 1) 8:30 A.M. - 8:45 A.M. Report of ACRS Chaiman (0 pen) 1.1) Opening Statement (DAW) 1.2) Items of current interest (DAW /RFF)
- 2) 8:45 A.M. - 10:15 A.M. McGuireNuclearPowerStation(0 pen / Closed) 2.1) Report of ACRS Subcommittee regarding proposed removal of the upper head injectionsystem(DAW /PAB) 2.2) Meeting with representatives of the NRC Staff and the Applicant
. (Note: Portions of this session will be closed as necessary to discuss Proprie-tary Infomation related to this Station.) 10:15 A.M. - 10:30 A.M. BREAK
- 3) 10:30 A.M. - 12:30 P.M. AdvancedReactorDesigns(0 pen / Closed) 3.1) Report of ACRS subcommittee regarding proposed liquid metal cooled reactors
' being developed by DOE (MWC/MME) 3.2) Meeting with representatives of the NRC Staff and DOE, as appropriate (Note: Portions of this- session will be closed as necessary to discuss Proprietary Information applicable to these designs.) 12:30 P.M. - 1:30 P.M. LUNCH
- 4) 1:30 P.M. - 2:15 P.M. Quantification of Health Effects in PRA (0 pen)
TT) Discuss letter prepared by the ACRS during the 311th ACRS meeting regarding PRA quantification of public health risk (HWL/DWM/RPS) 4.2) Meeting with representatives of the NRC Staff, as appropriate
-- - - - - - - _ - - - - - - - . _ . - . , - - - - - . - - - - - - - - - , . - - - - - , , - , , - - - , . - . - ~ . ----.,.m -
t 9 312th ACRS Meeting Agenda April 1, 1986
- 5) 2:15 P.M. .
3:45 P.M. Reactor Operations (0 pen / Closed) 5.1) Report of ACRS subcommittee regarding recent incidents and events at operating nuclear power plants (JCE/HA) 1 (Note: Portions of this session will be closed as necessary to discuss Proprietary Information or detailed security informa-tion related to the facility being discussed.) 3:45 P.M. - 4:00 P.M. BREAK
- 6) 4:00 P.M. - 6:00 P.M. Quantitative Safety Goals (0 pen) 6.1) Continue discussion of ACRS comments regarding proposed implementations of MRC quantitative safety goals (D0/RPS)
- 7) 6:00 P.M. - 6:45 P.M. FutureActivities(0 pen) 7.1) Anticipated Subcommittee Activity (MWL) 7.2) Proposed ACRS Activities (RFF)
I i (
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t 4 312th ACRS Meeting Agenda April 1, 1986 Friday, April 11, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.
- 8) 8:30 A.M. - 9:00 A.M. Reliability of Nuclear Components (0 pen) 8.1) Report of ACRS Subcommittee regarding a generic review of improvements in motor operated valves and problems associated with check valve aging and wear (CM/RKM)
- 9) 9:00 A.M. - 9:45 A.M. Meeting with NRC Commissioners - Premeeting Session (0 pen)
. 9:45 A.M. - 10:00 A.M. BREAK
- 10) 10:00 A.M.- 11:30 A.M. Meeting with NRC Commissioners (0 pen) 10.1) Peeting to discuss:
- 10.1-1) Scope and priorities associ-ated to ACRS activities *
(DAW /RFF)
. 10.1-2) GESSAR II and safety con-siderations for. future
- plants (D0/RKM) 10.1-3) Proposed Federal academy for training nuclear power plant personnel
- Note: The Commissioners may want to dis-cuss more specifically the anticipated ACRS participation in the review of TVA management problems / technical problems 11:30 A.M. - 11:45 A.M. BREAK
- 11) 11:45 A.M. - 12:15 P.M. Report of Congressional Hearing on 1986
, Ohio Earthquake (0 pen)
/ 11.1) Report of testimony to Congress-man Udall's Comnittee (DAW / CPS /RPS) 12:15 P.M. r 1:15 P.M. LUNCH
- 12) 1:15 P.M. - 2:30 P.M. . Human Factors (0 pen)
< 12.1) Report of subcommittee activities related to consideration of human factors in nuclear power plants (DAW /JOS/EGI) 12.2) Meeting with representatives of r / . NRC Staff, as appropriate l - - - , - - - . - , . . ,w -. .._,,,_,,_.,,._._,,,,_.,.__.m_ , . _ .-,_ ,.,_ -,.,_ m ,, ,...,.,_,_,,,,...,w_ ..-._,,,,,,m,--.m..--,,w,,,,., m,-.-.--,--y,
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'. f 312th ACRS Meeting Agenda April 1, 1986 I
- 6) 2:30 P.M. - 5:45 P.M. Quantitative Safety Goals (0 pen)
(BREAK - 4:30 - 4:40) 6.1) Continue discussion of ACRS comments regarding proposed implementations of NRC quantitative safety goals (D0/RPS)
- 13) 5:45 P.M. - '7:00 P.M. ACRS Subcommittee Activities (0 pen) 13.1) 5:45 P.M.-6:30 P.M.: Auxiliary Feedwater Systems and Resolution of USI A Report of ACRS Sub-committee activity regarding pro-posed NRC Action Plan to improve the reliability of auxiliary feedwater systems and status of USI A-45 resolution (DAW /PAB) 13.2) 6:30 P.M.-7:00 P.M.: LWR Standard Plant Design - Report of. ACRS Sub-committee regarding the requirements for licensing a standardized nuclear power plant design (CJW/HA) i
.t y > 312th ACRS Meeting Agenda April 1, 1986 Saturday, April 12, 1986, Room 1046, 1717 H Street, NW, Washington, D.C. ,
- 14) 8:30 A.M. - 9:00 A.M. Nomination of New ACRS Member (Closed) 14.1) Report of ACR5 Screening Panel for nomination of a candidate for appointment to the Committee (HWL/ALN)
(Note: This session will be closed to discuss information the release of which would represent a clearly unwarranted invasion of personal with 5 U.S.C. 552b(c)(6).) privacy consistent 2 j 15) 9:00 A.M. - 10:00 A.M. Decay Heat Removal (Closed) 15.1) Discuss primary blowdown system p(roposed GAR /PAB) for decay heat removal (Note: This session will be closed to discuss Proprietary Information related to this subject.) .
- 16) 10:00 A.M. - 12:00 Noon Preparation of ACRS Reports (0 pen / Closed) 16.1) Quantitative Safety Goals (D0/RPS) 16.2) McGuire Nuclear Station (DAW /PAB) 16.3) PRA Quantification of Public
, Health Effects (HWL/RPS)
- (Note: Portions of this session will be closed as necessary to discuss Pro-prietary Information.)
( , 12:00 Noon - 1:00 P.M.. LUNCH l l 116) 1:00 P.M. - 2:00 P.M. Continue Preparation of ACRS Reports
-(0 pen / Closed)
(Note: Portions of this session will be closed as necessary to discuss Pro-prietary Information.)
- 17) 2:00 P.M. - 2:30 P.M. Proposed change in ACRS Bylaws (0 pen) 17.1) Discuss proposed change in ACRS i Bylaws to provide for meetings l
of individual members with l individual Comissioners (DAW /TGM)
- 18) 2:30 P.M. - 3:00 P.M. Complete Items Considered During This Meeting (0 pen / Closed)
(Note: Portions of this session will be closed as necessary to discuss Pro-prietary Information.)
,; -p M$ l' g MINUTES OF THE 312TH ACRS MEETING APRIL 10-12, 1986 d
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-lI o WASHINGTON, D.C.
The 312th meeting of the Advisory Committee on Reactor Safeguards, . held at.1717 H Street, N.W., Washington,.D.C. was convened by Chairman D. A. Ward at 8:30 a.m., Thursday, April 10, 1986. [ Note: For a list of attendees, see Appendix -I. P. G. Shewmon was unable to attend the meeting. J. C. Ebersole did not attend the meeting on Saturday, April 12.] Chaiman D. A. Ward noted the existence of the published agenda for this meeting, and identified the items to be discussed.- He noted that the meeting was being held in confomance with the Federal Advisory Committee 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 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 Capitol Street, Washington, D.C. 20001.] y I. Chairman'sReport(0 pen) [ Note: R. F. Fraley was the designated Federal Official for this portion of the meeting.] ~ D. A. Ward mentioned that V. Stello, Jr. has now been formally appointed to the position of NRC Executive Director for Operations. He also indicated that the NRC will definitely . be moving and consolidating in an office building in White Flint, Maryland. The ACRS will probably now be relocated at -the Phillips Building in Bethesda instead of the East-West Towers as originally reported. F. J. Remick has been invited to give a talk of his views as they relate to ACRS activities at the annual Atomic Safety and Licensing Board Conference in June 1986. Chairman Ward announced that the Perry Nuclear Power Plant, Unit 1, has been issued a 5 percent low power license by NRR effective as of March 18, 1986. He noted that he, C.P. Siess, and P. Pomeroy, ACRS consultant, gave testimony at an April 8,1986 hearing on the 4 1986 Northern Ohio Earthquake and its effect on the Perry Plant. i II. McGuire Nuclear Power Station Proposed Removal of Upper Head i Injection System (0 pen) [ Note: P. A. Boehnert was the Designated Federal Official for this portion of the meeting.]
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s p 312th ACRS Minutes 2 July 7, 1986 D. A. Ward reported on the March 26, 1986 meeting of the ECCS Subconunittee on Duke Power Company's proposal to remove the Upper Head Injection System (UHI) at its McGuire Plant. He discussed the background of the inclusion of the UHI on Westinghouse Ice Condenser Plants. He indicated that the first ice condenser plant, D.C. Cook, when licensed 10 years ago, and was found to have difficulty in satisfying Appendix K requirements for ECCS performance without the placement of restrictive limitations on power peaking in the core (see Appendix IV). The reason for this involved the limiting case of a large break LOCA which would blow down into the ice condenser containment. The back pressure in the containment was calculated to be so low that vapor forms resulting in steam binding in parts of the reactor cooling system. The steam binding caused increased pressure drops in both the steam generator and the main coolant pump. This resulted in decreased flow into the core or a lower rate of reflooding, and the calculated peak-clad temperatures were higher than they would have been for similar conditions in a plant with a large dry containment. To counter this effect, for the next ice condenser plant (McGuire), the Licensee and Westinghouse agreed to add another piece of equipment to the ECCS system. An added accumulator would inject several thousand gallons of water directly into the upper head (UHI System) in the event of a large break LOCA. D. A. Ward described the complexity of the UHI System and some difficulties the Licensee, Westinghouse, and the NRC Staff have had in analyzing ECCS performance with the inclusion of the UHI System. Although all concerned finally concluded that the UHI System would perform satisfactorily, Duke Power's experience in operating McGuire with the UHI System has been fraught with operational problems involving maintenance of water levels and nitrogen pressures. l D. A. Ward explained that after some analysis with the latest ECCS evaluation model codes, Duke Power decided to petition the NRC for permission to remove the UHI System with the argument that new methods of analyzing the large break LOCA show that peak-clad temperatures are not excessive even without the functioning UHI . All Appendix K requirements can be met in the plant without the UHI System and without the necessity for unusual restrictions on core peaking values. The Staff has analyzed Duke Power's proposal. The ECCS Subcommittee has also reviewed the proposal and agrees that Duke Power has some good reasons for wanting to remove the system although they are not strongly safety related. The ECCS Subcommittee disagrees somewhat with Duke Power and the NRC Staff in that it believes that there are some safety related reasons for removing the UHI system. These involve the reduced complexity of the plant, particularly of the Emergency Cooling System (amount of ! piping in vulnerable locations and its vulnerability to failure) and the complexity of operating the plant both during normal and i off-normal situations. There is also concern that the UHI l Accumulator System provides another major scurce of noncondensable I nitrogen gas that could be injected into the reactor coolant system l and could conceivably interfere with good heat transport.
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312th ACRS Minutes L3 July 7, 1986 Subccmittee believes that there is a safety benefit as well as a practical operational benefit from removal of the UHI System. D. A. Ward indicated that the Subcomittee disagrees with the claims of Duke Power and the NRC Staff that removal of the UHI system will result in essentially no difference in the maximum peak-cladding temperature calculated for the McGuire plant in the event of a large break LOCA event. The McGuire plant, while. meeting the requ1ations without UHI, will have a large break LOCA performance quite similar to most all other PWRs in that there is a fairly high likelihood in best estimate analyses that a significant clad temperature peak will occur during the reflood phase. The UHI system apparently does mitigate the peak-cladding temperature in a large break LOCA situation, but that is really not necessary to meet the regulations. He indicated that the Subcomittee recommends that the full Comittee endorse the Staff's proposal to accept the Duke Power application for removal of the UHI System at McGuire. D. Okrent questioned the basis for the Subcomittee's recommendation as to whether it was based on factual infomation i or engineering judgment. He could not see a quantifiable safety benefit from removal of the UHI System. He suggested that insufficient information has been presented about the benefits or penalties of inclusion of the UHI system regarding small to medium LOCA scenarios. He did not see a technical justification to give blanket approval for physical removal of the system. He suggested that the Comittee recomend that Duke Power operate the plant in its next fuel cycle without the UHI system in operation but not physically removed from the plant. ~ C. Michelson pointed out that the internals associated with the UHI system are unique and quite different from the standard set of Upper Head Internals. They constitute a barrier between the upper head of the reactor vessel and the upper plenum of the vessel. This barrier then directs the flow down into the core when the UHI actuates. He expressed concern as to whether there is adequate understanding of the behavior of the UHI vessel from the viewpoint of hydraulic flow during blowdown. He wondered whether evaluation models accurately represent the situation with the UHI disconnected and the internals still present. He expressed concern as to how the disconnected UHI and present Upper Head Internals would affect the Westinghouse Reactor Vessel Level Indication System (RVLIS). He also expressed interest in how the standpipes which are part of the Upper Head Internals will be disconnected. C. Berlinger, NRC, reviewed the history of UHI removal proposals and submittals. He indicated that it was late 1984 and early 1985 i that the Licensee and Westinghouse entered into preliminary discussions with the NRC Staff to remove the UHI System. The - original submittal proposing a technical specification amendment for operation at 100 percent power with the UHI deleted was submitted in !!ay 1985. In October 1985, the Staff received the Safety Analysis that provided the basis for the UHI removal in the
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- e 312th ACRS Minutes 4 July 7, 1986 form of an analysis on a preliminary version of the BASH code. In early 1986, the Staff's preliminary draft SER identified several areas of concern regarding the BASH calculation. As a result, the Licensee submitted a revised Safety Analysis in March 1986 using a 1981 naluation model for non-UHI plants, a previously approved BART ccde.
Mention was made of difficulties with regard to maintenance of operations and operational problems associated with the UHI system. C. Michelson asked if other Westinghouse plants that have UHI are reporting similar difficulties. D. Hood, NRC, indicated that the TVA plants, Sequoyah, has reported and Watts Bar, anticipates similar problems as have been reported at the non-TVA Catawba Plant. The plan was to wait and see what happened on McGuire regarding their interest in removing the UHI system. The Committee discussed the flow complexity without the UHI Upper Internals with respect to the effects of removal of the UHI system. C. Michelson noted that there is a different time of injection flow with the UHI Internals present even though the system is disconnected. N. Lauben, NRC, indicated that the UHI Internals are still modeled in the current calculation for UHI because they are unique and the flow paths are unique. The UHI Upper Internals are accounted for in the calculations even though the UHI piping has been taken out. The Committee discussed fuel cladding data for large break LOCA scenarios assuming perfect mixing and imperfect mixing with the approved evaluation model. C. Berlinger discussed the impact of removal of the UHI system and the effects of nitrogen injection. C. Berlinger indicated that the overriding reasons for removing the system are basically operational and maintenance problems, and problems associated with operating experience that have occurred at McGuire and at other plants having this system. The real benefit from removal of the system is that you can eliminate or reduce the likelihood of a small break LOCA. This will result in reduction of the overall risk of this plant. While the UHI system was installed initially because of difficulties in satisfying large break LOCA acceptance criteria, the removal of this system, in fact, reduces the risk associated with the more likely or more probable event of a small pipe break. D. Okrent indicated that I. Catton, ACRS consultant, is a bit perplexed about the seeming lack of containment back pressure. C. Berlinger indicated that the fact that an ice condenser containment does produce lower containment back pressure has always been believed to contribute to lower reflooding rates. Therefore, it has always led to higher calculated peak clad temperatures using the evaluation models. D. Okrent asked if the proposal regarding the UHI system is to do something that is irreversible. C. Berlinger indicated that there are three options based upon the 1985 technical specification change request. One option is to do nothing to remove the system other than to valve it out. A second 1
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i,. 4 312th ACRS Minutes 5' July 7, 1986 IJ alternative 'would be to disconnect and cap the system and physically remove it. The third alternative is to do nothing and continue operating the . plant the : way it is.- The technical specification change request considered all-three cases. D. Okrent
- asked what ' the Staff - is proposing to approve. D. Hood, NRC, indicated that the proposal the Staff plans to approve would be to
. provide for' operation for one fuel cycle with the UHI system valved
- out and physically present. There would be nothing done that is
-irreversible. The proposal also provides for modifications to be -
c made to the cold leg injection accumulators when' the system is valved out before the plant starts up again.- That would ' include alterations in the liquid volume and the gas pressure for the cold
- leg injection and accumulators. After one year of operation, the components inside containment would be. removed from that unit. The lines would be capped off and physically removed.
! W. D. Reckley, Duke Power, discussed operational problems at McGuire involving the UHI system. He mentioned lost availability as a result of fill-and-vent procedures. He indicated that level instrumentation has been a big problem due to LOCA analysis assumptions (see Appendix VI). Mention was made of sight glass failures which occurred during a fill-and-vent procedure. G. A. Reed and C. Michelson did not think that the pressurizer relief tank and relief valve portion of the system with the sight glass was particularly well designed. C. Michelson did not think it met the ASME code. L j W. D. Reckley indicated that Duke Power considers the removal of the UHI system to produce a net increase in plant ' safety. D. - Okrent asked him to elaborate on that statement. W. D. Reckley i indicated that the benefit is not quantifiable but is based on the i fact that very little benefit is shown for the UHI and large break-LOCA analysis regarding the results of the peak-cladding temperature. Any benefit as far as preventing any other type of g' ' accident means' removal will be a net increase in safety. He noted that RVLIS, the reactor vessel instrumentation system at McGuire, [ is not affected by the UHI removal. The tap for RVLIS on the Upper Head comes from a spare control rod drive mechanism and is not connected to the UHI piping. Duke Power concludes that the ! elimination of UHI will provide significant operational benefits. Removal of. UHI will eliminate safety concerns that some people associate with the UHI system, such as breaks outside containment, j' and nitrogen injection for small breaks. It will also reduce occupational health affects through reduction of radiation exposure to workers. C. Michelson continued to express concern that the UHI internals would deleteriously affect the RVLIS. l D. Okrent noted that a strong thermal hydraulics presentation was made at the Subcomittee meeting. Yet the ACRS consultant experts - 1 in thennal hydraulics seemed dissatisfied with the information presented. D. A. Ward indicated that the consultants' responses j were not, however, particularly negative. W. Kerr asked if any of the consultants recomended against removal of UHI. D. A. Ward i indicated that they did not. In answer to C. Michelson, D. A. Ward i
312th ACRS Minutes 6 July 7, 1986 indicated' that the consultants recognized that the problems 4 involved departure from the original analysis,. the performance of the Plant with the UHI and the original licensing basis. He suggested that removal of the UHI should not increase the uncertainty in the calculation. H. W. Lewis noted that the idea. of pouring cold water into the Upper Head used to be very attractive. Yet, I. Catton argues for not disabling the UHI until next year. He wondered whether it was,
- in fact, untrue that pouring cold water . on the- core cools .it, whether the original design was flawed, or whether the requirements on the use of the system suggest that it is b'est to disable the system.
C. Michelson_ indicated that if the power plant went back to Westinghouse - internals configuration for non-UHI plants and , standard ECCS calculational models everything would be fine. The problem is that the water injection is to be eliminated but the internals retained and the standard ECCS calculational models still used. The Committee should be sure that those internals do not introduce unique hydraulic problems that the calculations might not take into account.- W. _ Kerr thought that Westinghouse would have difficulty dealing with this problem because they are forced to use , evaluation models which do not give a desired answer. C. Michelson agreed. N. Lauben indicated that the Staff has done as much as it can in the' current context of regulations to encourage the use of best' estimate calculations principally through SECY 83-472. He noted that there are many sensitivity studies that need to be done. To do appropriate sensitivity studies to quantify the uncertainty 4 with best- estimate calculations 'makes it quite expensive. 'J. C. Mark explained that I. Catton would very much like to see a comparison with and without the UHI system even though sensitivity analyses are not performed. The Committee discussed thermal hydraulic modeling of ice condenser p plants with and without Upper Head Injection. The TRAC, COBRA / TRAC, and RELAP 5 best estimate codes were discussed, as well as the BART and BASH evaluation models codes. D. Okrent pointed out that the sequence involving the loss of all AC power leading to a loss of the primary pump seals in a small LOCA is a potential scenario where the UHI might have a beneficial effect. B. McIntyre, Westinghouse Electric, acknowledged that it might buy some time after 20 hours. But, it is a tradeoff since for a small break, the pressure might not get down to the UHI setpoint (1000 psi). He noted that the UHI water is really of no impact on the large break LOCA sequence. B. McIntyre concluded that some significant advances in ECCS technology have been made since UHI
- i was first installed in 1974. However, even using currently approved ECCS mcdels, Westinghouse can show that McGuire meets all ,
i of the regulatory requirements of 10 CFR 50.46 without UHI. ) C. Michelson expressed concern regarding the accuracy of the RVLIS t after the disconnection of the UHI system and the retention of the 4 UHI internals. B. McIntyre explained that Westinghouse was l 1
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312th ACRS Hinutes July 7, 1986 7 confident that there would not be a problem with level indication, especially in the case of the transients where level indication would be important. D. Okrent indicated that in his opinion a strong thermal hydraulics case was not made at the Subcommittee meeting. He noted that the ACRS consultants in that area are not satisfied and that I. Catton has called for a reversible change which has been suggested by the Staff for next year. III. Reactor Operations (0 pen) [ Note: H. Alderman was the Designated Federal Official for this portion of the meeting.]
- 1. Implementation of TMI Action Item I.D.2 - Safety Parameter Display System at Operating Reactors G. Lapinsky, NRR, indicated that NRR sponsored a limited survey of operating plants to identify the status and quality of implementation of Task Action Plan, Item I.D.2, Safety Parameter Display System (see Appendix VII). SPDSs were found to be either malfunctioning or unavailable at three of the six plants surveyed. A substantial fraction of the systems did not fulfill the intent of NUREG-0737, Supplement 1 and were providing invalid or inaccurate information that could mislead users, and that at two of the plants operators refused to use the SPDS because it was so unreliable. W. Kerr suggested that the SPDS is an outstanding example of an NRC Staff-required design for a nuclear power plant. The Staff specified the requirements for the design, and the equipment was installed before the designs were ever approved. The ultimate purpose of the SPDS was never properly defined at the outset. D.
Okrent mentioned attempts by EPRI (which were well-focused) to develop a display for the operator in off-nonnal situations. W. Kerr noted that there really was no requirement for design, simply incentives to the utilities by the Staff to design an SPDS in a hurry and to install it in a hurry. G. Lapinsky suggested that this was a good faith effort on the part of industry which failed for a variety of reasons. W. Kerr asked if there were criteria used to judge SPDSs. G. Lapinsky indicated that the criteria generally used were the Standard Review Plan and engineering judgment. The reviews done were not compliance reviews. There was a general lack of awareness regarding software maintenance and sometimes it was found that plant management did not really accept the concept of SPDS. He also noted that there were no technical specifications regarding the operation of the SPDS. C. Michelson asked the purpose of the SPDS. G. Lapinsky indicated that it is to consolidate information for the operator in a convenient location so that related parameters can be viewec at the same time giving an overview of the plant
J 312th ACRS Minutes 8 July 7, 1986 safety status in a timely fashion. If actions are necessary, they can be appropriately initiated. C. Michelson wondered why the operators would even bother with the SPDS if they did not trust the displayed values. He noted how emergency operating procedures have been tied into the SPDS. He expressed his disenchantment in the fact that the SPDS concept had been a failure. G. A. Reed suggested that there might be root causes behind these poor performances such as the inability to translate the concept into hardware and real time instructions. G. Lapinsky indicated that an information notice with recommendations had been issued in February 1986 and the Regions attempted to inspect their particular plants for major problems. The findings were consolidated into a draft NUREG report which will be available by June, 1986. These resultant findings have caused the Staff to recommend that in-depth audits be conducted at all plants using the Standard Review Plan and Supplement 1 of NUREG-0737. The suggestion was made by W. Kerr that the subject be investigated further by the ACRS Instrumentation and Control Subcommittee.
- 2. Failure of Standby Liquid System E. Wise, Liquid IE, explained Control System that onValves Squib February)8,1986 (2 failedthe to Standby actuate during the surveillance test at the Vermont Yankee Nuclear Power Station, in part, due to failure of the local terminal box, but primarily due to the fact that the squib valves had incorrect pin-to-bridgewire grouping. Both trains of the Standby Liquid Control System failed.' One interesting aspect of this event was that throughout the 18 months during which both trains of the Standby Liquid Control System were unavailable, there was a pulse indication in the control room of circuit continuity. The Staff's immediate concern was that other plants could have these defective squib valves and if
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the problem occurs in the future, false indications in the control room could be misleading. E. Wise indicated that engineers involved in the manufacture of the squib valves confirmed that defective assemblies were shipped to the Vermont Yankee site. A further investigation indicated that no other defective squib assemblies were found. Some suspect squibs were distributed overseas, and the distributor was notified as well as the NRC Office of International Programs. An Information Notice was promptly published, and a Vendor Inspection Plan developed. J. C. Ebersole described the purpose of the bridgewires, and noted the manufacturer's attempt to be super conservative and provide redundancy since the four bridgewires were not put into parallel configuration but into two separate circuits. It invited a malconnection potential problem. E. Wise indicated that, when the valve is fired, the opening of the valve permits concentrated sodium borate to pass through the
312th ACRS Minutes 9 July 7, 1986 valve. A pump starts coincident to the valve firing and sodium perborate is pumped into the bottom of the reactor vessel to shut down the chain . reaction in the event of an accident. G. A. Reed asked about the origin of the incorrect wiring of the bridgewires. E. Wise indicated that the incorrect wiring was a human error, a misinterruptation of the wiring connection designations at the manufacturing facility in Florida. W. Kerr indicated that now that the Staff knows what the problem was, what does it intend to do about it? E. Wise indicated that the cure is to deliver the squib valve the , way it was promised. J. C. Ebersole indicated that it would be better to throw away the squib valves and put in a valve that would be in either a single circuit configuration or two parallel circuits. G. A. Reed agreed that such a design error should not be obviated by the use of new procedures because it places a potential additional burden on the operators.
- 3. Class 1E Station Battery Problems H. Bailey, IE, indicated that an extensive erosion (flaking) problem with the Class 1E Station Batteries at the Rancho Seco Plant was reported to the NRC Staff on March 28, 1986. This extensive erosion occurred inside the battery near the top of the plates between the plates and their supports. It was also found that the battery plate spacing was incorrect. The significance of this incident is that the Class 1E Station Batteries are not seismically qualified resulting in a potential for loss of all DC power during a seismic event. He indicated that the erosion problem was discovered by the battery vendor in February 1986. The present technical specifications at Rancho Seco would not reveal this problem.
Newer plants with standard technical specifications do have a definite procedure for inspection of the batteries for this type of problem, but some of the older plants do not have such a technical specification, including the requirement for specific surveillance to check for this situation. The vendor report was that erosion was not unusual for a " plate" type battery more than 15 years old. The batteries in question did not show significant erosion one year ago during a vendor inspection. Since the battery rack and cell-to-cell spacing was also found to be out of design specification, the vendor informed the Licensee that the batteries were not seismically qualified due to improper battery-to-end-rack spacing. Spacing has been corrected. He indicated that, as followup, the accuracy of the Licensee's battery surveillance program is under review by the NRC Regional Office. Concerns include the whole range of surveillance as well as erosion problems. Some Commonwealth Edison plants were reported to have the same type of battery and Commonwealth Edison is replacing them. Rancho Seco plans to replace these plate lead type batteries with a l 1ead calcium type battery before returning to power. C. Michelson and J. C. Ebersole expressed concern that older batteries in operating nuclear plants could structurally pass a seismically qualification test on a shake table. The
.w , 312th ACRS Minutes 10 July 7, 1986 Committee discussed the fact that an older lead calcium type battery may or may not fail the seismic qualification test on a shake table if it is naturally aged. It is. a random phenomenon.
- 4. Component Cooling Water System Problems D. Mcdonald, NRR, indicated that Florida Power & Light Company found during a Safety System Review that the discharge valves in the RHR heat exchangers at their Turkey Point Nuclear Generating Plant, Unit 3 were set at 30 percent open. During this Safety System Review on February 24, 1986, the Licensee could not find available documentation for that setting. There was lack of design documentation to assure v. hat there would be adequate flow through the RHR Heat Exchangers given the most limiting single failure. The result would be deleterious effects on multiple safety-related systems. The Licensee went to the nuclear steam supplier, Westinghouse Electric Company to determine what would be the single or minimum flow required and to determine if the setting was proper. What they found was that in 1972 Westinghouse supplied RHR heat exchangers to several facilities and identified at Indian Point a tube vibration problem in the RHR heat exchangers. They recommended that the flow be throttled at Turkey Point to assure that it would not create a problem. The Licensee was aware that the tube vibration problem was corrected in 1975, but they still could not find any documentation that certified tests were performed to assure that a minimum flow would occur, nor did they know the Plant's specific value for the minimum flow. Westinghouse indicated that the generic value that would assure adequate flow in the most limiting condition for the RHR heat exchanger would be about 4,000 gpm. Based on that infomation and the calculation showing that the valve settings would be somewhere between 2,000 and 3,000 gpm, the Licensee opened the valves to 100 percent flow pending verification by a test on Unit 4. Unit 3 was at power, but Unit 4 was in a refueling outage. Tests perfomed on Unit 4 showed that in the 30 percent position, flow through the RHR Heat Exchanger was approximately 2,500 to 3,500 gpm, less than the 4,000 generic value from Westinghouse. A test was performed with a single component cooling water pump and a single RHR heat exchanger, and all loads that would not normally be isolated during an accident condition. The loads assume minimum flow would be 4,000 gpm through the RHR Heat Exchanger. The other major load is through the Emergency Containment Coolers which is 2,000 gpm each for a total of 6,000 gpm. When the Licensee opened the valves to a 100 percent open, they got in excess of 8,000 gpm through the RHR Heat Exchangers. However, the required amount was 6,000 gpm total through the three emergency containment coolers.
Because of the flow imbalance that was discovered, the Licensee declared Unit 3 inoperable because of a declaration of the inoperability of the Emergency Containment Coolers and Unit 3 was brought to shutdown.
312th ACRS Minutes -11 July 7, 1986 The Licensee informed the NRC of component cooling water balance tests in an attempt to reconstitute the component cooling water design basis. It was indicated that the systen was adequately protected and balanced when it was initially set. The final settings on the valves were 35 and 38 percent open. In the process of doing the complete system flow balance, the balance for the Emergency Containment Coolers were also adjusted. D. Mcdonald explained that detailed team inspections by IE Headquarters and the Regional Office regarding problems with ' the auxiliary feedwater system at Turkey Point led to concerns about potential problems on other safety-related systems. As a result, the Licensee did a safety-related system review which identified 13 systems, including the Reactor Protection System and the Auxiliary Feedwater System. All were identified as safety related with the exception of Instrument Air. It was during the initial phase of this two-phase Safety System Review that the 30 percent setting on the throttle valves in the Compcnent Cooling Water System was observed and i that there was a lack of documentation to support why. The first phase of this Safety System Review identified such things as labeling, valve line-up, maintenance and other potential problems. Phase II of the Review is termed a reconstituted design basis consistent with the initial licensing and analysis of the plant. W. Kerr wondered if this planned design review might be likely to make things worse than they already are. D. Mcdonald thought that was not likely. He pointed out that there have been many modifications to the original design basis for these systems over time. This review will look at the original design basis and factor in all of the changes that have taken place.
, D. Okrent speculated on how the position of this valve would be documented or maintained after maintenance support tests have been performed. D. Mcdonald thought that, once these valves are set and properly balanced, unless there is a ,
problem with the valves, there usually is no periodic i maintenance on such a passive component (as with the rest of the piping system). D. Okrent asked if they might not be tested for the life of the plant. D. Mcdonald did not know. D. P. Allison, IE, indicated that typically manual valves are not disturbed for the life of the plant. D. Mcdonald suggested that if the valves are multi-purpose, they would certainly be tested periodically.
