ML19309D506
| ML19309D506 | |
| Person / Time | |
|---|---|
| Issue date: | 12/04/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1678, NUDOCS 8004100392 | |
| Download: ML19309D506 (140) | |
Text
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APPEf1DIXES TO MIfiUTES OF Tile 234TH ACRS MEETIfiG OCTOBER 4-5, 1979 i
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Chai nnan 's Report (0 pen to Publ i c)...................................... 1 A. Reviewers........................................................... 1 B.
Request from Rep. Udall for Opinion on a Proposed Hybrid Standard Nuclear Power Plant............................... 1 Request from Comissioner Gilensky for boments Regarding C.
Control of Xenon Following an Accident............................ 1 D.
Request from Comissioner Bradford Regarding Identification of NRC Regulations That May need Changing......................... 2 E.
Program Plan for the Investigation of Vent-Filtered Containment Conceptual Design for Light-Water Reactors............ 2 II. Meeting With Members of the NRC Staff Re. Sep 25, 1979 Transient At North Anna Power Station Unit 1 (0 pen to Public)................... 2 III. Executive Sessions (0 pen to Public)..................................... 2 A.
S ubcomi ttee Repo rts................................................ 2 1.
Priori ties for Research Projects................................ 2 2.
THI-2 Acci dent Bulletins and Orders............................. 3 3.
TMI-2 Accident Implications..................................... 5 4.
Wo l f C re e k...................................................... 5 B.
Membe r Req ues ts o f NRC Staf f........................................ 6 C.
Review of Regul atory Function and Process........................... 6
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D.
Comittee Position Regardir.g Proposed Plants for Which a Favorable ACRS CP Report has Been Written, but for Which a CP h as no t be en Is s ue d............................................ 6 E.
Con d u c t o f Membe rs.................................................. 6, 1.
Mr. Carbon......................................................6 F. Future Schedule..................................................... 6
- 1. Agenda..................................................'........ 6 2.
S ubcomi ttee A'c ti vi ti e s.......... '............................... 7 1
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i Table of Contents Minutes of the 234th ACRS Meeting G.
ACRS Reports and Letters 1.
Hybri d Standard Nuclear Pl ant Design.......................... 7 2.
Sys tema tic Eval uation P rogram................................. 7 3.
Xenon Emission Control........................................ 7 4.
Systems Interactions Study for Indian Point Nuclear Gene ra ti n g Sta ti on Uni t 3................................... 7 i
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M Appendixes to 234th ACRS Meeting October 4-5, 1979 APPENDIX 1 - Attendees.............................................. A-1 APPENDIX II - Future Agenda......................................... A-5 APPENDIX III - Schedule of ACRS Subcommittee Meetings and Tours..... A-7 APPENDIX IV - Request from Rep. Udall for ACRS Comments on Proposed Composite Standard Hybrid Nuclear Power Plants...... A-10 APPENDIX V - Request from Commissioner Gilinsky for Comments Regarding Control of Xenon Following an Accident..... A-16 APPENDIX VI - Request from Commissioner Bradford re Identification of NRC Regulations which Need Changes............... A-24 APPENDIX VII - Background Material re Sep 25, 1979 Transient at North Anna 1.................................... A-25 APPENDIX VIII - Sequence of Events During Sep 25, 1979 Transient at North Anna 1................................... A-52 APPENDIX IX - Letter from Mid-America Coalition for Energy Alternatives to NRC Commissioners Regarding Seismi c Design of Wolf Creek...................... A-59 APPENDIX X - ACRS Consultants' Report on Wolf Creek................. A-65 APPENDIX XI - Status of Plants Under Construction Permit (CP)
Review.............................................. A-67 APPENDIX XII - Letter to Rep. Udall re Proposed Hybrid Standard Nucl ear Pl an t Desi gn............................... A-69 APPENDIX XIII - Systematic Evaluation Program....................... A-74 APPENDI X XIV - Xenon Emi ss i on Control............................... A-76 APPENDIX XV - Systems Interactions Study for Indian Point 3......... A-78 APPENDIX XVI - Additional Documents Provided for ACRS' Use........... A-81 t
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- Fedund Lip 2ar / Val 4s. No.1st / Tmeadey, Octaher 2, les / Nonnes l
t power plant damip as a 6ssia far t'
- s regenetsons and. La some
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cases. to delineate techniques used by dose thsee seassons to proasce licensant the staff in evelaating specific problaans prop (rwtary irdarmation (a 1i ut De Casmesttee wd! hear and descame l
Fwther idermetten regenhus topium the repcrt of tas Saboossanttee on Three or postuletad anu-ess and to provide
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l to be discussed. wheder the awetag Mde Island N= clear Stetson Unit g (%G-guidance to appl. casts raar=rning certain information needed by the staff has been canceGed er reedenfuled,the
- 2) farplications and a; placation af this la hs review of We=hna or perman Cha trman's ruling on regruests for the experience to Operstsag f) censes for several BWR and PWR andeer pianta, and licenses.
j Re dragt sides.tensporarily identikd cppwturmty to preseat oral statements Members d the NRC Staff wel by its task ni=l=r. FP at3-4,la entitled i
c.nd the tune alksted theeefor een be cbtdnad by a prepaul telephone cm3 to partsctpats as approprises.
" Safety.Related Permanent Dewstering De Committee wdl hear and discans the Desgynated Federal Employee ear tbn report of tas Subcorramittee regardnes Bystems for Nudear Power Plants" and N
thb umeting. Dr. Andrew L Buses, impiar===tatmos of NRC Bunetma and is intended for Division 1. " Power Reacsore." It identmes==Malcal and
,l (talephone 7At/s%32s7) between etis Orders resultang fxan the anad=nt hydroksic agneenne design bases and y
a m. aad 12e p.m.. EDT.
which oocarted at TM1-a. Meanbers d o
itasas to be corundseed et this meetnas the NRC Staf will partacrpato as critena for pamanent dewataring Badiground halormatnam concernmag systems that are dep== dad upon to can be found in doctaments en fde and appropriata.
aarve safetyrelated p availatde for public inst =crxm at the Friday,OctM m his dran guide and sociated NRC Pubhc Document Roosa.171711
- Mam-2fMP.m./Ezecutive value/ impact statement are betng issued Strued. NW. Washington. DC amio and Session (QaanF he Cerassitas will.
to involve the pubDe in the earfy steges et the Governraent Publicatione==rhm discuss proposed ACRS===aats and of the development of a regulatory r
State I.kbrary d Pennsylvassa, recnrataendations regarding the NRC position in this area.ney have not Fdurmtton Buddseqq. Gunmonwealth and received comple9e stdf review, hete not Weinut Street, llarrisburg. PA 1712a, regulatory proccan.
flMP.m.-1Mp.m: Fam,4/ve Sessico been reviewed by the NRC Reguletory i
Deed """"- 25. tre' (Open)-ne Committee will discuna its Reqsarements Review Committee, and
).ka C. H@e, proposed annal report to the U.S-do not represent sa official NRC staff i
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Coagres4 far Calesdar Yeat 1s79 on the position.
ifs Dus. Em N m M and NRC Safety Research Progran Public r<====h are being solicited ne Committee wiD discuss proposed on both drafts, the guide (including any ACRS commanta en the use of degraded impkmaalation schedule) and the draft name coot m*"
conditions as a licensing basta.
value/ impact statement. Cosaments on W Caserviettee on W
%e Cosamhtee wd! discuss proposed the draft value/ impact saaternant should gm pg ACRS comments on the status d action be accompanied by supporting data.
Commise6cn; Revloed Motice et being taken to evaluate systems Commsets os both drafts should be sent 88#"9 interactions at the Zion Nudaar Station to the Secretary of the Commission, UA In accordance with the purposes of and the Indian Point Nuetaar Generating Nudear Regulatory Coensuosion.
sections 29 and 182b.of the Atomic Plant Unit 3.
Washingen.D.C rnm Attention:
ne Committee will discuss proposed Docketing and Servios Branch, by Energy Act (42 U.S C an9,2:'32 b.), the ACRS comments regarding the NRC-November 30.1979.
Advisory CommLatee on Esactor Safeguards wt3 hold a meeting on Systematic Evaluation Program fur Although a time limit is given for j
I October 4-6.1s'9. In Room 1046, in711 nudear plants.
comments on these drafts, comments and suggestions in connection with (1)
Street. NW, Washington, DC Notice of Saturday, Ocamber s, m items for indusion in guides currently tha meeting was published ou I:Mp.m -tM p.ma Secutive Seselo
being developed or (2) improvements in Septerober 19,1F79 (44 FR 543tiB). %e (Open)--he Commrttee will discess its all published guides are encouraged at schedule for conduct of thIs meeting is schedule for intere activities induding revised as noted below to accomrnodate consideratitm of methods to control egulatory guldes are available for the deferral of several matters origma4y menon estismons followmg a nudeur inspection at the Commf esion's Pubbe scheduled for considerstion.
Document Room.1717 H Street NW he revised egenda for the subject accident-Washington. D.C Requests for slasie meeting w1B be as foHows:
Dated. September as. Is7n copies of draft guides (which may be lohn C. Hoyle, reproduca d) or for placasient on an hrsday. October 4,1s75 A*isory cannw< tan Meney mentOAar autuamatic k% bet for %
8M o m..!?Mp m.: becutive I" D* * **""*"d " *"* "'I copies of future drdt pides in specdic Session (Optm)-.De Committee wf!]
same cooe mw+e divisions should be made in wnting to hear and d:scuss the report of the ACRS the U.S. Nudear Regulatory Chairman regarding miscellane".