- 5. Follow-up Discussion to Operating Events G. A. Reed expressed concern regarding two incidents at the McGuire Nuclear Station, Units 1 and 2. He was bothered a bit by the report at the Subcommittee meeting that McGuire made a start-up with two valves inoperable (which should have been operable). It was not specifically a technical specification violation, but a case of a desire on the part of the utility to increase the availability of the plant which might have i
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312th ACRS Minutes 12 July 7, 1986 compromised safety. J. C. Ebersole mentioned a second event at McGuire which involved a substantial prcblem with siltation or filling of the heat exchanger piping system. Clay reservoirs are silting up. He wondered whether this siltation process was being considered in the context of a silt deposit run-down or closure of the suction to this low lying intake. He noted that the Subcommittee expressed its concern regarding the fact that the Staff does not intend to look into the siltation matter any further. W. Kerr indicated that he thought that the details of these events at the plant were extremely interesting. He indicated, however, that he would be more interested in finding out about how the Staff intends to correct these problems. Also of importance would be the identificction of the potential generic significance of these events as far as operation as a whole is concerned. Somewhat len detail on what has happened and more emphasis on what the Staff is going to do about it would improve the balance of the presentations. D. Okrent agreed with W. Kerr that the ACRS needs to see more of what the Staff intends to do about these incidents despite the fact that the presentations are interesting and relevant. IV. Advanced Reactor Designs (0 pen) [ Note: M. M. El-Zeftawy was the Designated Federal Official for this portion of the meeting.] M. W. Carbon reminded the Committee that NRR is working with the Department of Energy in the review of advanced reactors. There is one activity in process in the gas-cooled area and two separate design activities in the LMR or liquid metal reactor area. T. Kane, NRC, indicated that NRR is negotiating with DOE internally regarding the resolution of budgeting constraints. At the present time, Staff activities on all advanced reactors are on hold. F. X. Gavigan, DOE Director of Advanced Reactor Programs, explained that the three designs, one a GCR and the two LMR designs, are competing with one another to achieve DOE's goal of a 50 mils per kilowatt hour passively safe (to the maximum extent possible) electrical generating option, competitive with contemporary ! alternative technologies. He described the proposed plant l configurations, utility problems and issues, and the characteristics of all three advanced reactor concepts. He indicated how DOE intends to fit the NRC's Advanced Reactor Policy Statement requirements into its objectives (see Appendix VIII). In discussing safety approaches, he addressed the matter of active devices, passive devices, and inherent processes as they applied to the three designs. J. Bruning, Rockwell International, discussed the Sodium Advanced Fast Reactor (SAFR) Design Program and its approach to safety and licensing. He discussed the SAFR program objectives which are to develop an LMR design to meet the electric industry market needs. The SAFR Program Team consists of Rockwell International and i __ - ._ _ - _ _ . - _ . - - - . . _ - - - - - . - . -- - - - - - - - -
312th ACRS Minutes 13 ' July 7, 1986 Bechtel Corporation. Rockwell has the overall respcnsibility of. system engineering, safety and licensing activity, . marketing and commercialization work. Combustion Engineering, another team member, provided support for licensing and commercialization as . well as specific responsibility for components. Bechtel has the responsibility for the conventional ' A/E activity. Significant-support from the national laboratories and base technology programs . have been forthcoming. He explained that SAFR goals are to assure public protection through the use of natural forces. Inherent design features provide investment - protection to the utilities through low plant damage probability and low accident-caused
. downtime.. The plan is to make maximum use of existing' technology, -
have (a 60 ayear plant capacity lifetime) andfactor greater than.80 low personnel- percent, radiation a long(life exposure less than 25 person-rem per year). The plant must be competitive with coal-fired generation, have less than a 4-year construction schedule and have a standard plant design that can be utilized on Financial risk to the the majority of: the existing U.S. sites. utility is to be limited by closely matching the capacity of the initial units to the growth in demand. It has to be a simple design, minimizing on-site work, and considering safeguards and security from the outset in the design effort (see Appendix IX). J. Bruning discussed the overall approach to licensing indicating that Rockwell International has initiated a dialogue with the NRC Staff which has reviewed many of the design features of the plant. A critical ingredient of the program is achieving an SER and a letter indicating the licensability of the plant at the end of the conceptual design stage before moving into the preliminary design phase. F. J. Remick asked if the licensing process is geared toward one or two step licensing. J. Brewning indicated that Rockwell International is orienting the activity to a one step j licensing process. J. Bruning discussed the major options Rockwell International had in developing an initial LMR design. He spoke of studies to 1 determine an optimum plant size and arrangement, an optimum fuel type, alternates for steam cycle and options available for safety ! systems. He noted that Rockwell International concluded that j utilization of large shipable assembly modules was the best compromise to meet current utility concerns and achieve a balance between economies of scale and economies of production. F. J.
- Remick asked to what extent plant maintainability and replacement i
of component parts have been considered for barge shipping. J. Bruning indicated that maintenance aspects of the plant have been included since its inception. The main components are removable for replacement. The reactor assembly is an all welded assembly, completely shop fabricated and shipped to the site. There is little nuclear grade welding done at the site. The goal is to minimize all nuclear grade construction activity at the site in order to achieve lower capital costs. G. A. Reed expressed concern regarding working space for personnel and access to the pipes and
< parts to maintain them. J. Bruning contended that access to all major pieces of equipment has been provided. H. Etherington asked r
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312th ACRS Minutes 14 July 7, 1986 if the reactor vessel will be inspectable throughout its life. J.
.Bruning indicated affirmatively.
J. Bruning discussed the inherently safe metal core configuration. F. J. -Remick. asked if the fuel requires full reprocessing capability. J. Bruning- indicated that the reprocessing can be j accomplished on . site. This 'is a part of the integral - advanced reactor program at Argonne National Laboratory and demonstrations
. have been done. -
The intent of the reprocessing effort will be to have .the plant loaded with uranium fuel at the start and no plutonium shipped offsite during the lifetime of the plant. G.'A. Reed expressed surprise at the selection of a superheated steam cycle with its more complicated design requirements and uncertainties regarding the steam generator. J. Bruning indicated that the steam generator is not deemed to have a safety impact. G. A. Reed thought it was not the right way to go if the plant is trying to achieve passive reliability. L J. Bruning described the basic reactor assembly as a small pool configuration which contains all of the radioactive material. He i noted that there are not penetrations into the reactor vessel. J. C. Ebersole asked if the plant stack will be tornado proof. J. Bruning indicated that it was. I R. T. Lancet, Rockwell International, discussed some of the safety characteristics of the SAFR Design. These include two independent decay heat removal systems, as well as passive natural convection decay heat removal. J. C. Mark asked about the design of the containment building. R.T. Lancet indicated that it is a low pressure containment structure of about 3 psi which is set by the tornado requirement. He indicated that there is no accident' postulated for this reactor that will produce a - pressure high enough to threaten the containment. D. Okrent asked what accident scenarios or combination of situations can be considered either incredible or of sufficiently low probability to be discounted as potential failures which would challenge the existing containment. R. T. Lancet indicated that the sodium piping is enclosed by another high pressure capable pipe with an inerted space between them. Only in case of the failure of the sodium and the guard vessel piping would pressurizing the building be considered but this failure is not considered credible. There is leak detection in the inerted space between the two sets of piping. D. Okrent asked if failure of the reactor vessel and the guard vessel was L considered. R. T. Lancet indicated that the guard vessel is considered containment so that by definition, failure of the guard vessel would be failure of containment. There would be loss of sodium and the core would become exposed and would melt. D. Okrent asked if such a scenario were judged probable, or in the range of interest, would the containment building design be modified. R. T. Lancet indicated that the probability of a failure of both the reactor vessel and guard vessel is considered of such a low probability as to be totally outside the range for consideration.
312th ACRS Minutes 15 July 7, 1986 R. T. Lancet described certain accident scenarios such as the transient overpower without scram event and the loss of heat sink without scram event. He noted that, to drar! the analogy with light water reactors in terms of concerns for accidents, the corresponding loss of cooling accident without scram for- water reactors is loss of flow. The equivalent for SAFR is the loss of power to the pumps and the coast down of the pumps. There is no credible way of losing coolant in a low pressure sodium reactor. J. C. Ebersole asked if Rockwell International had looked at the subject of security of the plant regarding sabotage. R.'T. Lancet indicated that quite good security for protecting the vessel and guard vessel from sabotage has been developed and no way has yet been identified that an operator and an intelligent individual can defeat the plant. R. T. Lancet discussed the loss of heat sink accident which is analogous to the transient without scram for a water reactor. He discussed a loss of pumps without scram event for the RAPS 0 DIE reactor, an experimental 20 megawatt liquid metal reactor in France. He also discussed an EBR-2 loss of flow without scram test from 100 percent power. R. T. Lancet discussed reactor vessel cavity cooling, and passive natural convection heat removal. It was noted that helina flowed in an annulus between the reactor vessel and the guard vessel. J. C. Ebersole asked if the helium was under pressure. R. T. Lancet indicated that it was not. D. Okrent asked what the criterion was on preconceived core melt or melting of largg portions of the core. R. T. Lancet indicated that the goal is 10 per reactor yeay for all events internal and external including seismic, and 10 per reactor year for internal events exclusive of seismic. D. Okrent asked if Rockwell International expects to maintain integrity of the reactor and guard vessel if a large scale core melt occurs. R. T. Lancet indicated that the answer to that question would probably be "No". D. Okrent then suggested that Rockwell International ought to think about containment performance under that situation. R. T. Lancet indicated that the scenario has been assessed as part 4 of the risk assessment in the PRA and, while its consequences are not catastrophic, they are not insignificant. R. T. Lancet discussed steam generator building layout, sodium leak collection and sodium fire mitigation. C. Michelson wondered from the viewpoint of sabotage whether it is possible to create a much more serious fire than the simple one that is being postulated. He asked if Rockwell International has considered the effect on safety-related equipment knowing that the steam generator building as well as the whole secondary system is considered a non-safety grade system. R. T. Lancet indicated that it has been considered and the primary concern was found to be impact on decay heat removal for an accident fire four times worse than the one postulated. There was about a one degree in temperature degradation in decay heat removal. Therefore, it has been analytically shown that there is no catastrophic impact. l l _ __ _ _ _ _ _ . - - _ _ . ~ _ -
y o 312th ACRS Minutes .16 July 7, 1986 N. Brown, General Electric, explained how the Power Reactor Inherently Safe Module (PRISM) was developed to address some of the
, problems in the nuclear industry, particularly those of scheduling, costs, difficulties .of _ handling complex systems in the field, and maintaining quality : assurance. He indicated that the plant is i multi-module and the nuclear safety-related portions are of such a
; size that they could be entirely factory-fabricated and shipped either by land or barge. Size was also coverned by the need to be able to do a productivity safety test in a full size module in an effort to generate information and confider.ce within. the -public regarding the safety built into the plant (see Appendix X). . He discussed a reference site ' layout which would ' consist of nine - ,
identical reactor modules. Three modules would feed a single steam generator and one turbine generator. The plant would be laid out with a single control room from which all nine modules and their. associated steam generator units would supply three turbines. The refueling system would consist of a multiple refueler which would , refuel one module at a time and transport the spent fuel to a ' central building for storage. F. J. Remick spoke of possible confusion in the single control room which is handling and may possibly bring some or all of the nine units to shutdown. N. Brown spoke of an all-digital, all-redundant ~ system control in the control room to be handled by three operators and one supervisor. The plant design calls for automatic startup from criticality and all feed systems run online automatically. It is an entirely automated system. N. Brown defined the high security portion of the plant and pointed out that the remainder is balance of plant including the control room. C. Michelson noted that the steam generators are outside the security area and he assumed that GE was not worried about sabotage on that aspect of' the plant or-sabotage-generated fires. N. Brown indicated that the. sturdiness of the heat removal systems that are safety related appears to be able to withstand the environments that are associated ~ with the kind of damage that would occur as a result of sabotage-generated fires. G. A. Reed noted that GE, as well as Rockwell, is approaching the concept of creating a safety envelope. It would concentrate on about four major safety parameters necessary to the safety of the core and forget about the balance of plant as well as the control room. He defined these four aspects as decay heat removal, containment, the reactor vessel, and some criticality systems. He wondered how the regulators would deal with this sort of approach. N. Brown explained that GE and Rockwell are depending upon demonstration of the inherent safety built into both the SAFR and PRISM designs in its' dealing with the regulators. N. Brown described the PRISM reactor assembly and structures. He described the PRISM main power system which consists of a nuclear steam supply system composed of one reactor module and one steam generator with three such modules tied to one turbine condenser. He defined the safety-grade boundary at the inlet and outlet sodium lines to the reactor and consisting of only the core intermediate heat exchanger and primary pump. J. C. Ebersole noted that the SAFR System was superheated. N. Brown indicated that GE has elected to use a saturated steam. cycle. The result will be a i r
312th ACRS Hinutes '17 July 7,_1986 simplification of controls as well as low flow capability. It will also provide additional margins on the temperatures. D. W. Moeller
. asked if GE has an occupational collective dose goal. N. Brown indicated that it is 20 person-rem per year for the nine module plant site.
D. Okrent took note of the design of the reactor vessel auxiliary cooling system which is to function as a passive heat decay removal system in an open confinement. N. Brown agreed that the containment is actually the guard vessel and in the PRISM system, i GE is ectually cooling the outside of the containment. D. Okrent l noted that it is this containment that is expected to be lost in the case of a large scale core melt. N. Brown agreed. D. Okrent pointed out that GE is proposing a concept whose safety philosophy is quite different from those that have been proposed on light water reactors. While it may be acceptable, it places greater emphasis on demonstrating a very low probability of a major event. Since there is no backup provided to yield another factor of safety of 10 to 100, one has to have great confidence in the PRISM concept. N. Brown agreed. J. C. Ebersole suggested that the sodium fire may become the worst accident for this design and be akin to the large loss of cooling accident for water reactors. N. Brown indicated that a sodium fire would have a different kind of consequence than a large LOCA would for a water reactor. GE can postulate fires from intermediate sodium leaks but the ability of the reactor vessel auxiliary cooling system to survive that fire environment because of the way it was designed leads to a situation where it will still cool under low heat. J. C. Ebersole suggested that it is improbable under the PRISM concept that the primary sodium will ever get intensely burned. N. Brown indicated that this is beyond the design basis threat. J. C. Ebersole asked if there are inert gas flooding systems in the buildings. N. Brown indicated that everything is to be inerted inside those areas where you would nonna11y have exposure to single breaks. In the case of double breaks, another inerting system has not been provided because such breaks have not been found to be a high enough probability event. N. Brown briefly mentioned the diversity in reactor shutdown systems. He indicated that there are two active shutdown systems which have a self-actuated mechanism which make them highly redundant and diverse. The failure to shut the reactor down is a very very low probability event. J. C. Ebersole asked if the diversity meant is two electrical systems driving one mechanical system. N. Brown indicated that there are two mechanical systems with diverse designs. A solid state multi-channel redundant electronics system provides diversity. The diversity is built into the software level and is a little bit different than looking at hardware diversity. N. Brown discussed the PRISM containment approach where the plant is operated as a sealed capsule with the reactor guard vessel being the containment.
, o 312th ACRS Minutes 18 July 7, 1986 N. Brown explained that one of the important thrusts involved with the PRISH concept is the effort to certify the design through the certification of the nuclear safety grade module. That module is small enough and is confined in its base so that an affordable full scale safety test can be performed to overcome uncertainties in the advanced LMR design. D. Okrent wondered whether this safety test would be able to address all specific uncertainties and he urged caution.
N. Brown explained that GE has proposed a change in the regulations to recognize a new optional approach (Safety Test) to design certification of advanced reactors. It will be called a new Appendix S and would be similar to Appendix 0 for certification but is intended only for final design. F. J. Remick asked if the long term aspects of the PRISM design assume reprocessing capability as does the SAFR Design. N. Brown indicated that it is a breeder reactor and GE is considering the use of reprocessing. However, the exact form of that reprocessing, whether it is integral onsite reprocessing or whether it is offsite, has not yet been included in the licensing status discussions. F. J. Remick expressed some skepticism regarding the plan to direct refueling operations from the control room. He suggested that it is a comonplace occurrence to have chaos in the control room during a refueling process. This will be especially important in this case where the turbine generators are controlled from the control room. N. Brown cited his experience with online refueling controlled from the control room and suggested that it can be a very quiet peaceful occurrence under the proper circumstances. V. Subcommittee Activities Related to Consideration of Human Factors in Nuclear Power Plants (0 pen) [ Note: J. O. Schiffgens was the Designated Federal Official for this portion of the meeting.] D. A. Ward explained that the Human Factors Subcommittee met for two days on March 19-20, 1986. The first day was an information session regarding the state-of-the-art in applying automatic monitoring control functions in the operation of nuclear power plants to reduce the burden on plant operators and enhance safety. On the second day, the Subcommittee reviewed the 1985 progress on the Human Factors Program Plan and plans for 1986 and 1987. There was also a briefing on the status of emergency operating procedures implementation. D. A. Ward explained that the early attitude of both the industry and the NRC toward computers was largely influenced by experience with large mainframe units and the associated unreliability of this type of computer in a process setting. In recent years, both the NRC and industry have become aware that computers specifically designed for monitoring and controlling processes can be l considerably more reliable than earlier experience indicated. In - fact, they may be extremely useful tools for aiding operators and I managing both routine and emergency operations in nuclear power ! l
'g 312th ACRS Minutes 19 duly 7,1986 l plants. He mentioned the fact that the Safety Parameter Display System (SPDS), which was mandated as an addition to each plant as a post THI backfit, served as an introduction for the nuclear power industry to . the use of modern process . computers. Some utilities have been successful in the use of -the SPDS while others have had' problems. Despite the introduction offered by the SPDS, the state of . application of computers in nuclear power plants is still not very advanced.- He noted that the NASA Ground Control Systems at the Kennedy Space Center are largely automated to make heavy use of process computers with considerable success, and NASA's design of a future space station calls for the setting aside of 10 percent of 4
the total development budget for software and related automatic control sy' stems. He spoke of the application of computers to the Savannah River Plant reactors which was initiated about 18 years ago and has proceeded through several generations of imprcvements. He discussed several other applications regarding the automation of monitoring and control functions (see Appendix XI). D. A. Ward mentioned reports describing the emergency operating-procedures developed for the Hatch Plant of the Georgia Power
- Company and procedures as applied at the Browns Ferry Plant of the
; Tennessee Valley Authority. He noted that the NRC has. required that each plant develop for itself Emergency Operating Procedures which are " symptom oriented" rather than " event oriented." The NRC provided general guidelines and specified a process for development
- of the procedures. ' The result has been a wide variety and spectrum of styles and types of procedures. These procedures run the gamut i between a flow chart format, a comprehensive and detailed approach which uses large logic diagrams or flow charts to provide the initial response to major plant upsets, to a single colunn instruction format. It is debatable which procedure format is better. The Staff approach has been to have the licensee develop its own Emergency Operating Procedures and then audit and monitor that process with the expectation that if the process is going well, good Emergency Operating Procedures will appear in the plants. The NRC Staff seems to have reached the decision that this approach has failed. They have monitored the process and often found the process satisfactory but when the actual procedures are i audited, major problems are found with the end product. The Staff '
recognizes the need for a more extensive auditing program of the actual final product. J. C. Ebersole explained that a natural evolutionary process is taking place. General guidelines have been distributed to the field and utilities have taken these guidelines
- to develop plant-specific Emergency Operating Procedures. What one must do is sample the field and find an adequate level and some
; norm from which one can define a baseline for standardization for 4
these procedures. D. A. Ward made the connection between the , adverse results regarding implementation of the SPDS and the i problems with the Emergency Operating Procedures. G. A. Reed suggested that the SPDS should have been developed more by
- engineering organizations than utilities and the Emergency Operating Procedures which are plant-specific should be developed with more input from the plant operators and less from the utility engineering groups.
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. 312th ACRS Minutes 20 . July 7, 1986 D. Jones ' discussed the-accomplishments of 1985 and the plans for 1986 regarding the seven program areas in the Human Factors Program Plan -(see Appendix-XII). .In the Staffing and Qualif.ications Program area, he indicated that the Staff has decided that it does < not need to license additional personnel besides operators. He mentioned ' two studies -of the Operational Qualifications or Educational Qualifications for Operational Personnel. lOne study, NUREG-4501, was ' an Assessment of Job Related Qualification for 4 Nuclear . Power Plant. Operations and . the second assessment was of Specialized Educational Programs for Licensed Operators. These - studies pointed out that at least half to two thirds of the knowledge requirements for an SR0 or R0 are plant-specific and must
- be handled by the utility. J. . C. Ebersole pointed out that.
physical characteristics of the plant _ represent an obvious
,. deficiency in the education of reactor operators. D. Jones
- indicated that this part of the program is being accomplished by-INPO and their job accreditation and training program, and also by NUMARC in its emphasis on the training of operators. C. Michelson indicated that one of the questions that comes up from time to time is how well an operator understands accidents beyond the design basis. He asked if the Staff had given some thought to what would constitute a'. training program appropriate to prepare an individual j to face hypothetical beyond design basis accidents. D. Jones indicated that the Staff is considering this matter. F. J. Remick noted that specified training for STAS does include
! beyond-the-design-basis scenarios. He suggested that a briefing on j this kind of training by a few utilities would be useful to the Comittee.
D. Jones discussed limits and conditions of shift work, including an operator feedback mailed survey (NUREG/CR-4139). He indicated that this study surveyed current attitudes toward STAS, shift work, , and staffing levels. The consensus of most of the operators was l that the STA was a waste of time. A deeper analysis showed that operators with a higher level of education tended to appreciate the STAS more than others (see Appendix XII). D. Jones discussed the second program area which included training
- regulation and guidance and the NRC Training Evaluation Program i (training review criteria and procedures for NRC Staff). D.
Ziemann, NRC, indicated that both the industry and the NRC have I been interested in integrating the STA into the shift crew. This
- effort to make the STA a part of the shift crew has resulted in better acceptance of the STA. G. A. Reed expressed concern i regarding the NRC's recent requirements for an STA as far as fomal
! degree requirements are concerned. He suggested that it might be a
- good idea to investigate the possibility of qualifying as STAS, aging shift supervisors who have served for 10 years as SR0s with j- good records. D. Jones made no further coment on this suggestion.
- D. Jones discussed Licensing Examinations from the standpoint of l the process, which is governed by the rulemaking (10 CFR Part 55 and Part 50) and the content of exams. F. J. Remick asked about ~ l the status of Part 55. D. Ziemann indicated that it is expected to
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312th ACRS Minutes 21 -July 7, 1986
, be submitted to the Commission. in April 1986. D. Jones noted that it is basically the same Rule that the Committee has reviewed on
- previcus occasions. The Comrtittee briefly discussed the contents
- -of licensing exams and the new computerized Exam Questions Bank at Idaho National Engineering- Laboratories to determine- the merit of questions, track information on - their performance validity, and
. reliability _of exam questions and make those examination questions
- available to the- public. In the Procedures Program Area - in the
!. Human Factors Program Plan, D. Jones indicated that the 3 Staff looked at Emergency Operating Procedures (EOP), upgrades, and
. post-implementation audits. An E0P inspection module for use by IE i inspectors was developed, and the Staff also looked at methods for evaluating alternative techniques and formats for procedures. ! Regarding the Man-machine Interface Program area, D. Jones-indicated that there are significant variations from .a human engineering point-of-view as to how local control stations are used, o;:erated and laid out. This will be a major item of business
!- in 1986. He indicated that the Staff recently conducted a study on i Computerized Annunciators Systems. G. A.~ Reed agreed that the 4
deficiencies in local control stations is a bothersome problem that , !- is akin to that which has ' occurred with SPDS. D. Jones. indicated . that the Staff has looked at a series of man-machine interface
< concepts for advanced techniques including simulator evaluation of a CRT displays. The Staff also published a Standard Review Plan for Human Factors. A new Regulation was also published on the Safety Indicators Program, an attempt to develop measures of safety ,
performance which could be used in the evaluation of plant , ! performance. Thi was also carried into an analysis of maintenance i factors. G. A. Reed asked if the Staff intends to pull together ! data to determine the possible impact of aptitude testing on the i performance of nuclear plants. D. Jones indicated that the Staff i las not done this but that it is an excellent idea. The Committee discussed tennination of human factors research in the NRC's Office of Research and its transfer to NRR. Brief mention was made of a 4 Mational Academy of Sciences Panel Study being funded by both NRR 1 and the Office of Research. NRR is to focus on audits. Note was i taken of the fact that all work on the Management and ,
- Organizational Program area under the Human Factors Program Plan '
has been suspended. The Commission's " Policy and Planning Guidance,1986" calls for a move toward performance-based programs i which revolve around improvements to SALP and the Development of [ Performance Indicators for use by IE. Mention was made of a new program area called " Human Performance," which was specifically set up to investigate areas of human error and related activities and to develop an improved human performance experience reportage. J. C. Ebersole asked if there is a move toward development of a set of
- guidelines or rules which would include consideration of SPDS which
! would lead one to a decision when to automate the function or when j one should do it manually. D. Jones mentioned the development in 8 l 1985 of Visual Display Evaluation Criteria. He indicated that it ! is this area of artificial intelligence that the Staff is l exploring, but the plan is not yet well developed. b
9 K, 312th'ACRS Minutes 22 . July 7, 1986 V I .- Meeting with the NRC Commissioners-(0 pen)
-[ Note: ~ Commissioners present: N. J. Palladino, Chairman, T. M.
Roberts, J. K. Asselstine, and F. M. Bernthal.]
! 'A.; Anticipated ACRS Activities and Resources in Light of Budget
! Reduction I
-D. A. Ward discussed. the scope and priorities for. ACRS activities (see Appendix XIII). He mentioned a fairly deep budget cut in administrative and clerical services provided to ACRS members. . He mentioned a less deep cut that will have to be made in the technical staff combined with the decrease in allotments for travel and related expenses. This will result '
- in a decrease of subcommittee meetings, which have averaged i about 110 per year in recent years, down to about 85 in FY
- '87. He mentioned the division of ACRS program activities into six mission areas
, . Operating Nuclear Power Plants - 30 percent
. Generic Issues and Unresolved Safety Issues - 28 percent
. Safety Research - 9 percent
. High Level Waste - 12 percent
! . Control of Radiation Exposures - 3 percent j . Future Plant Designs - 6 percent
- He noted that 12 percent has been set aside for-contingencies.
The attention to operating plants represents a significant increase from a year _ or two ago and will include one or two l' NTOL' reviews anticipated in the coming year, as well as ACRS - review of TVA management efforts. Commissioner Bernthal thought that it would be important for ACRS members to visit
- nuclear power plants more frequently if they are to adopt a
!- broader role in plant operations and evaluation of operating procedures. He also thought it of importance to have members on the Committee that have actual operating experience. G. A. Reed pointed out that it might not be practical to have ACRS members sitting through a shift in a control room and might
- actually be disruptive and counterproductive. Routine
- monitoring of nuclear plants ought to be performed by resident inspectors and IE personnel. D. A. Ward noted INPO audits of i operations in a plant and the fact that some ACRS members have attended those evaluations. F. J. Remick mentioned his l experience with various visits which averages about three ,
plant visits per year. He thought these visits were extremely
! valuable. D. A. Ward mentioned that the ACRS plans to spend six percent of its effort in fiscal 1987 on the review of ; future plant designs. He noted that there is no Staff effort j budgeted for these activities which involve both advanced 4 reactor concepts and advanced LWR designs. He expressed i
concern as to whether it was really wise to zero out the l effort in this area. Chairman Palladino suggested that the i cutback was on research in these areas. M. W. Carbon ! suggested that the cutback was also in NRR personnel working l' l
' =
312th ACRS tiinutes 23 July 7, 1986 with DOE on the advanced liquid metal and gas reactor activities. Commissioner Bernthal indicated that he was aware of a letter sent by NRC personnel to DOE indicating that the I Agency would not be able to do the work that DOE requires. He I thought that Congress ought to be aware of this fact that the NRC is not even meeting the minimal internal administrative requirements for reviewing advanced reactors. He thougSt that the situation .is unacceptable and Commissioner Asselstine agreed. Comissioner Asselstine suggested that the ACRS provide advice or coments on this activity end, in particular, an approach or program the agency should have on advanced reactors to help fc;us Comission attention on what resources can usefully be provided in this area. D. A. Ward mentioned areas in which the ACRS does not intend to spend any effort. These included low level waste, the regulatory process, fuel fabrication and mill tailings and significantly fewer resources on unresolved safety issues. Chairman Palladino suggested that the NRC Chairman and the ACRS meet periodically to discuss such changes in allocations of ACRS resources so that the Comission can appreciate the efforts of the ACRS to cope with Agency reductions. B. Scope of ACRS Activities Regarding TVA D. A. Ward described the nature of the proposed ACRS review of the TVA reorganization and restart of affected plants (see Appendix XIV). Chairman Palladino asked if the ACRS is dovetailing its plans for the most efficient interaction with the NRC Staff review efforts. D. A. Ward indicated that the Comittee certainly intends to do that. Comissioner Bernthal wondered whether the ACRS would be able to make a timely contribution to the problems that TVA has had. He cited that important coments by several ACRS members on the Indian Point Special Proceeding were not submitted to the Comission until more than 6 months after the final decision was made. G. A. Reed indicated that he has been very concerned since joining the ACRS regarding the impact of late information or late positions taken by the Comittee. He spoke in favor of a system for tracking " fast moving" issues. Comissioner Bernthal agreed that if the ACRS wants to influence public policy, Comittee positions need to be available when the decisions are made. Information that is not furnished during the decision making process is nice for the record but will
, not marketedly influence Comission policy.
J. C. Ebersole spoke of his 38 years of work experience with TVA explaining that employees worked in highly compartmentalized areas. There was no overview and no integral central control. Comissierer Asselstine thought that the review approach of the ACRS was reasonable and that the Comittee may well be able to make a contribution regarding some of the problems TVA is facing. Comissioner Roberts thought that the ACRS should not be looking into the i
, , 6o 312th ACRS Minuties - 24- July 7, 1986 TVA matter. He noted that the Panel on ACRS Effectiveness recommended that the Comittee concern itself with broad-technical policy questions and not focus on individual utilities. Chairman Palladino thought that it was appropriate for the ACRS to provide' advice on safety issues regarding TVA and, if in the process of looking at safety issues, there is interaction of management or maintenance policies / procedures, D 'the Committee ought to continue its review to the extent that-those issues are important to providing the Comission with advice on safety issues. Comissioner Asselstine agreed. The Comittee discussed the ability or . inability of the ACRS to
- deal with management problems and management's contribution to safety. M. W. Carbon indicated that, in his personal opinion,
- safety in the operation of a nuclear power plant is governed L by the attitude of management. Comissioner Asselstine agreed. Comissioner Roberts thought that the ACRS should t
examine how the NRC regulatory process broke down and allowed i- the TVA situation to occur in the first place. Both Comissioners Bernthal and Asselstine agreed that this was a i good idea. 1 i C. Responses to Recomendations of Panel on ACRS Effectiveness t Chairman Ward referred to the March 20th letter on the Committee's responses to the Effectiveness Panel Recommendations.- He thought that any coments from the
- Comission regarding any of these responses would be very i
useful since it may have some effect on what the ACRS does in , the next few years. Chainnan Palladino asked if the Comittee plans to implement a term of two years for the ACRS Chairman. D. A. Ward indicated that this is his second year but it is only on a trial basis. The Comittee discussed the use of . ACRS members as individual experts as opposed to a collegial l body advising the Comission. C. P. Siess suggested that the individual ACRS members are always available to offer their i expertise to an individual Comissioner at his request. He questioned, however, whether it was appropriate for the ! Comission as a unit to ask an individual member for advice.
- Chairnan Palladino pointed out that the value of the ACRS is
- that it deliberates and exchanges views among a variety of disciplines to come up with a collegial opinion. C. P. Siess pointed out that you may not get the best advice out of a ,
collegial opinion but you will very rarely get bad advice. D. ! , Okrent noted that it is unfortunate that the Comittee has ! difficulty developing consensus opinions on difficult issues. j As a result, on many of the difficult issues the Comittee
- , never develops consensus opinions but puts them off. He cited
! the Indian Point issue which was referred to earlier where the ! Comittee did not provide collegial advice. The Comittee i however was not faced with a request from the Comission to j force a consensus. He also noted the Comittee has been j discussing what safety features would be appropriate for i advanced LWRs and has not made substantive progress. He j suggested that it might be worthwhile to have a Comission
-,% s 312th ACRS Minutes 25 July 7, 1986 recuest -to provide significant additional pressure 'on the
' Committee to reorder its priorities and force Committee coment. Comissioner Bernthal thought that the advanced LWR.
area would be one particular area in which the ACRS could influence the direction of public policy. D. Moynihan Proposal for National Training Academy 2 G. A. Reed explained that Senate Bill 16, the National Academy for Power Plant Personnel Training Act of 1985, was introduced by Senator Moynihan to provide " comprehensive" and " standard" J training for operators of U.S. nuclear plants. The proposed training seems to be addressed to licensed reactor operators, licensed senior reactor operators, supervisors, technicians, engineers and other personnel in nuclear plant-related positions. It proposed centralized and on-site training and g requires that the Academy's personnel must be hired'at nuclear plants. He expressed his opinion that this National Academy approach would be disruptive and noncontributory to a high , standard of nuclear power plant performance. He suggested
- that the efforts of INP0 and the individual utilities spurred on by NRC Regulatory requirements from the Three Mile Island event and lessons learned have created a sound program for
- nuclear power plant training even though some aspects are too j repetitive. Nuclear power plant training of plant personnel may be the strongest aspect of the nuclear regulatory scene today. He noted that the National Academy for Training is unnecessary and mislocated as a centralized facility. It is repetitive of educational institutions already available and
' not in a good position to provide the on-site training which is most useful. The National Academy concept is wrong for the nuclear power nonstandardized plant situation. He concluded i that already available institutional training, INPO guidance and accreditation, and the utility on-site training schools are quite adequate for plant personnel. He recognized a deficiency in aptitute testing during the initial selection process for trainees. He pointed out that much of the problems with reactor operations incidents today involve subtle design-oriented problems and are not primarily people-L problems as previously thought. Comissioners Bernthal and Asselstine were in favor of degree requirements and saw a need i to upgrade the area of engineering theory to deal with beyond-1 design-basis events. They also cited the poor performance in requalification exams administered by the NRC. G. A. Reed , repeated that it is design vulnerabilities that are causing the problems that lead to the need for backup procedures of l all kinds. He suggested that it might be worthwhile to have a National Academy for power plant designers. Comissionc ! Asselstine noted that there is a problem in that the direction ! ! that the Commission seems to be taking is not to pursue major , design changes that would substantially reduce core melt probability for either existing plants or even for future designs, but to put more of the burden on the operators. As a i result, one is left with a need to augment the knowledge and i
i Y .s
?/ } ,
312th ACRS Minutes 26 July 7, 1986 skills of the operators so they can mitigate the consequences of potential accident' scenarios. The Comissioners. discussed the irrelevancy of some questions on the requalification exams for reactor operators. . F. J. Remick suggested that when nuclear plant training becomes , truly perfonnance-based with, learning objectives, and the NRC l examinations use these learning objectives, many of the problems associated with the requalification exams will be solved. He agreed with many of the statements of G. A. Reed but thought there definitely was a need for real professionalism in reactor operation. D. Okrent questioned whether top and middle level management at nuclear plants was cognizant of the questions associated with severe accidents. He suggested that an orientation for utility top and mid-level management would be 'sorthwhile. H. W. Lewis spoke in favor of theory in the educational experience. He suggested that the theoretical education teaches one the ability to deal with problems not seen before, and that this is particularly useful in many instances in the operation of nuclear plants. G. A. Reed suggested that one can overcome the great complexities in the technology through the use of a designated representative system similar to' that used by the FAA. Chairman Palladino
~
agreed that many important points were made by G. A. Reed and other ACRS members but he thought it still important that the ACRS continue as it has done in the past to help address important designLconcepts. He defin tely thought there was a place for the NRC regarding encouragement in that direction. E. ACRS Review of GESSAR II D. A. Ward noted that the Committee had written a letter to the Comission, dated January. 14, 1986 on the topic of GESSAR II which pointed cut significant differences between the ACRS position and that taken by the NRC Staff. He indicated that the Committee welcomes an opportunity to discuss this subject at a future meeting since the Canmittee is concerned that the Staff position is becoming hardened. Comissioner Asselstine indicated that one of the questions that came to his mind from the ACRS letter on GESSAR was with respect to what kind of a system would be a rational system for looking at future designs. He indicated interest in ACRS concents on that subject. Comissioner Bernthal suggested that when the topic-of GESSAR is discussed, the ACRS give some thought to whether there is any real value to the Comission offer of a 5 year approval for the GESSAR design. He thought it not likely to see any of those plants built in the U.S. in the next 5 years.