Commission. Washington. D.C 20555, matters reletmg io ACRS activities Draft Regulatory Guide;laamance and Attention Director, Division of h Canmime vnt dncree gW Avaltability TechnicalInformation and Document ACRS comnets enhcanewndaum ne Nudear Regu!stery f'-ime Contrd. Telephone requests cannot be regarding the NRC regulatory proceea.
has lasmad fur pubbc cassament a draft of accommodated.
guides are f.mp m.-7mp m;Dreutive Sesseee a new guido planned for hs Regalatory no c pW t en saion (Open)-The Cmmnittee wdl discwes its Culde Sestes togetbar wedi a drpft d tbs prtrosed annual report to the Ua associated valms/ impact statement.nds a[ proval la not required to reprod Congrees for Calualar Yeer 19r9 es the senes has been devdoyed to dancnise h Conuaittee wtB discuse proposed and make acadable to the public (5 U.gc sas(s3 NRC Safe +y Reseerth Program snethods aweptable to the NitC staf of Dated as endnens.ud. Aihn seah day of ACRS comments regoedmg the lanpI====eing specafic puuts d de Septendser mee.
devdopment of a omenpoolse a 4===
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SCHEDUI.E AND OLTTLINE FOR DISCUSSIW 234Td ACRS MEETIE OCTCBER 4-6, 1979 R SHIMTON, DC H Street, NW, Washington, DC_
n ursday, October 4, 1979, Room 1046, 1717 Executive Session (open) 1.1) 8:30 A.M. - 9:00 A.M. - Chair-man's Report Discuss 1.2) 9:00 A.M. - 12:30 P.M. -
proposed ACRS comments on the nu-clear regulatory process LUNCH 12:30 P.M.. - 1:30 P.M.
Executive Session (Open) 2)
1:30 P.M. - 7:00 P.M.
2.1) 1:30 P.M. - 3:30 P.M. - Discuss proposed ACRS Annual Report to Congress on the NRC Safety Re-search Program 3:30 P.M. - 5:00 P.M. - Discuss 2.2) proposed ACRS response to in-quiry from Congressman Udall re-garding a nuclear composite plant design 5:00 P.M. - 6:30 P.M. - Report 2.3) of ACRS Subcomnittees on 'IMI-2 Implications and plants with pending operatirq licenses I
6:30 P.M. - 6:00 P.M. - Report of 2.4)
ACRS Subcommittee on NRC/IMI-2 Bulletins and Orders a
H Street, NW, Washington, DC Friday, October 5, 1979, Room 1046, 1717 Executive Session (Open)
- 3) 8:30 A.M. - 12:30 P.M.
3.1) Discuss proposed ACRS comments on the nuclear regulatory pro-cess UJNCH 12:30 P.M. - 1:30 P.M.
Executive Session (Open)
- 4) 1:30 P.M. - 6:30 P.M.
4.1) 1:30 P.M. - 2:30 P.M. - Discuss proposed ACRS Annual Report to Congress on the NRC Safety Re-search Program 4.2) 2:30 P.M - 4:30 P.M.- Discuss proposed ACRS comments on the nuclear regulatory process re-garding consideration of de-graded conditions as a licensing basis 4.3) 4:30 P.M. - 5:30 P.M. - Discuss proposed ACRS comments on analy-sis of systems interactions at the Zion Nuclear Plant and the-Indian Point Nuclear Generatirg Station Unit 3 4.4) 5:30 P.M. - 6:30 P.M. - Discuss proposed ACRS comments on the NRC Systematic Evaluation Pro-gram 9
9 mm...
H Street, IM, Washinton, DC Saturday, October 6, 1979, Room 1046, 1717
'Ihe ACRS meeting will be recessed.
8:30 A.M. - 1:30 P.M.
Executive Session (Open)
- 5) 1:30 P.M. - 2:30 P.M.
5.1) Schedule for Future Activities 5.1-1) Response to com. Victor Gilinsky regarding redoc-tion of xenon emissions following an accident 5.1-2) ACRS position regarding review of nuclear plants for which ACRS reports on construction have been completed but CP's have not yet been issued 5.1-3) ACRS review of seismic design of Wolf Creek nu-clear plant (tentative) 5.1-4) Anticipated Subcommittee Activities 5.1-5) Anticipated full Committee activities 5.1-6) Complete discussion of iters considered at this meeting l
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December 4, 1979 I
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234W ACRS MEETING OC'IDBER 4-5, 1979 WASHDCTON, DC j
g the Advisory Comittee on Reactor Safeguards, held at 1717 H St. N.W., Washington, DC ws convened at 8:30 a.m., %ursday, Oc
%e 234th meeting of 1979.
Mr. Iewis was not present (Note:
For list of attendees, see Appendix I.
on Friday, October 5, 1979.]
d We Chairman noted the existence of the published agenda for th (FACA) and the Government in the the items to be discussed.
ance with the Federal Advisory Committee Act He noted that Sunshine Act (GISA), Public Laws92-463 and 94-409, respectively.
no requests had been received from members of the pub portions of the meeting would be available in the NRC's Public Docume ments.
1717 H Street N.W., Washington, DC within approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I. Chairman's Report (Open to Public)
Fraley was the Designated Federal Dnployee for this
[ Note:
Raymond F.
portion of the meeting.]
A.
Reviewers _
and Moeller as reviewers for the Mark We Chairman named Messrs.
234th ACRS Meeting.
Request from Rep. Udall for Opinion o'n a ProksedHybridStandar B.
Nuclear Power Plant Udall of a letter fran Rep. ferris K.
he Gairman noted receipt requesting the Committee's opinion regarding a proposed hybrid j
standard nuclear power plant (see Apperdix IV).
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Request from Commissioner Gilinsky for Comments Regarding Con C.
of Xenon Following an Accident of a memorandtn from Commissioner l
We Gairman noted the receipt the Committee's comments regarding the control Gilinsky requestingof xenon followire an accident at a nuclear power p dix V).
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MINIITES OF WE 234W ACRS MEETIN3 OCIVBER 4-5, 1979 D.
Request from Commissioner Bradford Regarding Identification of NRC Regulations h at May Need Changing he Chairman noted receipt of a memorandum'from Comissioner Bradford regardirq identification of NRC regulations that may need changing and the adequacy of the regulatory process for implementation of ACRS recomendations in this area.
The Committee deferred until the 235th ACRS Meetirg (November) answerirg this memorandum.
E.
Program Plan for the Investigation of Vent-Filtered Containment Conceptual Design for Light-Water Reactors The Chairman noted receipt from the NRC Staff of a Sandia draft
- report, SAND 79-1088, Program Plan for the Investigation of Vent-Filtered Containment Conceptual Design for Light Water Reactors, with an accompanyirg request for Comittee coment on the report.
It was the consensus of the Committee that the vent-filtered containment I
design is of sufficient importance that the Comittee should review the matter in' detail; therefore, the matter was deferred to a future meeting when adequate time could be devoted to it.
II. Meeting With Members of the NRC Staff Re. September 25, 1979 Transient at North Anna Power Station Unit 1 (Open to Public)
(Note:
Ragnvald Muller was the Designated Federal Employee for this portion of the meeting.]
(For background material, see Appendix VII.)
M. Wilber, NRC Staff, briefed the Comittee on the sequence of events that occurred during the Septenber 25, 1979 transient at North Anna l
l Power Station Unit 1 (see Appendix VIII).
i In answer to a question, M. Wilber agreed to provide the Comittee with information regarding the level of readings of radiation instru-j ments in the auxiliary building, and the degree of margin that was left on these instruments.
He noted that the NRC Staff doer not expect the Licensee to prepare an Abnonnal Operating Report.
III. Executive Sessions (Open to Public)
[ Note:
James M. Jacobs was the Designated Federal mployee for this portion of the meeting.]
A.
Subcomittee Reports 1.
Priorities for Research Projects Mr. Siess, Subcomittee to-Chairman, discussed the "first-cut" on priorities for research projects as suggested by members of the ACRS Staff and ACRS Fellows.
he Committee agreed that it 2
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MINtfrES OF WE 234W ACRS MEETING OC1OBER 4-5,1979 I
should attempt to set priorities for the researen projects and include these priorities in the forthcoming annual report.
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It was also the Comittee's consensus that the "first-cut" as presented, did not provide an adequate spread of priorities by l
which decisions could be made.
Mr. Okrent recomended that the appropriate subcomittee chair-man suggest priorities for those projects in their own areas using the same format as the "first-cut", and if they can, also stegest priorities for other areas. Rese stqgested priorities should be available for the November 6, 1979 Reactor Safety Subecmmittee Meeting, or they can be forwarded to Mr. Seiss at home within one week after the 236th ACRS Meeting (December).
In addition, the appropriate subcommittee chairmen should submit drafts of their respective chapters for the annual RES report to Mr. Seiss at home.
2.
1NI-2 Accident Bulletins and Orders Mr. Mathis, Subcommittee Chairman, noted that the Subcomittee j
had met on October 2, and that he had invited members of the NRC Staff to rediscuss the items covered at the subcommittee meetirg with the full Comittee.
D. Ross, MC Staff, discussed the progress the 1MI-2 Bulletins and Orders Task Force is making with the reactor vendors in resolving the issues identified as a result of the 1MI-2 acci-dent.
He said that with respect to boilirg water reactors, he believes good progress is being made. Both system analyses and guidelines have been received from General Electric, and it is hoped that the NRC Staff review will be completed by November.
It is intended that specific plant procedures will be reviewed by January 1980. With respect to Westinghouse plants, he said that the final draft of the review of the Westinghouse response is underway, and publication is expected by late October.
He said that the Westinghouse owners group report is also near publication.
The NRC Staff's upgraded auxiliary feedwater requirements are not complete. %e NRC Staff hopes to have all of its requirements out to Westinghouse and Westinghouse-plant owners within a week.
Following the review of the response to the requirements, the NRC Staff will carry out a plant by plant review. With respect to Combustion Engineering plants, the work 1s approximately two to four weeks behind that on Westinghouse plants.
We EC Staff expects to have its operator training guidelines completed by late Decenber 1979, or early January 1980.
By the end of October, all plant specific auxiliary l
feedwater requirements will be published.
He noted that there are still some open items:
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MINtfrES OF THE 234DI ACRS MEETING OCTOBER 4-5, 1979 e small break ILCA analysis, o power operated relief valve analysis e loss of all feedwater analysis, e loss of main coolant punp, and e analysis of the effect of break isolation.
Other open items include requirements for tripping reactor coolant pumps following activation of high pressure coolant injection from a low pressure signal, and when the action of the high pressure coolant injection system should be terminated.
D. Ross noted that the bulletins issued by the imC Staff are advisory in nature only, and are not really enforceable.
Mr. Plesset noted his opinion that Appendix K calculations may not be conservative for small break IDCAs.
Mr. Ebersole pointed out that there is greater reliability of auxiliary feedwater systems in BWRs than PWRs after system blow-down has taken place.
Mr. Shewmon requested that the ACRS Staff determine what is being done to assure the automatic operation (improve reliabil-ity) of power-operated relief valves (PORVs).
With respect to Babcock and Wilcox plants, D. Ross said that t
the short-term orders are currently complete, and industry has reviewed them.
Subnissions from the first B&W plant, Rancho Seco, are being reviewed and if they are found to be acceptable, will be used as a guide for other B&W plants. The IRC Staff is considering carrying out a study of plant stability.
D. Ross said that the NRC Staff is starting to work with all plant vendors to determine or improve the adequacy of core cooling under abnormal conditions.
I Mr. Mathis noted that the next subconnittee meeting will take place at the Westinghouse lead plant (perhaps Salen) in late November or December.
)
Mr. Ebersole requested that the NRC Staff find out whether the l
21/2 in. control roonM>perated stean dunp valve provided on I
the pressurizer of Arkansas Nuclear One Unit 2 is unique to that plant, or is provided on all late-nodel Combustion Engineering plants.
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MINtTTES CF 'lHE 234'IH ACRS MEETING O m m 4-5, 1979 3.
TMI-2 Accident Implications Mr. Okrent, Subconsnittee mairman, noted that a planned review of Diablo Canyon, Zimer, and Westirghouse Ice Condenser plants scheduled for Subcomittee review at the October 3 Subcomittee meeting was deferred because the NRC Staff was not then ready to discuss these icsues.
In place of these planned discussions, the Subcommittee considered 'IMI-2 implications on BWRs and seismic matters, briefly, and considered in detail long-term lessons learned. We NRC Staff's Probabilistic Assessnent Staff
(
discussed hydrogen control, and the effects on MSH-1400 proba-bilities if hydrogen is controlled. It was concluded that there would not be much gain in risk reduction.
It is believed that controlled filter-vented contairunent might provide a big gain in risk reduction. Wis matter will be considered in detail at the next Subcomittee meeting. Were was some discussion of proba-bility of Class-9 Accidents. Short-term recomendations include hydrogen control in containment for all IMRs, and consideration of provision for containment filtered-venting for the relief of overpresssure. The Subcommittee also briefly discussed the concept of a single standard nuclear plant design.
Mr. Mark requested that the ACRS Staff obtain information on the time-history of the pressure spike recorded during the TIM-2 accident, location of the pressure sensors vis a vis the hydrogen " explosion" etc., to help develop a better understandirg of the nature of the event.
4.
Wolf Creek Mr. Etherington, Subcomittee Gairman, noted that the Subcom-mittee's attention has been directed to the seismic design basis for Wolf Creek by a letter to the Commissioners from the Mid-America Coalition for Energy Alternatives (see Appendix IX). He said that a report has been received frczn J. C. Maxwell, ACRS consultant, on this matter (see Appendix X).
We Committee agreed to defer further action regarding the con-dition of the concrete base mat of the Wolf Creek Generatirg Station until the NRC Staff Safety Evaluation Report is avail-able. %e ACRS Staff was directed to obtain the services of Dr.
Pomeroy, ACRS consultant, to further evaluate the adequacy of the seismic design of Wolf Creek.
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OC'ICBER 4-5, 1979 MINtfrES OF WE 234U1 ACRS MEETING B.
Member Requests nf NRC Staff Mr. Kerr recommended that the Members check with the ' INK:hnical Secretary at the end of each meeting to determine whether the requests that they had mada during the meetig are both pertinent and clear. Wese requests would then be forwarded to the IRC Staff in writing.
C.
Review of Regulatorv Function and Process he Comittee continued its discussions regarding the NRC regula-tory function and process that was begun at the 233rd ACRS Meetirq.
Mr. Bender agreed to rewrite the draft report on the regulatory process with assignments to rewrite specific chapters as follows:
e Chapter 10, NRC Spokesman, Mr. Okrent, o Chapter 14, Legal Influences, Messrs. Bender and Okrent, independently, and e Chapter.17, Emergency Control and Radionuclide Cleanup, Mr. Moeller.
D.
Committee Position Regarding Proposed Plants for Which a Favorable ACRS CP Report Has Been Written, but for Which a CP Has Not Been Issued Mr. Okrent recomended that the Committee defer taking a position regarding a further review of those plants for which a favorable ACRS CP Report has been written, but for Wich a CP has not been issued, until the position of the NRC Staff has been discussed.
(A listing of the plants so affected is contained in Appendix XI.)
E.
Conduct of Members 1.
Mr. Carbon Mr. Carbon informed the Comittee that he has been requested to be interviewed by the firm of Yankelovich, Skelly, and Wite, Inc. on the general subject of the status of energy issues, not on nuclear safety.
%e Committee offered no objection to this interview.
F.
Future Schedule 1.
Agenda I
We Comittee agreed on a tentative agenda for the 235th AGS Meetiry (November) see Appendix II).
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OCitBER 4-5, 1979 MINUTES OF WE 234W ACRS MEETING 2.
Subcomittee Activities A schedule of future subcomittee activities was provided the Members (see Appendix III).
G.
ACRS Reports and Letters 1.
Hybrid Standard Nuclear Plant Design We Comittee considered the concept proposed by Rep. Morris K.
in his letter of September 13, 1979 (see Appendix IV)
Udall regarding a hybrid standard nuclear plant design, and prepared a reply to Rep. Udall (see Appendix XII).
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2.
Systematic Evaluation Program We Comittee completed its review of the Systematic Evaluation Program begun during the 233rd ACRS Meeting (September), and prepared a report to the Comissioners regardirg the scope and pace of this program (see Appendix XIII).
3.
Xenon Emission Control We Comittee responded to an inquiry fran Commissioner Gilinsky (see Appendix V) regarding control of xenon emissions which may occur during nuclear plant accidents (see Appendix XIV).
Systems Interactions Study for Indian Point Nuclear Generating 4.
Station Unit 3 The Comittee prepared a memorandum to the Executive Director for Operations regarding the systems interactions sttz!y reques-ted for the Indian Point Nuclear Generating Station Unit 3 in the Committee's report of July 13,1978 (see Appendix XV).
234th ACRS Meetirg was adjourned at 8:05 p.m., Friday, October 5,1979.
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234th ACRS Meeting Meeting Dates:
October 4-5, 1979 APPENDIX I ATrDEES ADVISORY COMMITTEE ON REACTOR SAFH3UARDS Max W. Carbon, Chairman Milton S. Plesset, Vice-Chairman Myer Bender Jesse C. Ebersole Harold Etherington William Kerr Stephen Lawroski Harold W. Lewis J. Carson Mark William M. Mathis Dade W. Moeller David Okrent Jeremiah J. Ray Paul G. Shewmon Chester P. Siess ACRS STAFF Raymond F. Fraley, Executive Director Marvin C. Gaske, Assistant Executive Director James M. Jacobs, Technical Secretary Herman Alderman John H. Austin Andrew L. Bates David E. Bessette John Bickel Paul A. Boehnert Sam Duraiswamy Elpidio G. Igne David H. Johnson William Kastenberg Morton W. Libarkin Richard K. Major Thomas G. McCreless John C. McKinley Robert E. McKinney Ragnwald Muller Gary R. Quittschreiber Jean A. Robinette Richard P. Savio John Stampelos Peter Tam Hugh E. Voress Harold Walker Gary Young Dorothy Zukor A-i
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NRC ATTENDEES
October 4, 1979 4
i Nuclear Regulatory Research Nuclear Reactor Regulation Inspection and Enforcement l
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A. Schwencer Howard Wilber E. L.. Jordan i
P. J. Morrill
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PUBLIC ATTENDEES 234TH ACRS MTG.
October 4, 1979 S. R. Phelyn, EEI R. A. DeLorenzo, WPPSS T. D. Martin, NUTECH R. H. Leyse, EPRI R. P. Smith, McGraw-Hill R. Borsum, B&W S. Harris, EEI Kunihiu Ota KEPC0 J. Ahledas, VEPC0 W. Harrell, Va. Electric & Power M. Lord, Newsweek F. Stetson, AIF A. Sasehico, McGraw-Hill K. Layer, BBR J. Schermbeck, Self S. Lewis, McGraw Hill
10 4-3
i i
i PUBLIC ATTENDEES
October 5, 1979 R. E. Schaffstall, KMC, Inc., Reston, Va.
Saul Harris, Edison Electric Inst., Landover, Md.
Roger P. Smith, McGraw-Hill, Freehold, NJ Kunihiro Ota, KEPCO,1725 K St., NW, Wash., DC R. Borsum, B&W, Derwood, Md.