~
VII.QuantificationofHealthEffectsinPRAs(0 pen) [ Note: R. P. Savio was the Designated Federal Official for this portion of the meeting.]
, e s 312th ACRS Minutes 27 July 7, 1986 D. theA. Ward 311th briefly(discussed meeting March,1986) thewhich poll taken of ACRS resulted members after in the non-release i of the ACRS report entitled "ACRS Comments on Prcbabilistic Risk Assessment Quantification of Public Health Risks." He noted that this was done be::ause D. W. Moeller had concluded that the report may have contained errors and requested additional discussion on this matter before issuance of an ACRS report. D. A. Ward stated that H. W. Lewis had suggested that not issuing the original letter may have been a violation of proper procedures and that this issue would be discussed later in the meeting. Dr. S. Yaniv, NRC, discussed early and late somatic effects from ionizing radiation and the health effects models used in the Reactor Safety Study (WASH 1400). . He indicated that the Reactor Safety Study followed the BIER I (1975) report in estimating health effects. A linear dose response relationship was used. Upper bound risk coefficients for latent cancer fatalities assumed a plateau period of as much as 30 years for cancers affecting the lung, gastro-intestinal tract and bone. He suggested that this model was too optimistic and a lifetime risk is now assumed. S. Yaniv indicated that in 1980 the National Academy of Sciences BIER III report was published, and that it included a review of data on ionizing radiation health effects to 1978. It used three models which assumed either a linear, quadratic, or linear-quadratic dose response both for relative, as well as absolute risk models. The BIER III report separated leukemia and bone cancers from solid tumors. S. Yaniv indicated that additional data for the period 1982 and 1983 have now become available. A critical review was reconnended of the health effects models, and an Advisory Group with several working groups was established to accomplish this task. In the Advisory Group Report, both mortality and morbidity were discussed. Prior studies had not discussed morbidity. It was concluded that " the relative risk model is the best model for assessing the risks from solid tumors. The latest information from the Hiroshima and , Nagasaki bomb victims was not included in the Advisory Group ~ Report, because these data are currently being re-evaluated. S. Yaniv indicated that an ad hoc group of the National Institute of Health is developing new tables for assessing causation relative to cancer mortality and morbidity. Both the Advisory Group and the National Institutes of Health Ad Hoc Group are using the same data and have decided to use the same models with similar dose factors. The absolute linear-quadric model is being used for leukemia and bone, breast, lung and gastro-intestinal cancers. He noted that a , c very extensive review of this work is now in progress which is international in nature (and includes the Federal Republic of \ Germany and Great Britain). Meetings are planned to review and connent on the models now in use. The latest review of the Hiroshima and Nagasaki data has indicated that the dose contribution from neutrons is lower than previously estimated. As a result, the quantitative assessment of the health effects for a
, t .
312th ACRS Hinutes 28 July 7, 1986 given dose from low LET gama radiation r:ay go up by a factor of two or three. S. Yaniv indicated that the National Research Council has called for a revision of the BIER III report (to be called BIER V). The National Academy of Science has solicited funds for this effort but may not obtain support from the NRC. H. W. Lewis noted the slowness of the NRC to use health affects data as up-to-date as BIER III in licensing applications. He pointed out that on the recent Shearon Harris Plant review, only data from the BIER I
, .c report was used.
D. W. Moeller indicated that the letter written at the April 1986 x ACRS meeting implied that the BIER III report contained the best data available and that it did not. It implied that the absolute redel should be used in all cases, and this is not so. He also pointed out that the letter implied that the NRC is moving too slowly, and it appears that the NRC is probably doing the best that it can under the circumstances regarding use of the best data available and stimulating foreign work in this area. H. W. Lewis refused to discuss or debate the April letter because of the procedural flaw in withholding issuance of the letter. He asked when the Staff will start using BIER III in its licensing actions. S. Yaniv did not respond. VIII._QuantitativeSafetyGoals(0 pen) [ Note: R. P. Savio was the Designated Federal Official for this portionofthemeeting.] The Comittee discussed the results of the March 28 Special Meeting with the NRC Comissioners. C. Michelson, W. Kerr and D. Okrent were of the opinion that a Safety Goal Policy without implementation was not much of a policy. Mention was made of the specific wording for qualitative and quantitative safety goals requested by Commissioner Zech. J. C. Mark expressed the opinion that a rigid application of cost-benefit analysis may not work as far as the safety goal is concerned. He thought that the Comission should decide on a goal of 10-4 per reactor year mean core melt frequer.cy or less and order that, in general, all plants s be modified if they do not meet this requirement, irrespective of
, 's the costs. H. W. Lewis thought that the Commissioners were coming to a position at the March 28 meeting that an implementation plan for the safety goals should be a Staff document endorsed by the Comission. He asserted that the Comission Safety Goal Policy ought not to include outright numbers. M. -W. Carbon agreed that the policy statement should be very careful regarding numbers and possibly only infer values. C. P. Siess suggested that there are advantages to a deterministic approach regarding the Safety Goals.
N He thought that Comission-mandated Safety Goals would be most
, effective.
^ "
The Committee discussed the substance of a report to be issued at the 312th meeting. It agreed on the need for a Safety Goal Policy. ? ,
, r c 312th ACRS Minutes 29 July 7, 1986 One of the members made a motion that the ACRS endorse the statement that the risks from nuclear power ought to be "significantly" less than the risks from competing technologies (i.e. , coal). The motion was not carried. The motion was made that public policy should not compare the two technologies (nuclear vs. coal) on the basis of risk alone. This motion carried unaminously. D. Okrent indicated that implementation of the Safety Goals might take three possible forms: a) The Comission could issue a policy and then ask the Staff to develop an implementation plan; b) The Comission could hold back the Policy Statement until an implementation plan is developed; c) The Comission could issue a Policy Statement while an implementation plan is being developed but is still being fomulated. The Committee was infomally polled regarding its views on these alternatives and nearly twice as many members were in favor of the Comission issuing a high level Policy Statement without any reference to an implementation plan than thought the Comission ought to issue a Policy Statement only when - a plausible implementation plan is available. H. W. Lewis thought that the safety goals should be used as a guide for the NRC. Everything the NRC does regarding regulation would be implementation. D. A. Ward suggested that the Policy Statement should state that the majority of the population of currently operating nuclea plants meets the safety goal. He suggested that there be a 10~g per reactor year mean core melt frequency guideline as .well as a containment perfomance guideline as they both relate to the safety goal. If those guidelines are met, he thought there was a good likelihood that the safety goal would be met. C. P. Siess and D. Okrent expressed concern regarding not including a containment performance guideline in the safety goal. They cited the fact that the SAFR and PRISM and the advanced GCR are aiming for a target of a minuscule core melt probability and no containment. These three advanced concepts would not meet the safety goal from the standpoint of health effects unless the defense-in-depth concept were invoked. IX. Reliability of Nuclear Components (0 pen) [ Note: R. K. Major was the Designated Federal Official for this portion of the meeting.] C. Michelson reported on the April 1, 1986 meeting of the Reliability Assurance Subcomittee whose primary subject of discussion was the reliability of motor-operated valves. He explained that the Subcomittee -began its investigation of check valves, as well as air-and pressure-operated valves and intends to look at a full spectrum of such valves focusing on aging effects. He noted that the Subcomittee heard briefings from valve vendors, utilities, EPRI, and the NRC Staff, among others. Staff activity is mostly being conducted in the Office of Research regarding the development of guidance documents for in situ testing of valve
, e . .
312th ACRS Minutes 30 July 7, 1986 operability. NRR has no current commitments in the area of valve reliability except where they involve problems at operating plants. IE activities are minimal although they are now beginning to look at check valves. AE00 is in the process of compiling useful input on valve reliability. EPRI's involvement is concerned with the development of a microprocessor for control and monitoring of valves. The M0 VATS diagnostic system for valve operability was also reviewed. C. Michelson indicated that B. Brooks of EPRI reported on EPRI's research regarding a microprocessor control system that incorporates diagnostic capability. This microprocessor is being developed for EPRI by the Foster Miller Company with a test installation planned at a Duke Power Company nuchar plant. An interim report is expected at the end of 1986. He noted that the Subcomittee members adopted a wait-and-see attitude regarding the worthiness of such close monitoring of valve operability. Some Subcomittee members were skeptical of the merits of the EPRI effort. C. Michelson indicated that research being conducted at the Oak Ridge National Laboratory (ORNL) is attempting to identify aging characteristics and monitor them over time. They are investigating valve operators and have also evaluated the M0 VATS system. ORNL researchers have concluded that the M0 VATS system is not easy to use; it is difficult to install and remove. Also part of ORNL's efforts have been workshops to develop better cooperation with the nuclear industry. C. Michelson indicated that the Subcomittee was reasonably comfortable with the attention to valve operability given by the NRC Staff and thought that industry efforts were adequate. The Subcomittee had no specific recomendation at this time. C. Michelson indicated that there are few data on the issue of check valves and no results yet frem any IE investigations of
- failures at operating plants. He indicated that the Reliability Assurance Subcomittee has been combined with the Equipment I
Qualification Subcomittee and C. J. Wylie appointed as chairman. This new Subcomittee will take over the valve operability effort. C. J. Wylie suggested that the EPRI work is actually overkill. It does provide a useful tool but a better approach to this subject would be more training of the operating people. C. Michelson defended the EPRI work as perhaps the only way to treat the issue of valve overrun. 1 i X. Report of Congressional Hearing on 1986 Ohio Earthquake (0 pen) i [ Note: R. P. Savio was the Designated Federal Official for this i portion of the meeting.] l l D. A. Ward indicated the attendance of himself, C. P. Siess, and l ACRS consultant, P. Pomeroy, at the April 8,1986 hearing of the Subcomittee on Energy ano the Environment of the Comittee on
, c. .
312th ACRS Minutes 31 July 7, 1986 Interior and~ Insular Affairs of the United States House of Representatives chaired by Congressman M. Udall. He indicated that the meeting was conducted with members from several panels giving , statements followed by extensive questioning from Congressmen Dennis Eckert and John F. Seiberling. Chairman Palladino discussed the NRC's evaluation of the effects of the January 31st earthquake on the Perry Plant, noting that a low power license for Perry, Unit i 1, was issued on March 19, 1986 and H. Denton, .the Director of NRC's Office of Nuclear Reactor Regulation, found no reason to step in and withdraw the low power license. D. A. Ward discussed the pertinent information provided as 1 - testimony by the ACRS . (see Appendix XVI). In summary, it was indicated that the .Comittee was satisfied that -the Perry Plant could operate safely. The Committee did recommend a seismic monitoring program near the suspected injection wells . to be conducted over the next couple of years. Reference was made to a statement by - Congressman Seiberling that he is satisfied with nuclear plant safety if there is adequate concern for safety , expressed by the NRC, the utility and others. D. A. Ward mentioned a report regarding microearthquakes not occurring in the vicinity of the suspected injection wells. He indicated that he had obtained a copy of this report and asked P. Pomeroy to study the report. It was indicated that no decision on Perry, itself, was made during-this Subcommittee session, but the ~ Subcomittee indicated that it would report back to the full , Congressional Committee on its findings. XI. Auxiliary Feedwater Systems and Resolution of USIA-45 (0 pen) r [ Note:- P. A. Boehnert was the Designated Federal Official for this portion of the meeting.] , D. A. Ward referred the Committee to his reports on the March 18, 1986 Decay Heat Removal System Subcommittee meeting on the status of Task Action Plan A-45 and a subsequent Subcommittee meeting held on March 26, 1986 regarding Auxiliary Feedwater System Reliability in Older Nuclear Plants (see Appendix XVII). XII. LWR Standard Plant Design (0 pen) [ Note: H. Aldennan was the Designated Federal Official for this portion of the meeting.] C. J. Wylie presented a summary of the March 12, 1986 meeting of the Standard Plant Design Subcommittee. Topics discussed at the meeting were FAA Certification, GE Power Worthiness, industry perceptions, DOE views and an update on the related NRC Policy Statement. He mentioned EPRI's Advanced LWR Program which has a goal of seeking regulatory stability by cooperating with the NRC to resolve outstanding safety issues in advance. Some 700 issues have been identified and plans to follow up on these issues will be the
; subject of another Subcommittee meeting. G. A. Reed suggested that 4
, 4 :s 312th'ACRS Minutes 32 July 7, 1986 4
EPRI's approach is .the same direction without a fresh look. It does not consider anything new or advanced. ! C.: Wylie mentioned a briefing by the Atomic Industrial Forum Study Group on Standard Plants which emphasizes. current options such as FDAs and- PDAs and stresses building on existing experience toward the one. step license concept. A design certification and construction license would be. combined in the one step licensing approach. A plant safety report would be submitted with the , license application. -.The design engineering would be essentially i complete and would describe the complete plant. The NRC would issue a design certification good for 10 years which would be renewable. He mentioned the FAA Office of Air Worthiness briefing which described the design approval process for new aircraft. The FAA delegates this work to - designated vendor-manufacturing representatives. C. Michelson indicated that he saw little similarity between certification of a standard nuclear plant design J
- and the FAA Air Worthiness - Certificate. D. A. Ward pointed out
, that the PRISM concept, an advanced LMR, is proposed as a prototype operating ' plant. C. P. Siess noted that the FAA people work with aircraft designers from the beginning and this is not the case in the nuclear industry. C. J. Wylie speculated on the Subcomittee's followup of the EPRI program. 'R. .Hernan, NRC, indicated that NRC input regarding the EPRI program might be available for a late June 1986 Subcommittee meeting. XIII.FortSt.Vrain(0 pen) [ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.] C. P. Siess reported on a Subcomittee Meeting at the Ft. St. Vrain plant held at Plattville, Colorado on_ April 2,1986 to discuss technical problems addressed during the recent extended outage, to discuss problems associated with regulation of a gas-cooled nuclear power plant, and to discuss management changes made as a result of the Licensee's independent assessment of management controls. C. P. Siess indicated that all presentations made during the Subcomittee. meeting were made by Public Service of Colorado. Mention was made of the continuing problem of moisture intrusion in the helium circulator bearing and the consideration by the utility
- to changing to magnetic bearings.
C. P. Siess mentioned some of the problems that Ft. St. Vrain plant has had recently. These included moisture, corroding of the graphite moderator at high temperatures, and several significant
- events recently which included the failure to scram of six control rod drive mechanisms. He indicated that the licensee has refurbished all 37 control rod drive systems by changing the material to Inconel. This resulted in a tripled occupational exposure, with the collective dose totalling 36 person-rem.
Another problem mentioned was with boron balls for secondary reactivity control. Also mentioned was corrosion in
l c . 312th ACRS Minutes 33 July 7, 1986 l l steel tendons in the prestressed concrete containment which is l under study in a major surveillance program. ! C. P.-Siess mentioned regulatory issues that are being considered.
- 1) Environmental qualification of electrical equipment regarding a main steam line break. The solution to this problem has been a steam line rupture detection system; 2) The question of attempting to apply light water reactor criteria to an HTGR. The Design Basis Accident (DBA) in an HTGR is a stem line break, different from the LWR DBA; 3) A cable spreading room problem regarding fire protection. The solution to this problem is a dedicated heat removal system, (independent of anything else in the plant) to cool the core. Another question was raised by the Staff regarding decay heat removal in other areas of the plant. C. P. Siess indicated that the latter two questions might be the subject of another Subcomittee meeting.
C. P. Siess informed the Comittee that D. Eisenhut of HRR has said that all LWR criteria must be met by Ft. St. Vrain. The utility must apply for exemptions if they wish to avoid certain LWR criteria. C. P. Siess mentioned a performance . enhancement program, a management effort to correct a serious morale problem among employees. Seminars being conducted by an outside human factors organization have had some effect at the plant. D. W. Moeller wondered about talk he had heard about potential decomissioning of Ft. St. Vrain. C. P. Siess indicated that this information is untrue. XIV. Executive Sessions (0 pen) [ Note: R. F. Fraley was the Designated Federal Official for this portion of the meeting.] A. Subcomittee Assignment
- 1. Status of Emergency Operating Procedures During the 312th ACRS meeting, the Comittee heard reports of its Subcomittees on Reactor Operations and Human Factors on two topics of concern to the Committee:
(1) Status of implementation of Safety Parameter Display Systems (SPDS) in operating plants. (2) The status of implementation of new " symptom-based" Emergency Operating Procedures (EOPs) in operating plants. The SPDS problem was assigned to the Subcomittee on Instrumentation and Control Systems. The Subcomittee on Human Factors will follow the issue of " symptom-based" E0Ps. (Note: Following the 312th meeting, it was
. ~. s ,
312th ACRS Minutes 34 July 7,~1986 proposed. that the development and use of operating procedures related to.use and adequacy of the SPDS will be the respcnsibility of the Subcommittee on Operating' Procedures. The responsibility for " criteria for preparation of normal, abnormal, and emergency operating procedures" should also be assigned to the Subcomittee on Operating Procedures rather.than the Subcomittee on Human Factors.)
- 2. Assignment of the Instrumentation and Control Systems
, Subcomittee The Comittee reviewed the proposal by the Duke Power Company to remove the upper head injection system (UHIS) portions of the emergency core cooling systems at the McGuire Nuclear Station, Units .I and 2. During the discussion, C. Michelson expressed concern that the Westinghouse RVLIS level instrumentation might be damaged during blowdown because of the presence of the UHIS in-ternals after the UHIS is disconnected. The matter was assigned to the Subcomittee on -Instrumentation and Control Systems which was asked to report its findings to
- the full Comittee within six months.
- 3. Fort St. Vrain C. P. Siess reported on the subcomittee meeting that was .
held on April 2,1986. He reported on the modifications and studies the Licensee has undertaken to control moisture ingress from the circulator bearing water system. J. C. Ebersole had suggested that the Comittee urge development of a circulator design that would
, correct this problem. No specific Comittee action was taken at this time. The Subcomittee plans to meet again in the near future to further consider the problems at Fort St. Vrain.
B. Reports, Letters, and Memoranda
- 1. ACRS Comments on Ouantitative Safety Goals The Comittee completed its report to the Comissioners of its continuing review of the proposed NRC safety goal
' policy. Additional coments by H. W. Lewis were appended.
- 2. ACRS Comments on Quantification of Public Health Risks i
l' The Comittee prepared a report to the Comissioners regarding the need for guidance in environmental and probabilistic risk assessments for characterizing the l public health risks to nearby population groups due to routine and potential accidental releases of radionuclides from nuclear power plants. Additional
-q=.
312th ACRS 111nutes 35 July 7, 1986 coments by H. W. Lewis, and by D. W. Moeller and D. Okrent were appended.
- 3. Proposal by Duke Power Company to Operate the McGuire Plant Without the Upper Head Injection System U The-Comittee prepared a letter to the ED0 of its review of the proposal by the Duke Power Company to operate the
, McGuire Plant, Units 1 and 2, without the Upper Head Injection System portions of the Emergency Core Cooling-System. C. J. Wylie did not participate in .the deliberations regarding this report because of an apparent conflict of interest.
- 4. NRC Review of Advanced Reactor Designs The Comittee prepared a report to the Comissioners of its recomendations regarding continued NRC review of one advanced gas-cooled reactor (GCR) design and two advanced liquid-metal reactor (LMR) designs which are being de-veloped by D0E.
- 5. Move to Bethesda H. W. Lewis introduced a draft report regarding the consolidation of the NRC at White Flint and the interim move of the ACRS to Bethesda, another Maryland suburb.
. The Comittee decided against sending another letter to the Chairman (previous correspondence sent -March 18, 1986) and agreed to fom a delegation (3 members) to discuss this matter in person with the NRC Chairman and Chairman pro-tem within two weeks.
- 6. Proposed Change in ACRS Bylaws Time did not permit a sumary report of the ACRS Management Group meeting held on April 9, 1986 or a discussion of the proposed changes to the ACRS Bylaws regarding Section IX -- Conduct of Members, the ACRS Bylaws, providing for meetings of individual members with individual Comissioners. This matter will be. discussed during the 313th ACRS Meeting (May 8-10,1006) along with a proposed revision to Section V of the ACRS Bylaws regarding the handling of ACRS reports by the ACRS Chairman after a meeting has ended.
C. Future Agenda
- 1. Future Agenda The Comittee agreed on tentative agenda items for the 313th ACRS meeting, May 8-10, 1986 (see Appendix II).
312th ACRS Minutes 36 July 7, 1986-
- 2. Future Subcomittee Meetings
'A schedule of future subcomittee activities was distributed to members (see Appendix III).
D. Nuclear Reactor Safety Infomation Exchange Meeting H. W. Lewis reported to the Committee on his attempt to invite representatives of the National Academy in the U.S.S.R. as ob-servers to the foreign technical meeting at the Wingspread l site near Racine, Wisconsin. Concerns by Dr. Birkhofer of the Federal Republic of Germany have been satisfied. H. W. Lewis
- noted that the French have not yet responded to the notion to inviting Soviet observers and the Japanese have not yet-indicated whether they plan to attend the meeting.
E. Appointment of New Members The Comittee discussed background infomation on an additional list (to supplement those applicants discussed during the 311th ACRS meeting) of applicants to fill the vacancy on the ACRS that arose when R. C. Axtmann left the Committee in 1985. Additional background data were requested on two prospective applicants 'before the Committee will be able to move closer to closure on this matter. G. A. Reed
- expressed concern during the discussion regarding the fact that candidates should be infomed of the forbidden list of securities and the divestiture provision in the rules of the NRC. Further discussion on this matter is scheduled for the 313th ACRS meeting (May 8-10,1986).
F. PRA Quantification of Public Health Risk A Comittee report regarding- the need for guidance in characterizing dose categories from calculations performed 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 was withheld for further discussion after the adjournment of the 311th ACRS meeting (March 13-15, 1986). Members expressed concern over potential inaccuracies or errors in the report. H. W. Lewis, l the member who introduced the draft report, objected to the procedure used to defer transmittal of this report to Chaiman Palladino and denied that there were any errors or inaccuracies in the report. The ACRS Executive Director, R. F. Fraley, agreed to review the ACRS Bylaws regarding this
- matter and introduce such Bylaw changes as are considered appropriate.
L I
, .. : g 3122h ACRS Minutes 37 July 7, 1986 G. Safety Parameter Display System and Emeroency Operating Procedures A proposed -letter on this subject was presented by J. C.
Ebersole but time did not permit its discussion which has been deferred to the 313th ACRS meeting. The 312th ACRS Meeting was adjourned at 3:05 p.m., Saturday, April 11, 1986. l
i 1 P l i f APPENDIXES TO 312TH ACRS MEETING APRIL 10-12,1986 66fs-ay/g i
---~----*pn ,-ws.m------,w w- n, ,, - ----,_, w w,e m --o---,- - - , - , - , --
APPENDIX I NRC ATTENDEES AT 312TH ACRS MEETING
/N NRC ATT
312TH ACi Thursday, April 10, 1986-4 0FFICE 0F NUCLEAR REACTOR REGULATION C. E. Rossi- 'G. Lapinsky J. J. Watt
! D. Hood
- j. J. Wilson
, R. Hernan N. Lauber M. B. Spangler
- j. T. L. King J. E. Rosenthat D. E. Sells D. G. Mcdonald R. Eckenroda S. Weiss W. G. Kennedy 0FFICE OF INSPECTION.AND ENFORCEMENT H._ Bailey E. Weiss D. P. Allison J. Petrosing .
i l < i 0FFICE OF NUCLEAR REr,ULATORY RESEARCH i l J. Martin t P. Lonneman, NRC f. i 1 I er
. s 2 t i INVITED ATTENDEES' 312TH ACRS MEETING i. i Thursday, April 10, 1986 DUKE POWER COMPANY , R. L. Gill i P. M..Sbraham i W. D. Reckley WESTINGHOUSE ELECTRIC CORPORATI,0N M. Y. Young i J. Kabadi
- M. Beaumont l B. McIntyre GENERAL ELECTRIC COMPANY N. W. Brown
! F. E. Tippet
~
i . { ROCKWELL ! R. T. Lanlet . J. Brunings ' t DEPARTMENT OF ENERGY F. Gavigan A. C. Millunzi ^ i l i l l f,b
PUBLIC ATTENDEES i 312TH ACRS. MEETING , Thursday, April 10, 1986 L. .D. Buxton- Sandia National Labs ! J. D. Trotter, NUS Corpo. ! J. hvemi, QATEL D. Airozo, McGraw-Hill < M.C. Bryan, TVA R. Misiaszek, Stone & Webster i' J. Trotter, NUS C. J. Pindnia, Serch Licensing Bechtel
- D. S. flack, ORAU
! R. -J. Stevens, Florida Power & Light
, P. F. Riehm, KMC i H. M. Fontecilla, Virginia Power l'
I i a t i + l i f p3
.-.. . . . . -. ..- . .-..- ...- - . - - . . --.. . - - . . _._..~.- -. . --.- ,. I t
NRC ATTENDEES 312TH ACRS MEETING
- Friday, April 11, 1986 i ;- 0FFICE OF NUCLEAR REACTOR REGULATION R. Hernan j D. B. Jones
- - M. B. Spangler
}- D. Scaletto l 0FFICE OF NUCLEAR REGULATORY RESEARCH i 4 p: J. P..Jenkins' ,
- :0FFICE OF EXECUTIVE DIRECTOR FOR OPERATIONS
- M. Taylor F
1 i l b l
**gy-w'yy
1 l s PUBLIC ATTENDEES l-312TH ACRS MTG. i e 3
- Friday, April 11, 1986 J. Herring, Bechtel
! M. Remley, Rockwell' International J. Nurmi, QATEL , K. Campbell, NUS M. Beaumont, Westinghouse i L. Toth, Gasser Associates R. B. Leachman, Lockheed Missiles & Space Company . P. Muenzen, Shaw, Pittman H..Fontecilla, Virginia Power ! P. F. Riehm, KMC i D.'Goss, Public Service of Colorado f i j j i n l I l I 4 I l l ?
APPE7fDIT II: ' FUTURE AGENDA I' ') AF
/ FUTL MAY ACRS MEETING Meeting with NRC Commissioners to discuss the proposed 2 hrs NRC quantitative safety goals and ACRS review of GESSAR II Meeting with representatives of the FRG regarding 1-3/4 hr radwaste handling and disposal to discuss:
. How were their radionuclide release limits (or dose limits) established for LLW sites; for HLW sites?
. How do these standards compare to each other and to those for other nuclear facilities (nuclear power plants and chemical processing plants)?
. How do they conduct a performance assessment to assure compliance with these standards? Do they use models?
If so, how are they developed?
. Do i. hey require environmental monitoring for a waste disposal facility: (a) during operation; (b) post closure? If so, what does such monitoring entail?
(o NRC Safety Research Program -- Discuss scope, format, i hr and schedule for providing input for the ACRS report to the Commissioners on the proposed NRC Safety Research l l Program and Budget for FY 1988-89 1 USI A-17, Systems Interactions -- ACRS comments requested 3 hrs on a proposed generic letter i Decay Heat Removal -- Briefing by G. A. Reed of a bleed and 1 hr feed system proposed for the advanced Westinghouse pressurized water reactor Report of ACRS Subcommittee on Waste Management regarding: li-2 hrs (1) Salvaging of contaminated smelted alloys (2) Modeling strategy for HLW performance assessment (3) Quality assurance for HLW geologic repositories (4) Research efforts for both HLW and LLW including:
. setting priorities for HLW research, results of modeling workshop, interna-p tional programs and cooperative agreements, and natural analogs (v)
(5) The NRC LLW program as required by the recently passed Low Level Radioactive Waste Policy Amendments Act of 1985 [h
9 ACRS comments / recommendations have been A) ( V requested / proposed regarding Items 1, 2, 3, and 5 noted above Safety Parameter Display System and Emergency Operating 3/4 hr Procedures -- Discuss proposed ACRS report New ACRS Member -- Discuss nomination of panel of can- 3/4 hr didates for consideration by the Commission Report of Subcommittee on Thermal Hydraulic Phenomena i hr regarding the NRC's proposal to revise 10 CFR 50.46 and Appendix K NRC Long-Range Plan -- Discuss proposed ACRS guide for the defer preparation of a long-range agency plan San Onofre Nuclear Generating Station, Unit 1 -- Review defer plant restart following an incident in which AC power to July / was lost and resulting water hammer damaged the main August feedwater system Millstone Nuclear Power Station, Unit 3 -- Review of defer PRA per ACRS report on OL to June South Texas Project, Units 1 and 2 -- OL Review defer Davis-Besse Nuclear Power Station, Unit 1 -- Evaluation of defer review and approval process regarding the design of the to June auxiliary feedwater system and the operation / inspection of the Davis-Besse Plant TVA Reorganization -- ACRS briefing regarding proposed defer management changes in suppor,t of TVA nuclear plants to June B&W Water Reactors -- Review of long-term safety of defer B&W nuclear power plants to June Future ACRS Activities i hr Additional Subcommittee Reports Subcommittee on Severe (Class 9) Accidents regarding i hr review of rebaselining studies for four or five reference ' plants as part of NUREG-1150 Subcommittee on Scram Systems Reliability regarding i hr implementation of the ATWS Rule Thermal Hydraulic Subcommittee report regarding proposed i hr changes to 10 CFR 50.46 and Appendix K, ECCS Evaluation Models ACRS Management Group - Report of April 9 and May 7 meetings I hr k _. . __ . .
., e' ' APPENDIX TII 7 "
ACRS SUBCOMMITTEE MEETINGS ACRS SUBCOMMIl ' Waste Management, April 24 and 25, 1986, 1717 H Street, NW, Washington, DC (Merrill), 8:30 A.M., Room 1046. The Subcommittee will review various topics 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 natural analogs, 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. Shewmon MILLERS Dr. Carbon CARLYLE Dr. Carter ANTHONY Dr. Kerr LOMBARDY Dr. Foster ANTHONY Dr. Mark LOMBARDY Dr. Steindler NONE Dr. Remick NONE Thermal Hydraulic Phenomena, April 29 and 30,1986,1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M., Room 1046. The Subconmittee will: (1) continue its review of the NRC's proposal to revise 10 CFR 50.46 and Appendix K, (2) continue discussions on defining the thermal hydraulic C safety issues of most importance that need to be addressed in the future, and (3) discuss the NRC Staff's review of the W BASH ECCS Code. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the nights of April 28 and 29: Mr. Michelson DAYS INN Dr. Catton DUPONT PLAZA Mr. Ebersole CARLYLE Mr. Schrock ANTHONY Mr. Ward NONE Dr. Sullivan NONE Dr. Tien ANTHONY Severe (Class 9) Accidents, May 1, 1986, AMFAC Hotel, 2910 Yale Blvd., SE, Albuquerque, NM, (Houston), 8:30 A.M. The Subcommittee will review rebaselining studies for four or five reference plants; part of NUREG-1150 study. On May 2 a tour of the Sandia containment model has been arranged for Subcomittee members and consultants (A.M. only). Attendance by the following is anticipated, and ! reservations have been made at the AMFAC Hotel (505/843-7000) for the nights of April 30 and May 1: Dr. Kerr Dr. Catton Dr. Carbon (5/1 only) Dr. Corradini Dr. Mark (5/1 only) Mr. Davis Dr. Okrent (5/1 only) Dr. Plesset Dr. Siess O
-. R
..- te.
REVISED Scram Systems Reliability, May 6,.1986, 1717 H Street, NW, Washington, DC . (Boehnert), 8:30 A.M., Room 1046. The Subcommittee will continue its review of the ATWS Rule implementation effort. Attendance by the following is anticipated, , and reservations have been made at the hotels indicated for the night of L May 5: Dr. Kerr LOMBARDY Mr. Ward NONE Mr. Ebersole CARLYLE Mr. Wylie DAYS INN Dr. Lewis HYATT Mr. Davis HOLIDAY INN Safety Research Program, May 7, 1986, 1717 H Street, NW, Washington, DC (Duraiswamy), 8:30 A.M. - 12:00 N00N, 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 Commission on the related matter. Attendance by the following i is anticipated, and reservations have been made at the hotels indicated for the night of May 6: Dr. Siess ANTPONY Dr. Moeller CARLYLE Dr. Carbon CARLYLE Dr. Okrent ANTHONY 4 Dr. Kerr LOMBARDY Mr. Ward NONE I Dr. Mark LOMBARDY Mr. Wylie DAYS INN , Mr. Michelson DAYS INN Safety Philosophy, Technology, and Criteria, May 7, 1986, 1717 H Street, NW,
- Washington, DC (Savio),1
- 00 P.M., Room to be detemined. The Subcommittee will review the NRC Staff's proposed resolution of USI A-17, " Systems Interactions in Nuclear Power Plants." Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of May 6:
9
. Dr. Okrent ANTHONY Mr. Michelson DAYS INN Mr. Ebersole CARLYLE Mr. Ward (p/t) NONE Dr. Kerr LOMBARDY Mr. Wylie DAYS INN 313th ACRS Meeting, May 8-10, 1986, Washington, DC, Room 1046.