Suzanne R. Phelps, EEI, McLean, VA R. H. Leyse, EPRI, Rockville, MD Paul E. Knapp, TVA, Muscle Shoals, AL
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APPENDIX II DRAFT #2:RFF 235th ACRS MEETING NOVEMBER 8-10, 1979 O
V
~
TKJRSDAY 8:30 - 8:45 Ch:f.r.:n s Spo rt (MWC )
8:45 - 12:30 Regulatory Process (MB) 12:30 - 1:30 Lunch 1:30 - 4:30 Diablo 4: 30 - 6:30 Ice Condensors 6:30 - 7:30 Resolution of ACRS Generic Items FRID4Y 8:30 - 12:30 Regulatory Process (ME) 12:30 - 1:30 Lunch 1:30 - 2:00 Review of Task Action Plans (S. Hanauer) 2:00 - 2:45 Westinghouse / Salem 1 - steamline break (Report by NRC Staff) 2:45 - 3:45 Regulatory Guide 1.97 (CPS) 3:45 - 6:45 NUREG-0630 (HE) 6:45 - 7:15 SC Report on Siting Policy Task Force SATURO* Y - 8: 30 4:00 8:30 - 10:00 Regulatory Process (ME) 10:00 - 10:45 Diablo 10: 45 - 11:45 NUREG-0600 11:45 - 12:00 Regulatory Guides 12:00 - 1:00 Lunch 1:00 - 2:00 Letter to Bradford R:00 - 3:00 SC Reports on Long Term Lessons Learned /Kemeni Report 3:00 - 3:15 ATWS 3:15 - 3:45 Wol f Creek l
3:45 - 4:00 Miscellaneous L
1
007 0 3 1979 ENCLOSURE ACRS FUTURE AGENDA ACRS MEETING TYPE OF REACTOR SER ISSUE PROJECT REVIEW VENDOR DATE November None l
December 4
Floating Nuclear ML W
11/1/79 Plant l
GETR Restart after GE 9/27/79 show cause order Geology and Seismology only J anuary THI-l Restart After B&W 12/3/79 Refueling t
F ebrua ry LaSalle 1 & 2 OL GE 1/2/80 San Onofre OL CE 1/2/80 i
Watts Bar OL W
1/2/80 March i
None l
1 O
+$
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UNITID STATES E ','.. / #[ g NUCLEAR REGULATORY COMMISSION a
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, ADVISORY CO'AMITTEE ON '4E ACTOR S AF EGUARDS t '!4 waswincion -2 c.2osss
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. APPENDIX III vctober 5,19h9 ACP.S Me:bcrs
,SO!EDL1E OF ACRS St~r,CO:M ITEE PEETINGS,A';D TOUP.S The follo ing is a list of tourc and Subco: ittee rectin;s cur-rently scheduled, subject to the approval of the Advisory Coe-uittec."an:gement Officer.
If you are listed and cannet attend a =eeting, or if you are not listed but vould like to attend, please advise the ACRS Office as soch as possible.
Most hotels currently being used by ACP.S Me=bers in the de n-to..m ',.'athin, ton and Ecthesda arcas require a guarantecd reser-
~ vatien if arrivcl is scheduled after 6:00 p.:.
Ta11ure to use a roo: under these conditions involves forfeiture of the cost.
Plea u advise the ACP.S Office as soon as possible if you cannet attend a =ectin; for which you are scheduled so that reservc-tions can be cancelled in ti e to avoid this.
e l Y, M. W. Libarkin Assistant Executive Directo:
for Project Keview cc: ACP.S Technical Staff M. E. Vanderholt B. Dundr R. F. Traley I
e M. C. Cas1:e 3'
9 w
J. Jacobs
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OCTOBER O
16-17 Rad. Effects & Site Eval. (RM - DM, JE*, 50 17-18 ECCS (AB) - MP, JE*, HE 26 Lacrosse BWR (JCM/PT) - WK, SL 30 TMI NUREG-0600 (RM) - HE, MC, JE, *, MP 31 Waste Maaagement (PT) - SL, DWM, WK, JCM, JR, WM NOVEMBER 5
TMI-2 (accident) Implications re Nuclear Power Plant Desig'n (RM4) - 00 W 5
Metal Comp. (EI) - PGS, MB, HE 6(a.m.) & 7(p.m. )RSR TGM) - D0, HE, SL, CM, DM, MP, CS, WK(11/7 only) 6(p.m. ) & 7(a.m. ) Reliability & Prob. Assessment (GRQ) - D0, JCM, MP, WK, HL 7
Reg. Act. - (50) - CS, HE, JR, MB, JE O
8-to 235th acas seet4#9 14 GE Test Reactor (San Francisco, CA) (EI) - WK, MB, 00 15-16 Extreme External Phen. (Los Angeles, CA) (RS/TGM) - 00, MC, HE, JCM, WM 16 Fluid Dyn. (San Francisco, CA) (AB/SD) - MP, HE 17 FNP (Los Angeles, CA) (GRQ) - DWM, HE, 00, SL, MP, PS, WM 29-30 Advanced Reactors (Albuquerque, NM) (RS/TGM) - WK, MC, CM, MP, PS
- Note dual schedule for 10/17 19 - 2
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l DECEMBER
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4 TMI-2 Implications (Tent.)
.4 Reliab. & Prob. Assess. (GRQ) - D0, f2, JE, HL, JCM 5
Reg. Act (SD) - fIS, HE, JR.
6-8 236th ACRS Meeting 13 Power & Elect. Syst. (GRQ) - WK, JE, JCM, JR
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JANUARY I
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RSR (TGM) - 00, CPS, HE, WK, SL, JCM, MP, PGS i
10-12 237th ACRS Meeting l
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APPENDIX IV
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REQUEST FROM REP. UDALL FOR ACRS O.:'.'" '**:.'.**.F COMMENTS ON PROPOSED COMPOSITE STANDAR
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HYBRID NUCLEAR POWER PLANTS
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Chairman Dr. Max Carbon, Advisory Committee on Reactor Safeguards Nuclear Regulatory Commission Washington, D.C.
20555
Dear Dr. Carbon:
Three Mile Island raises questions as to The accident at design is so whether the diversrty in nuclear power plant as to overwhelm the ability of those involved in the In regulatory process to sufficiently analyze each plant.each plant has it great g) addition, it appears that strengths and weaknesses with no one design as safe as it
("
would be were it to include certain features from other in some circumstances the desire to keep designs.
- Also, certain costs down seems to have led to decisions to omit features that would increase the safety of a particular reactor.
In order to reduce safety hazards resulting from diversity features in design and failure to incorporate in each plant the that maxim.:e safety, I am considering recommending that be amended to required that after some Atomic Energy Act no construction permit be issued (e.g. January 1, 1985) for plants designed in accordance with a general date Such a design developed by the DOE and approved by the NRC.
except design might be a hybrid of existing light water reactor but would in any designs (or it might be a gas reactor) event be based on existing technology, and would not require An essential element significant research and development.
of the basic design would be that changes could be incorporated into it-as necessary.
6 4-n
41[
Dr. Max Carbon Page 2 I would appreciate the views of the ACRS on this subject.
(1)
What would be the advantages and dis-In particular:
advantages of a regulatory system in which the only plants eligible for licensing would be those built in accordance (2)
To what extent would such a with the DOE design?
design reduce the number of items on the NRC's list of high (3)
How much priority unresolved generic safety issues? time would be required to s of a hybrid of current designs?
the Subcommittee might have the benefit of In order that the Committee's comments when we consider amendments to the Atomic Energy Act, I would appreciate receiving your response by October 31.
Thank you for your assistance.
Sincerely, t.
.r. xh.L J.-
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/d?.RIS K. UDALL w
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4 The Honorable John M.
Deutch Under Secretary, Departmcnt of Energy Uashington, D.C.
20585
Dear Mr. Secretary:
The accident at Three Mile Island raises questions as to uhether the diversity of power reactor design is so great as to overwhelm the ability of the regulatory process to sufficiently analyze each nuclear poper reactor.
In i
addition, it appears that each particular reactor has its own peculiar strengths and weaknest.es with no one design j
as safe as it would be were it to include certain features O
from other designs.
Also, in some circumstances the desire to keep costs down seems to have led to decisions to omit certain features that would increase the safety of a particular reactor.
In order to reduce safety ha:ards resulting from diversity of design and failure to incorporate in cach plant features that maximize safety, I am considering recc = cnding that the Atomic L'ncrgy Act be amended to require that af ter some date (e.g. January 1, 1985) no construction permit be issued cncept for plants desigr.ed in accordance with a general design developed by the DOE anf appreved by the URC.
Such a design might be z. hybrid of e::isting light water reactor designs (or it might be a gas reactor) but would in any event be based on existing technology, and would not require significant resesrch ar.d development.
An essential element of the basic design vould be that changes could be incorporated into it as necessary.
S b/A Ir
_ _ _.. ~. _ _
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Tne !!onorable John M. Deutch i
I would appreciate the views of the Department of Energy l
on this subject.
In particular:
(1) What would be the advantages and' disadvantages of a. regulatory system in which the only reactors eligible for licensing would be those built in accordance with the DOE design? (2) To what extent would such a design reduce the number of items on the NRC's-list of high priority unresolved generic safety issues?
( 3.) Is there a group within a DOE laboratory qualified to produce such a design? (4) Hou much time would be required to specify an optimum design consisting.of a m
hybrid of current light water reactor designs?
In order that,the Subcommittee might have available the-DOE views when we consider amendments to the Atonic' f
Energy Act, I would appreciate your response by 2
October ~31.
1 c; Thank you for your assistance.
Sincerely,
,l$!' h N l
HORRIS'$. Y
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The' Honorable Joseph Hendrie Chairman, Nuclear Regulatory Commission Washington, D.C.
20555
Dear Chairman Hendrie:
The accident at Three Mile Island raises questions as to whether the diversity in nuclear power plant. design is so great as to overwhelm the ability of those involved in the regulatory process to sufficiently analyze each plant.
In addition, it appears that each plant has its own peculiar ~
strengths and weaknesses with.no one design as safe as it Q
would be vere it to include certain features from other
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V designs.
Also,'in some circumstances the desire to keep costs down seems to have led to decisions to omit certain features that would increase the safety of a p, articular reactor.
In order to reduce safety hazards resulting from diversity in design and failure to incorporate in each plant features that maximize safety, I am considering recommending that the Atomic Energy Act be amended to required that after some date (e.g. January 1,1985) no construction permit be issued except for plants designed in accordance with a general design develo".ed by the DOE and approved by the NRC.
Such a design might be a hybrid of existing light water reactor designs (or it might be a gas reactor) but would in any event be based on existing technology, and would not. require significant research and development. An essential element of the basic design would be that changes could be incorporated into it as necessary.
j L2 U LA L3 L2 L3 Wy C.
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The Honorable Joseph Hendrie Page 2 I would appreciate the Commission's views on this subject.
In particular:
(1)
What would be the advantages and j
disadvantages of a regulatory system in which the only 1
plar.ts eligible for licensing would be those built in l
accordance with the DOE design?
(2)
To what extent would j
such a design reduce the number of items on the NRC's j
list of high priority unresolved ger.aric safety issues?
j (3)
How much time would be required to specify an optimum 1
design consisting of a hybrid of current designs?
In order that the Subcommittee might have the benefit of the Commission's comments when we consider amendments to the Atomic Energy Act, I would appreciate receiving your i
response by October 31.
Thank you for your assistance.
Sincerely, A*
o MORRIS K. UDALL Chairman t
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APPENDIX V REQUEST FROM COMMISSIONER GILINSKY FOR COMMENTS REGARDING CONTROL OF XENON i
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FOLLOWING AN ACCIDENT
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9, NUCLEAR REGULATORY COMMISSION f,,,,f WASHINGTON, C. C. 205$5 N
September 12, 1979 OFFICE oF THE COMMISSION E R ELCitytv ADY150RT COMMIr!IL o 3 ^"3 U 3 "
Dr. Max W. Carbon Chairman Advisory Committee StP 1219/9 on Reactor Safe-gg g
U.S.NuclearRegu!uards 0 lI I2 l I4d 'I 5d, atory "it 18 91 lI 1l i 8 Com=1sslon y
Washington, D.C.
20555
Dear Dr. Carbon:
I thought you might be interested in the attached report on r.enon emission from the DII-2 nuclear plant.
I would be interested in any comments you have,especially on the O
las'. part of the report which discusses a v
scheme for reducing xenon emissions after accidents.
Sincerely, s
(}
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Victor Gilinsky
Attachment:
Ltr to Gilinsky fm Pollack dtd 9/5/79
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MICHIGAN STATE UNIVERSITY im ec ecce
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September 5,1979 The Honorable Victor Gilinsky
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Commissioner U
U.S. huclear Regulatory Com.ission g!
%s 1717 H Street, N.W.
Washington, D.C.
20555
Dear Dr. Gilinsky:
Here is my report in response to the questions you asked me to look into l3 concerning Xe emissions from the accident at the Three Mile Island Nuclear Station.
In order to get this information I had discussions with many scientists and engineers mostly of NRC, but also with others te whom I was refer, red.
I also read relevant NRC reports and other related material.
What I've tried to do is to synthesi:e this information into a coherent picture cf the problem of I33 Xe emission.
Tne report is ordered under three headings each of which is a question or a few questions.
I. How much Xe was emitted as a result of the TMI accident? How are the 133 estimates of Xe emission made? How reliable are the estimates?
Tne direct answer to these questions is that at least three independent I33 estimates of total Xe emissions have' been made by people. The estimates are 6
6 6
2 x 10, 2.4 x 10, and 13 x 10 Ci.
The estimates are rough ones at best, i.e.
they make use of rough approximations, and are probably only reliable to within
~
factors of two or three.
However, they represent. responsible sensible work and are probably the best estimates one can make under the circumstances.
Backcround_:
Xenon-133 is produced in the fuel of a reactor by fission of 235 U
during normal operation.
The TMI reactor had run for 140 equivalent full 133 power days at the time of the accident and the total amount of Xe in the g- / )
I 6
inventory at that time was 140 x 10 Ci. This number comes from the origin code, O
computer ceicu,atioa of the raeioect4ve 4setones present es a rese>t of pro-duction minus decay, and is generally believed to be reliable.
If all of this l3 were ecitted at the beginning of the accident we'd get the maxir:um amount Xe 6
133 of Xe emitted, i.e. an upper bound is 140 x 10 Ci.
We have good evidence from DOE environmental samples (helicopter and ground 33 That surveys) that the main radionuclide emitted during the accident was Xe means that some Xe13~' got free from the fuel elements and into the air, probably by the pathway:
fuel elements to coolant water, then out of the reactor building into the auxiliary building, then through the filters and into the air via the stack.
During normal operation the emissions from the reactor are measured at the stack by a combination of batch spectral analysis, which measures the relative amounts of the various radionuclides, and a stack monitor, which measures the s
gross number of curies being emitted.
Unfortunately the stack monitor went eff scale about two or three hours after the accident started and it stayed off s cal e.
Had that not happened, i.e. had there been a suitably high intensity monitor in the sta:k, we could have used the stack monitor to estimate the total, emissions.
Since this can't be done we have to use more indirect, less reliable I found out about three of these and they are described briefly with means.
their results below.
A.
Estimate from measured doses and meteorolocical data.
This estimate is described in NUREG-0555 and I discussed it in some detail with Dr. F.J. Congel of the NRC Radiological Impact Section.,
Essentially what was done here was to take the doses as measured by thermor luminescent dosimeters in place before and during the accident and to combine l
these doses with meteorological dispersion factors to get estimates of the strength 6
I33 of the source of the radiation.
This gives an estimate of 13 x 10 Ci of Xe h*l
3 emitted.
O The re eosimeter eete for this est4 mate come fre= the Teleerne isotopes dosimeters in place at 20 onsite and offsite locations.
The raw meteorological data come from a tower, on the island, which reads wind magnitude and speed and the air temperature at various elevations. These kinds of data are gathered routi nely.
The meteorological data is analyzed by a meteot ology group and they calculate tables of x/Q, the meteorological dispersion fact:or.
This factor is the ratio of the steady state radioactive concentration et a point in the field
( ). in Ci/m ) to the steady state rate of emission of radioaactivity from the stack 3
source (0 in Ci/sec). Thus the ratio X/Q is kind of a response function, similar to Green's functions used in formal physics.
These factors depend in general on direction and distance from the source.
l3 probably the main source of uncertainty in this estirw.te of Xe emission is due to meteorological factors.
In general values of 1K. are thought to be pV good to a factor of two or three out to a few miles; beyond that the accuracy is worse.
At TMI the uncertainty was fairly large due to Tight winds.
Another pocential source of uncertainty is in detemining the doses; to which the TLD's have been exposed.
However the National Bureau of Standards has made a preliminary study cf tne calibrations and that appears to be 0.K. (i.e. tne uncerta.inty is about +25% and -30t).
6 l33 My conclusion is that this value (13 x 10 Ci of totali Xe emission) may be off by a factor of two or three mainly due to meteorological uncertainties.
B.
Estimate from delayed crab samoles.
This estimate was made by Dr. Andrew P. Hull of Brookhaven National Labora-l tory and the work was reported at a recent Health Physics society Meeting.
133 He obtained data on the actual Xe concentrations in about five grab samples of tne effluent taken at the stack.
The samples were taken between April 6-10 several days after the accident, which was on March 28.
He plotted these data as
~l
d i
l a function of time and extrapolated back to t = 0 using a straight line fit.
He 5
O aea su m ee #9 tae tot 1 xe' " emiss4oa eae so:
the va,ue 2 x 20 Ci.
The main strength of this method is that it relies on direct measuremen*a of Xe coming from the stack.
The main weaknesses are: first, as mentioned above, 13 the data wePe taken well after the accident and second, there is probably no real reason to believe that a straight line extrapolation back to the time origin is an accurate representation of what actually happened.
C.
Estimate from a remote monitor.
This esthate was described to me by Dr. Carol D. Berger, a health physicist at Oak Ridge flational Laboratory.
It is independent of the previous estimates 6
133 and gave a value of L.4 x 10 Ci of Xe emitted.
Dr. Berger's technique was to use readings from a monitor, in tne auxiliary.
Tnis building, tnat saw fro: a distance of 40 feet a duct tnat feeds the stack.
monitor saw part of what went up the stack all the time and it never went off scale, as did the main stack monitor.
She calibrated the remote monitor against the main monitor during the short time that the main monitor wts still on scale.
Associated with this calibration is an uncertainty of a factor of two, but there are other indications that the. Calibration may have been much better than that.
There also is some uncertainty introduced into this method by the necessity cf This distribu-assuming the distribution of radioactive gases in the effluent.
tion was taken fro: the inventory at the moment of shutdown, a reasonable choice.
In some ways this is, in principle, the most reliable of the available estimates.
I learned in the course of this work that Keith Woodard, a consultant D.
of the utility, estimated the total Xe emission to be 15 - 20 x 10 Ci. I[-
l33 6
unfortunately do not know what information he used as a basis for his estimate.
Cou,e we determine tae t=tel xe' 3 e=ission from Txt ia =ther wars:
O II.
I know of two other ways which in principle mignt be used to deterTnine
5-d b
the Xe emission.
One way depends on conservation of Xe and the other depends 13 1 Xe which is adsorbed on the filters that precede the stack.
I33 133 A.
Determinine total Xe emitted by conservation of Xe We know frc:r. the core inventory at tne time of the accident that there were l32 initially 14D x 10 Ci of Xe present.
If no more fission or production.of Xe 6
133 133 occurred after this time, then the total amount of Xe present at a time t.t 133 (days) after the accident is 140 x 106,et Ci, where 1 = 0.13153 day (Xe has a half-life of 5.27 days).
The idea is to measure the total Xe still in the reactor (containment volume, core, coolant water) and cmr. par e this with the Whatever is missing has escaped out of the reactor into the environment.
above.
The difficulties with making this work, are:
(a)
Xe has a short half-life se that measurements must be made shortly af ter the accident before it has decayed away to subthreshold amounts; even then it is a dif'icul t measurement, and 13"2 (b)
It is necessary to measure radiation from Xe within the reactor where there is so much other radiation present.
6 13 As a practical example:
Suppose for illustration that 20 x.10 Ci of Xe had 6
3 Ci of Xe escaped at the time of the accident and we were thus left with 120 x 10 contained on March 28.
By August 20, wnen the present study was, undertaken, that would have decayed to 120 x 106,W 45 0.53 Ci still contained witnin the reacter if no more had escaped.
This means, in order to detect. whet.her any Xe escaped in-
^
133 itially it would be necessary,by August 20,to measure the total Xe present to better than about + 0.1 Ci out of a total of 0.6 Ci all against the large back-I ground of other radionuclides in the reactor.
That's probably rasch too hard to dc.
There is an interesting application of this principle to the TMI accident, and 133 an associated puzzle.
On June 29 NRC engineers measured a Xe concentration,cf The volume of this 1.5 x 10-2 pCi/ml in the gas space of the containmer.t volume.
vessel O
sas space is 2 x 10' cubic feet the voiume of weter i# the cea=>ic=es:
is 525,000 gallons, the solubility of Xe in water is about 0.1, 79-a /
J-1 a
and t.t between 3/28 and 6/29 is 93 days.