Babcock and Wilcox (B&W) Reactor Plants, May 19, 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-term safety of B&W reactors, includ- , ing the implications of operating experience on the adequacy of B&W plant de- ! signs. The Subcomittee will also be briefed on the NRC Staff's Incident Inves-tigation 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. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of May 18: Mr. Wylie DAYS INN Mr. Reed DAYS INN O Mr. Ebersole Mr. Michelson CARLYLE DAYS INN Mr. Ward NONE s d* E - _ _ __ _ _ _ _ .- . #1)_ _ _ - _ _ _ _ _ _
- ___ m _
APR 12 m [m} U REVISED i Ad Hoc Subcommittee on TVA, May 20, 1986, 1717 H Street, NW, Washington, DC (Savio), 8:30 A.M., Room 1046. The Subcommittee will discuss TVA reorganization 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 May 19: Mr. Wylie DAYS INN Mr. Remick NONE Mr. Ebersole CARLYLE Mr. Ward NONE Mr. Michelson DAYS INN Dr. Hagedorn NONE Mr. Reed DAYS INN Regulatory Policies and Practices, May 27, 1986, 1717 H Street, NW, Washington, DC The Subconrnittee will discuss the IIT Process after the Tssu(Quittschreiber) ance of the San Onofre . and Davis-Besse IIT reports and the report of the Ad Hoc Independent Review Group on the Davis-Besse Incident. Attendance by the
, following is anticipated, and reservations have been made at the hotels indicated for the night of May 26:
Dr. Lewis HYATT Mr. Wylie DAYS INN Mr. Michelson DAYS INN Dr. Siess ANTHONY Dr. Remick NONE D South Texas Units 1 and 2, May 28 and 29, 1986, Bay City, TX (El-Zeftawy). The Subcommittee will review Houston Lighting and Power Company's application for an operating license. Lodging will be announced later. Attendance by the following is anticipated: Dr. Mark Mr. Michelson Mr. Ebersole Dr. Siess Reactor Operations, June 2, 1986, 1717 HStreet,NW, Washington,DC(Alderman),
. 8:30 A.M., Room 1046. The Subcomittee will review recent events at operating plants. Lodging will be announced later. Attendance by the following is anticipated:
Mr. Ebersole Mr. Reed Mr. Michelson Dr. Remick Dr. Moeller Mr. Wylie Severe (Class 9) Accidents, June 3, 1986, 1717 H Street, NW, Washington, DC (Houston), 1:00 P.M. - 5:00 P.M., Room 1046. The Subcommittee will review a final draft of NUREG-0956, " Reassessment of the Technical Bases for Estimating Source Terms." Attendanc: by the following is anticipated, and reservations have been made at the hotels indicated for the night of June 2: Dr. Kerr LOMBARDY Mr. Bender NONE Dr. Okrent ANTHONY Dr. Catton NONE Dr. Shewmon MILLERS Dr. Corradini NONE Dr. Siess ANTHONY Mr. Davis NONE Mr. Ward NONE
. . hl
APR 12 m g l O N l Safety Research Program, June 4, 1986, 1717 H Street, NW, Washington, DC (Duraiswamy), 8:30 A.M., Room 1046. The Sub:omittee 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. Attendance by the following is anticipated, and reservations have been made at the hotels indicated for the night of June 3: Dr. Siess ANTHONY Dr. Moeller (p/t) CARLYLE Dr. Carbon (p/t) CARLYLE Dr. Okrent ANTHONY Dr. Kerr LOMBARDY Dr. Remick (p/t) NONE Dr. Mark LOMBARDY Dr. Shewmon MILLERS Mr. Michelson DAYS INN Mr. Ward NONE Mr. Wylie (p/t). DAYS INN Long Range Plan for the NRC, June 4, 1986, 1717 H Street, NW, Washington, DC 1 (Major), 8:30 A.M. (A.M. Only), Room 1167. The Subcomittee will continue discussions related to long range planning for the NRC. Portions of this meeting may be closed to discuss internal ongoing draft documents. Attendance by the i following is anticipated, and reservations have been made at the hotels indicated for the night of June 3: j Dr. Carbon LOMBARDY~ Dr.Remick(p/t) NONE Dr. Lewis HYATT Mr. Wylie (p/t) DAYS INN Dr. Moeller (p/t) CARLYLE
- l 314th ACRS Meeting, June 5-7, 1986, Washington, DC, Room 1046.
Westinghouse Reactor Plants, June 26, 1986 (tent.'), 1717 H Street, NW, Washington, DC (Houston), 8:30 A.M. - 12:00 P.M.), Room 1046. The Subcommittee will continue discussion and comment on NRC Staff actions taken with respect to . the SONGS-1 waterhamer/ loss of AC power event. This will be a follow-up Subcommittee meeting to the February 12, 1986 meeting on the same subject. Lodging will be announced later. Attendance by the following is anticipated: Mr. Reed Mr. Michelson Mr. Ebersole Dr. Shewmon Mr. Etherington Mr. Ward Dr. Kerr Mr. Wylie Dr. Catton ' i !O l 71 U
APR is as
-m REVISED
( V) Auxiliary Systems, June 26, 1986, 1717 H Street, NW, Washington, DC (Duraiswamy), 1:00 P.M. - 5:03 P.M., Room 1046. The Subcommittee will discuss: (1) the status of the Appendix R compliance, (2) differing technical views among the Staff, (3) proposed resea ch and associated budget for FY 1988 and 1989 in the fire protection area, (4) updates on the progress being made in the Sandia experimental program on fire protection. Lodging will be announced later. Attendance by the following is anticipated: Mr. Michelson Dr. Shewmon Mr. Ebersole Mr. Wylie Mr. Reed Davis-Desse, June 27, 1986, 1717 H Street, NW, Washington, DC (Alderman), 8:30 A.M., Room 1046. The Subconmittee will review start-up activities for Davis-Besse. Lodging will be announced later. Attendance by the following is anticipated: Dr. Remick Mr. Reed Dr. Carbon Dr. Siess Joint Occupational & Environmental Protection Systems and Auxiliary Systems,
/7 June 27, 1986, 1717 H Street, NW, Washington, DC (Schiffgens/Duraiswamy),
8:30 A.M., Room 1167. The Subcommittee will: (1) review a draft AEOD report on the effects of ambient temperature on I&C Systems, (2) be briefed on the status of various control room HVAC Systems problems and the Staff's control room habitability improvement effort, (3) discuss with the Staff the 1 mrem /yr
" cutoff" dose rate for the calculation of collective population doses, and (4) be briefed on the Staff's evaluation of the Shearon Harris Chilled Water Systems.
Lodging will be announced later. Attendance by the following is anticipated: Dr. Moeller Dr. Shewmon Mr. Michelson Mr. Wylie Mr. Ebersole Dr. First Dr. Mark Mr. Healy Mr. Reed Mr. Kathren Reliability Assurance, July 8, 1986, 1717 HStreet,NW,Washignton,DC(Major), j 8:30 A.M., Room 1046. The Subcommittee will review the final resolution of USI A-46, " Seismic Qualification of Equipment in Operating Plants." Lodging will be announcqd later. Attendance by the following is anticipated: Mr. Wylie Mr. Michelson Mr. Ebersole Dr. Siess 1 J l I g.- { A-
9 [b REVISED I2 % Extreme External Phenomena, August 5 and 6,1986 (tentative),1717 H Street, NW, Washington, DC (Savio), 8:30 A.M., Room 1046. The Subcomittee will conduct a workshop on what is currently known regarding the importance of seismic risk to nuclear power plants (emphasis on seismic hazard). Lodging will be announced later. Attendance by the following is anticipated: 1 Dr. Okrent Dr. Lewis (all other ACRS Members, as available) Wingspread International Conference, October 19-23, 1986, Racine, WI (McCreless). Representatives from the ACRS, RSK, GPR, and Japan will exchange information on nuclear reactor safety. Decay Heat Removal Systems, Date to be determined (late May), Washincton, DC (Boehnert). The Subcommittee will review NRR's Action Plan to adc ress concerns with the reliability of certain plants' AFW systems. Attendance by the following is anticipated: Mr. Ward Mr. Reed s Mr. Ebersole Dr. Catton Mr. Michelson Mr. Davis S)ent Fuel Storage, Date to be determined (May/ June), Washington, DC (Alderman). T1e Subcommittee will continue its review of 10 CFR Part 72 and Monitored Retrievable Storage (MRS). Attendance by the following is anticipated: Dr. Siess Dr. Remick Dr. Kerr Dr. Shewmon Dr. Moeller Metal Components, Date to be determined (June, tentative), Pittsburgh, PA or Charlotte, NC (Igne). The Subcommittee will review the status of. NDE 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 Gas Cooled Reactor Plants, Date to be determined (June),1717 H Street, NW, Washington, DC (McKinley) (i day). The Subcommittee will review the appli-cability of NRC requirements for equipment qualification and cable testing to Fort St. Vrain, an HTGR. Attendance by the following is anticipated: O Dr. Siess Mr. Ebersole Mr. Michelson Mr. Reed Mr. Wylie h . _ - - - -
REVISED UM Thennal Hydraulic Phenomena, Date to be determined (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. Michelson Dr. Catton Mr. Ebersole Mr. Schrock Mr. Etherington Dr. Sullivan Mr. Reed Dr. Tien Mr. Ward 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. Dillun Dr. Lewis Mr. Kassner Mr. Michelson Mr. Rodabaugh Dr. Okrent Mr. Thompson Mr. Ward Structural Engineering, Date to be deter::.ined (June / July), Albuquerque, NM (Igne). The Subcommittee will visit and review containment integrity and Category I structures, facilities, and programs. Attendance by the following is anticipated: Dr. Siess Dr. Shewmon Mr. Ebersole Mr. Bender l Dr. Kerr Dr. Pickel l Dr. Okrent t l Decay Heat Removal Systems, Date to be determined (July), Washington, DC I (Boehnert). The Subcommittee will begin its review of NRR's proposed resolution l position for USI A-45, " Shutdown Decay Heat Removal System." Attendance by the following is anticipated: Mr. Ward Mr. Reed Mr. Ebersole Dr. Catton Mr. Michelson Mr. Davis Seabrook Units I and 2, Date to be determined (late summer /carly fall), Washington, DC (Major). The Subcommittee will review the application for a full power operating license for Seabrook 1 and 2. Attendance by the following is anticipated: Dr. Kerr Dr. Moeller Dr. Lewis Mr. Michelson
- f. _ _
REVISED II % p Reliability and Probabilistic Assessment, Date and location to be detennined (Savio). The Subcommittee 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 Westinghouse Reactor Plants, Date to be detarmined, Washington, DC (El-Zeftawy). The Subcommittee will begin the PDA review of the Westinghouse Advanced Pressurized Water Reactor (RESAR SP/90). Attendance by the following is anticipated: Mr. Ebersole Mr. Ward Mr. Etherington Mr. Wylie Mr. Michelson Mr. Davis Dr. Shewmon O O
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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, Fo' ster, Steindler PURPOSE: To review several High-Level ar.d Low-Level Radioactive Waste topics:
~
(1) Modeling. Strategy for HLW oerfor1rance assessments; (2) Quality Assurance,' addressing safety issues of geologic repositories; (3) the NRC LLW - program; (4) several research ef forts, including international programs ~and - cooperative agreements, results of modeling workshop, and setting priorities for HLW research, and natural analogs; and (5) the Salvaging of Contaminated Smelted Alloys. LOCATION: WASHINGTON, DC BACKGROUND: _ What action is requested; by what date is it needef NMSS Division of Waste Managerrent and RES Waste Management Branch have requested ACRS oversight. review on the above Nmeds top'es. What will be done at this meeting? See Purpose. :
\
What would be the consequence of pestf onitig this meetina? None, except for Item (5) above, which C$missioner Bernthal requested the ACRS to review (see. Item 3 below). , 4 ( PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: _
- 1. NMSS Background Paper on Quality Assurance Program, 2/86; NRC Review Plan: Q/A Programs, June 1984; and.four papert presented by NRC/NMSS Staff at Am. Scty, for Quality Control Topical; Conference on Nuclear Water Management Q/A, 1/20/86. l
- 2. Revised Modeling Strategy document for HLW Perfortance Assessment, July 1985. )
- 3. Memo for 0. Merrill from M. Knapp re: Briefing on LLW to ACRS WM Subc.
(w/ Attachments),(undated),.
- 4. International Waste Management Research, briefit;g materials (undated).
- 5. Natural Analog, briefing starials (undated).s l
- 6. Staff Requirements, memo, S. Chilk 'o V. Stello'3, dtd.1/30/86,
Subject:
SECY-85-373, " Denial of DOE Request for Exemption to Permit Salvaging Contaminated Smelted Alloys," dtd 11/25/85. .
- 7. Memo for Comissioners fm J. Zerbe, OPE,
Subject:
OPE Coments on SECY-85-373.
- 8. NUREG-0518, Draft Environmental Statement, dtd. October 1980.
- 9. Memo for ACRS from R. Minogue requesting ACR1 review of Smelted Alloys topic, March 17, 1986.
- 10. Status Report memo from O. P$rril'1 to 0.' Hoeller for April 24 & 25 ACRS WM Subcommittee meeting, dtd.,4/11/86. ,
O a-
s '3 SCHEDULE OF ACRS SUBCOMMITTEE MEETING l
\
b DATE- SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS APRIL 29 & 30, 1986 THERMAL HYDRAULIC. PHENOMENA (B0EHNERT)Michelson, Ebersole, Ward Cons.: Catton, Schrock, Sullivan, Tien PURPOSE: To: (1) continue the review of the NRC's proposal to revise 10 CFR 50.46 and Appendix K, (2) continue discussions on defining the thermal hydraulic safety issues of most importance that need to be addressed in the future, and (3) discuss the NRC Staff's review of the W BASH ECCS Code. 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 CRGk ~1n 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. .
%s
a 3 SCHEDULE OF ACRS SUBCOMMITTEE MEETING
- J v
.i.
DATE- SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS
- n i MAY 1 & 2, 1986 SEVERE (CLASS 9) ACCIDENTS (HOUSTON) Kerr, (SEVERE ACCIDENT PHENOMENA) Carbon (5/1 only),
,y i s Mark Okrent(5/1 (5/1only),),
only
'~ Siess
\ ;
, s Cons.: Catton, Corradini, Davis i- Plesset
}
PURPOSE: To review risk rebaselining studies for five reference plants; part of NUREG-1150 study; also Tour containment model for severe accident performance studies. LOCATION: ALBUQUERQUE, NM MAY 1 - AMFAC Hotel MAY 2 - Bus tour at Sandia (A.M. only) F % BACKGROUND: Q What action is requested; by what date is it needed? x Review baseline risk calcualtions for five refence plants (Surry, Peach Bottom, Sequoyah, Zion, and Grand Gulf); tour containment model. What will be done at this meeting? Determine the adequacy of the risk calculations. What would be the consequence of postponing this meeting? I I Could delay the issuance of NUREG-1150 if Subcommittee coments require extensive revision. PERTINENTPUBLICATIONSANDTHEIRAVAILABILITY: , 3 l RES will provide some preliminary documents. t O
~. . . - . _ _ - .,. - _. , _ _ . . - . _ ._. _
O SCHEDULE OF ACRS SUBCOMMITTEE MEETING O DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MAY 6, 1986 SCRAM SYSTEMS (BOEHNERT) Kerr, RELIABILITY Ebersole, Lewis, Ward, Wylie Cons.: Davis 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 O What will be done at this meeting? See Purpose What woulo be the consequence of postponing this meeting? No significant consequences. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: To'be provided on a timely basis. O
.-,4_ /?
/O SCHEDULE OF ACRS SUBCOMMITTEE MEETING V
DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MAY 7, 1986 SAFETY RESEARCH PROGRAM (DURAISWAMY) Siess, (8:30 A.M. - 12:00 P.M.) Carbon, Kerr, Mark, Michelson, Moeller, Okrent, Ward, Wylie PURPOSE: To 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 Commission 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 Connission, 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 the early part of May).
O > w
C} SCHEDULE OF ACRS SUBCOMMITTEE MEETING i.J DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBER _S MAY 7, 1986- SAFETY PHILOSOPHY, TECHNOLOGY, (SAVIO) Okrent (1:00P.M.) ' AND CRITERIA Ebersole, Kerr, Michelson, Ward (p/t) Wylie
-PURPOSE: The Subcomittee will review the NRC Staff's proposed resolution of USI A-17,
" Systems Interaction in Nuclear Power Plants."
LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? A review of the NRC Staff's proposed resolution of USI A-17 by the ACRS at the ilune 5-7, 1986 ACRS meeting What will be done at this meeting? Develop recomendations fer full Comittee Actions. What would be the consequence of postponing this meeting? Delay of NRC issuance of proposed resolution for public coment. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. CRGR package on USI A-17 and NRR Offices' coments.
- 2. CRGR coments, if available.
O - _ _ 22
1 h SCHEDULE OF ACRS SUBCOMMITTEE MEETING 4 s , DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MAY 19, 1986 B&W WATER REACTORS (MAJOR) Wylie, Ebersole, Michelson, Reed, Ward PURPOSE: Consideration of the B&W Owners Group plans to reassess the long-tem 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 l safety of their plants. Comittee coments on the ' course the review is taking might I be appropriate. What will be done at this meeting? r See Purpose. , What would be the consequence of postponing this meeting? Loss of timeliness. Committee would become out of phase with NRC Staff and B&W Owners 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 Committee on 2/26/86).
- 2. Awaiting a course of action regarding the reassessment of B&W plants by the f Owners Group.
O
- s. >>
e l SCHEDULE OF ACRS SUBCOMMITTEE MEETING
-DATE. SUBCOMMITTEE MEETING STAFF ENGR. & MEMBEPS MAY.20, 1986' AD HOC TVA (Savio) Wylie, Ebersole, Michelson, Reed, Remick, Ward:
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? l 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 comments to Commission. I What would be the consequence of postponing this meeting? No comments to the Commission by May/ June full Comittee meeting. ! PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: 1 , 1. TVA Nuclear Performance Plan, Volume 1, Corporated dated 3/10/86. 2 .' Staff SER (if available). I
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O O SCHEDULE OF ACRS SUBCOMMITTEE MEETING 2 DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MAY 27, 1986 REGULATORY POLICIES AND (QUITTSCHREIBER) Lewis, PRACTICES Michelson, Remick, Siess, Wylie PURPOSE: To discuss the IIT Process after the issuance of the San Onofre and Davis-Besse IIT reports and the report of the Ad Hoc Independent Review Group on the Davis-Besse Incident. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Review the San Onofre and Davis-Besse IIT reports and the Ad Hoc Review Group Reports on Davis-Besse to determine the effectiveness of the IIT review process. 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 comments for possible full Comittee actions. 1 What would be the consequence of postponing this meeting? None PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. San Onofre Unit 1 IIT report.
- 2. Davis-Besse IIT report.
- 3. AdHocIndependentReviewGrouponDavisBesseIncident(tobeissued).
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l OV SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS MAY 28 & 29, 1986 SOUTH TEXAS 1 & 2 (EL-ZEFTAWY) Mark, Ebersole, Michelson, Siess t 4 ' PURPOSE: To review Houston Lighting and Power. Company's application for an OL. LOCATION: BAY CIT.Y, TX
; BACKGROUND:
What action is requested; by what date is it needed? Issue an ACRS " full power" OL letter; at the June 1986 ACRS meeting. What will be done at this meeting? i Subcommittee OL review in time for Comittee consideration at the June 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 in April. i i i e 1 pp. p.3
SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS JUNE 2, 1986 REACTOR OPERATIONS (ALDERMAN) Ebersole, Michelson, Moeller, Reed, Remick, Wylie PURPOSE: To review recent events at operating plants. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Periodic review of events, select a limited number for June full Comittee review. What will be done at this meeting? See Purpose. What would be the consequence of postponing this meeting? Failure to keep abreast of current events. i PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: t
- 1. Status report and schedule will be prepared prior to meeting.
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A' SCHEDULE OF ACRS SUBCOMMITTEE MEETING U DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS JUNE 3, 1986 SEVERE (CLASS 9) ACCIDENTS (HOUSTON) Kerr, Okrent, (1:00 P.M. - 5:00 P.M.) -Shewmon, Siess, Ward Cons.: Bender, Catton, Corradini, Davis PURPOSE: To review final draft of NUREG-0956, " Reassessment of the Technical Bases for
- Estimating Source Terms."
LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? The Committee provided comments on 12/12/85 regarding a draft of NUREG-0956. RES has considered the comments in preparing their final report. They want to discuss the resolution of the comments prior to issuing the final report. What will be done at this meeting? Review final draft of NUREG-0956. What would be the consequenc'e of postponing this meeting? Would delay the issuance of NUREG-0956, tentatively scheduled for July 1,1986. t PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. NUREG-0956, " Reassessment of the Technical Bases for Estimating Source Terms," Draft of Final Report available in late May 1986.
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l 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, Wylie*
- Part time 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 Commission, by June 1986, on the proposed NRC Safety Research Program and Budget for FY 1988 and 1989. l What will be done at this meeting? ) See Purpose. What would be the consequence of postponing this meeting? Timely completion of the report will be delayed. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. Proposed Safety Research Program and Budget for FY 1988 and 1989.
- 2. Budget Review Group mark (if available).
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS JUNE 4, 1986 LONG RANGE PLAN (MAJOR) Carbon, Lewis, (A.M.) Moeller*, Remick*, Wylie*
- Part time i
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 Committee; 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. 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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L N SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS JUNE 26, 1986 (tent.)- WESTINGHOUSE REACTOR PLANTS (HOUSTON) Reed, (8:30 A.M. - 12:00 P.M.) Ebersole, Etherington, Kerr, Michelson, Shewmon, Ward, Wylie Cons.: Catton PURPOSE: To continue discussions and comment cm NRC Staff actions taken with respect to the SONGS-1 waterhammer/ loss of AC power event. Follow-up Subcommittee meeting to February 12, 1986 meeting on same subject. LOCATION: WASHINGTON, DC BACKGROUND: l 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 comments for full Committee action. j 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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l SCHEDULE OF ACRS SUBCOMMITTEE MEETING {V DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS JUNE 26, 1986 AUXILIARY SYSTEMS (DURAISWAMY)'Michelson, (1:00 P.M. - 5:00 P.M.) Ebersole, Reed, Shewmon, Wylie PURPOSE: The Subcommittee will discuss: (1) the status of the Appendix R compliance, (2) differing technical views among the Staff, (3) proposed research and associated budget for FY 1988 and 1989 in the fire protection area, and (4) updates on the progress being made in the Sandia experimental program on fire protection. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? None What will be done at this meeting? See Purpose. What would be the consequence of postponing this meeting? None PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: . Will be identified later. O
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' JUNE 27, 1986 DAVIS-BESSE (ALDERMAN) Remick, (RESTART) Carbon, Reed, Siess PUPPOSE: The Subcommittee will review start-up activities for Davis-Besse.
LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Comittee letter; July full Committee meeting. What will be done at this meeting? Review start-up activities for Davis-Besse. - What would be the consequence of postponing this meeting? O' Minimal. Start-up of Davis-Besse projected for October 1986. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. SER expected May 1986.
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SCHEDULE OF ACRS SUBCOMMITTEE' MEETING
.G DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS JUNE 27, 1986 JOINT OCCUPATIONAL & (SCHIFFGENS/DURAISWAMY)
ENVIRONMENTAL PROTECTION tioeller, Michelson, . SYSTEMS / AUXILIARY SYSTEMS Ebersole, Mark, Reed, Shewmon, Wylie Cons.: First, Healy, Kathren PURPOSE: To: (1) review a draft AEOD report on the effects of ambient temperature on I&C systems, (2) be briefed on the status of various control room HVAC systems problems and the Staff's control room habitability improvement effort, (3) discuss with the Staff the 1 mrem /yr " cutoff" dose rate for the calculation of collective population doses, and (4) be briefed on the Staff's - evaluation of the Shearon Harris Chilled Water Systems. LOCATION: WASHINGTON, DC BACKGROUND: i What action is requested; by what date is it needed? The Subcommittees want to clean up numerous items that have been carried over from , previous meetings to see what Staff actions have or have not been taken. There is no deadline associated with these. i What will be done at this meeting? i A ' report for the Committee will be prepared on these items. The Subcommittees may want to draft a letter on the AE00 report. What would be the consequence of postponing this meeting? l Probably none. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: To be provided later. 1 O l 1
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SCHEDULE-0F ACRS SUBCOMMITTEE MEETING 7 4 DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS JULY 8, 1986 RELIABILITY ASSURANCE (MAJOR) Wylie, Ebersole, Michelson, Siess PURPOSE: To review the final resolution of USI A-46, " Seismic Qualification of Equipment in Operating Plants." LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Final ACRS approval of USI A-46; if appro?riate present to the full ACRS in July. What will be done at this meeting? O Review final resolution of USI A-46. Staff will want a recommendation for full Committee review of the final resolution. What would be the consequence of postponing this meeting? Loss of timeliness. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: Prior to meeting the results of public comments and CRGR review, as well as, final resolution of A-46 should be available. i l O
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, SCHEDULE OF ACRS SUBCOMilTTEE MEETING
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DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS AUGUST 5 & 6, 1986- EXTREME EXTERNAL PHEiiOMENA (SAVIO) 0krent, Lewis, et al. PURPOSE: To conduct a workshop on what is currently known regarding the importance of seismic risk to nuclear power plants. It is expected that the emphasis will be on the seismic hazard. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed?
- To conduct an ACRS reevaluation of the current ACRS position on seismic risk.
What will be done at this meeting? Invited experts will give presentations on the various issues associated with the estimation of seismic risk. This information will be used in the estimation of O seismic risk. What would be the consequence of postponing this meeting? A major EPRI report on seismic hazard for the United States will be available in July
- 1986. The August 586 date would allow some time for this report to be reviewed by the ACRS before this meeting. The Tuesday and Wednesday before the August ACRS meeting have been chosen in the hopes of having the maximum number of ACRS members attend.
The subject is an important one and one which the ACRS has agreed to discuss as soon as practical. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: To be provided later.
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b i SCHEDULE OF ACRS SUBCOMMITTEE MEETING a !- DATE- SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS 5 0CTOBER'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: ) i What action is requested; by what'date is-it needed? N/A i I What will be done at-this meeting? , 'See Purpose. i What would be the consequence of postponing this meeting? i l Wingspread is available only on certain days. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: ! Exchange papers will be prepared after agenda planning is completed in April. i 1 a , t
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q EC+DTJdE & #tTi SNE'DHTEE PEETING DATE SUb;0Nf6 ict ?EIME 53E ENC 4. & MEMBERS TO BE DETERMINED GEGI +EW !EtW!1t SYSTEMS {iOFJiMERO Ward, (late MAY) Ihetsulie, Michelson, Peet Cans.. :: Catton, Davis PURPOSE: The Subcommittee w5H revEew WW s. Act!:m Man to addness camcerns with the reliability of .certain p'innts' W 5,ystems. LOCATION: WASHINGTON, DC BACKGR0llND: Wh_at action is requested; by what date fs it TieededT l Meeting needs to be scheduled fn coordratfne with submittal of Action Plan to CRGR. What will be done at this meettng? See Purpose. b What would be the consequence of postpatrim; tnis weeting? See above. Review is scheduled for ACE cons 5rfmt' ion et June AUR5 neeting. PERTINENT PUBLICATIONS AND THEIR AVAILAER3Th i j 1. Action Plan package to be provided Wm. waHatbe. 4 i 6 4 O L
.l X SCHEDULE OF ACRS SUBCOMMITTEE MEETING
( )
.\ J DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED SPENT FUEL STORAGE (ALDERMAN) Siess, Kerr, (MAY/ JUNE) Moeller, Remick, Shewmon
- PURPOSE
- To continue review of 10 CFR Part 72 and Monitored Retrievable Storage (MRS)
LOCATION: WASHINGTON, DC BACKGROUND: i What action is requested; by what date is it needed? Review background to write Committee letter on 10 FR 72 and MRS. 10 CFR 72 hasn't gone out for public comments yet -- probably about May 1986. What will be done at this meeting? Detailed review of 10 CFR 72 and discuss MRS. What would be the consequence of postponing this meeting? Date hasn't been established yet. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: l O
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i, 3 SCHEDULE OF ACRS SUBCOMMITTEE MEETING i- . DATE- SUBCOMMITTEE MEETING -STAFF ENGR. & MEMBERS L TO BE DETERMINED METAL COMPONENTS (IGNE)Shewmon, (JUNE) Etherington, Lewis, (Tentative) Michelson, Okrent, Ward ! Cons.: B. Thompson, W. Shack
- 4. 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 needed? l
- ~
The Subcommittee requested that it keep abreast of the progress of NDE of cast . stainless steel materials. Early spring,1986 -- befcre 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. ! Wh'at would be the consequence of postponing this meeting? l Nothing. Will not be able to intelligently coment on GDC-4 broad scope rule i regarding the detection and sizing of flaws in cast stainless steel. 1 PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: i
- 1. Status Report on NDE of cast stainless steel, WOG Materials Subcommittee (Tentative
- Schedule, June.
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5 SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED GAS COOLED REACTOR PLANTS (MCKINLEY) Siess, (JUNE) (i day) Ebersole, Michelson, Reed, Wylie PURPOSE: To review the applicability of NRC requirements for equipment qualification and cable testing to Fort St. Vrain, an HTGR. LOCATION: WASHINGTON, DC BACKGROUND: What action is requested; by what date is it needed? Neither the NRC nor the licensee has requested ACRS action. The licensee currently has an exemption to the NRC equipment qualification requirements which permits operation at power levels up to 35%. This exemption expires on May 31, 1986 after which the plant will have to shut down. The Subcommittee wocid like to better understand the NRC's bases for choice of the harsh environment as compared to water 'O reactors. The NRC is requiring that certain cable be examined to meet the requirements of Appendix R. The Subcommittee would like to see if this matter is amenable to resolution by the same techniques as used in the SEP plants. What will be done at this meeting? See Purpose. What would be the consequence of postponing this meeting? Perhaps the continued use of inappropriate requirements,'perhaps unnecessary expense to the utility with the premature decommissioning of FSV. PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: Tobefrovided. . i O e
!O SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED THERMAL HYDRAULIC PHENOMENA (B0EHNERT) Michelson, (JUNE / JULY) Ebersole, Etherington, Reed, Ward Cons.: Catton, Schrock, Sullivan, Tien i
PURPOSE: To review GE's application for use of the SAFER /COREC00L ECCS code on BWR NJP plants. LOCATION: WASHINGTON, DC j 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? !O s See Purpose. What would be the consequence of postponing this meeting? l Possibly lose coordination with Staff review. PERTINENT PdBLICATIONS AND THEIR AVAILABILITY:
- 1. GE Licensing Topical Report on SAFER /COREC00L (will be provided when issued - early
- June ?).
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p SCHEDULE OF ACRS SUBCOMMITTEE MEETING d 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. - l
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SCHEDULE OF ACRS SUBCOMMITTEE MEETING U^ ! 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 2 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. I 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, i !O i
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p .;. > t' , s (-' a, cs ( i SCHEDULE OF ACRS SUBCOMMirTEE MEETING d i g. DATE + h0BCOMMITTEEMEgING- STAFF ENGR. & MEMBERS TO BE DETERMINED DECAY HEAT REliOVAL SYSTEMS (BOEHNERT) Ward, (JULY) Ebersole, Michelson, 4 3 Reed Cons.: Catton, Davis e , PURPOSE: BeginreviewofNRR'spropMedbesolutic>npositionforUSIA-45," Shutdown Decay Heat Removal System." e(s ( LOCATION: WASHINGTON, DC l s
',i s BACKGROUND: sU
. s What action is requested; by wha,c date is it nescid? ,
Review of NRRs proposed resolution.' position. 'Need in July to track NRR's resolution schedule. < What will be done at this meeting? 5 I (
\- See purpose. t .
. ,i What would be the consequence of pos+.poning this meeting?
Could become critical path on NRR/CRGR ieview schedJ1'e. s< s , ( PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:
- 1. USIA-45resolutionpackagewilkbeprovidedwhenissuedbyNRR.
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^x SCHEDULE OF ACRS SUBCOMMITTEE MEETING 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? i 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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i s y SCHEDULE.0F ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED RELIABILITY AND PROBABILISTIC (SAVIO) Okrent, Kerr, ASSESSMENT Ebersole, Lewis, Mark, Michelson, Siess, Ward, Wylie ( s 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; the 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. s_ What will be done at this meeting? Review of the Millstone 3 PRA for infomation. ,l What would be the consequence 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 (distributed).
- 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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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE . SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED WESTINGHOUSE REACTOR (EL-ZEFTAWY) Ebersole, PLANTS Etherington, Michelson, (CLOSED) Shewmon, 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 naeting? 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: l 1. RESAR SP/90 Standard Plant Design (50-601). l
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- 0 cY APPENDIX IV REPORT OF MARCH 26, 1986 ECCS SUBCOMMITTEE MEETING h
/ April 8,1986 TO: ACRS MEMBERS FROM: D. A. WARD, CHAIRMAN b>t%%wJdL A C-CC.S, Report from the March 26,1986 Meeting ECCS Subcommittee on Duke Power Company's Proposal to remove the UHI System at its McGuire Plant.