From these data I calculate that on l33
'29 there were present about 850 Ci of Xe in the gas containment volume, and a negligible 3 C1 of Xe in the water volume.
This means that, if we assume l33 escaped or was added to the containment building after the accident, l33 that no Xe there were about 850 e 3 = 170 x 106 I33 Ci of Xe present initially. That l33 implies that essentially all of the Xe escaped from tne fuel into the contain-ment building.
In fact 170 x 10 Ci is even somewhat larger than the origin code 6
The pu :le is that working from this same data, NRC engineers have con-v alue.
I do not under-ciuded that half the noble gas was released from the fuel rods.
stand the discrepancy between our conclusions.
l33 5.
Deteminine total Xe emitted by analysis of filters.
The idea nere is that all of the Xe that comes out of the stack first passes througn HE A filters and charcoal filters. The charcoal filters are 2" The q thick beds of carbon and some of the Xe remains on the charcoal filters.
b mechanism for tnis is probably surface adsorption of Xe and subse:;uent diffusion into the bulk carbon.
In principle it might be possible to tell something about how much Xe l3 passed througn the filters from measurements of how much Xe they retained.
However I believe that this cannot with oresent knowledge tell us anything ;uanti-tatively useful about the Xe emission. The difficulty is that we would have to understand the interaction between Xe and carbon very well, then we'd have Did to know the entire history of the filters since they were exposed to Xe:
they ever get wet? Did f resh air blow over them? Did they ever get warn?
etc.
1**
These are the kinds of things that determine the connection between how much Xe ~~
the filters contain now and how much Xe passed through them.
I understand from Mr. John T. Collins that some of his group of engineers O
are working with the charcoal filters from the TMI - 2 reactor.
h 2
3 emissions and doses will be measured III. What can be done so that Xe e
O e'4 >1x 4e ex <#:2re ecc4eents?
In retrospect the best single way to insure that these emissions are reliably monitored is to install a stack monitor that will not go off scale during high This may require a series of monitors which work from radioa:tivity emissions.
low concentrations to high ones yith overlapping ranges.
There is a closely related question: How can the population doses due.to Xe emissions be measured more accurately? This can probably most conveniently I33 be done by increasing the number of TLD's that are routinely in place, by more dense distribution of their locations, and by using TLD's tnat are well calibrated 3
for the ga ma (81 kev) and beta (350 kev) emissions of Xe How can Xe "
emission be prevented in any future accident?
IV.
inis is difficult to answer optimally sin:e the pathway for the emission isn't yet clear.
iiowever, the following may work: Xe has boiling point and O
triple point temperatures of,respe:tively,165 K and 161 K.
Probably if one previoed surfaces which were cooled to liquid nitrogen temperatures and over had to pass in close proximity, tnen a large fraction of the Xe U3 whi:h the Xe The same kind :f condensing plates
- culd De condensec and held until it de:ayed.
would probably work witn Kr whose boiling point and triple point temperatures 0
There are h few practical problems in i
are,respectively,120 K and 116 K.
For example it would implementing this but the method should be straightforward.
33 emissions were high probably suffice just to cool charcoal filters when the Xe f
and let the Xe be condensed and retained on them.
Report submitted by
[
~
i Gerald L. Pollack O
er=fessor, ehys4cs oepartment
[$2 Clap %
UNITED STATES o
[ , s.,
g NUCLEAR REGULATORY COMMISSION E
W ASH 4N CTON, D.C. 20655
%....+#
September 28, 1979 OFFICE OF THE APPENDIX VI cOMSSIONE R REQUEST FROM COMMISSIONER BRADFORD RE IDENTIFICATION OF NRC REGULATIONS WHICH NEED CHANGES 11El10 TO: Max W. Carbon, Chairman Advisory Committee on Reacto afeguards FRQ4:
Peter A. Bradf ord
SUBJECT:
IDENTIFICATION OF NRC REGULATIONS WHICH NEED CHANGES In an April 20, 1978 memorandum, the Commission requested the ACRS 'and the staff to take certain actions to implement the recomendations of the report, " Follow-up on ACRS Letters." The Commission later considered the need for additional procedural guidelines relative to Commission involvement in ACRS advice items on which Commission action might be appropriate. The guidance drafted by the Office of Policy Evaluation included the following item:
2.
"When ACRS advice indicates a need or desirability of changes in regulations, or in any procedures requiring Commission-level
('
consideration, the matter should be brought to the Commission's timely attention, together with any staff views and recommendations with respect to appropriate Commission action."
The Commission decided that specific guidance was unnecessary, in part because the staff has to identify any significant regulatory changes for Commission approval.
Recently, however, the question has been raised whether the lack of a formal procedure for obtaining ACRS views on NRC regulations needing changes has inhibited Committee recommendations in this area. Therefore, I would appreciate the Advisory Committee on Reactor Safeguards' views on whether or not the lack of a specific procedure for identifying rules and regulations which need revision has inhibited the Comittee.
In addition, please identify any rules and regulations which you believe need to be addressed promptly in order to ensure public health and safety.
cc:
Chairman Hendrie Commissioner Gilinsky Commissioner Kennedy Commissioner Ahearne Lee V. Gossick Samuel J. Chilk Al Kenneke Len Bickwit G
e*
APPENDIX VII BACKGROUND MATERIAL RE. SEPT 25, 1979 TRANSIENT AT NORTH ANNA 1 MEMORAllDUM FOR:
Chairman Hendrie Comissioner Ahearne Comissioner Dradford Corrnissioner Gilinsky Comissioner Kennedy THRU:
Lee V. Gossick, Executive Dire v for Operations s
FROM:
Harold R. Denton, Director, Office of Nuclear Reactor Regulation
SUBJECT:
TRANSIENT AT NORTH A14NA As requested by Comissioner Gilinsky, we have prepared a description (Enclosures 1 and 2) of the North Anna transient that occurred on September 25, 1979. This description is based on the findings of the IE-NRR team that went to the site on the evening of September 25; and is considered as preliminary. We have asked Vepco to provide detailed infonnation (Enclosure 3). An investigation
.1 l, by IE is in process, and the results will be provided by them when available.
O
,w Harold R. Denton! Director Office of Nuclear Tteactor Regulation
- v;;t~-
Enclosures:
.f?y),1. A Description of Events Occurring at the North Anna Q..J Nuclear Plant Unit No.1, on September 25, 1979
.4 V.
2.
Assessment of HPI Tennination Criteria from North Anna Unit No. 1. Transient 3.
Letter to Vepco dated September 28,1979 re North Anna Unit No. 1 Distribution
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// 2 C
ENCLOSURE 1 A DESCRIPTION OF EVENTS OCCURRING AT THE NORTH ANNA NUCLEAR PLANT - Villi NO. ONE AU ON SEPTEMBER 25, 1979 At 5:44 a.m. on September 25, 1979, the operators at the North Anna Nuclear plant - Unit No. I observed that the drain cooler dump valve for the fifth point feedwater heater was cycling abnormally. The cycling was believed to be due to a tube rupture inside the 5B drain cooler. The drain cooler dump valve then apparently failed closed causing extraction steam condensate to back up into the fifth point heater. This ccndensate backup caused a turbine trip by design.
An explanation for why a tube rupture inside a feedwater heater can lead to a turbine trip can ba obtained from a discussion of simplified flow diagrams.
Although the enclosed diagrams are not specifically for the North Anna Plant, they are sufficient to provide an understanding of the scenario.
Figure 1 illustrates a Simplified steam Cycle.
Steam is produced in the steam generator and sent to one high pressure and three low pressure turbines.
A majority of the steam flow produces useful work in the turbine and is condensed in the condenser. At various points in the turbine, steam is extracted to preheat the feedwater. This regenerative heating process recuces thermal shock to the steam generator and increases the overall plant efficiency. The steam from the turbine (extraction steam) preheats the feedwater in feedwater neaters (Figure 2).
The extraction steam which is condensed in the feedwater heaters flows opposite to that of the feedwater. That is, condensed steam (condensate) is pumped from the condenser through a series of feedwater heaters to each steam generator (Figure 3). At the same time, the condensed extraction steam (drains) is pumped in the opposite direction into the condenser through the same feedwater heaters. The condensate feedwater goes through the tube side of a feedwater heater while the drains run through the shell side (Figure 4). A rupture of a iy tube in the feedwater heater (or drain cooler) allows the two fluids (tondensate and drains) to mix and causes level to increase.
If level increases tec much, water backs up into the turbine resulting in failure of the turbine blading. To prevent this, the North Anna turbine was tripped on high feedwater heater level.
//- GLb
At tne time of turbine trip, the reactor was operating at 78% power with no dissolved boron in the coolant system.
The reactor tripped by an interlock (P-7) which requires reactor trip when a turbine trip occurs above 15% power (reduces the overpressure transient on sudden loss of load).
Normal plant response results in a transfer of electrical power, a reduction of coolant 0
system temperature (Tave) to its no-load setpoint (547 F), and a reduction in pressurizer level to its no-load setpoint (20%). The operators, in the process of carrying out the reactor trip "immediate actions," noted that T ave 0
was continuing to decrease below its normal value (547 F) for a trip. One operator, in scanning his panel, observed that a turbine bypass valve was indicating fully open with the other valves (seven) fully shut. This indicated a stuck open valve. The turbine bypass valves allow main stes., flow to " bypass" the turbine and pass directly into the condenser (Figure 5). These valves open on a turbine trip to " cushion" the shock of sudden loss of heat removal (the turbine). The valve that stuck open is cabable of passing about 5%
steam flow.
Since the reactor is tripped (no longer producing power), a very rapid cooldown of the primary plant occurs.
Two minutes af ter the turbine trip, the operator shut the main steam isolation valves, stopping the steam flow through the stuck open bypass valve. The rapid cooldown caused a sharp decrease in primary pressure and pressurizer level. The pressure decrease actuated the safety injection system at 1765 psig.
The net result of the actuation signal was to start an additional charging pumo and three auxiliary feedwater pumps. A manual action by the operators at the instant of safety injection was to trip the reactor coolant pumps (required by IE Bulletin 79-06C). Tripping of the reactor coolant pumps along with the addition of cold water by auxiliary feedwater pumps and charging pumps resulted in a further primary system cooldown and shrinkage.