Attendance at the subcommittee meeting included Mr. Reed and myself, and also Consultants Catton, Schrock, Sullivan, Tien and Davis. Back about 10 years ago when the first ice condenser plant, DC Cook, was being licensed, there was found to be difficulty in satisfying Appendix K requirements for ECCS performance without putting restrictive limitations on the so-called FQ value, which is an expression powei peaking in an operating reactor core. The reason for this is that in the limiting case of large break LOCA with blowdown into the ice condenser containment, the backpressure, that is the containment pressure, is so low, that vapor forms and there is a steam binding effect in parts of the reactor ] . cooling system. 'Ihis meant that flow to the core was calculated to be delivered less J j rapidly than othenvise. The steam binding caused increased pressure drops in both the steam generator and the pump. With this decreased flow into the core, or lower rate of reflooding, it was found that calculated peak clad temperatures (PCT) were higher than they would be for similar conditions in a plant with a large dry containment. To counter this effect in the next ice condenser plant which was Duke Power's McGuire, the licensee and Westinghouse agreed to add another piece to the ECCS system. This is another accumulator, which would inject several thousand gallons of water directly into the upper head, rather than into the cold leg in the event of a large break LOCA. 'Ihis is the so-called Upper Head Injection system. It consists of a tank containing several thousand gallons of borated cold water and another tank filled with nitrogen under pressure of about 1500 psi (?) to push this cold water directly into the upper head when the RCS pressure drops. To - accomplish this injection, it is necessary to have, in addition to these tanks, about four 8-inch lines connected into special nozzles in the upper head. In addition to that, the upper head intemals are modified somewhat to provide for improved [. g. ff
2 April 8,1986 l ACRS MEMBERS l distribution of flow from the upper head down into the fuel core. The regular large cold leg' accumulator is still maintained in the system and still functions in the traditional way. With this system, including both the traditional cold leg accumulator and the Upper Head Injection system, it was then calculated that maximum peak clad tempemtures would not be excessive, and that in fact, the McGuire reactor could be operated with peaking factors more or less up to the normal range permitted in the other similar Westinghouse reactors. There were some difficulties, however, in analyzing the situation. %, TRAC, RELAP, RETRAN, and other codes that had been developed to analyze ECCS performance (vintage of 10-12 years ago) had difficulty in accurately calculating the performance and the details of the UHI system which now provided for post-blowdown reflood of the core from the top as well as the normal post-blowdown reflood from the bottom by the cold leg accumulator. However, after a lot of discussions and, I gather, considerable agonizing, the applicant, Westinghouse and the NRC staff all reached the conclusion that the Upper Head injection system would perform satisfactorily and that normal peaking factors could be used in operation of the McGuire units. This, I believe, was based largely on O informed judgments rather .:an being very directly a result of the code calculations. So all was well. However, Duke's experience in operating McGuire plant with the Upper Head Injection system has not all been well. It's not that there have been any indications that the system would not perform its job to cool the core in a hypothetical large break LOCA event. Rather, it is that the system is a practical, operating " pain in the neck". There are several reasons for this, most of them center around the complexity of this added system and in particular, the four 8-inch jumper lines that are flanged into nozzles in the upper head and have to be removed when the upper head is removed for refueling. That and certain operational problems, maintaining water levels and nitrogen pressures, etc. in these additional accumulators have led Duke Power to address the question internally whether it be better off with the 1 system removed. In essence, whether it was a mistake to put it in, in the light of present' knowledge. Part of present knowledge is that the codes in 1985 are more capable of analyzing ECCS performance than were the codes of 1975 vintage. This applies to the so-called evaluation model codes or EM codes, but also in 1985 we have more tools in the form cf so-called best estimate (BE) codes that help better understand the plant characteristics and the phenomena that occur in large break LOCA with ECCS
\ action.
L M
3 April 8,1986 ACRS MEMBERS i I So after some analysis, Duke has petitioned the NRC for permission to essentially remove the UHI system with the argument that the new smatter methods of analyzing the large break LOCA show that peak clad temperatures are not excess even without the UHI functioning and that, in fact, all Appendix K requirements can be met in the plant without the UHI and without the necessity for unusual restrictions on core peaking values. The Staff has analyzed Duke's proposal. It has taken about a year for this analysis to be completed; there was some difficulties in doing it. The Staff has now concluded that Duke's petition is valid and the Staff intends to issue ^^ an SSER giving their OK to the proposal. The ECCS Subcommittee has reviewed the proposal and agree that Dukgs some good reasons for wanting to remove the system. Is gcr.: :CIhe reasonifor removal are not strongly safety related, sc n agree ht it is simply up to D* what Sy want en an However, the Subcommittee disagrees somewhat with Duke and with the Staff in that we think there are some safety related reasons for removing the system. These are the reduced complexity of the plant and particularly of the ECCS system, both the amount of piping that is there and there in critical locations and vulnerable to failure, and the complexity of operating the plant both in nonnal situations and in any sort of emergency. In addition, there is concern that the UHI accumulator system is another major source of a non-condensible gas, in this case nitrogen, that could be somehow, in some complex accident scenario be injected into the reactor cooling system and conceivably interefere with the good heat transport that would be needed to keep the core cool. The calculations by both the licensee and the Staff show that the nitrogen won't be a problem in relatively straightforward and traditional LOCA scenarios. However, the Subcommittee has concern that it could be a problem in other sorts of unexpected and not specifically analyzied scenarios. So, the bottom line is we think there is a safety benefit as well as a practical, operational benefit in removing the system. That brings us down to whether there is a safety penalty in removing the system. Duke claims, and the NRC staff supports them in their claim, that with both EM analysis and BE analysis, there is essentially no difference in the maximum peak clad temperature calculated for the McGuire plant and without the UHI. The Subcommittee does not disagree, but there is another additional wrinkle: When the large break LOCA event is calculated by EM, there are two clad temperature peaks p that occur. One occurs within the first few seconds during the blowdown phase of Q the accident and a second occurs about 40 seconds to a minute into the accident,just
.A? - Ed
4 April 8,1986 ACRSMEMBERS g O before reflood begins to cool the core again. With the EM models both these peaks are approximately the same magnitude and provide about the same reasonably , comfortable margin from the 2200'F limit on peak clad temperature. This sort of performance is typical for PWRs. With EM model calculations you get two peaks, ' one in blowdown and one in reflood, both of about the same magnitude. However, ' there is a difference when BE calculations are made. That is, in non-ice condenser plant for most conditions, you tend to see the same two peaks, although there are some combinations of parameters where the second peak doesn't occur, or is much smaller than the blowdown peak. With McGuire with the UHI functioning (as it is l as present) most of the calculations show that the second peak will not occur. Although under some assumptions about more power and peaking, etc., the second peak, that is the peak during reflood, does occur. However, the calculations for McGuire with the UHI removed, show a much stronger tendancy, I guess it could be expressed as over the spectrum of possibilities, there is a much higher likelihood
. that a second peak will occurjust before the reflood takes hold.
l This puts McGuire without the UHI meeting the regulations, but more like non-ice condenser plants in that there seems to be a fairly high likelihood in best estimate l analyses that a significant peak clad temperature will occur during the reflood phase. l This means that McGuire without the UHI will have a large break LOCA
~
performance quite similar to almost all other PWRs. However, McGuire with UHI functioning, tends to have a peak clad temperature profile which, for the majority of the spectmm of possibilities, does not include a second peak during the reflood phase. So here we have the interesting sort of delimma in that removing the UHI system from McGuire will possibly make some marginal negative difference in the ability of the overall ECCS system to cope with a hypothetical large break LOCA but will, in fact ,make McGuire without the UHI more like the majority of other PWRs, that is, non-ice condenser PWRs without the UHI. , So the bottom line is that the UHI really seems to help things a little bit in large break LOCA situation but it is help that really isn't needed. One interesting additional issue I think we have to consider is that the second peak during reflood is of particular interest in that it occurs at the time when the pressure in the reactor cooling system is very, very low. This means that there must be a higher probability for fuel clad ballooning and bursting during a second peak, than during the first peak even though the two peak temperature values are the same. So the UHI helps protect against this higher likelihood of hypothetical clad failure. However, we also have to consider that the hypothetical fuel bursting is just in a d' P
a 5 April 8,1986 ACRS MEMBERS 4 very limited number of assemblies, a statistically small number of rods. And, that in fact although clad burst would be an unhappy event for the plant owner, it is not obvious that there is anyparticularly important direct contribution of risk to public health and safety. So finally , the Subcommittee makes the recommendation that we endorse the Staffs proposal to accept the Duke application for removal of the UHI system at McGuire. The question of whether " conservative" is necessarily " safest" is also germane to this issue. During subcommittee discussions, the licensee observed that normal practice calls for tuning the setpoints of the accumulators (i.e., adjust water level and nitrogen pressure) to provide an optimized response to a LOCA using EM analysis. It seems possible that safety would be better served if plant systems were adjusted to operate optimally for the BE scenarios. Regulations irrregulatory practice do not address this, od O I l l O
- ,e r>
APPENDIX Y '
.NRC UHI PRESENTATION BY C. BERLINGER
- 'O NRR STAFF PRESEniAiiore su suna
SUBJECT:
PROPOSED UPPER HEAD INJECTION REMOVAL / ISOLATION FOR MCGUIRE UNITS 1 & 2 APRIL 10, 1986 5 BY CARL BERLINGER CHIEF, REACTOR SYSTEMS BRANCH DIVISION OF PWR LICENSING-A 492-7415
- O e
I e e e O
,.9. s3
y 4A L-- .. - - m a.,. L i l O i OUTLINE MCGUIRE UHI REMOVAL / ISOLATION o LICENSEE PROPOSALS o SAFETY ANALYSIS o NRR UNDERSTANDING 0F UHI REMOVAL BENEFITS I o UHI REMOVAL EFFECTS ON SBLOCA, ' TRANSIENTS, AND CONTAINMENT o CONCLUSIONS 5 4 4 O j
UHI REMOVAL PROPOSALS AND SUBMITTALS
'- MAY 9, 1985 - PROPOSES AMENDMENTS TO CHANGE TECHNICAL SPECIFICATION FOR OPERATION AT 100% POWER WITH i UHI DELETED OCT, ?, 1985 TRANSMITTAL OF SEPTEMBER 1985 DOCUMENT "MCGUIRE NUCLEAR STATION, SAFETY ANALYSIS FOR UHI ELIMINATION" (BASH CODE)
OCT, 14, 1985 REQUESTS FOR ADDITIONAL TECHNICAL SPECIFICATION CHANGE REGARDING ECCS COLD LEG ACCUMULATOR PARAMETERS. l DEC. 17, 1985 SUPPLEMENT 2 TO PROPOSED TECHNICAL SPECIFICATION CHANGES TO PROVIDE FOR INTERIM OPERATION WITH UHI ISOLATED. (AS IS, UHI LOCKED OUT, OR UHI EXTERNAL REMODELLED) DEC, 23, 1985 RESPONDS TO STAFF REQUESTS REGARDING RADIOLOGICAL ASPECTS JAN, 14, 1986 CLARIFIES TECHNICAL SPECIFICATION CHANGES l AND BASES - MARCH 17,1986 REVISED SAFETY ANALYSIS USING BART INSTEAD OF BASH e V
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- g. H
SAFETY ANALYSIS o APPROVED CODE "BART" 1981 VERSION W EVALUATION MODEL FOR LBLOCA WCAP 9220-P-A REV, 1, 1982 o " WESTINGHOUSE EMERGENCY CORE COOLING SYSTEM EVALUATION (9 MODEL APPLICATION TO PLANTS EQUIPPED WITH UPPER HEAD INJECTION", WCAP- 8479, (PROPRIETARY), JANUARY, 1975, e 4 O r? N
IO TABLE 15.6.4-4 LARGE BREAK LOCA RESULTS FUEL CLADDING DATA PEPFECT MIXING CD = 0.8 CD = 0.6 CD = 0.4 DECLG DECLG DECl.G RESULTS PEAK CLAD TEMPERATURE (*F) 2018 2175* 1791 PEAK CLAD TEMPERATURE LOCATION (FT) 6.5 6.5 6.75 LOCAL ZP/H O REACTION (MAX), (%) 2.6 5.1 1.9 2 LOCAL ZR/H 0 LOCATION, (FT) 6.5- 5.75 7.0 2 TOTAL ZR/H2O REACTION (%) (0.3 <0.3 ( 0.3 HOT ROD BURST TIME, (SEC) 76.1 83.0 108.0 HOT R0D BURST LOCATION, (FT) '6.0 5.75 7.0 TABLE 15.6.4-5 IMPERFECT MIXING CD = 0,6 DECLG PESilLTS 7 PEAK CLAD TEMPERATURn. (*F) 19%6 PEAK CLAD' TEMPERATURE (FT) 7.0 LOCAL ZR/H O REACTION (MAX), (%) 2.6 2 LOCAL ZR/H O LOCATION, (FT) 7.0 2 TOTAL ZR/H 0 REACTION, (%) < 0.3 7 HOT ROD BURST TIME, (SEC) 68.1 HOT ROD BURT !.0 CATION, (FT) 7.25 8 ,9. ??
O O O LARGE BREAK LOCA ESULTS REL CLADDING DATA
= 0.8 0.6 0.4 C = 0.6 LIMITS Cm_CLG "DE Cn_
"DE =__CLGCn "DE =__CLG hCLG 10 CFR 50.%
(mx st) RESULTS PEAK CLAD TEMPERATtRE (*F) 1865 1895 1863 2132. 2200 PEAK CLAD TEMPERATURE LOCATION (FT) 6.75 6.75 6.75 6.50 LOCAL ZR/H,% REACTION (MAX), (%) 2.53 2.12 2,16 5.05 17.00 LOCAL ZR/lh0 LOCATION (FT) 5.50 6.00 5.50 6.50 TOTAL ZR y REACTION (%) 0.3 0.3 0.3- 0.3 1.0 HOT ROD BURST Tim, (SEC) 61.4 62.2 88.8 63.0 Oxx.ABLE Q . hot ROD BURST LOCATION, (FT) 5.50 6.00 5.50 6.00 t.x
l 0 - O O i EVALUATION MODEL i . j . i ! N E , E ,, - Mt _/ CD = 0.8 DECLG Max.8' cc O f/ V N j
~
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'UHfCD = 0.6 DECLG
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- O I
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- -t- :::::: .; :o iI.6l :t n Maximum Hot Rod Clad Temperatures e
4 W
SMALL BREAK LOCA (::) o NOTRUMP - STAFF APPROVED EM WCAP-10071 A AUGilST l 1985 o COMPARISON UHI N0 UHI 1 PCT 1499'F 1488 SIZE 6 IN, DIA, 3 IN, DIA, TRANSIENTS o ONLY TRANSIENTS DEPRESSING PRIMARY SYSTEM PRESSURE BELOW 1200-1300 PSIA RANGE CONSIDERED I.E. () INADVERTENT OPENING OF S.G, RELIEF OR SAFETY VALVE AND STEAM SYSTEM PIPING FAILURE 1 o DNB DESIGN BASIS MET IN ALL CASES I I e O s -7. d'- -
CONCLUSIONS
~. \-- 1. LBLOCA ANALYSIS PERFORMED WITH APPROVED EM (1981 VERSION V EM) o RESULTS WITHIN.10CFR 50.46 LIMITS o SLIGHTLY MORE MARGIN THAN WITH UHI BECAUSE:
A) UPDATED FLECHT. CORRELATION B) W UHI EM REQUIRED CONSERVATISMS TO COMPENSATE FOR DATA AND MODELLING UNCEP.TAINTIES-UHI BENEFITS NEVER REALIZED
- 2. SBLOCA AND TRANSIENTS DO NOT EXCEED LIMITS WITH OR WITHOUT UHI - NO EFFECT 3, STAFF AGREES THAT UHI:
- 1) ADDS TO OPERATIONAL, MAINTENANCE, AND INSPECTION PROBLEMS, 2) INCREASES LIKELIHOOD OF SMALL BREAKS, LEAKS, AND FAILURES, 3) IT INCREASES RADIATION EXPOSURE OF OPERATING l PERSONNEL. THEREFORE THE STAFF AGREES WITH THE LICENSEE THAT UHI SHOULD BE DELETED,
- 4. BASED ON MORE REALISTIC ANALYSES WE SEE SOME SMALL l RELATIVE BENEFIT OF UHI DURING BLOWDOWN AND DURING l REFLOOD. HOWEVER, CLADDING TEMPERATURES IN BOTH CASES ARE SUFFICIENTLY LOW TO HAVE A SMALL EFFECT ON PREDICTED RELEASES, l
O r-o . C
APPENDIX VI ' REMOVAL / ISOLATION OF THE UPPER . ' HEAD INJECTION SYSTEM
+
O i-1 REMOVAL / ISOLATION OF THE :
- UPPER HEAD INJECTION SYSTEM i
l APRIL 10, 1986 lO BY W. D. RECKLEY DUKE POWER COMPANY l O es -.
INTRODUCTION DURING FIRST SEVERAL CYCLES OF OPERATION OF THE MCGUIRE NUCLEAR STATION, PROBLEMS WERE ENC 0UNTERED WITH THE UPPER HEAD-INJECTION SYSTEM. AS A RESULT-0F THE PRO-BLEMS AND THE ADDITIONAL DELAYS ASSOCIATED WITH NORMAL - UHI RELATED ACTIVITIES, DUKE POWER UNDERTOOK A DETAILED EXAMINATION OF: HISTORY OF UHI RELATED PROBLEMS AND DELAYS
. EXPECTED CONTINUATION OF UHI PROBLEMS AND DELAYS -
IMPACT OF THE ELIMINATION OF UHI ON PLANT SAFETY
. BENEFITS OF THE ELIMINATION OF UHI i
O
O UHI OPERATIONAL PROBLEMS
- 1) NORMAL DELAYS ASSOCIATED WITH THE CONNECTION &
FILL / VENT OF UHI PIPING AFTER RELOADS /RCS DRAINDOWNS d
- 2) COMPONENT PROBLEMS A) RUPTURE MEMBRANE
. B) LEVEL INSTRUMENTATION C) CHEMISTRY D) SIGHT GLASS FAILURES
- 3) EXPECTED LOST AVAILABILITY IS APPR0XIMATELY 10 HOURS AFTER EACH OUTAGE PLUS ADDITIONAL LOSSES DUE TO OCCASIONAL PROBLEMS WITH SYSTEM OPERABILITY O
- s. o
.O INCENTIVES FOR UHI ELIMINATION ELIMINATES UHI RELATED DELAYS AND PROBLEMS REDUCES FREQUENCY.OF FORCED PLANT MODE CHANGES ELIMINATES NEED TO DEDICATE RESOURCES FOR UHI RELATED-PROJECTS
+
REDUCES PERSONNEL RADIATION EXPOSURE O . RESULTS IN MCGUIRE DESIGN WHICH IS SIMILAR TO STANDARD WESTINGHOUSE DESIGNS
. RESULTS IN REDUCTION OF OVERALL PLANT COMPLEXITY, PROCEDURES, MAINTENANCE, ETC, O
;-j. c3
\v' SAFETY ANALYSES LBLOCA ANALYSES PERFORMED WITH APPROVED EVALUATION MODEL AND RESULTS SATISFY ALL REGULATORY REQUIRE-MENTS SBLOCA ANALYSES PERFORMED WITH APPROVED EVALUATION MODEL AND RESULTS SATISFY ALL REGULATORY REQUIRE-MENTS NON-LOCA TRANSIENTS INVOLVING UHI INJECTION , (STEAMLINE BREAK) ANALYZED WITH APPROVED METHODOLOGY t
AND RESULTS SATISFY DESIGN CRITERIA CONTAINMENT RESPONSES EVALUATED AND EXISTING ANALYSES SHOWN TO BE BOUNDING l I
^
U CONCLUSIONS ELIMINATION OF UHI PROVIDES SIGNIFICANT OPERATIONAL BENEFITS ELIMINATION OF UHI REMOVES POTENTIAL PROBLEMS RELATED TO UHI COMPONENTS WHICH MAY INTRODUCE SAFETY CONCERNS i PLANT OPERATION WITHOUT UHI CONTINUES TO SATISFY APPLICABLE REGULATORY REQUIREMENTS O . LBLOCA ANALYSES PERFORMED WITH CONSERVATIVE EVALUATION MODEL AND SANDIA TRAC MODEL DEMONSTRATE THAT PLANT RESPONSE WITHOUT UHI IS SIMILAR TO PLANT RESPONSE WITH UHI CONSIDERING UHI RELATED CONCERNS AND RESULTS OF SAFETY ANALYSES, UHI ELIMINATION REPRESENTS A NET REDUCTION IN PLANT RISK O s ,s
)
i.
~
O f SCHEDULE
. ' PLAN TO OPERATE: UPCOMING CYCLES (MCGUIRE 1. CYCLE 4, MCGUIRE 2 CYCLE 3)'WITHOUT OPERABLE-UHI SYSTEM L
- PLANT OUTAGE SCHEDULES:
. UNIT 2: MARCH 14 - MAY 17
. UNIT 1: JUNE 3 - AUGUST 19 1 I C h i l O __s,x- _:
's i' APPENDIX VII RECENT SIGNIFICANT EVENTS
'n
,c-0 Agenda fe Meeting on Aprii lu, Asoo n
2:45 p.m. t. - Room 1045,' H Street, r > . RECENT SIGNIFICANT EVENTS Presenter / Office Date Plant Event telephone P3 3/28/86 Rancho Seco Class 1E/htationBattery H. Bailey, IE .2. Problems- - 492-9006 3 2/8/86 Vt. Yankee Failure of Standby Liquid E. Weiss, IE Y
' Control System 492-9005 1
s Implementation of TMI Action G. Lapinsky, NRR 12/16/86 ------ Item I.D.2.- Safety Parameter 492-8166
~f Display System at Operating Reactors ,
3/3/86 Turkey Point Component Cooling Water System Protilems D. Mcdonald, NRR 492-4777
/h i
t
- g. > z.
RANCHO SEC0 - CLASS 1E STATION BATTERY PROBLEMS MARCH 28,1986 (H? BAILEY, IE) PROBLEM: EXTENSIVE EROSION (FLAKING) INSIDE BATTERY NEAR THE TOP 0F-THE PLATES BETWEEN THE PLATES AND THE PLATE SUPPORTS.
- BATTERY SPACING IS-INCORRECT.
SIGNIFICANCE: CLASS'1E STATION BATTERIES NOT SEISMICALLY QUALIFIED; POTENTIAL FOR LOSS OF ALL DC POWER DURING SEISMIC EVENT. DISCUSSION: , EROSION PROBLEM DISCOVERED BY BATTERY VENDOR (GNB) IN FEBRUARY 1986. PRESENT TECH SPEC SURVEILLANCES WOULD NOT REVEAL THIS PROBLEM. a
- VENDOR REPORTS THIS EROSION IS NOT UNUSUAL FOR A "PLANTE" TYPE BATTERY MORE THAN 15 YEARS OLD. BATTERIES DID NOT SHOW SIGNIFICANT EROSION 1 YEAR AGO DURING VENDOR INSPECTION.
BATTERY-TO-RACK AND CELL-TO-CELL SPACING ALSO FOUND TO BE OUT OF DESIGN SPECIFICATIONS. VENDOR INFORMED LICENSEE THAT BATTERIES WERE NOT SEISMICALLY QUALIFIED, DUE TO IMPROPER BATTERY-TO-END-RACK SPACING. SOME SPACING HAS BEEN CORRECTED. FOLLOWUP: ADEQUACY OF LICENSEE S BAT rJY SURVEILLANCE PROGRAM IS UNDER REVIEW BY REGION V. CONCERNS INCLUDE THE RATED LOAD TEST AT EACH REFUELING AS WELL AS EROSION PROBLEM. SOME COMMONWEALTH EDISON COMPANY PLANTS WERE REPORTED TO HAVE THE SAME TYPE BATTERY AND CECO IS REPORTED TO BE REPLACING THEM. RANCHO SEC0 PLANS TO REPLACE THE "PLANTE" (LEAD) TYPE BATTERIES WITH A LEAD-CALCIUM TYPE BATTERY. 0 9 - . _ =
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I . VERMONT YANKEE - FAILURE OF STANDBY LIQUID CONTROL SYSTEM f-tBRUARY 8, 1985 (ER C-WtlSS,lt)
- PROBLEMS SLCS SQUIB VALVE FAILED TO FIRE DURING SURVEILLANCE TEST
~
PLANT WIRING IN TERMINAL B0X QUESTIONABLE SQUIB VALVES HAD INCORRECT PIN-TO-BRIDGEWIRE GROUPING CONTROL ROOM INDICATION OF " CIRCUIT CONTINUITY" WAS FALSE SIGNIFICANCE OTHER PLANTS COULD HAVE DEFECTIVE SQUIB VALVES IF PROBLEM OCCURS IN FUTURE, FALSE INDICATION IS MISLEADING CIRCUMSTANCES MANUFACTURING MOVED FROM NEW YORK TO FLORIDA IN 1983
" TRAVELER" MAY HAVE CAUSED CONFUSION
- CONAX ENGINEER AT VT YANKEE ON FEB 19 CONFIRMED ASSEMBLIES DEFECTIVE j POTENTIALLY AFFECTED LICENSEES NOTIFIED BY REGIONS SOME SUSPECT SQUIBS DISTRIBUTED OVERSEAS NO OTHER DEFECTIVE ASSEMBLIES IDENTIFIED FOLLOWUP IE INF0 NOTICE ISSUED FEBRUARY 21, 1986 IE INFO NOTICE SENT TO INTERNATIONAL PROGRAMS IE VENDOR INSPECTION PLANNED ,
k a
. . - /*
4 POTENTIALLY SUSPECT ASSEMBLIES Primer Part Number , Assembly Plant 3 of Pieces Serial Numbers 4 1617-139-01 Vt. Yankee 6 (defective) 5522557 1617-139-01 Dresden 6 546-551 1621-240-01 Shoreham 7 , 635-640, 668 1621-240-01 Duane~ Arnold 6 669-674 1621-240-01 Susquehanna 19 675-681, 686-697 1621-240-01 Limerick 10 699-708 1621-240-01 Piigrim 3 659-661 NRC Regional representatives have contacted the above facilities by telephone. O llm
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IMPLEMENTATION OF T.A.P ITEM I.D. 2, SAFETY PARAMETER DISPLAY SYSTEM, ( .A_T OPERATING REACTORS - DECEMBER 16, 1985 ( (GEORGE LAPINSKY, NRR) PROBLEM: . SPDS FOUND TO BE MALFUNCTIONING / UNAVAILABLE AT 3 0F 6 OPERATING REACTORS SURVEYED l - SIGNIFICANCE: SUBSTANTIAL FRACTION OF SYSTEMS DO NOT FULFILL THE INTENT OF NUREG-0737, SUPP, 1, "T0 AID... IN DETERMINING THE SAFETY STATUS OF THE PLANT" INVALID AND INACCURATE INFORMATION COULD MISLEAD USERS SOME OPERATORS REFUSE TO USE SPDS BECAUSE OF P00R RELIA-BILITY DISCUSSION: REVIEW 0F 0.R SAFETY ANALYSIS REPORTS WERE ONLY TO IDENTIFY SERIOUS SAFETY QUESTIONS l BY MID-85 MANY PLANTS HAD OPERATIONAL SPDSs THAT HADN'T BEEN AUDITED l SAIC AND HFEB SURVEYED SIX 0,R,s TO DETERMINE THE STATE OF THE INDUSTRY i - RESULTS SUGGEST THAT IN-DEPTH AUDITS SHOULD BE DONE AT
.- ALL PLANTS 1O 7
_ - - - - _ - - i M ? - - . - - _ - - ._
F .j FOLLOW-UP: i INFO NOTICE l CONSIDERING TEMPORARY INSTRUCTIONS 70 REGIONS - PLANNING TO AUDIT l I f f.- 9 i i l'
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O o VISITS REVEALED MAJOR PROBLEMS NOT APPARENT FROM SARs ABCDEF ______________________9 LOWRELIABILITY/ AVAILABILITY 9 9 INADEQUATE V & V O9 9 9 POTENTIALLY HISLEADING 9 9 9 - P00R OPERATOR ACCEPTANCE 9 9 9 s OO O9 ? P SOFTWAREMAINTENANCE MANAGEMENT SUPPORT 9 9 PROBLEM MAJOR PROBLEM , SOME CONCERN e k
PLANT B:. OBSERVED INVALID DISPLAYS COULD MISLEAD AND CONFUSE OPERATORS AND EMERGENCY RESPONSE PERSONNEL i
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O W O RECOMMENDATIONS IMMEDIATE ACTIONS:
- Information Hotice
- Regions inspect for major problems
- Heno to all division directors forwarding audit reports to have PR send to plants for appropriate action
?.g SHORT TERM ACTIONS: w - Consolidate findings into IIREG LONG TERM ACTIONS:
- Conduct audits of every plant '
- Establish Tech Specs for SPDS availability 2
d IMPLEMFFTATION OF SPDS AT OPERATING PEACTORS O l 4 GEORGE LAPINSKY ROOM SL6
- PHILLTPS BLDG.
WASHINGTON, D. C. 20555 PHONE: . 492-8166 i OTHER KNOWLEDGEABLE CONTACTS: SY WEISS - 492-7100 PICHARD ECKENPODE - 492-7896 l l a i iO i
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TURKEY POINT UNIT 3 - LACK OF DESIGN DOCUMENTAT10N TO SUPPORT VALVE POSITIONS IN COMPONENT
~
COOLING WATER SYSTEM )
~ '
FEBRUARY 24, 1986 (D. MCDONALD, NRR) . l . 1
- , PROBLEM
CCW DISCHARGE VALVES THROTTLED TO 30% SINCE START OF
- PLANT OPERATION (RHR HEAT EXCHANGER)
SAFETY SIGNIFICANCE: CCW SYSTEM CANNOT MEET DESIGN BASIS FLOW WITH SINGLE FAILURE EFFECTS MULTIPLE SAFETY-RELATED SYSTEMS i- CIRCUMSTANCES: l - UNIT 3 AT 100% POWER AND UNIT 4 IN REFUELING ERROR IDENTIFIED DURING LICENSEE SAFETY SYSTEM REVIEW ) PHASE 1 (JANUARY 31, 1986)
- UNIT 4 TESTS PERFORMED, 1 CCW PUMP (MARCH 3 & 4) RHR DISCHARGEVALVES30%OPENAND100%OPEN DECLARED UNIT 3 INOPERABLE-AND COMMENCED SHUTDOWN e t FOLLOWUP:
INFORMED NRC 0F CCW FLOW BALANCE TESTS R-11 AND NRR PERSONNEL ON-SITE TO ASSESS TEST PROCEDURES AND RESULTS PRIOR TO STARTUP WESTINGHOUSE CONFIRMS 2500 GPM MINIMUM REQUIRED FLOW RHR
; HEAT EXCHANGERS
! - RECONSTITUTE CCW DESIGN BASIS, (SAFETY SYSTEM REVIEW PHASE 2)
GENERIC ASPECTS: 0THER WESTINGHOUSE PLANTS (1972 CONCERNS AND CORRECTIONS) O
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} FPL SAFETY SYSTEM REVIEW PHASE 2 SELECTED SYSTEMS: ~
INSTRUMENT AIR - EMERGENCY POWER INTAKE COOLING WATER - EMERGENCY FILTERS COMPONENT COOLING WATER - CONTAINMENT ISOLATION SAFETY INJECTION - VITAL AC/DC CONTAINMENT SPRAY - REACTOR PROTECTION SYSTEM
- EMERGENCY CONTAINMENT COOLERS - AUXILIARY FEED WATER
- MAIN STEAM ISOLATION PROCESS:
REVIEW SYSTEM BOUNDARIES KEY SYSTEM DOCUMENT APPLICABILITY EVALUATION REVIEW LICENSING CORRESPONDENCE AND COMMITMENTS REVIEW DESIGN AND ACCIDENT ANALYSES ESTABLISH SYSTEM PERFORMANCE CRITERIA USING DESIGN BASIS, ANALYSES, AND COMMITMENTS O - EVALUATION SYSTEMS TESTING VERIFY CONSISTENCY BETWEEN SYSTEM DOCUMENTS AND THE DESIGN BASIS: DRAWINGS PROCEDURES TECHNICAL SPECIFICATIONS 0-LIST VENDOR DOCUMENTS RESOLVE INCONSISTENCIES AND MODIFY SYSTEM AS REQUIRED PRODUCT: RECONSTITUTED DESIGN BASIS CONSISTENT WITH LICENSING COMMITMENTS AND ANALYSES CONSISTENCY BETWEEN DESIGN BASIS AND AS-BUILT DRAWINGS l - VERIFICATION OF SYSTEM PERFORMANCE
'NOT INCLUDED IN PHASE 1 EFFORT
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FPL SUPPORT EFFORT f'] U SAFETY SYSTEM REVIEW PHASE 2 SYSTEM MANAGERS: 14 ENGINEERS (ADDED) . TECH STAFF: SYSTEM ENGINEERS (UNDER REVIEW) PERFORMANCE ENGINEERS (UNDER REVIEW) FPL SUPPORT - SITE: 6 ENGINEERS (UNDER REVIEW)
- 2 LICENSING ENGINEERS (ADDED)
FPL SUPPORT - CORPORATE: 2 CIVIL ENGINEERS (ADDED) 3 ELECTRICAL ENGINEERS (ADDED) 5 MECHANICAL ENGINEERS (ADDED) O'
- 3 igg ENGINEERS (ADDED) 3 LICENSING ENGINEERS (ADDED) l 4 CLERICAL (ADDED) 1 RELIABILITY ENGINEER (ADDED) 9 BECHTEL - SITE:
50 ENGINEERS (ADDED) SAFETY ENGINEERING GROUP - SITE: ADDITIONAL ENGINEERS (UNDER REVIEW) l
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS ON 4
' s l DOE ADVANCED REACTOR PROGRAMS l
l 4
; APRIL 10, 1986 i
) FRANCIS X GAVIGAN i DIRECTOR OF ADVANCED REACTOR PROGRAMS i 4 1 l O O O
ADVANCED REACTOR PROGRAM o OBJECTIVE DEVELOP A LOW COST, PASSIVELY SAFE, ELECTRICAL GENERATING OPTION COMPETITIVE WITH CONTEMPORARY ALTERNATIVES, N o SCOPE , k LMR'S AND llTGR i 1 O O O
. PROPOSED PLANT CONFIGURATIONS, (RATINGS ARE APPR0XIMATE)
SAFR PRISM HTGR PLANT ELECTRICAL OUTPUT (MWE) 1,400 1,250 560 TG UNITS PER PLANT 4 3 2 REACTORS PER PLANT 4 9 4 ' 3 REACTORS PER TG UNIT 1 3 2 CONTROL ROOMS PER PLANT 1 1 1 REACTOR THERMAL RATING (MWT) 925 430 350 TG ELECTRICAL RATING (MWE) 350 420 280 EQUIVALENT ELECTRICAL RATING 350 140 140 PER REACTOR (MWE) O O - O
UTILITY PROBLEMS / ISSUES o PUBLIC ACCEPTANCE , o LICENSING PROCESS AND UNCERTAINTY o Sil0RTER AND MORE PREDICTABLE. CONSTRUCTION TIME h D o COST AND FINANCIAL RISK UNCERTAINTIES > I o IMPROVED PLANT CAPACITY FACTORS ,
'o FINANCING / INSTITUTIONAL ALTERNATIVES o~ SHARED OWNERSHIP RISKS O O O
ADVANCED REACTOR CHARACTERISTICS o SMALL/ MODULAR o PASSIVELY SAFE o b CERTIFIED AND STANDARDIZED PLANT CL o SHOP FABRICATION k
.O PRE-ASSEMBLED - BARGE OR RAIL TRANSPORTABLE o COST COMPETITIVE o SHORT CONSTRUCTION TIME i
0 HIGH PLANT CAPACITY FACTORS l i i
~
O O O
NRC ADVANCED REACTOR POLICY STATEMENT REQUIREMENTS o RELIABLE, SIMPLE HEAT REMOVAL o EASY MAINTAINABILITY SYSTEMS o '\ FEWER SUPPLEMENTAL SAFETY o REDUCED PERSONNEL EXPOSURE (p FEATURES
- d o LONGER TIME CONSTANTS o MULTIPLE BARRIERS o SIMPLIFIED SAFETY SYSTEMS o INCREASED STANDARDIZATION AND FABRICATION o UTILIZE INilERENCY, RELIABILITY, o EXPERIMENTALLY VERIFIABLE REDUNDANCY, DIVERSITY, SAFETY FEATURES INDEPENDENCE o RELIABLE B0P EQUIPMENT O O O
+_____________________+
l SAFETY APPROACllES l
+_____________________+
+_________________+ +___________________+ +-_____________________+
l ACTIVE DEVICES l l PASSIVE DEVICES l l INilERENT PROCESSES l
+_________________+ +___________________+ +______________________+
- PUMPS - CONTROI. ROD EXPANDER - DOPPLER COEPPICIENT f\
- VALVES - GAS EXPANSION MODUI.E - GRAVITY Is. ' '
- DIESEL GENERATORS - IlYDROMETER - IIEAT CONDUCTION,
}b CONVECTION, AND
- BLOWERS AND COOLERS - GERMER LATCll RADIATION'
- CONTROL ROD DRIVES - CURIE MAGNET - TIIERMAL EXPANSION
- SWITCIIGEAR - OUT MOTION LIMITERS .- BOILING
- BUOYANCY OP llOT PLilIDS
- CURIE POINT
- IIEAT CAPACITY
- DIFFUSION AND DEPOSITION
- IlEAT TRANSFER
- CROSS SECTIONS O O O
DOE ADVANCED REACTORS ASSURE PUBLIC SAFETY l l 0 WITHOUT OPERATOR ACTION i ! o WITHOUT OPERATION OF ACTIVE, POWERED SAFETY SYSTEMS l l %
; (F I .
l l I I l O O O
ADVANCED REACTORS ACCOMPLISilMENTS TO DATE o LMR's/HTGR - MIDWAY THROUGH CONCEPTUAL DESIGN o NRC INTERACTION SCHEDULES ESTABLISHED AND UNDERWAY - POSITIVE EXPERIENCE o COSTS - PRELIMINARY ESTIMATES COMPETITIVE WITH LWR's AND C0AL o SAFETY - ENHANCED PASSIVE CORE SHUTDOWN AND HEAT REMOVAL PROCESSES INCLUDED O R&D NEEDS. EVOLVE FROM CONCEPTS - FOCUSES THE SUPPORTING R&D PROGRAM o UTILITY INTERACTION - ACTIVELY PURSUED - POSITIVE TO DATE O O O
- 5 l @
a AGENDA i t; E l-o ti i 53 j: de er - l WB 28m i- . - .
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i. N V l 5 1 O m MN SSMOSSTISW Z T Rocketdyne Divistori l-1 ~ i ? *
- SAFR PROGRAM OBJECTIVES .
l ... j \ -,; -l i i: ) 1 l
- ASSESS ELECTRIC INDUSTRY MARKET MARKET NEEDS l2 -
- DEVELOP INHERENTLY SAFR LMR DESIGN TO MEET !
lK l 's MARKET NEEDS 1: !k
- ESTABLISH BASIS / APPROACH FOR COMMERCIAL INTRODUCTION j i
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- RockwellantemsAlonel N f- -
Rocketdyne Division 15-i
SAFR PROGRAM TEAM
' Rockwell jg intemational noo. wye. ce.w.e seasessisessh ' .. es . .
e SYSTEM ENGINEERING / INTEGRATION r e REACTOR SYSTEM - l 9 REACTOR VESSEL AND INTERNALS e BUILDING AND STRUCTURES L e HEAT TRANSPORT SYSTEMS -
# HEAT EXCHANGERS
- 8 POWER CONVERSION SYSTEMS e SHUTDOWN HEAT REMOVAL i 8 CONTROL ROOM DESIGN SYSTEM e AUXILIARY SYSTEMS b b
# LICENSING SUPPORT e REFUELING SYSTEMS -
e LICENSING SUPPORT \ e MARKET / COMMERCIALIZATION e SAFETY AND LICENSING e MARKET / COMMERCIALIZATION e SUPPORT SUPPORT }% e MARKET / COMMERCIALIZATION s ASSESSMENT , O ____________________ ANL ETEC HEDL LANL ORNL Al WAESD l e CORE ANALYSIS
- PREHEAT DESsGN
- FUEL CYCLE eCORE e SHIELDING eSTEAM e PlPING l ENERGETICS ANALYSES ERATOR SUPPORT
- TRANSIENT e CONTROLS! e CORE DATA ANALYSES SUPPORT o PHYSICS e MATERIALS / STRESS e SHRS THERMAL / l e PLANT EXPERIENCE e FUEL CYCLE e NaTECHNOLOGY
- RIS.( ASSESSMENT SUPPORT e INSTRUMENTATION e SHRS ANALYSES
- CORE DESIGN
- Na FIRE /NDHX DATA DESIGN SUPPORT l e P: PING ANAL.YSIS
- SAFETY PRA SUPPORT e STRUCTURES e NA AEROSOL ANALYSES e SASSSUPPORT l e NONNUCLEAR SUPPORT ANALYSIS e THERMAL / I&C l
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=4^croas e SWRPS ANALYSIS Rockwellintemational Rocketdyne Division sessessisess)esessessenses 85-011-77-51 A
l, SAFR PROGRAM LOGIC i t , , INSTITUTIONAL ADVISORY l COMMITTEE /C k 2' i '--T r-- ./ ' ) M ,~ / D (' N ' 9 DESIGN EVALUATION / PLANNING FOCUSED ON USER NEEDS ,.
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\ ) p e DESIGN MEETING USER NEEDS I 'd ' '
l SAFR / e LICENSING PLAN ~ POIN / \, . _ m NRC LICENSING DESIGN .? " POSITION .
' " '. COMMITMENT e ORGANIZATION /
MANAGEMENT PLAN
- BASE UTILITY NRC
- PROGRAM INTERACTIONS DIALOG e FINANCIAL PLAN R&D e SUPPLY PLAN i
g T e POTENTIAL 3 8 CUSTOMERS ! S ! 3
$ o hNb Rockwelllntemational hS$MS e
Rocketdyne DMsion cowousteoes amoneenwea.sec. b e j
SAFR GOALS - e ASSURED PUBLIC PROTECTION
- ESTABLISH SAFE CONDITIONS WITH NATURAL FORCES e AVOID NEED FOR FAST OPERATOR ACTION e OBVIATE NEED FOR EVACUATION i' s '
- ASSURED INVESTMENT PROTECTION
{ e LOW PLANT DAMAGE PROBABILITY e EXCEED ASME-SC-C <100/ YEAR l e LOW ACCIDENT-CAUSED # EXCEED ASME-SC-B <10 / YEAR i DOWNTIME 4 e MINIMUM DEVELOPMENT RISK e USE EXISTING TECHNOLOGY
- HIGH CAPACITY FACTOR e >80%
- LONG PLANT LIFE o >60 YEARS g
- LOW PERSONNEL RADIATION e <25 MAN-R/ YEAR
, 12 EXPOSURE
[ kO l Flockwelllntemational Rocketdyne Division seasmesvoonfessoussenuse 15-95-0 l . 0 e - e
1 SAFR GOALS i
- COMPETITIVE ENERGY COSTS e <50 mils /kWh i e SHORT ONSITE CONSTRUCTION * <4 YEARS
! SCHEDULE e WIDELY USABLE STANDARD PLANT * >75% OF U.S. SITES
- e LIMITED FINANCIAL RISK e MINIMlZE TIME TO INCLUSION OF
!'i INVESTMENT IN RATE BASE
!- e MINIMlZE UNCERTAINTY IN PREDIC- ..
- , TlON OF CAPACITY REQUIREMENTS /
j,] PLANT COST, SCHEDULE Jg e CLOSELY MATCH CAPACITY ADDI-l TIONS TO GROWTH DEMANDS , e MINIMlZE COMPLEXITY OF MANAGING e SIMPLIFY DESIGN i PROJECT e MINIMlZE ONSiTE LICENSE-RELATED i WORK
; e EFFECTIVE SECURITY AND e CONSIDER IN DESIGN FROM i
SAFEGUARDS MEASURES OUTSET , RockweNbtemational Rocketdynt DMsion m) m m 85-AU6-43-133B i e G - -- G -
SAFR KEY FEATURES REQUIRING DEMONSTRATION FIRST PLANT NON- (LOW - NUCLEAR POWER ' ISSUE TESTS TREAT EBR-il FFTF TESTS) DRACS PERFORMANCE 1986 1985 COMPLETED 1995 RACS PERFORMANCE 1985/1987 1986 1995 SELF-ACTUATED SHUTDOWN 1976/1990 1987 1995 SYSTEM (SASS) k
\
i RBCB PERFORMANCE 1983/ <*' 1990 INITIAL STARTUP AND 1995 CHARACTERlZATION , N ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS) j EVENTS 3 LOSS OF FLOW WITHOUT 1989 1985/ 1990 1995 SCRAM 1986 } 4 LOSS OF HEAT SINK 1986 1995 WITHOUT SCRAM , h TRANSIENT OVERPOWER 1989 1995 1 g WITHOUT SCRAM Na b Rockwellintemational Rocketdyne Divisim seaunesteess)esseessasauss 7741 1
l l SAFR FIRST PLANT SCHEDULE l 85 86 87 88 89 90 91 92 93 '94 95 96 ACRS i LETTER
! l SER/ STD SER- SSER STD 1
LICENSABILITY FES RULEMAKING/
,2 PLANT CP/ PLANT LETTER PDA DRAFT. OL FDA CODIFIED ',
ES FDA r [{ , , , , , ,, ,, , , , , , , s '"^ " CONCEPTUAL DESIGN DES N DETAll DESIGN CONSTRUCTION T TING PS D S FETY ST ANT TEST PLANT I FSAR ! PLAN PSAH SITE SPECIFIC FSAR ' AND ER - 2 1 l . i hh Madg;*=' ======)====== l W - - - - -- G 85-09-151-28A G
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- e l Q l 5 INLET TEMP.
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LOHSWS - NORMALIZED POWER AND FLOW i i i 1.2 l l l , i l 1 .
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f 1500 - l ' E 15 y - g h
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5 POWER ! m o i I I i I 8 0 200 400 000 soo 1000 d TIME (s) ! b Rockwellintemational Rocketdyne Division commesison%mesusemens i
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'l l SODIUM BOILING POINT -
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900 20 - N POWER . g _ g FLOW D- -- n Y 0 1 I I I I '7-~C 700 3 0 1 2 3 4 5 6 7 8 TIME (min) g RocketdyneDivision. - g _ e= G 9 9
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B/A FUEL STORAGE ! 11 MINIMUM 16.8 3.1 COOLING STACK MAXIMUM 25.4 12.4
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6 l1. TOTAL I
- 551 150 rm A-FRAME l '
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EXCHANGES- " I 1 l k, Q W PEAK ASSEMBLY TEMPERATURES (*F) k F/A STUCK IN REACTOR PORT 463 - 7; I F/A STUCK IN FTC 319 ; '
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l F/A IN INTERIM STORAGE 450 ' i i i '1 l
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FUEL STORAGE g COOWG AIR W h IVS TO A-FRAME CORE TO IVS EXCHANGES EXCHANGES 1 Rockwellintemebonel ' Rocketdyne Diviolon I i
CAVITY COOLING DURING RACS OPERATION COLLECTOR (EXTENDED l SURFACE) ' FLOWING AIR , 4 *, *
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CAVITY COOLING YV i . l 1. 5.~. - AIR FLOW (Ib/s) 26 . T. E'-}
- AIR INLET TEMPERATURE (*F) 100 'l '
REACTOR W *. ~ . AIR OUTLET TEMPERATURE (*F) 259 - VESSEL 9; , , CONCRETE TEMPERATURE (*F) <150 .
~~
is 'T i CONCRETEi- GUARD VESSEldCOLLECTOR ANNULUS 3 l{ .
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- 2.j AIR FLOW (Ibm /s) ,
75 ) ' s - ',, ,.),'f,. AIR OUTLET TEMPERATURE (*F) 237 GUARD VESSEL TEMPERATURE (*F) 818 A> SODIUM - -
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REACTOR VESSEL
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* : I HOT POOL TEMPERATURE (*F) 1191 I ..N.
! COLD POOL TEMPERATURE (*F) 1,097
*.a s d CORE OUTLET TEMPERATURE ("F) 1210 -
=<.6 TIME TO REACH MAX TEMPERATURE (h) 45 j GUARD VESSEL e .- HEAT REMOVAL RATE (MWt) 3.74 I
[ POWER LOSS DURING NORMAL 1 HEllUM. ..k'i OPERATION (MWt) , 1.6 3 j *W;/ , .. i INSULATION '4 1
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- e e 85-015-151-1 4 :
- DECAY HEAT ACCOMMODATED WITH MARGIN l
l I I I I I I I g 1600 - a BOlWG ec ?
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' MAJOR PLANT l
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) g 1400 - [ COOLING -- f t 3 - SHUTDOWN FOR 5 1200
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3 NORMAL 1 8% I i l I i i 1 0 10 20 30 40 TIME AFTER REACTOR SHUTDOWN (h) E; Y 4 4 7 g - (D
- Rockwellintemational j Rocketdyne Division sensassison)seesseessames 9 O~ G --
I l s SODIUM-WATER REACTION PRODUCTS " l SYSTEM (SWRPS) l l l STEAM i 2 6 A IGNITEF t l NaIN l' QI i STACK STEAM I '- GENERATOR 1 41 e U RUPTURE jL Na OUT DISC e3 !.:
'M dhh CENTRIFUGAL SEPARATOR rw =
l}H)l ! RUPTURE DISC jf ' f ., j . l ( ) SWRPS/ DRAIN TANK j! i, Rockwelllntemettonal 1 Rocketdyne Division seannesmos) - 85-JY22-92-315
PEAK IHTS PRESSURE DURING SWR CAN BE ACCOMMODATED 350 300 - l 250 - i Q.
- g200 -
. 4
, 8 N tu
$ 150 -
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- 1 GUILLOTINE TUBE RUPTURE
- 50 - -
FAULTED PRESSURE PULSE l ALLOWABLE ~ 855 psi 0 6 j 1 -50 I I I I I I I I l T 0 50 100 150 200 250 300 350 400 450 500 h TIME (ms) - Rockwellintemational j Rocketdyne Division seanousness)esessessannes I O O O
Oc ROOF EXHAUST FANS FOR A T NAL 'E TING l PANELS AS REQUIRED. l E' IHTS I PUMPI l MAIN STEAM LINES i lL.O. "EXP f DO NOT ENTER SGB
;p TK ,
n CONCRETE s
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. FLOOR / - -
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- INSULATION l l VAULT ,
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, to VAULT. (OPTIONAL l ! A AGE TO SWRPS l ,,,,,,, g ,j , N Q waucm y-- -
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e e e , Rocketdyne Division t
SODIUM FIRE ACCOMMODATED 220 , i GAS 200 - , l 180 - ~ CEILING Q I L UPPER WALL I E N i h 160 - D i I \ ! e , l } 3 b LOWER WALL - k
- 140 ~
1 I 120 - 1 j i n i y 100 t I j % 0 1 2 3 m TIME (h) ! Hockwellintemational Rocketdyne Division seanovsmess)essisessenause O O O .
i PASSIVE, LOCALIZED SAFETY APPROACH ASSURED DECAY HEAT REMOVAL
'i G NATURAL CONVECTION DIRECT REACTOR AUXILIARY AIR A 0 TU AL CONVECT O REACTOR AIR COOLING UTDOWN SYSTEM (RACS) 1 ****? ' S REDUNDANT, DIVERSE I C REACTOR SHUTDOWN hA '
r S INHERENT, SELF-ACTUATED SHUTDOWN I f f5 C>cq { __ l 2 g e INHERENT, SELF-LIMITING FEEDBACKS j
.. /, - - -
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W.), - i Is USER FRIENDLY hm*# M - ! G ALL RADIOACTIVE ~** I "
- j lg#
MATERIAL IN SIMPLE i l ENCLOSURE 9 THERMALINERTIA p. g [ '- @% : ALLOWS LONG PERIODS *: a - '
' FOR OPERATOR ACTION-- ['
I F - SODIUM FIRE RISK AVOIDED ,
*; 8 G INERT ATMOSPHERE l * *[.
O GUARD VESSEL / PIPES M [* 4: . N.' *;*' [ l'V#$ 9 IN STEAM GENERATOR e BUILDING - TESTED FIRE T SUPPRESSION FEATURES m g RockweN intemational gg
@g g
1 SAFETY GOALS SATISFIED WITH MARGIN l n ASSESSED RISKS LESS THAN LWR ESTIMATES 10 - LATENT RISK GOAL (0.1% OF ALL CANCER FATALITIES) ' iiT 10-5-- p PROMPT RISK GOAL
=!
2 / (0.1% OF ALL , ACCIDENT FATALITIES) 104~ Y////////////////////////////////E////// ///A { h //S//////A
! I [O 10-7. -
O O l $O O = WASH-1400 RESULTS M E 10-s- - O O = OTHER LWR RISK l Q ASSESSMENT RESULTS ,
$ $ = 1-UNIT SAFR PLANT y,
) $ 10 - O g = 4. UNIT SAFR PLANT
! 3 h
I 10 - i ' t I O E 10-11-- 1 Q -l 10-12 l l l l l h 10-11 10-10 10-8 10-8 10-7 10-6 10-5 104 i $ LATENT FATALITY RISK (PER YEAR, WITHIN 50 MILES) ! o d RockwellIntemstional l Rocketdyne Division seanousveess)sessesessames l O - O . _ . . _ - - . _ - - . O-__ _ _.______ - -_.
i . SAFR CONFORMS WITH NRC ADVANCED REACTOR POLICY STATEMENT . NRC PROPOSED POLICY SAFR FEATURES ' 1. HIGHLY RELIABLE, LESS COMPLEX SHUTDOWN, # TEMPERATURE-TRIGGERED MAGNETIC" LATCH" AND DECAY HEAT REMOVAL SYSTEMS; CONSIDER- SHUTDOWN I PASSIVE SYSTEMS # TWO NATURAL CIRCULATION DECAY HEAT ' REMOVAL SYSTEMS
# SELF-LIMITING RESPONSE TO ATWS EVENTS
- 2. MINIMAL NEED FOR SUPPLEME TAL SAFETY 9 1400 MWe SAFR 1E POWER == 2% OF 350 MWe ORBRP FEATURES TO ENSURE SAFETY # SAFR ISOLATION VALVES < 50% OF TYPICAL : WR
,s # ONLY 6 REACTOR SCRAM SIGNALS
~
- 3. LONGER TIME CONSTANTS AND SUFFICIENT G LARGE POOL HEAT CAPACITY PROVIDES 30 h TO t.
INSTRUMENTATION FOR DIAGNOSIS AND SERVICE CONDITION D/42 h TO BOILING ! MANAGEMENT PRIOR TO SAFETY SYSTEM e FAULT TOLERANT DIAGNOSTIC SYSTEM 1 Kw CHALLENGE OR EXPOSURE TO ADVERSE , j CONDITIONS
- 4. SIMPLIFIED SAFETY SYSTEMS WITH FEW G PASSIVE INHERENT HEAT REMOVAL - CORE AND OPERATOR ACTIONS. FEW COMPONENTS STORED FUEL DECAY HEAT, REACTOR COVER, AND ;
i FACILITATES OPERATOR COMPREHENSION AND CAVITY RELIABLE SYSTEM FUNCTION O AUTOMATIC AND INHERENT REACTOR SHUTDOWN l 9 INHERENTLY SAFE RESPONSETO ALL CREDIBLE + 1 EVENTS ; 5 . MINIMlZE POTENTIAL FOR SEVERE ACCIDENTS S INHERENT AND DIVERSE SHUTDOWN
; $ AND THEIR CONSEQUENCES BY INHERENCY, e INHERENT AND DIVERSE DECAY HEAT REMOVAL i j g RELIABILITY, REDUNDANCY, DIVERSITY, AND e INSENSITIVE TO BOP TRANSIENTS T INDEPENDENCE IN SAFETY SYSTEMS S ACCOMMODATION OF ATWS EVENTS
{ # HETEROGENEOUS CORE 'l Rockwellintemational -
%ssamassasssa ocketdyne Division 15-67-0
SAFR CONFORMS WITH NRC ADVANCED REACTOR POLICY STATEMENT. NRC PROPOSED POLICY SAFR FEATURES
- 6. PRO'.*lDE RELIABLE EQUIPMENT IN BOP REDUCING G PROVEN ONCE-THROUGH STEAM, GENERATOR CHALLENGES TO SAFETY SYSTEMS 9 REllABLE FEEDWATER SYSTEM l
9 RELIABLE REDUNDANT CONTROL SYSTEM AVOIDS j REACTOR SCRAMS FOR BOP EVENTS I 7. EASILY MAINTAINABLE EQUIPMENT AND l 9 PROVISIONS FOR MAINTENANCE AND INSPECTION COMPONENTS PROVIDED IN DES!GN l 8. REDUCED POTENTIAL RADIATION EXPOSURES TO 9 O/l/M EXPOSURE <25 MAN-REM / YEAR PLANT PERSONNEL g
- 9. PROVIDE IN-D5PTH DEFENSE WITH MULTIPLE 9 HIGH QUALITY FUEL CLADDING 9\
BARRIERS TO RADIATION RELEASE AND ASSESS 9 LOW PRESSURE PRIMARY SYSTEM BOUNDARY
- POTENTIAL FOR MITIGATION OF SEVERE WITHOUT VESSEL PENETRATIONS ACCIDENTS $
9 LOW LEAKAGE CONTAINMENT ) O PRA INCLUDES FULL SPECTRUM OF EVENTS T N i
- 10. INCREASED STANDARDlZATION AND SHOP O SHOP-FABRICATED REACTOR ASSEMBLY FABRICATION TO MINIMlZE POTENTIAL FOR FIELD # PRIMARY SYSTEM COMPONENTS FIT UP IN SHOP j CONSTRUCTION ERRORS. NO NEW DIFFICULTIES j IN TRANSPORT, INSTALLATION, AND 9 BARGE-SHIPPABLE REACTOR ASSEMBLY
! MAINTENANCE # SHOP FABRICATED NSSS CELLS
- 11. DESIGN FEATURES THAT CAN BE PROVEN BY # MOST FEATURES PROVEN, e.g., POOL CONCEPT, CITATION OF EXISTING TECHNOLOGY OR BY DRACS, STEAM GENERATOR COMMITMENT TO SUITABLE DEVELOPMENT 9 SAFETY TEST PLAN COVERS REMAINDER, e.g.,
PROGRAM RACS, SASS, ATWS, RBCB 5 k Rockwellintemational nacketoyn.oivisen sessousness)sessassammes 2-65-2 85-O11-151-95 O O O
NRC INTERACTIONS ARE OF MAJOR VALUE ' TO ADVANCED REACTOR EFFORT .
- HIGH PAYOFF AT LOW COST :
- PRE-PSID MEETINGS ARE AIDING EARLY RESOLUTION OF s ISSUES i
' i j
- SAFR IS IN COMFORMANCEWITH DRAFT NRC POLICY
! y STATEMENT ON ADVANCED REACTORS -,
'\
- UTILITIES ENDORSE PSID/SER/LICENSABILITY LETTER 1 APPROACH
- EARLY LICENSABILITY ASSESSMENT BY NRC IS A i
PREREQUISITE TO ANY LMR TRANSITION TO MAJOR PRIVATE SECTOR SUPPORT ! k . noc sn )
$g.kw.ukem g ,sg
s l WE STRONGLY RECOMMEND CONTINUING REVIEW l OF ADVANCED REACTOR DESIGNS AT PREVIOUS r NRC STAFF LEVELS AND ISSUANCE OF A Q' l POSITION STATEMENT AT END OF "' ! CONCEPTUAL DESIGN l 1 l . 8 S l @ ==. =- @ -->--
NEm3 a := lE 4 EFING TO ADVISORY COMMITTEE s R ON REACTOR SAFEGUARDS g APRIL 10,1986 m e---" - - - - - _s _ _ _ . , .,,. __
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l4 # \< p INITIATE SAFETY TEST (00E, NRC.IN0U:TRY, UTILITY) 1 T ! N \ e CONCEPTUAL DESIGN,36 MONTHS j j 3d V e OBJECTIVES:
- CHARACTERISTICS RESPONDING '
@ / TO MARKET NEEOS g - COST COMPETITIVE g, - ENHANCE 0 SAFETY I
~
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-HIGH AVAILABILITY I p -COMMERCIAL ACCEPTABILITY 00E INITIATES COMPETITIVE PROGRAM .' )
! GE INITIATES INNOVATIVE DESIGN PROGRAM ! e OBSTACLES TO NUCLEAR EVIDENT s e SOLUTION -NEW APPfl0ACH -SAFETY TEST
- l I 9 i
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' l l INLET PLENUM h l' '
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j j lTl' h . i INTERMEDIATE yllldl { RVACS COLLECTOR CYLINDER HEAT EXCHANGER (2) W I i of/ N '=k 6- : 4 . REACTOR CORE
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REACTOR VESSEL MI)- W s 83 I . d
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s i APPENDIX XI HUMAN FACTORS SUBCOMMITTEE REPORT L ) C Meeting No. Agenda Item 12 312 Title HUMAN FACTORS (HF) Authors David A. Ward, Chairman, ACRS Human Factors Subcommittee - attachment 1 K. L. Gimmy, ACRS Consultant on Human Factors - attachment 2 List of Documents Attached num u u u m e m u** m **
- 1. David A. Ward, Chmn. muen mmemummuu HF Subcommittee memo of unmu **u m m m**uun I I 4/8/86 Subj: Report of men umumunuuun l
Meeting of March 19-20, 1986 , , , , ,,,,
- 2. K.L.Gimmy,kCRSConsultant ************* ********
memo of Mar. 24, 1986, * * * " * * * " * * * * **** Commentson HF Subcommittee ******** Meetings of March 19 and 20 *** ** ***** ****
***+**** ******
******** ***t*** ******
l nstructions to Preparer From Staff Person
- 1. Punch holes
- 2. Paginate attachments J. O. SCHIFFGENS
- 3. Place copy in file box AY '-
. 1
/-
April 8,1986 TO: ' '- ACRS MEMBERS FROM: D. A. WARD, CHAIRMAN, SUBCOMMITTEE ON HUMAN FACTORS W j l l REPORT OF MEETING OF MARCH 19-20,1986 Attending the meeting were ACRS Members Moeller, Michaelson, Wylie and myself and ACRS Consultant Gimmy. l This two-day meeting had three purposes. The subject the first full day was spent was an information session on one called the state-of-the-art in applying automatic monitoring and control functions in the operation of nuclear power plants. We heard from representatives from Staff and several representatives from the i industry. The second topic was a review of the NRC's Human Factors program plan by the Staff. l I'll summarize what we learned on the first and third topics and then a representative of the Staff will review the present Human Factors Progress Plan for you. At the present time, I do not see a necessity for an ACRS report on any of these topics at this month's meeting. The third topic concerned the implementation of emergency operating procedures in actual plants. On the first day of the beeting We heard 11 presentations on the state-of-the-art in automating monitoring and control functions. In an introductory presentation from the Staffit was stated that the early attitude of both the industry and the NRC toward computers was largely influenced by experience with big main frame computers and with the unreliability of this type of computer in a process setting. However, in recent years, both the NRC and the industry have become aware that computers specifically designed.for monitoring and controlling processes can be , considerably more reliable than the earlier experiences indicated, and in fact, may O . . l 4 1s #T7~ / a t m if. 3 ,
/
ACRS MEMBERS 2 April 8,1986 O be extremely useful tools for aiding operators in managing both routine and emergency operations in nuclear power plants. It is interesting that the SPDS (Safety Parameter and Display System) which has been mandated as an addition to each plant as a post-TMI backfit seems to have served as sort of an introduction for ' the nuclear power industry to the use of modern process computers. enere was not really a specific requirement by the NRC that SPDS should be computer based but the characteristics desired for the SPDS and the availability of process computers that could have the appropriate characteristics has led really every licensee to use some form of digital computer as the heart of his SPDS system. 4
< However, despite the introduction to the computer age offered by the SPDS, the state of applications of computers in nuclear power plant control is not very advanced. We heard a report from Staff people who had meetings with NASA at the Kennedy Space Flight Center. Originally NASA made very little use of process computers in their ground control systems (as opposed to on-board systems). It has gradually, however, through the years, made more and more use until now the ground control systems at Kennedy Space Center are largely, what you might call, i automated and make heavy use of process computers with, they believe, i considerable success. They also reported that in NASA's design of a future space station, they have set aside 10% of the total development budget for the development of software and related automatic control systems.
/ We also heard a report from Leo Beltracchi of the Staff who sometime ago visited the Savannah River Plant and was briefed there by our Consultant, Mr. Gimmy, on l
, the application of process computers in the monitoring and control of the Savannah River Reactors. This approach was initiated about 18 years ago and has gone through several generations ofimprovements. First there was only monitoring, then some close-loop control of the process was added. He next step was adding scram system safety circuits based on computers, a backup scram system, and most recently, addition of an alarm diagnosis system which is a form of an " expert system". Application of computers in today's nuclear power plant control is about at the stage where the Savannah River Plant reactors were in about 1970.
/ We also heard a report on methodology that was developed for so-called function allocation. This is a systematic approach for deciding which monitoring and control functions will be assigned to the human operator and which will be assigned to computers or other automatic control systems.
LO l w +
ACRS MEMBERS 3 April 8,1986 - O We also heard a report from Professor Barclay Jones from the University of Illinois on research he is doing in what he calls the process of"embeded training". By err.beded training, he means combining the training functions for a nuclear power plant operator with his normal operational functions by providing them with a control system for the reactor that is part simulator. To most of us, this seemed to , be like a radical, perhaps even dangerous concept, but Professor Jones is looking at it conceptually and he sees this as a possibility for fully automated control rooms of a future generation. We also heard a report from Westinghouse on the approach they are taking on advanced control room design which is going to extensive use of plant computers : and automation of the many process functions. Their approach is to move in very deliberate step.s and their intent is not use for the fulll potential of the technology for near-term designs. However, the French, with similar plants, seem to be taking rather radical steps toward quite complete automation of the control rooms. We also heard a report from Rom Duffey from EPRI who described the work that is going on at EPRI. This program is all directed toward application to the present population of plants, and not for future plant designs. He says that the utilities are ' O. very supportive of the research they are doing; in fact, each element of the research program has what he calls a host utility which is particularly involved in nuclear development work. Finally, we heard a couple of good tutorials really from Mr. Bill Bertch and Mr. i George Niederauer of Energy, Incorporated. They described in practical fashion the application that artificial intelligence and particularly expert systems will have in nuclear power plant operation. Then do not expect artificial intelligence to ever give close-loop automatic control of a process; for example, of the nuclear power plant process. Instead, it would automate 'certain observation, interpretation and decision analysis in nuclear power plant operation, leaving the final decision and actions to the human operator. Next, we heard reports describing Emergency Operating Procedures developed for use at two specific plants. The first was on behalf of the Hatch Plant of the Georgia Power Company. They described and demonstrated a " flow chart" format. This is a very comprehensive and detailed approach to emergency operating procedures which uses large logic l diagrams, or flow chans, to provide the initial response to any major plant upsets. L The purpose of the procedures under the flow chart was described as being to bound 6-/fS b
ACRS MEMBERS 4 April 8,1986 the problem and stop the spread of degradation of the plant equipment or systems or for the process. Each flow chan ultimately leads through a logic path to a preferred recovery procedure in what is called "End Path Manuals". The second type of procedure we heard about was the so-called single column instruction format. These procedures were described to us and demonstrated as applied at the Browns Ferry Plant of the Tennessee Valley Authority. As you have heard before, the NRC has required that each plant develop for itself some sort of emergency operating procedures which are " symptom oriented" rather 4 than " event oriented". The NRCjust provided general guidelines and specifies a - process for development of the procedures. 'Ihe result is that there is a very wide variety and spectrum of styles and types of procedures in use at the many operating plants in the country. The two that we heard described may not be at the extreme ends of the comprehensive versus simplicity spectrum, but they are probably representative of something close to the ends of the spectrum. So far as which type of procedure is superior, I don't know, and I don't think
- anyone in the NRC Staff has taken a panicular position on this. Certainly, each of i
the utilities believes that their system of procedures is adequate. Whether they think s their own system is, in either case, the best approach that can be taken, I don't know. The Staff approach to emergency operating procedures has been to insist that each licensee develop a process for developing its own EDPs and that the NRC staff then would audit and monitor that process with the expectation that if the process was l going well, good emergency operating procedures would somehow appear in the 1 plants. The NRC seems to have reached the decision that this approach has failed. They have monitored the process and often times found indications that there was a satisfactory process, but when they audited actual procedures, they found major , problems with the all-important end product. Apparently, the Staff recognizes the l need for more extensive auditing program of the actual final product, the emergency operating procedures. o O l r9 HA l
TECHNICAL DIVISION
~
'/ SAVABNAH RIVER LABORATORY CC: J. O. Schiffgens March 24, 1986 TO: D. A. WARD, Chairman ACRS Subcommittee on Human Factors
. FROM: K. L. GIMMY, Consultant on Human Factors COMMENTS ON HF SUBCOMMITTEE MEETINGS OF MARCH 19 AND 20
, A_UTOMATION AND COMPUTER APPLICATIONS The scheduled presentations gave an excellent overview of the state of the art. The NRC staff has a positive view of in-plant computer systems. From my contacts, I think the group under S. Weiss is very much up to date on the strengths and weaknesses of computer-based systems.