Pressurizer level went off-scale (less than 0%) for 2 minutes and then began a rapid increase as the charging pumps restored level. NGte that the off-scale pres-surizer level does not necessarily indicate an empty pressurizer.
As pressurizer level increased, system pressure also increased. At 2335 psig the power-operated relief valve (PORV) began to cycle open and shut for about 13 minutes to maintain system pressure below the safety valve setpoint. Two important. points can be noted by examining a flow diagram of the pressurizer
/"
and its connections (Figure 6). The PORV lifted by operator compliance with the l
0-27
g.
requirements of IE Bulletin 79-06A.
Item 7.B of the bulletin requires that high pressure injection (charging pumps) remain on for 20 minutes if actuated on a low pressure condition.
(They were.) This continued flow from the charging pumps, performed by the operator in accordance with the bulletin requirements, resulted in cycling the PORV and discharging to the pressurizer relief tank inside containment.
The second point is that without reactor coolant pumps, normal pressurizer spray flow could not be generated. Without spray, nonnal pressurizer pressure control was not availabh.
After operating high pressure injection for 20 minutes, the safety injection signal was reset, injection" flow was reduced, and auxiliary pressurizer spray (normally isolated upon initiation of safety injection) was initiated to reduce system pressure.
A reactor coolant pump was restarted to restore forced circulation and the plant was being returned to a normal condition per the Long Term Ooerator Actions of Procedure 1-EP-3 (Main Steam Line Rupture). Step 4.29 of that procedure provides the steps necessary to restore charging and letdown. The subsequent sequence of events, still under investigation, led to filling of the VCT and relief from a VCT relief valve to a high level waste drain tank (Figure 8). The drain tank vents to a waste gas header, but the header connection was found to be disconnected O
with the waste tank being vented directly to atmosphere.
The VCT was therefore vented directly to atmosphere inside the auxiliary building.
Two additional observations were made while at North Anna:
1.
A small release of activity occurred within the containment when the reactor coolant pumps were started. At first it was thought that the-rupture disk on the pressurizer relief tank had been breached; however, a visual inspection showed no leakage around the relief tank.
Some leakage was observed at the pump seals. This seal leakage is believed to be the source of containment activity and may have occurred as a result of pres-sure oscillationt on the pump seals.
!O
4 2.
The temporary open-ended piping shown in Figure 8 may have been installed during construction and not removed.
Spikes in Auxiliary Building activity sd have been seen on previous occasions with the source of leakage apparently never discovered.
Further, the tempora j piping installed is typical of the type used for initial testing, not for continued operation.
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ENCLOSURE 2 P
ASSESSMENT OF HPl TERMINATION CRITERI A FROM NORTH ANNt UNIT NO.1 TRANSIENT As described in Enclosure 1, the North Anna Unit No. I turbine trip transient of September 25, 1979, involved the initiation of safety injection because of excessive system cooldown from a stuck-open condenser steam dump valve.
Because the stuck-open du9p valve was quickly isolated, the primary system 0
rapidly achieved the 50 F subcooling required by the HPI termination criteria (set forth in Bulletins79-05A & B).
The HPI was not terminated until it had been in operation for a minimum of 20 minutes as also required by the bulletin criteria. Maintaining the HPI on for 20 minutes resulted in repressurizing the primary system to the PORV setpoint, and cycling the valve open and closed 1
for a period of about 13 minutes, until the HPI was turned off.
The rate of system repressurization from HPI was also enhanced because the reactor coolant pumps were manually tripped by the operators when the reactor tripped and safety injection initiated by low pressurizer pressure. Tripping the pumps under these conditions is required by Bulletins79-05C and 79-06C to protect the core during small breaks. As discussed in Enclosure 1 normal pressurizer spray was not available to condense the steam in the pressurizer to accommodate the liquid inflow due to HPI. While this led to an earlier PORV lif ting, it may have prevented the system trom becoming water solid (and thus allowing the PORV to pass solid liquid). Lack of pressurizer spray did not allow the pressurizer steam to condense, and thus when the PORV pressure setpoint was reached, the valve opened and relieved only steam. By relieving steam rather than water, the potential for valve. damage was significantly reduced.
The staff has evaluated the PORV lifting due to HPI operation for 20 minutes (as required by the bulletins). While we agree that it is not desirable for this valve (or the safety valves as well) to lift unnecessarily, the fact that it lif ted did not put the plant in an unsafe condition. No primary coolant was released to con.ainment as a result of the valve lifting and the continued HPI flow ensured adequate core cooling.
From this transient, it was shown that requiring HPI operation for as long as 20 minutes may not be O
/)-2 7
2-necessary, and that shorter periods of operation would also be acceptable.
0 In particular, as long as a sufficient subcooling margin is maintained (50 F l
as specified in the bulletins' to assure no void formation in the primary system, HPI termination can be allowed to prevent system overfill.
I During the staff review of the proposed B&W emergency guidelines for small brea6 s, B&W was requested to reexamine the requirement for a minimum of 20 minutes of HPI due to the possibility of system repressurization.
B&W's evaluation concluded that the 20 minutes of HPI operation criterion was not needed and was eliminated from their emergency guidelines. Most C-E plants do not have HPI systems which can repressurize the primary system.
Some Westinghouse plants have HPI systems that could repressurize the primary system to the PORV setpoints. Both the Westinghouse and C-E emergency pro-cedures are under review, including their HPI termination criteria. We intend to notify operating W plants, through the owner's group, that sequences like North Anna can lead to lifting of the POP.V.
It is not clear at this time thether the pronesed E criteria muld Save avoided such lifting.
It is clear that a subcooling criteria (e.g., 50 F without a 20-minute rule would have permitted HPI termination before PORV lifting.
1 O
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1 ENCLOSURE 3 s %,
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6, UNITED STATES y yt,#..-(((h-NUCLEAR REGULATORY COMMISSION 5 Q %,.~ #' '
E W ASHIN G T O N. D. C. 20555
'% ' '.. [. /
Alli4 f
September 28, 1979 a.
Docket Nos. 50-338 50-339 i
Mr. k'.
L. Proffitt Senior Vice President - Power Virginia Electric and Power Company Post Office Box 26666 Richmond, Virginia 23261
Dear Mr. Proffitt:
During a meeting with you the morning of September 26, 1979, we reviewed with you a number of areas where we expected to require additional infor-mation from you concerning the safety injection event at North Anna Unit 1 on September 25, 1979.
Please provide written responses to the enclosed information requests. This information should be provided as soon as the necessary work can be done by your staff but no later than 30 days prior to your planned return to power from the current plant outage, which we understand you are scheduling for about 12 weeks from now.
It is expected that any actions you take will be fully reflected in the design and procedures to be implemented on North Anna Unit 2.
Sincerely,
/
/
l N
A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Enclosure:
Request for Information -
l cc: w/ enclosure See next page LO I
A-Yo
l i
Mr. W. L. Proffitt O
v4rs4"4 t' ectr'c "e "o er cc=a "> seate= der 28. 1979 cc: Mr. Anthony Gambardella Mr. Richael S. Kidd Office of the Attorney General U. S. feuclear Regulatory Cc= mission 11 South 12th Street - Room 308 P. O. Box 122
-Pichmond, Virginia 23219 Spotsylvania, Virginia 22553 Richard M. Foster, Esquire Directcr, Technical Assessr.ent Divish l
1230 A Pearl Street Office of Radiation Pro;rr:s ( AK 459)
. Denver, Colorado 80203 U. S. Environmental Protection Agency Crystal Mall #2 liichael W. Maupin, Esquire Arlingten, Virgir.ia 20460 6
Hunton, Williams, Gay and Gibson P. O. Box 1535 U. S. Envircnmental Protection Agency Richmond, Virginia 23212 Regien !!! Office ATTt;:
EIS COORDINATOR Mrs. June Allen Curtis Evilding 412 Owens Drive 6th and Walnut Streets i
Huntsville, Alabama 35801' Philadelphia, Pennsylvania 19106 t
fir. James Torson Alderman Library 501 Leroy Manuscripts Department Socorro, fiew Mexico 87801 University of Virginia Charlottesville, Virginia 22901 1
4 t'.rs. Margaret Dietrich Route 2, Box 568 Mr. Edward Kube O
Gordonsville, Virginia 22042 Board of Supervisors Louisa County Courthouse Ellyn R. Weiss, Esquire P. O.. Box 27 Sheldon, Harmon, Roisman and Weiss Louisa, Virginia 23093 1725 I Street, fi.W., Suite 506 l
Washington, D. C.
20006 Mr. J. B. Jackson, Jr.
t Commonwealth of Virginia Mr. James C. Dunstan Council on the Envirernent State Corporation Commission 903 tiinth Street Office Building Commonwealth of Virginia Richmond, Virginia 23129 Blandon Building Richmond, Virginia 23209 Mr. Paul W. Purdom Environ ental Studies Institute Mr. A. D. Johnson, Chairman Drexel University Board of Supervisors of Louisa County 32nd and Chestnut Streets Trevillians, Virginia -23170 Philadelphia, Pennsylvania 19104 O
i
i s
l' r. ',:. L. Proffitt
'~/ '/
(-
Virginia Electric and Power Company September 25, 1975 cc:
Alan 5 Ecsenthal, Esquire g',g';;2lkb.$[\\
Attmic Safety and Licensing Appeal Board
\\
U. 5. !;uclear Regulatory Commission pM, p Washington, D. C.
20555
\\G. ;:'c l'ichael C. Farrar, Esquire 19 0
.s s
-.xO Atomic Safety and Licensing Appeal Board
(
\\ '!) ~)\\iv s
V. 5. Iluclear Regulatory Commission Washington, D. C.
20555 Dr. John H. Buck Atomic Safety and Licensing Appeal Board U. 5. f;uclear Regulatory Commission Washington, D. C.
20555 Atomic Safety and Licensing Board Panel U. 5. f.'uclear Regulatory Commission Washington, D. C.
20555 v
0 cw
'v
NORTH ANNA UNIT NO. 1 Sl" DOCKET NO. 50-338
'~'
RE0 VEST FOR ADDITIONAL INFORMATION RELATED TO THE SEPTEMBER 25,1979 EVEN 1.