I, of course, disagree with M. Paradies' conclusion that computers'should be " precluded" from' diagnosing events. I think this could be a very strong application; particularly as some Emergency Operating Procedures are already evolving toward computer-lixe flowcharts. I thought Mr. Bertsch gave an excellent presentation on Expert Systems. Two of his points were especially relevant: d Expert systems ensure that you examine facts in a logical manner and not overlook some component. Thus, domain experts routinely use their own systems, just to be sure they haven't forgotten something on a particular analysis.
- 2) Expert systems incorporate each new fact as it appears.
They don't take a mindset. - ry. /T U ATT L
l 1 Many people still feel uncomfortable with software-based plant systems. Computers do have problems. But, many of their worrysome traits have a parallel in conventional analog instruments. The difference is that we've learned to live with the analog failures. The following table illustrates my point: SOFTWARE-BASED EQUIVALENT. ANALOG
, SYSTEM PROBLEM SYSTEM PROBLEM
- l. Failure modes unpredictable 1 Recorders can fail upscale, downscale, or stay the same.
Relay-races unpredictable.
- 2. Sof tware fix caused problem 2. Replaced 20 MFD. capacitor elsewhere, with 40 MFD. and changed response time-constant.
- 3. Sof tware problem caused 3. Loss of instrument air, massive data failure. Loss of DC buss.
- 4. Strange behavior possible. 4 Ground-loops, feedback to sensors.
- 5. Poor documentation causes 5. Out-of-date prints cause errors, errors.
j BUMAN FACTORS PROGRAM PLAN ! A disappointing session. The NRC Branch Chief obviously doesn't like this assignment or the way resources have been divided. The program seems to involve much interaction between NRC groups but I was unable to determine who is responsible for the overall schedule. EMERGENCY OPERATING PROCEDURES l The industry presentations were good examples of the extremes one can encounter in EOP's. The presentation by Georgia Power (or rather, their consultant) was impressive. I think the flow-chart procedure will work. I think it also helps establish "who's in charge" during an emergency and even helps him verbalize his orders i properly. Also, we saw the product - not some nebulous plan to make the product (EOP). The TVA presentation was a good example of " standard" procedure technology but with one glaring flaw - the numbering system. While the STA talked-through the procedure effortlessly, a less-prepared I person would have to follow the numbering system. Remember, all f7 nJ
"~
- forward references were by number. And, what numbers! 5.2.4.6, if not applicable go to 7.3.1.7.8. A single typographical error would take you to the wrong place. The question here is not di.d they-follow their " approved plan", but can anyone follow the procedure .l under stress? l I think it's important to audit the-final product. Every major l company has a word-wizard who can turn out a high-sounding plan on l demand, De it a QA plan, EOP plan, or Environmental plan. l Frequently, the author cincerely believes in his plan and can defend it knowledgably. The question is, does anyone else in the company support the plan and is it being followed? Auditing will tell.
I i 1 b s j
~ - APPENDIX XII HUMAN FACTORS PROGRAM PLAN: ,, NRR STAFF PRESENT O ACRS
SUBJECT:
HUMAN FACTORS PROGRAM PLAN: PROGRESS AND PROSPECTS DATE: APRIL 11, 1986 PRESENTER: DANIEL B'. JONES PRESENTER'S TITLE / BRANCH /DIV: PROJECT MANAGER: DHFT/HFIB ,p PRESENTER'S NRC TEL. NO~: . x24879 SUBCOMMITTEE: O (7. / M
9 O STAFFING AND QUALIFICATIONS MINIMUM QUALIFICATIONS REQUIREMENTS LICENSING OF ADDITIONAL OPERATIONS PERSONNEL -- NO ACTION EDUCATIONAL QUALIFICATIONS FOR OPERATIONS PERSONNEL NUREG/CR-4501 NUREG/CR-4411 O ' l < RE6ULATORY GUIDE 1-8 -- PERSONNE,L QUALIFICATIONS AND TRAINING SUPERVISOR (SRO AND STA) EFFECTS ON CREW PERFORMANCE IN , SIMULATOR NUREG/CR-4280 , e
- NO DIFFERENCE IN CREW PERFORMANCE
.r. n /
O STAFFING AND QUALIFICATIONS 1 l LIMITS AND CONDITIONS OF SHIFT WORK l l l
~*
SHIFT SCHEDULING AND DVERTIME (NUREG/CR-4284)
- BASIS TO REVISE NRC POLICY ON OVERTIME
- GUIDANCE ON ROUTINE 12-H00R SHIFT OPERATOR FEEDBACK -- MAILED SURVEY (NUREG/CR-4139) j - EFFECTIVE WAY FOR NRC TO ACQUIRE INFORMATION:
RE: OPERATORS l
- PERSONNEL ATTITUDES: RE: STA SHIFTWORK, OVERTIME, AND
~
STAFFING r < 'O I - ~. - . - .. .-.. . pp / '7 % '
d TRAINING TRAINING REGULATION AND GUIDANCE POLICY STATEMENT ON TRAINING AND QUALIFICATIONS
- INP0 MOA ON ACCREDITATION TEAM SKILLS TRAINING -- NUREG/CR-4258
- INSTRUCTIONAL SKILLS EVALUATION -- NUREG/CR-4344 l
NRC TRAINING EVALUATION PROGRAM TRAINING REVIEW CRITERIA AND PROCEDURES (FOR NRC STAFF) O C , Wp #
,, ..w- .
,- 1 h.
t, l a LICENSING EXAMINATIONS O PROCESS l 5 RULEMAKING: 10 CFR PART 55 AND PART 50 s l; - R. G. 1.8 -- PERSONNEL QUALIFICATIONS
't ~
- R. G. 1.134 -- MEDICAL EVALUATION
" ~
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. - R. G. 1.149 -- SIMULATORS FOR EXAMS '
Il IMPROVING SIMULATOR EXAM PROCEDURES II SIMULATION FACILITY EVALUATION I CONTENT PWR KNOWLEDGE AND ABILITY CATALOG (NUREG-1122) ' EXAMINERS HANDBOOK (NUREG-1121) COMPUTERIZED EXAM QUESTIONS BANK
- ~ . .
9
'O V
/
O PROCEDURES E0P UPGRADE AND POST-!MPLEMENTATION AUDITS
- REVISED SRP 13.5.2, "0PERATING AND MAINTNANCE PROCEDURES" E0P INSPECTION MODULE (FOR ltE)
METHODS FOR EVALUATING ALTERNATIVE TECHNIQUES AND FORMATS 1 . l l l
, A .. /0$ - _ . . . -~,4--..- - ._ - _ , . . _ - _ , _ - , , - _ - _ _ _ _ _ _ . - _ _ - -. _ . . _ _ _ .
MAN-MACHINE INTERFACE (MMI) O
- MMI FOR EXISTING DESIGNS DEFICIENCIES IN LOCAL CONTROL STATIONS (NUREG/CR-3696)
ANNUNCIATORS q COMPUTERIZED ANNUNICATOR SYSTEMS (NUREG/CR-3987) ALARM REDUCTION TECHNIQUES O - ami eoa aovaacto Tecitao'oates . SIMULATOR EVALUATION OF CRT DISPLAYS (NUREG/CR-3767) EXPERIMENTAL ASSESSMENT OF AN EXPERT SYSTEM (NUREG/CR i
- DECISION MAKING UNDER STRESS (NUREG/CR-4040) ,
GUIDELINES FOR EVALUATION OF VDUs (NUREG/CR-4277) i HUMAN PERFORMANCE IN NON-DESTRUCTIVE TESTING (NUREG O nn.= .-
/
MAN-MACHINE INTERFACE (MMI) (CONT'D) l REGULATORY DOCUMENTS SRP 18.0 -- HFE ; SRP 18.1 -- CONTROL ROOM SRP 18.2 -- SPDS s e 4 e e O l
~
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- s. /'? D i
I
3 (J MANAGEMENT AND ORGANIZATION (M40) ESTABLISH REGULATORY POSITION - ors SAFETY INDICATORS (NUREG/CR-3737) s EMPERICAL ANALYSIS OF MAINTENANCE FACTO REVIEW OF INP0 SAFETY EVALUATION MATERIALS ASSESS PROCEDURES FOR OLs GUIDELINES AND WORKBOOK FOR ASSESSING ORGANI ADMINISTRATION FOR OLs (NUREG/CR-4125) ~ e o e f s 9 *
,,m.-.G;9 -:=-l!?L- . . . . --
_ _ . _ _ _ - _ . . . . - - - - - - - - - - - - - - ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ^ " ~ ~ ~ ~ ~ ~ ' ' ' ' ~ " " ~ . _ - - . _ . - -
O PRINCIPAL DIFFERENCES BETWEEN REVISION 1 AND 2 0F THE . HUMAN FACTORS PROGRAM PLAN BUDGET CUTS RES HAS TERMINATED HUMAN FACTORS RESEARCH OTHER THAN THE PRA-RELATED HUMAN RELIABILITY WORK AND THE NRR BUDGET IS REDUCED ()
- CONFORMANCE WITH THE COMMISSION'S PPG AND POLICY STATEMENTS:
i . SHIFT FROM EMPHASIS ON REQUIREMENTS DEVELOPMENT TO AUDITING AND ENCOURAGING INDUSTRY SELF-IMPROVEMENT l REVISION 2 IS AN NRR PLAN ASSIGNED TO DHFT; REVISION 1 COVERED ALL ED0 0FFICES REVISION 2 IS AIMED AT A BROADER AUDIENCE: DETAILED PROGRAM SCHEDULES AND MILESTONES ARE OMITTED O
4 HFPP RESOURCES
~
B. TECHNICAL ASSISTANCE BUDGET . FY 1985 FY 1986 i NRR' RES TOT. NRR RES TOT.
- 1. STAFFING & QUALIFICATIONS 125 115 240 0 -
0
- 2. TRAINING 445 0 445 50 -
50
- 3. OPERATOR LICENSING 695 476 1171 413 -
413
- 4. PROCEDURES 300 560 860 34 -
34
- 5. MAN-MACHINE INTERFACE 0 1032 1032 660 -
660
- 6. MANAGEMENT a ORGANIZATION 250 '115 365 0 -
0 l
- 7. HUMAN RELIABILITY 0 1100 1100 0 -
0
- 7. (NEW) HUMAN PERFORMANCE - - -
0 - 0
- 8. NAS PANEL 43 132 175
! TOTAL 1815 3398 5213 i y ,1 . ,) d' '
I, TRAINING (MAINTENANCE AND TRAINING BRANCH) l BACKGROUND .. POST-TMI RECOMMENDATIONS AND THE NUCLEAR WASTE POLICY ACT OF 1982, SECTION 306, LED STAFF TO PROPOSE A TRAININGRULE(SECY-8k-76,76A,76B) COMMISSION ELECTED TO PROMULGATE POLICY STATEMENT (FR 11147, MARCH 20, 1985) IN LIEU OF RULE TO ALLOW INDUSTRY TWO YEARS TO IMPLEMENT THE INP0 ACCREDITATION PROGRAM CURRENT AND PLANNED ACTIVITIES EVALUATE IMPLEMENTATION OF POLICY STATEMENT POST-ACCREDITATION REVIEWS l PERFORMANCE-BASE INSPECTIONS OBSERVE INP0 TEAM VISITS AND ACCREDITATION BOARD O
,- :,. .w
\
() I. TRAINING (CONT'D) l SALP 4 ARCHIVAL PERFORMANCE DATA INTERIMPROGRESSRbPORT(MARCH 1986) i
- i
! - ACCREDITATION STATUS (MARCH 1, 1986)
.310 0F 610 PROGRAMS AT 54 SITES " READY FOR ACCREDITATION" PROBLEMS IDENTIFIED WITH " ANALYSIS" AND LEARNING OBJECTIVES, ELEMENTS NRC AND INPO DISCUSSING IMPROVEMENTS FINAL EVALUATION REPORT AND RECOMMENDATION FOR i
FUTURE ACTION (MAY 1987) REVISE SRP SECTION 13.2 BASED ON 1986 DECISION i
- O
,, o o
4 III, PROCEDURES (HUMAN FACTORS ISSUES BRANCH) d SCOPE: ALL OPERATING PROCEDURES. MAINTENANCE PROCEDURES COVERED BY M&SPP CURRENT AND PLANNED ACTIVITIES: . EMERGENCY OPERATING PROCEDURES (E0Ps) GUIDANCE FOR NORMAL PROCEDURES (0P) AND ABNORMAL OPERATING PROCEDURES (AOP) , ,. HUMAN FACTORS DEFICIENCIES REGULATORY ANALYSIS TO DETERMINE IF REGULATORY GUIDANCE JUSTIFIED j TRANSITIONS BETWEEN PROCEDURES: OPERATORS HAVE DIFFICULTY IN TRANSFERRING BETWEEN PROCEDURES. DRAFT GUIDANCE ON PROCEDURE INTERFACES BEING DEVELOPED, VALUE-IMPACT (V/I) ANALYSIS OF E0P UPGRADE: RES COMPLETING RETROSPECTIVE V/I STUDY OF THE CHANGE FROM EVENT-BASED TO FUNCTION-BASED E0Ps. l 1 RISK' BASED VALUE ASSESSMENT AND COST DATA TO SHARPEN FOCUS OF FUTURE DEVELOPMENT WORK IN PROCEDURES, l
- . - _ _ _ _ _ ' F :' '
l
Ill. PROCEDURES (CONT'D) STRESS OF ACCIDENTS ON OPERATORS: RES STUDY WITH RECOMMENDATIONS. HFIB WILL, WHEN l RESOURCES PERMIT, EVALUATE WHAT FOLLOW-UP ACTIONS ARE' APPROPRIATE. RELATED RESEARCH EPRI ADVANCED OPERATOR AIDS / EXPERT SYSTEM HALDEN STUDY OF COMPUTERIZED PROCEDURES i I O i _.--.__.___--.--_._,._..__.________.._-.g
*O
IV. MAN-MACHINE INTERFACE (HUMAN FACTORS ISSUES BRANCH) BACKGROUND , HFIB IS TRACKING THE EVALUATIONS BY THE DIVISIONS OF-LICENSING 0F CONTROL ROOM DESIGN REVIEWS AND SAFETY PARAMETER DISPLAY SYSTEMS. 'A!; IE INFORMATION NOTICE HAS BEEN ISSUED ON SPDS PROBLEMS. -
^
.P00R AVAILABILITY f
UNRELIABLE DATA AND ALARMS POOR OPERATOR ACCEPTANCE AND/0R INTEGRATION INTO E0Ps WE WILL EVALUATE THE NEED FOR ADDITIONAL REGULATORY ACTION CURRENT PLANS i LOCAL CONTROL STATIONS COMMONLY HAVE HF PROBLEMS ! CONDUCIVE TO HUMAN ERROR. WE PLAN TO SCOPE THE PROBLEMS, EVALUATE THE V/I INCENTIVES FOR REGULATORY OPTIONS. O pp . p6 5
IV, MAN-MACHINE INTERFACE (CONT'D)
. w OPERATOR OVERLOAD DUE TO ANNUNCIATORS: .. .
THE NRC HAS YET TO RESOLVE THE PROBLEM 0F T00 MANY ALARMS / ANNUNCIATIONS DURING SIGNIFICANT UPSET EVENTS, RESEARCHONADVANCEDANN0NCIATORSYSTEMSHASBEENDONEBY RES, IS UNDERWAY AT HALDEN AND EPRI. THE STAFF WILL - FORMULATE OPTIONS FOR RESOLUTION AND PREPARE REGULATORY
. ANALYSES.
ADVANCED CONTROLS, OPERATOR AIDS, AND DISPLAYS: THE STAFF IS TRACKING DEVELOPMENTS IN BOTH THE NUCLEAR AND OTHER INDUSTRIES. WE ARE GEARING UP FOR SAFETY EVALUATIONS OF LICENSEE PROPOSALS TO MINIMIZE THE REGULATORY DISINCENTIVES AND REWARD THOSE LICENSEES WHO PROPOSE TO IMPLEMENT ADVANCED CONTROLS AND DISPLAYS, l l l lO ,
V. STAFFING AND QUALIFICATIONS (HUMAN FACTORS ISSUES BRANCH) i CURRENT AND PLANNED ACTIVITIES , 4 I
- POLICY STATEMENT ON ENGINEERING EXPERTISE ON SHIFT
- ISSUED, OCTOBER 1985 COMMISSION PAPER ON INDUSTRY PLANS FOR IMPLEMENTATION 1 TO BE DEVELOPF.D BY JULY COMMISSION REQUESTED AND RECEIVED A DRAFT ADVANCE NOTICE OF RULEMAKING REQUIRING DEGREES OF SR0s IN FEBRUARY l
l l POLICY STATEMENT ON SHIFT WORKING HOURS .
- AT THE COMMISSION'S REQUEST, THE STAFF IS EVALUATING l
THE NEED TO UPDATE THE CURRENT POLICY STATEMENT AND THE TWO GENERIC LETTERS 82-12 AND 82-16. O l
#. f y n ")
-. 5.._.. _ ________
h
- O V. STAFFING AND QUALIFICATIONS (CONT'D) l
, . UPDATES ARE IN THE WORKS FOR REG. GUIDE 1.114, " GUIDANCE l 10 OPERATORS AT THE CONTROLS AND TO SENIOR OPERATORS..." AND THE CORRESPONDING SRP SECTION 13.1.2, "0PERATING ORGANIZATION" TO CONFORM WITH THE STAFFING RULE 1 10 CFR 50.54(M)(2)(III). I
- O e
i g 4 4 lO j t a
! f7 ) ,' i'
1 1 O VI. MANAGEMENT AND ORGANIZATION (HbMANFACTORSISSUESBRANCH) ALL WORK ON MANAGEMENT AND ORGANIZATION UNDER THE HFPP IS SUSPENDED: . THE COMMISSION'S " POLICY AND PLANNING GUIDANCE, 1986" ; CALLS FOR A MOVE TOWARD PERFORMANCE-BASED PROGRAMS.
. THE NRC IS PLANNING TO BE MORE RESPONSIVE TO ACTUAL INDUSTRY PERFORMANCE IMPROVEMENTS TO SALP ARE UNDERWAY
- DEVELOPMENT OF PERFORMANCE INDICATORS, f]. }?
O VII. HUMAN PERFORMANCE (HUMAN FACTORS ISSUES BRANCH) , A. DEVELOPMENT OF IMPROVED HUMAN PERFORMANCE EXPERIENCE REPORTAGE CONCERN: CAUSITIVE FACTORS OF HUMAN ERRORS (AND NOTABLE SUCCESSES) ARE ESCAPING REPORTAGE, l 1. HFIB IS WORKING'WITH AEOD TO DEVELOP A PROTOCOL FOR
. INCIDENT INVESTIGATION TEAMS TO GUIDE THE
! INVESTIGATION OF THE ROOT CAUSES OF UNUSUALLY POOR I (OR GOOD) OPERATOR PERFORMANCE DURING THE INCIDENT,
- 2. HFIB PLANS TO WORK WITH AEOD ON LER REPORTAGE GUIDANCE, AND WITH IE AND THE REGIONS ON AUGMENTED ,
i INSPECTIONS, l i 3. WE PLAN TO AUDIT AND EVALUATE THE TREATMENT OF HUMAN i PERFORMANCE IN INP0'S SEE-IN PROGRAM AND THE NEW HUMAN PERFORMANCE EVALUATION SYSTEM,
f
;u ,
O VII HUMAN PERFORMANCE (CONT'D) , t
\
;4 ,
s B. RISK AND PRECURSOR STUDIES ( CONCERN:
, L~
WE NEED BETTER VALUE/ IMPACT TOOLS TO ASSESS PRIORITIES, BUDGETS, PERFORM REGULATORY ANALYSES, BACKFIT RULE,
. . PLANS:
- )
WORKINGWIYHDRA0 ASSISTANCE,WEPLANTOREVIEW ! SELECTED PRECURSORS AND PRAs TO ASSESS THE EXISTING HF REGdLATIONSFORRISKRELEVANCE,IDENTIFYGAPSINTHE REGULATIONS, AND PREFABRICATE MODELS OF THE DIFFERENCES ( IN RISK TO BE EXPECTED OF CHANGES IN HUMAN RELIABILITY l IN EACH OF THE CONTEXTS WE D0 (OR SHOULD) REGULATE, FOR USE IN PRIORITIZATIONS, TO FOCUS DEVELOPMENTAL PROGRAMS, AND IN V/I STUDIES FOR STANDARDS DEVELOPMENT, O
- ___nt__
O OPERATOR LICENSE EXAMINATIONS (OPERATOR LICENSING BRANCH) G0AL . ENSURE THAT NRC-ADMINISTERED EXAMINATIONS GIVE A CONSISTENT, VALID MEASURE OF R0 AND SRO COMPETENCE FOR SAFE OPERATION CURRENT AND PLANNED ACTIVITIES CONTINUE DEVELOPMENT OF CATALOGS OF R0 AND SRO KNOWLEDGE AND ABILITIES BASED INITIALLY ON THE INPO JOB-RELATED ANALYSES , (1) VALIDATION BY STAFF AND SUBJECT MATTER EXPERT REVIEW (II) ASSESS NEED FOR PLANT-SPECIFIC CATALOGS TO SUPPLEMENT GENERIC CATALOGS HANDBOOK FOR EXAMINERS WILL BE PILOT TESTED AND IMPROVED AS NECESSARY, E.G., CONSTRUCTION OF SIMULATOR SCENARIOS EXAMINERS ARE BEING TRAINED ON FUNCTION-BASED PROCEDURES THROUGH SPECIAL AND REFRESHER TRAINING COURSES O fq7. ,2 / dl
O OPERATOR LICENSE EXAMINATIONS (CONT'D) . l UPDATES ARE PLANNED FOR . l 10 CFR 55, "0PERATORS' LICENSES" -
. R.G.1.1149, " NUCLEAR POWER PLANT SIMULATOR FOR USE IN OPERATOR LICENSE EXAMINATIONS" i
! - R.G. 1.8, " QUALIFICATION'AND TRAINING OF PERSONNEL FOR
, NUCLEAR POWER PLANTS" R.G. 1.139, " MEDICAL EVALUATION OF NUCLEAR FACILITY PERSONNEL REQUIRING OPERATOR LICENSES" l
l l l l l i I r} ?/ 3 l
APPENDIX XIII ACRS CHAIRMAN STATEMENT ON ACRS ACTIVITIES O Scope and Priorities for ACRS Activities Two significant influences have caused us to study and to develop a rather explicit plan for the scope and content of future ACRS activ-ities.
- 1. The changing nature of the regulatory scene -- few Ols, many operating plants. In this, we have had help from the Effectiveness Panel we appointed last year (more on that later).
I
- 2. The rather drastic reduction in our manpower and resources budgeted for FY 1987. 0ur "FTE" allotment is being reduced from 54 of two
,O years ago to 42 in 1987. i This FTE total includes members and consultant 5and the technical staff and administrative staff under Ray Fraley. Commensurate cuts have been made in budget allowances for travel and related expenses. This is highly critical for members; none of us live in Washington.
- I'd like to first describe how we plan to make cuts and control to the
. 1%er budget and then discuss what we plan to work on and what we plan to drop.
Our first cut is a fairly deep one in administrative and clerical services provided to the Members. We have already reduced the amount of l documentation and sorting of documentation supplied to the Members and l g y/ V ,
2-will make further reductions in the comprehensiveness of technical records of our activities. This has two negatives: i
- we will be more dependent on the NRC Staff 6nd other sources for raising issues; less on our own initiatives
- less support for Members means a decrease in their individual efficiency and effectiveness Our second, less deep, cut will be in our technical staff. Combined with a decrease in allotments for travel and related expenses, this !
makes necessary a decrease in our program activities. I can best express this as a decrease in the number of subcomittee meetings from about 110 per year in recent years to about 85 in FY 87. We believe we can maintain the level as high as 85 by more efficient scheduling (more back-to-back) of meetings. [Ishouldcommentthatthesechanges--less support for Members, more inconvenient scheduling of meetings for ! Members have a down side. This is not the world's most attractive job, l asitis.] Traditionally, it is the subcommittees that furnish the program leader-ship, in each specific area or set of issues, for the overall ACRS. Thus, the most straightforward way to control and prioritize the use of our more limited resources is to control the subcomittee activities. l O l l g ,2 / T
l x We have divided our program activities into 6 " mission areas." And ! beginning now and extending into FY 1987 we intend to devote approxi-mately the following portions of total ACRS effort to each: Operating reactors 30% Generic issues and USIs 28% Safety research 9% High-level waste 12% Control of radioactive exposure 3% Future plant designs 6% j Contingencies 12% 100% let me say a few words about some of these: The attention to operating reactors represents a significant increase from a year or two ago. In addition to some NTOL reviews, TVA, and our bimonthly meetings with your staff on operating events, we have formed new subcommittees.
- Systematic Assessment of Operations
- Plant Operating Procedures ' 6 0 f
/
g , ;> l : '
A V - Inspection and Enforcement Programs Whether we spend effort on future plant designs will depend to a large extent upon whether the agency staff intends to. I understand there is no such intent at present. The ACRS has concerns about whether this is wise -- we may coment later. We've set aside resources for contingencies -- reaction to events, response to Comission requests or Congressional requests. Histori-cally, such activities have consumed more than the 12% we have budgeted. You may . find us sometimes to be more sparing in our responses to future requests. We intend to spend no ACRS effort on:
- LLW (although Dr. Moeller may be able to squeeze some out)
- the regulatory process
- fuel fabrication, mill tailings We will be spending substantially fewer resources on:
USIs, GIs, investigation and exploration of equipment problems and other issues that individual Members would like to pursue O the safety research program y) ,2 ! D
APPENDIX XIV , ACRS REVIEW 0F TVA REORGANIZATION AND RESTART ACRS Review of TVA Reorganization and Restart The Commission has recently asked that we describe the nature of the proposed ACRS review of TVA's reorganization and restart plans. I wrote a letter to Chairman Palladino on April 7, 1986 on the subject; let me summarize that. We understand, from preliminary briefings and extensin d:::tmentation, that a number of safety issues have been identified at TVA -- by the NRC Staff and by TVA itself. These include EEQ, operational readiness, design control, QA, safety concerns of employees and others. A comon thread among these seems to be, again as acknowledged by both nd the O Staff, problems in the management and organization of TVA efforts. There' are concerns, apparently, at both the plant and corporate levels. The ACRS is interested in reviewing information about these problem
- areas and about the possible causes. We believe we have a responsibil-l ity to the Comission to provide it with our perspective on any safety problems that have existed, the causes, and the effectiveness of actions and programs proposed to correct the problems. ,
I-The review we plan is, I believe, consistent with our charter and, certainly, is within the envelope of issues that we have reviewed many times in the past. For example, in every OL review the Comittee must consider the applicant's FSAR and the SER provided by the Staff. Important chapters and sections in these documents are concerned with: (?- ? / '~
Management and technical support organizations O Operating organizations Qualifications of nuclear power plant personnel Administrative procedures Our reports to the Commission have often included advice or coments related to these matters. In a report of four or five years ago for a particular OL application, we stated:
" management has not been successful in putting together the team of experienced and qualified personnel we believe will be necessary."
"An extraordinary effort will be required to prepare the management and staff for operation."
"A more concerted effort is needed to build an integrated organization."
g "Better use should be made of sources other than the Applicant's 3 organization and its contractors" to provide professional expertise for its safety review comittees. The training program "has suffered from a lack of professional O direction." 4
/'. ,3 / O
All of these concerns were addressed by the Commission, the Staff, and the applicant and extensive changes were made in the applicant's organi-zation over the course of two or three years before their OL was granted. My understanding is that these changes were improvements and operation of the plant has been successful. As you know, the ACRS does not have the resources to make in-depth, first-party investigations. Instead, we depend on second-hand informa-tion -- reviews of documentation, and interactive meetings with appli-cants, licensees, staff, and consulting experts. This is what we plan for our TVA review. After several delays, out initial subcommittee meeting is planned for May 20 with a possible full Comittee meeting in June. Depending on what we hear, and on the program and progress by TVA and the Staff, there may or may not be another meeting or meetings following that. f f O
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~g
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J- Advisory Group Health Effects Models WASH-1400 SOMATIC EFFECTS Victor P. Bond, M.D.; Brookhaven National Laboratory Marvin Goldman, Ph.D.; University of California, Davis Leonard D. Haq[lton, M.D.; Brookhaven National Laboratory George G. Hutchison, Ph.D.; Harvard School of Public Health Clarence C. Lushbaugh, M.D.; Oak Ridge Associated Universities hg Roger O. McClelJan, D~.'V.M.; Inhalation Toxicology Research Institute
' Harry R. Maxon, M.D.; University of Cincinnati College of Medicine Samuel C. Morris, Sc.D.; Brookhaven National Laboratory ;
}J Lp Eugene L. Saenger, M.D.; University of Cincinnati College of Medicine l
~~% Leonard A. Sagan, M.D.; Palo Alto Medical Clinic -
Maurice F. Sullivan, Ph.D.; Battelle Pacific Northwest Laboratories 3 o Niel Wald, M.D.; University of Pittsburgh School of Public Health 5 Joseph A. Watson, Ph.D.; University of Pittsburgh School of Public Healti h 3, 55 *4 ; 2NETIC EFFECTS gzgg o iooo James F. Crow, Ph.D.; University of Wisconsin. ((35 E James V. Neel, M.D.; University of Michigan yE Dean R. Parker, Ph.D.; University of California, Riverside { William L. Russell, Ph.D.; Oak Ridge National Laboratory m 5 R d
4 + WASH-1400 i f TABLE VI 9-2 UPPER BOUND RISK COEFFICIENTS FOR LATENT CANCER FATALITIES l i Risk ; j Age at
- Time of Latent Plateau Coefficieng/
(deaths /10 Period (years) Type of Cancer Irradiation Period (years) yr/ rem)
- Leukemia In utero 0 10 15 0-9.9 2 -
25 2 10+ 2 25 1 Lung 10+ 15 30 1.3 ! Gastrointestinal tract: Stomach 10+ 15 30 0.6 Rest of ali-l}sa j mentary canal 10+ 10+ 15 15
'30 30 0.2 0.2 Pancreas
]I ' f 1, Breast 10+ 15 30 l.5(a) !b Bone 0-19.9 10 30 0.4 i lU 20+ 10 0 30 10 0.2 15 (b) ! All other In utero 0-9.9 15 30 0.6(c) 10+ '15 30 1 (d) l (a) Includes males and an assumed 50% cure rate. ! (b) "All other" includes all cancers except leukemia. (c) "All other" includes all cancers except leukemia and bone. l (d) "All other" includes all cancers except those specified in table. l i l l l i
] WASH-1400 i i , TABLE VI 9-4 EXPECTED LATENT CANCER (EXCLUDING THYROID) DEATHS PER MILLION MAN-REM OF EXTERNAL EXPOSURE 4 Expected Deaths Type of Cancer per 10" Man-rem Leukemia 28.4 Lung 22.2 i I jg Stomach 10.2 i'n Alimentary canal 3.4 l ,' 4O ! 1., Pancreas 3.4 ,,r
" Breast 25.6 l
l Bone 6.9 l 1 All other 21.6 l 1 ! Total (excluding thyroid) 121.6 l ) 1 Thyroid 13.4 l l 1 l
i ' s I. 1 ) NUREG/CR-4214 1 i h l i i l J i x_ _ i 9 1 s m R l ib. ib i ..q : ) ! l-LATENT PERIOD,1-l: PLATEAU,p -l ! t i l' Til01 "51NCE DOSE, 7 ? i l
- i Figure 2.3 Risk as a function of time since dose under an
! absolute risk model. t
l l l l l l l l c I i l WASH-1400 I i j TABLE VI 9-7 DOSE-EFFECTIVENESS FACTORS l i ! Total Dose Dose Rate (rem per day) b (rem) <1 1-10 >10 lM Je ! 1
<10 0.2 0.2 0.2
- 10-25 0.2 0.4 0.4 lV 25-300 0.2 0.4 1.0 1
4 v 4 i 4 ' r
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_ R E e V t EY I o M/ ID T 9 3 n) A 6 0 TA L 1 4 eI ER bI F E I I1 R o L tR O I _ ST wE U oB L OE E l E UR T y D NU U 8 ob IS L 7 o D O O 5 5 YA M TO S 1 T BR NP B ( N OX A O CE DR I EE T CP) C L C E DEE J E _ NUS O V 6 1 8 IL0 R O I 0 2 AF P T T 2 5 S '. X A 2 _ E IE-E K S E R L E TGS I U R _ IAN I R SD LRO I OA _ AES I PR TVR X _ AAE R E0 1 F(P I _ E E E _ RNN B L T EOO G U 7 7 0 CII N L 6 1 NTL I O 7 AAL S S 1 CII B DM A FA
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. NUREG/CR-4214
.J-
~ Process Osed in Model Development i i o Advisory Group formed on basis of nominations solicited from 300 experts. l o Wcrking Groups selected on basis of recommendations of Advisory Group. The group leaders . sere: I a Early effects - Dr. Fletcher Hahn fg j .y I* Cancers - Dr. Ethel Gilbert !b Thyroid ef f ects - Dr . Harry Maxon lb i 7 Genetic effects - Dr. Seymour Abrahamson I l o Health offects nodels were developed.by the Working j Groupa. i i ) o They were reviewed by the Advisory Group and by an independent review group. f i l 1 l l
NUREG/CR-4214 f~% ! ( MEMBERS OF ADVISORY GROUP ( Seymour Abrahamson, Ph.D. University of Wisconsin William J. Bair, Ph.D. Battelle Pacific Northwest Laboratories Michael A. Bender, Ph.D. Brookhaven National Laboratory Victor P. Bond, M.D., Ph.D. Brookhaven National Laboratory Richard C. Cuddihy, Ph.D. Lovelace Inhalation Toxicology Research Institute Keith Eckerman, Ph.D. Oak Ridge National Laboratory Jacob I. Fabrikant, M.D., Ph.D. University of California and - Lawrence Berkeley National Laboratory Marvin Goldman, Ph.D. University of California George B. Hutchinson, M.D., M.P.H. Harvard School of Public Health Dade W. Moeller, Ph.D. Harvard School of Public Health Edward P. Radford, M.D. Radiation Effects Research Foundation Eugene L. Saenger, M.D. University of Cincinnati College of Medicine Warren K. Sinclair, Ph.D. National Council on Radiation Protection and Measurements Niel Wald, M.D. University of Pittsburgh Edward W. Webster, Ph.D. Massachusetts General Hospital Shlomo Yaniv, Ph.D. U.S. Nuclear Regulatory Commission i ix Y1'
NUREG/CR-4214 OVERVIEW OF MODELS FOR LATE SOMATIC EFFECTS Effect Mortality Morbidity Leukemia X - Bone Cancer X - Breast Cancer X X 4 Lung Cancer X X Gastrointestinal Cancer X X Thyroid Cancer X X Skin Cancer - X Other Cancers X X Leukemia - In Utero X - Other Cancers - In Utero X - Benign Thyroid Nodules - X l 0 _f__________________
~
4 NUREG/CR-4214 i Table 2.0 Central Esttestes (Vith Upper end tower Sounde) ror Lifett alske of Mortality seeuttant 1 From Low-LET Exposure Received at Low Dose Rates ( < 0.05 Cy Per Day ) Based on the Linear i Term of the Linear Quadratic Function Number of Deaths Teere of Lif e test (Fer 10 For Cy) (Per 10 Per Cy) EII'** Lower Central Upper Lower Central Upper Bound Eatteste Bound Sound Estleate Bound l } Cancere Ove to Other Then g Utero Esposure l Laukeste 5 14 48 168 SOS 1682 1 Sone 0.2 1 2 7 22 75 j treset & 60 87 97 955 1452 Lung 5 20 138 100 288 1971 klI.s,N Castrotnteettnel 9 57 189 222 661 2202 Thyroid 7 7 7 203 203 203 , other 5 29 96 124 378 1260 i j ,. Cancere Due to ', 'A h Utero Eaposure ) Leukeete 1.2 1.2 3 80 80 200 l Other 1.2 1.2 3 80 80 200 i i l 'l I l ll
NUREG/CR-4214 Table 2.3 Models and Farameter Values for Central Estimates of Sometic Itieke for Individuale Coefficients Latency Plateau Minimum Age Mortality Horbidity o B Index Effect Type of Model . 14 Leukeata Absolute, linear- 2 25 n/a 2.24 x 10 n/a 0.3 0.47 quadratic 15 sone cancer Absolute, linear- 2 25 n/a 1.00 x 10-5 n/a 0.3 0.47 l quadratic l Breast Cancer
- Relattae, linear, 10 = 30 451 452 1.0 0.0 l 16 l
non-age-specific l 10 - 40 181 18% 0.3 0.47 17 Lung Cancer stelative, linear-quadratic l 10 = n/a 39Z 39% 0.3 0.47 18 Castroin- Relative, linear-quadratic testinal Cancerb 19 Thyroid Absolute, linear b[ Cancerc ,d.e ag ..p eggge \ ' ' gender-specific. a,s 18 5 = n/a 2.5x10j 2.5 x 10-' l.0 0.0 18 5 = n/a, 1.25 x10 1.25 x 10-' l.0 0.0 a*> %) Absolute, linear- ,4 y 20 Skin Cancer = n/a n/a 2.0 x 10 0.3 0.47 quadratic 10 N 21 Other Cancers f Relative, linear- 10 = n/a 202 20% 0.3 0.47 quadratic 22 Leukemia Absolute, linear 0 12 n/a 2.50 x 10-3 n/a 0.4 0
- g utero 8 Other b Absolute, linear 0 10 n/a 2.80 x 10~ n/a 0.4 0 23
. M uteroE 24 Benign Thyroid Absolute, linear llodulesCoeie age-gender-opecific
, a,5 18 10 = n/a n/a 9.3 x 10~' 1.0 0
. e,
- 18 10 = n/a n/a 4.7 x 10 1.0 0
) O NUREG/CR-4214 1
1 ( l Table 2.1 Summarv Mortalityetand theIncidences Model Dead.bto Determine Upper Sound. Centret, and 1.over Bound Lifetime Etek Estimate for t 3.tek Estimatten Model Effect Upper Bound Centret laver Sound t Cancera Due to Other Then h Utere Exposure Leukente Use absolute 11mest eettmate Modify
- upper bound by central Modify upper bound *by estimets reduction factore in tower bound reduction factore Table 2 4 in Table 2.4 Breast Use age-specific relative Use non-age-opecifte relative Modify non-age spectite absolute
) 1tneer eettmate lineer settmate by lower bound i 11neer estimate reduction factore in Tebte 2.6 Lung Use relative 11ater estimate Modify relative linear estimate Modify abeelste lineer estimate based on a stok coefficient based on a elek coef f tetent of by tower bound reduction factore of 371 per Cy IBt per Cy by centret eettmate in Table 2.4 reduction f actors in Table 2.4 i ,% Castrotn- Use relative linear estimate Modify upper bound by central Modify absolute lineer eetteste teettnal eetteste reduction factore in by lower bound reduction factors x1 y Table 2.4 in Table 2.4
! Thyroid C Use absolute Itnest estimates Use absolute Itnear eetteste Use ebeslute !!near estimate 3
5 Skin Use abeelste lineer eetteste Modify upper bound by central Modify upper bound by lower Q estieste reduction factors in bound reduction factore in Table 2.6 Table 2.6 l i "M other Cancers Use relativelineer eettestee Modify *ueser bound by centret ' Modify absetute Itnear estimate I g estimate reduction factore la bv tower bound reduction factero Table 2.4 in Table 2.4 l: Use abootute linest eettoete Senign Thyroid Use ebeslute 11meer eettmate Use ebeelste lineer estimate d Nodulee Cancers ove To Use absolute it- er eett-tes Use ebeelut. eett-te. .uiti. U.e centret esti-tes l i h Utere Esposure ; , , ; plied by 0.4 k t i
*The linear estimates referred to are given in Table 1.2 (morte11ty) and Table 2.3 (incidence).
yet convenience. "11aeer Itfettne rtok eettestes based on the ebeslute (relative) risk model" are referred to es
" absolute (relative) Itnear eettestee."
"3331 to assumed to be as ef fective,as enternal radietton for the upper bound thyroid cancer one third as effective for the central estimate, and one tenth as ef fective for the lower bound (see section 2.4.6).
d3 331 to assumed to be as ef fective se enternal radiation for the upper bound thyroid nodules, and one fif th as ef fective for the centret estimate and lower bound (see section 2.4.6). i l e j
O NUREG/CR-4214 6-m - 7 5-e BASE ~ 7 + INCREMENT O - v r- 4-7 l< 3 - w - , , o -
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$ 1-cc -
y O 20 40 60 80 TIME SINCE DOSE (YR)
' Figure 2.4 Risk as a Function of Time Under a Relative Risk Model.
d Data shown are for gastrointestinal cancer and for a I' person receiving i Gy to the lower large intestine at age 0-1 year. Points are plotted at the beginning of the 5-year intervals to which they apply. p'). , ). I $
i
' NUREG/CR-4214 6 ' ' '
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l I + BREAST CANCER ! I - Y OTHER CANCERS F 4 W 3-O l i - IL o 2- ,1 ~
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, hd 0 91-m y -
- o N : == = ===_- 5 5 5 5 5 5 5 5 5 a 5 5 5 5 5 5 5 5 5 0 20 40 60 80 100 AGE (YR)
Figure 2.5 Baseline risk of death from four types of cancer (gastrointestinai, l lung, breast, and other) as a function of age. Data are from the Vital Statistics of the United States, 1978.
REPORT OF THE NATIONAL INSTITUTES OF HEALTH AD HOC WORKING GROUP TO DEVELOP RADI0 EPIDEMIOLOGICAL Responding to the Congressional Mandate under Section 7(b) of the Orphan Drug Act of January 4, 1983 (PL 97-414) January 4, 1983 Published by the Office of the Director National Institutes of Health
- e NIH Publication No. 85-2748 U.S. DEPARTHENT OF HEALTH AND HUMAN SERVICES Public Health Service National Institutes of Health l
i i l l i 4
i AD HOC WORKING GROUP TO DEVELOP RADI0 EPIDEMIOLOGICAL TABLES l ROSTER OF PEMBERS , J. Edward Rail (Chairman) Deputy Director for Intramural Research National Institutes of Health Shannon Building, Room 122 Bethesda, Maryland 20205 (301) 496-1921 Oddvar F. Nygaard, Director Gilbert W. Beebe Clinical Epidemiology Branch Division of Radiation Biology National Cancer Institute Department of Radiology Landow Building, Room 8C41 Case Western Reserve University Bethesda, Maryland 20205 2058 Abington Road (301) 496-5067 Cleveland, Ohio 44106 (216) 844-3539 David G. Hoel* Director of Biometry and Arthur C. Upton Risk Assessment Program Professor and Chairman National Institute of Environmental Department of Environmental Medicine and Health Sciences New York University Medical Center PO Box 12233 550 First' Avenue Research Triangle Park, NC 27709 New York, NY 10016 (919) 541-3441 (FTS 8-629-3441) (212) 340-5280 Seymour Jablon Rosslyn S. Yalow i National Research Council Senior Medical Investigator 2101 Constitution Avenue Veterans Administration Medical Center Washington, D.C. 20418 130 West Kingsbridge Road - (202) 334-2825 Bronx, NY 10468 (212) 579-1644
. Charles E. Land Radiation Epidemiology Branch Victor H. Zeve (Executive Secretary)
National. Cancer Institute Special Assistant Landow Building, Room 3A22 office of the Director Bethesda, Maryland 20205 National Cancer Institute (301) 496-6600 Building 31, Room 4B47 Bethesda, MD 20205 (301) 496-5854
- Address during 1983-84:
Board of Directors Radiatien Effects Research Foundation 2-5 Hijiyama Park, Minami-ku 4 Hiroshima-shi, 730
.1apan j - - - . - - - . _ - . . , _ _ _ , _ _/ L if J _ _ _ _ _ __
Principal Developments Since' Publication of BEIR III
- 1. Reassessment of radiation doses received by atomic bomb survivors
- 2. Additional radioepidemiological data
- a. Hiroshima and Nagasaki survivors
- b. People irradiated for medical reasons i
- 3. Publication of Radioepidemiological Tables - National
- Institute of Health
- 4. Mental retardation from irradiation in_ utero
- 5. National Academy of Sciences proposes to update BEIR III 1
O l O 4
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I, APPENDIX XVI J
- - REPORT OF CONGRESSIONAL HEARING h { ON 1986 OHIO EARTHQUAKE Meeting No. Agenda Item Handout No.
312th 11 Title Report of Congressional Hearing on 1986 Ohio Earthquake Authors G. R. Quittschreiber List of Docunents Attached ******* ************* Memo to ACRS Members from Gary Quittschreiber******** on " Highlights of Congressman Udall's Subcommittee Hearing on the Janaury 31, ******* 1986 Earthquake Near the Perry Nuclear ******** Plant," dated April 10, 1986 ***** ******** l ******** ********* l i 1 ******************************************** Instructions to Preparer From Staff Person
- 1. Punch holes .
- 2. Paginate attachments G. R. Quittschreiber j
- 3. Place copy in file box J2?: ?
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4
# ~q,, UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION
~$ I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wasHwaTow, o. c. rosss
_o g Aoril 10, 1986 MEMORANDUM T0: ACRS Members FROM: w! try#RT Quittschreiber, Chief, Project Review Branch No. 2
SUBJECT:
HIGHLIGHTS OF CONGRESSMAN UDALL'S SUBCOMMITTEE HEARING ON THE JANUARY 31, 1986 EARTHQUAKE NEAR THE PERRY NUCLEAR PLANT The Subcommittee on Energy and the Environment of the Committee on ! Interior and Insular Affairs of the United States House of Representa-tives met from 9:45 a.m. to 2:30 p.m. on April 8,1986, in the Longworth O House Office Building to hear testimony from representatives of the Commission, ACRS, CEI, USGS and others, concerning the Janucry 31, 1986 earthquake in the northeastern Ohio, near the Perry Nuclear Power Plant. ! Congressman Eckert and Seiberling, both from Ohio, were in attendance. Congressman Udall made a very brief appearance. l , l The meeting was conducted with several panels, with panel members giving a statement, followed by extensive questioning from Congressmen Eckert
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and Seiberling. Acting Chainnan Seiberling opened the meeting noting that the NRC and - i ACRS both agreed that the January 31, 1986 northeastern Ohio earthquake caused no damage at the Perry Nuclear Plant. He commented that this meeting gives the Subconnittee an opportunity to review the safety of the Perry Nuclear Plant. NRC Testimony of NRC Chairman Palladino and Robert Bernero Chainnan Palladino discussed the NRC's evaluation of the effects of the January 31st carthquake on the Perry Plant, noting that a low power license for Perry Unit I was issued on March 19, 1986, but that the l Commission has not issued a full power license. Three contested issues ! are before the AS& LAP, including a motion to reopen the record. Mrs Bernero summarized the NRC Staff's technical findings with regard to i i O the January 31st earthquake and discussed the items to be confirmed prior to authorizing Perry, Unit 1 a full power license. _ .3
4 Udall's Hearing on Perry Earthquake April 10, 1986 e i O i In response to lengthy questions by Congressman ^Seiberling and Congress- , man Eckert, the following pertinent items were mentioned: ! l'
- Mr. Minogue noted that NRC funding of approximately 21 million
-dollars plus another 21 million dollars from the USGS is being authorized to investigate eastern seismicity. Minogue felt this amount is adequate for the purpose intended. He felt that the USGS
, should provide continuous long-tem monitoring, not the NRC. Mr. Bernero said that at this time there seems to be no depth, location, or time relationship with well injection and the January 31st earthquake epicenter. Phylis Sobel added that it will be difficult to ever make any determination regarding well injection having any relationship with the earthquake. A determination will be made over the next couple of years with the help of USGS and l ' John Carroll University. In the short-term, a decision will be L made within the next few months for a full power license. Ms. Sobel would not give a probability of occurrence of exceedance of an SSE-type event at the Perry site, but noteg that s p sites in the eastern states have been estimated at 10- to 10- per year. i Chairman Palladino said that the NRC's seismic requirements have been developed over the years using the best experts and best information available.
- Mr. Bernero said that for the January 31st earthquake at high frequencies there was some exceedance of the seismic response re-e quirements at the Perry site. He said this is not uncommon since the response envelope is set at 84%. He noted that a longer range question is whether to reexamine and maybe make a " modest correc-tion" to the response spectrum envelope for high frequencies.
Bernero said a follow-up study of the effect of high frequencies on equipment will be required as part of the ' continuing measures, and they are following ACRS Consultant Pomeroy's suggestions.
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Commissioner Palladino explained the procedure to reopen Appeal Board Hearings at Perry, Unit 1, if it is found to be necessary,
- s Mr. Malsh noted that ASLB hearings were complete at the time of the F January 31st earthquake and might not be reopened. If reopened, hearings could take several months and may or may not affect issuance of a full power license.
Commissioner Palladino stated that the Comissioners did not play any role in the issuance of the Perry, Unit 1, low power license. ' l If the earthquake had involved severe damage, the Commissioners likely would have interjected. i -
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4 'p Udall's Hearing on Perry Earthquake April 10,1986
- Commissioner Asselstine said there is a need to understand what caused the January 31st earthquake before the full power license on
.. Perry, Unit 1, is granted. He felt that what was being done is a ,
start to provide that information, but he does not want an "artifi- l cial timetable".
- Mr. Bernero'said that if a magnitude of 5.3 event had occurred during Perry operation a few more systems would have tripped, but he would have still expected a safe shutdown. Mr. Denton added that most probability studies show loss of offsite power during seismic events as the biggest problem and that events of 2 or 3 times the SSE would be needed to produce a high likelihood of core melt.
- Mr. Steffano said that with regard to an ACRS comment in an earlier report concerning restructuring the quality assurance area, the ;
quality assurance electrical contracting area was revamped after a Stop Work order in about 1978. !
- Mr. Bernero said that 57 instrument valves were found mispositioned .
during an inspection following the Perry earthquake. These were , small unnumbered valves which h, ave been known to be problem areas with regard to positioning. Finding these mispositioned valves had
- no association with the earthquake since they would likely have been found during system turnover even if the earthquake had not occurred.
! ACRS Testimony of ACRS Chaiman Ward Mr. Ward summarized the March 17, 1986 ACRS Report on the Perry Nuclear Power Plant, Unit 1, with regard to the January 31st earthquake. He mentioned Ms. Hiatt's, the Ohio Citizens for Responsible Energy, letter to the ACRS which was received after the Committee had prepared its report. He noted that she called attentior, to a " capable fault" exist-ing near the Perry Plant. Dr. Pomeroy did review Ms. Hiatt's letter for the ACRS and did report to the Committee in a March 25 letter. Mr. Ward mentioned the ACRS recommendation of its July 13, 1982 report concerning the need for studies to evaluate the margins available to accomplish safe shutdown. He noted that the NRC and EPRI have established such programs which are in progress and that CEI has assured the Committee that its seismic margin study for the Perry Plant will be completed and that needed modifications will be made before the resumption of op-erations following the second fuel load. In response to numerous questions from Congressman Seiberling and Congressman Eckert, the following pertinent points were mentioned: O
- Dr. Pomeroy stated that at the time his letter was written, infor-mation indicated two fault planes and that aftershocks were on one fL yk
Udall's Hearing on Perry Earthquake April 10, 1986 of those two planes. Now there is information on aftershock motion
.with locations differing by different groups. He said differences
.. in direction are associated with the different locations of the aftershocks.
- Mr. Ward noted that the ACRS made some recommendations similar to those recommended by Dr. Poo roy and that programs have been committed to.
- Mr. Ward stated that the ACRS disagrees with the NRC "quite often" and is "yes quite independent". Mr. Ward noted that the ACRS is not totally independent since it is part-time and "we find our- '
selves skeptics". The ACRS does not have the resources to do its own complete review. He noted the ACRS is a part of the system to protect the public health and safety and has no axe to grind with~ either the NRC or the utilities.
- In response to questions from Congressman Eckert to Dr. Pomeroy concerning his opinion of whether Perry, Unit 1, should be licensed to operate before some studies were performed. Dr. Siess felt that the question to Dr. Pomeroy was unfair, since Pomeroy was called O upon by the ACRS to give expert opinion on seismology, not a balancing of public risk.
- Mr. Ward said he was reasonably confident that the ACRS had all the information needed to write its repor,t.
- Dr. Pomeroy stated that a relationship of seisancity has been associated near injection wells. In this case, there has been no l such association between the location of the wells and the location of the January 31st earthquake epicenter. He added that the USGS will pursue such a relationship. -
- Dr. Pomeroy said the 200 year history of earthquakes in this area is sufficient to predict earthquakes of this size, but may not be sufficient to predict larger earthquakes.
. TestimonyofM.Edelman,CEI,C.Penzone,CraftsofNgheastOhio,and -
Dr. Ahmad, Ohio University Mr. Edelman discussed the CEI inspections and evaluations at the Perry Plant following the January 31st earthquake. He felt that well in-jection did not impact the earthquake. Presently one-third of the core is loaded. Mr. Penzone testified concerning the excellent craftsmanship, the O, cooperative spirit of the utility, and the appreciation of the nearby communities for employment. g)D l
Udall's Hearing on Perry Earthquake _ April 10, 1986 l Dr. Ahmad (difficult to understand his testimony) stated that he feels l that injection into the wells was definitely related to the January 31st l qvent. He also said that tne NRC knew about some faults as early as ' 1979, but said they were unknown. He accused the statistics". HesaidDr.Sobelsstatisticsof10gRCof"usingphony of exceeding the SSE are " baloney". Ahmad added that a new hydro-seismicity theory by Fennington and Davis have shown that using gas company statistics in the Lubbock Texas area shows that well injection does trigger earthquakes of low pressure zones adjacent to high pressure zones. He suggested that the NRC and Dr. Pomeroy did "not do their homework," when they stated that they did not believe well injection had any relationship to the January 31st event since the earthquake location could not be associated to the area near the point of well injection. The status of the Penn-ington and Davis paper was not clear, Ahmad mentioned that it would be published shortly in a seismic journal, but also noted that it had been in existence for a period of time.
- Congressman Seiberling suggested that Dr. Pomeroy and Dr. Sobel be given copies of the Pennington and Davis paper and be asked to provide comments in writing to the Subcommittee.
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- In response to questions from Congressman Seiberling, Dr. Pomeroy said he did bring the Ahmad question up at a ACRS Subcommittee meeting concerning whether associated earthquakes could occur away from the wells. Pomeroy said he could provide 10-15 different hypothesis as to what caused the earthquake. He added that he was somewhat familiar with Ahmads studies but does not know whether i they are applicable.
- Ms. Sobel said the NRC is aware of the Pennington and Davis studies and will provide a written statement to the Subcomittee.
Testimony of Mr. Hagen, Ohio State Representative of the 61st District and Ms. Hiatt, Ohio Citizens for Responsible Energy Ms. Hiatt felt that the NRC and CEI were gambling with lives, have committed a blatant violation of the regulatory system, and made a pitch for reopening licensing board hearings. Mr. Hagen said at the time of the January 31st earthquake he called NRC and was told pipe flanges were damaged and concrete was cracked at the Perry Plant. Later.he was told no damage. He questioned the reliability of the agencies which are supposed to protect the people. In response to questions from Congressman Seiberling and Congressman Eckert the following points were discussed: . ,O gA ~ '
l Udall's Hearing on Perry Earthquake April 10,1986
- Mr. Hagan was concerned that economics were more important than the i safety of the people. He would like to see the plant converted to coal for safety reasons.
- Ms. Hiatt'said her main concern was with regard to the Mark III 15 psi containment in a severe accident.
- Mr. Seiberling said that after many years on this Comittee he is satisfied with nuclear plant safety if there is adequate concern by the NRC, utility and others for safety.
Testimony of R. Wesson, USGS and Dr. Seeber, Columbia University Mr. Wesson said the USGS is investigating the possibility that the January 31st earthquake was caused by well injection, but that it is unlikely they will ever know for sure. He said that the pressure at the well 7 miles from epicenter was sufficient to trigger an earthquake if a fault was present; however, he said that the probability that the injection triggered the event is low. Dr. Seeber discussed the distribution of historical earthquakes in O
\ northeast Ohio. He said the January 31st event was small, about 1 kilometer across, and that small faults cannot be easily identified. He said the fault occurred on a lineament which reactivates possibility of an active lineament. Seeber said the NRC seismic research program as funded is a major mistake by cutting funding of earthquake monitoring in eastern states.
In response to numerous questions from Congressman Seiberling and Congressman Eckert the following significant points were mentioned: (
- Mr. Wesson said active monitoring near wells is necessary and if seismic activity extends out in the future this would change his opinion.
- Both Wesson and Seeber said they were familiar with the Pennington and Davis work. Wesson said the Davis and Pennington study has injection and extraction in the same areas. He did not see how that model could be applied to the January 31st earthquake.
- Mr. Wesson said the January 31st earthquake was "really a crummy little earthquake" and although it was a magnitude 5 it was of very short duration and not much of a test. He felt there was a large margin difference between the event and what it would have taken to damage Perry.
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- Mr. Wesson said the NRC requirements are based on western U.S. type tQ earthquakes and that finding cable faults in the eastern U.S. is difficult.
l yy. , ') s 4 . '
Udall's Hearing on Perry Earthquake .7- April 10, 1986 i
- Mr. Wesson said that with regard to historical data, a 5.5 magni-tude earthquake is the largest event known to have been triggered
, by well injection,
- Dr. Seeber suggested that the existing level of HRC seismic re .
search is adequate but'that he was concerned that funding was being shifted away from universities toward the USGS. He also felt more should be shifted from the western states to the eastern states. No final conclusions or statements were made at the end of the meeting. cc: ACRS Technical Staff O O O 4 9 4 A.)
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I
'm ' _ APPENDIX XVII f' DHRS SUBCOMMITTEE REPORTS i
Meeting No. Agenda Item Handout No. 312TH' 13.1 1 Title DECAY HEAT REMOVAL SYSTEMS SUBCOMMITTEE REPORT - RESOLUTION OF USI A-45 AND AUXILIARY FEEDWATER SYSTEMS Authors D. WARD t List of Docunents Attached
) ********* ****** ,/ ********** *******
l *********** ********
*******o**** *********
1 t ******** ************************ Instructions to Preparer From Staff Person
- 1. Punch holes PAUL BOEHNERT
- 2. Paginate attachments
- 3. Place copy in file box
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l April 8,1986 O TO: ACRS Members i FROM:- D. A. Ward, Chairman , ACRS SUBCOMMITTEE ON DECAY HEAT REMOVAL SYSTEMS
. Report from March 18,1986 Meeting In attendance for the ACRS: Ebersole, Michelson, Ward and Consultants Catton and Davis.
We heard a report from the Staff on the status of Task Action Plan A-45. The final draft resolution of A-45 is now scheduled for October,1986. That is to be a resolution package to the director of NRR. Following CRGR review, they expect to L issue a resolution package for public comment in March of 1987. However, the program manager indicated that this may very well be an optimistic estimate of the resolution dates. Although I have to admit to impatience with what seems to be an ever receding date for resolution, I also have some sympathy with the Staff and with
% the Project Manager because it is a very complex and comprehensive issue. He resolution of A-45 is going to have rather wide impact throughout the agency and the industry.
If you will recall, the analytical approach being taken under the Action Plan is with Sandia National Lab as prime contractor, the decay heat removal systems of 6 or 7 plants are being analyzed using PRA methods. These 6 or 7 plants are intended to be representative of the entire population of plants in the country. He intent is that anaysis of these plants willindicate whether there is need to improve the reliability
- of performance of decay heat removal system in existing plants, and if there is such a need, to establish whether cost effective means to make the improvements can be identified. A-45 does not address the design of future plants. It might have an impact on such design, but the action plan is not intended to provide direct input to the design of future plants.
I would like to relate a couple of items of interest from the meeting. One, the PRA methodology being used in the analysis is conventional, in that it addresses only the operating modes of plants. It does not specifically address the reliability of decay ' heat removal from plants during shutdown modes. There is experience that EPRI/NSAC has reported on which indicates significant unreliability of decay heat
removal systems in the shutdown modes. The reasons for this are several, but include the fact that during these modes, less attention tends to be paid to 4
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ACRS MEMBERS 2 April 8,1986 4 O maintaining the shutdown removal systems. That is, Tech Specs are relaxed, operating procedures are relaxed and perhaps attitudes are somewhat sue relaxed. While the maximum potential risks are lower in the shutdown modes, if controls are relaxed this could more than compensate for the lower potential. The result could be that risk is actually larger. The Subcommittee perceived that the Task Action Plan was not paying much attention to this, but was concentrating on the traditional PRA analyses of the plants in their operating modes. The Subcommittee believes that the Task Action Plan should include at least a qualitative analysis of experience in the shutdown modes so that an eventual resolution package can be properly balanced to include analysis of decay heat removal reliability in all modes of plant operation. Our Consultant, Pete Davis, raised what I think is an important point about the need to recognize what might be orthogonal requirements in BWRs for improving decay removal systems while at the same time maintaining the ability to deal with the ATWS. In general, the decay heat removal system improvements concentrate on more and more ways to pour water into the core under unusual and extreme conditions.' You might recall that the ATWS mitigation in BWRs depends on proceuures to control the water level to just the right (and really rather a low) level. Enough to keep the core cool, but also a low enough level so that significant boiling (7 occurs to decrease moderation and keep power low. I think it is reasonable to say that there is a conflict between these two issues; it is probably more of a ATWS problem than it is an A-45 problem, however. The concept of using a PRA methodology at the beginning of this program was that existing techniques would be used. PRA would be only a tool for doing the analysis and there was not going to be any attempt to improve the PRA technology. 'Ihere is
. a limited budget for this program and it has its own mission. I think that has been a l reasonal le approach, but as we listen to the results of anlayses of plants, our l Subcommittee members continue to have problems with the PRA methodology and l
concern that many important things are left out of even the best PRAs. For example, the operation and reliability of various support systems contrasted with the primary reactor heat removal and safety systems. Again, I think this is true, there are problems with PRA, but again, this is not uniquely an A-45 problem, but is an issue that should be recognized and perhaps dealt with in other aspects of PRA work. The bottom line of the two analyses presented to us at this me ting, one for a BWR-4, the Cooper Plant, and the other for a PWR, the Tur oint Plant, continues Q V to be similar to those we've heard about for previously studied plants. The core
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., l ACRS MEMBERS 3 April 8,1986 l
melt probabilities as calculated by traditional PRA methods are somewhere in the range upward from 10 4 per reactor year, but that there are few easily identifiable fixes to provide cost effective reductions in core melt frequency. For the PWRs, crediting the feed and bleed process does a little bit of good so far as the calculated numbers are concerned, but not a great deal. It continues to look like any sort of dedicated add _on decay heat removal system will not be justifiable through the PRA and cost benefit analysis. This is certainly not the final conclusion of the Task Action Plan study, but this is the trend of results that we are seeing in this effort. O
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O April 1,1986 TO: ACRS MEMBERS FROM: D. A. WARD, CHAIRMAN DECAY HEAT REMOVAL SYSTEM SUBCOMMITTEE Report on Meeting on March 26,1986 - Auxiliary Feedwater System Reliability Attendance at the Subcommittee meeting included Mr. Reed and myself, and also Consultants Catton, Schrock, Sullivan, Tien and Davis. In December oflast year, the Committee wrote a letter to the NRC staff exhorting them to develop a better defined program for addressing a perceived problem of the v relative unreliability of the auxiliary feedwater systems in a group of older plants. If we recall the history of requirements for design of auxiliary feed systems, there . has been some evolution toward more rigorous requirements. Prior to the SRP
- (1975), " good engineering practice" was required for AFWS design. After1975, i AFW systems were to be " safety related". All plants were then reviewed against requirements spelled out in NUREG 0737. Finally, the SRP was revised for post-TMI NTOLs to require " unavailability" no greater than 104 to 10-5 per demand. (What is intended by the range is unclear.) There is some residual concern that a group of early plants may have feedwater systems that are not as good as present standards indicate they should be.
The staff has designated this concern as Generic Issue 124, Auxiliary Feed System j Reliability. Before the reorganization of NRR, the old Division of Safety
- Integration had prepared a draft regulatory analysis to resolve the issue. With the l reorganization, they transferred this to the Division of Risk Analysis and Operations. It was classed as nearly resolved when it was transferred. There was an early report on the proposed resolution late last year. That triggered the ACRS letter of this past December.
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ACRS MEMBERS 2 April 8,1986 At the March 26 meeting, the Staff reported that the resolution is proceeding, with a schedule calling for an ACRS review of the technical proposal for resolution in May of this year. It then will go out for public comment. However, there may be a bit of a problem. Although we just heard a status report on March 26, the preliminary indications are that the current methods of cost benefit analysis don't lead to a quantitative estimate of benefits that will support significant improvements in these systems. The perceived deficiency of most of these older plants is that they have two-train auxiliary feed systems. The obvious, or at least one obvious, means for improvement systems is to provide a third train, whether a replicate or ogwith diverse drive or flow paths. C However, the Staff reports that the cost benefit analyses that they are doing do not ypha17 the need for addition of a third train. What this seems to mean is if the Staff does indeed come out with a recommendation, that these plants should add a third train to improve reliability, or if the ACRS wants to make such a recommendation,that recommendation will have to be based on something other than the present methods of cost benefit analysis and PRA. O i i \
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APPENDIX XVIII ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE
- 1. Memorandum, J. MacEvoy, ACRS Fellow to F. J. Remick,-Foreign Safety Goals, March 5, 1965
- 2. Memorandum, M. Norman Schwartz, ACRS Technical Secretary to R. F.
1 Fraley, March 28, 1986, Commission Meeting - Advisory Committee on j Reactor Safeguards (ACRS) Meeting on Safety Goals, April 11, 1986 I '3. Memorandum, R. Savio, ACRS Staff Engineer to ACRS Members, (Sort of individual member's comments and the July 17, 1985 and March 19, j 1986 ACRS recommendations) April 11, 1986
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