Perforn an engineering analysis of the Septer.ber 25, 1979 event, identifying all of the significant initial conditions, and provide a comparison with f:sents analyzed in Chapter 15 of the North Anna FSAR especially with regard to assumed initial conditions, boundary conditions and system failures.
2.
Identify all plant procedures (number, title, dt..e) used at various phases of the event. Provide a summary descrit'ng the extent to which these procedures were used and why this use wa: considered appro-priate.
3.
Describe, specifically, the criteria that you used to determine that natural circulation was achieved following reactor coolant pump trip.
Identify the instrunents relied upon and their readings.
Provide the schedule by which formal written natural circulation cool-down procedures will be available to plant operators.
4 Provide a description of the reactor coolant pump seal performance during the event.
5.
Provide an analysis of the actual pressurizer level, including the
( minimum level reached. 6. Indicate how many times the power operated relief valve on the pressurizer cycled and what indications were aveilable to the operator (e.g., quench tank level, tail pipe temperature, valve position indication,etc.). Explain the apparent second cycling of the PORY approximately 25 minutes after the first interval (see pressurizer pressure strip chart record). 7. Quantify the mass. loss.through the PORV and explain how this was 1 determined. 8. Considering the nature of how plant parameters (e.g., pressure temperature, pressurizer level, etc.) varied and were displayed to the operator during the event, indicate how and when you would have decided to terminate the high pressure injection based on the HPI termination criteria recommended by Westinghouse. \\ g G b 9
o 9. Indicate incore tempt ature readings taken cur ng the event. Provide details as _to magnitude, time, and location.
- 10. Describe the extent to which you consulted with Westinghouse during and immediately following this event.
- 11. Provide a detailed chronology of significant events during the period from the initiating event at approximately 0544 hours through return of activity in the auxiliary building to below MPC limits.
12. Your alarm typewriter printout indicates tha: for at least one hour prior to the turbire trip, the containment wmp level continually cycled to the alarm setpoint. Why? 13. Identify the number of VEPC0 personnel working in the control room during the first 30 r;nutes of the event. 14 State whether the site emergency plan was activated in any fom.
- 15. Describe the extent to which the events which occurred at North Anna had been simulated at the Surry Simulator and demonstrated to North Anna operators, prior to September 25, 1979.
Discuss the training which North Anna operators have received at the Surry Sinulator on tripping RCPs, natural circulation and Bulletin 79-065 HPl requirements. 16. Explain. in detail the uncontrolled reactor coolant activity release directly to the auxiliary building. a. Include why it happened, how it was detected, its release path (s), how long it continued, the amount, type and form (liquid, gaseous, particulate) of activity. released, personnel exposures at the site and potential dose rate at the site boundary and beyond. b. Given the inadvertent operator error, equipment failure, or combination thereof, involved on September 25, 1979, state whether an uncontrolled release would have been prevented had the piping to the process vent from the high level waste drain tanks been installed as called for in the. plant design rather than in the as-found conditions. c. If the answer to b above is in the negative, propose a design ' modification that will prevent a future uncontrolled release of activity outside containment. O //-Vf
O 3 17. The FSAR indicates that a high level in the VCT will alare in the control room and divert the letdown stream to the boron recovery systen. Did any part of this alarm and diversion system activate prior to or during the time that the VCT relief valve was open (assumed to be approximately 10 minutes)? Would such activity sN\\ g g f' .gew$p, affect the release in 16 above? w m f' ) i l O I I l l l O i
-'j,5" "'c, v c,, UNITED STATES [ '{.-q '( j NUCLE AR REGULATORY COMMISSION "J WASHINGTON, D, C. 20555 (d
- c, h'.,,'40 E
p '%'C"../ September 28, 1979 Docket Nos. 50-338 50-339 Mr. W. L. Proffitt Senior Vice President - Power Virginia Electric and Power Company Post Office Box 26666 Richmond, Virginia 23261
Dear Mr. Proffitt:
During a meeting with you the morning of September 26, 1979, we reviewed with you a number of areas where we expected to require additional infor-mation from you concerning the safety in,jection event at North Anna Unit 1 on September 25, 1979. Please provide written responses to the. enclosed infomation requests. This infomation should be provided as soon as the necessary work can be done by your staff but no later than 30 days prior to your planned return to power from the current plant Q outage, which we understand you are scheduling for about 12 weeks from b now. It is expected that any actions you take will be fully reflected in the design and procedures to be implemented on North Anna Unit 2. Si ncerely, / NWh A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Enclosure:
Request for Information - cc: w/ encl osure See next page /9-Y(
ljf (U]L i ]d\\ h{ @ gy Tg gI L fir. W. L. Prof fitt c September ,1979 -( Virginia Electric and Power Company !!r. Aichael S. K'idd cc: Mr. Anthony Gambardella U. 5. tiuclear Reculatcry Cc? mission i Office of the Attorney General 11 South 12th Street - Room 308 P. O. Box 128 Richmond, Virginia 23219 Spotsylvania, Virginia 22553 Richard M. Foster, Esquire Director, Technical Assessrent Divisi Office of Radiation Pro;rans (AW 459) 1230 A Pearl Street U. S. Env;ronmental Prctection Agenc;. Denver, Colorado 80203 Crystal !!all #2 Michael W. fiacpin, Esquire /.rlington, Virginia 20460 Hunton, Williams, Gay and Gibson U. S. Environmental Protection Acenes P. O. Box 1535 ~ ~ Richmond, Virginia 23212 Regicq III Office ATTri: EIS COORDINATOR Curtis Building Mrs. June Allen 6th and Walnut Streets 412 Owens Drive Huntsville, Alabama 35801 Philadelphia, Pennsylvania 19106 Alderman Library tir. James Torson Manuscripts Department 501 Leroy University of Virginia Socorro, New Mexico 87801 Charlottesville, Virginia 22901 firs. Margaret Dietrich "r te*>re x=8e O aoute 2. sox 558 Gordonsville, Virginia 22042 Board of Supervisors Louisa County Courthouse P. O. Box 27 Ellyn R. Weiss, Esquire Sheldon,' Harmon, Roisman and Weiss Louisa, Virginia 23093 1725 1 Street, N.W., Suite 506 Washington, D. C. 20006 Mr. J. S. Jackson, Jr. Commonwealth of Virginia Council on the Enviroment Mr. James C. Dunstan State Corporation Commission 903 Ninth Street Office Building Commonwealth of Virginia Richmond, Virginia 23129 Blandon Building Ric hmond, -Vi rginia 23209 Mr. Paul W. Purdom Envirorcental Studies Institute Mr. A. D. Johnson, Chairman Drexel University Board of Supervisors of Louisa County 32nd and Chestnut Streets Trevillians, Virginia 23170 Philadelphia, Pennsylvania 19104 &n
a O. ". r.rgiria Electric and Power Company W. L. Frc'fitt September 28, 1979 Alar 5. Pctenthal, Escuire cc: Atenic Safety ar.d Licensing Appeal Board U. 5. i;ucl ear F.egul atory Ccamission
- ashington, D. C.
20555 Michael C. Farrar, Esquire Atcaic Safety and Licensing Appeal Board U. S. ::uclear Regulatory Commission Washington, D. C. 20555 Dr. John H. Buck Atomic Safety and Licensing Appeal Bocrd U. S. tiuclear Regulatory Commission Washington, D. C. 20555 Atomic Safety and Licensing Board Panel U. S. fiuclear Regulatory Commission Washington, D. C. 20555 0 4 l '8 ) 4-
O " oats ^""^ u"it "o 1 DOCKET NO. 50-338 REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE SEPTEtiBER 25,1979 EVENT 1. Perforn an engineering analysis of the September 25, 1979 event, icentifying all of the significant initial conditions, and provide a ccmparison with events analyzed in Chapter 15 of the North Anna FSAR especially with regard to assuned initial conditions, boundary conditions and system failures. i 2. Identify all plant procedures (number, title, dr.e) used at various phases of the event. Provide a summary describ ng the extent to which these procedures were used and why this use wa: considered appro-priate. 3. Describe, specifically, the criteria that you used to determine that natural circulation was achieved following reactor coolant pump i trip. Identify the instruments relied upon and their readings. Provide the schedule by which formal written natural circulation cool-down procedures will be available to plant operators. 4 Provide a description of the reactor coolant pump seal perfomance during the event. l 5. Provide an analysis of the actual pressurizer level, including the minimum level reached. 6. Indicate how many times the power operated relief valve on the pressurizer cycled and what indications were available to the operator (e.g., quench tank level, tail pipe temperature, valve position indication,etc.). Explain the apparent second cycling of the PORY approximately 25 minutes after the first interval (see pressurizer pressure strip chart record). 7. Quantify the mass loss through the PORV and explain how this was i determined. 8. Considering the nature of how plant parameters'(e.g., pressure temperature, pressurizer level, etc.) varied and were displayed to the operator during the event, indicate how and when' you would ' have decided to terminate the high pressure injection based on the HPI termination criteria reconmended by Westinghouse. L.
9 FA]D 3) 7D ' M i D O N M_u 2 o Ju a O Indicate incore temperature readings taken during the event. 9. Provide details as to magnitude, time, and location. Describe the extent to which you consulted with Westinghouse during ~ 10. and immediately following this event. Provide a detailed chronology of significant events during the period 11. frcm the initiating event at approximately 0544 hours through return of activity in the auxiliary building t,o below MPC limits. Your alarm typewriter printout indicates the for at least one hour 12. prior to the turbine trip, the containment amp level continually cycled to.the alarm setpoint. Why? Identify the number of VEPCO personnel working in the control room during, 13. the first 30 minutes of the event. State whether the site emergency plan was activated in any form. 14 Describe the extent to which the events which occurred at North Anna 15. had been sinulated at the surry Simulator and demonstrated to North Anna operators, prior to September 25, ic79. Discuss the training which North Anna operators have received at the Surry Simulator p on tripping RCPs, natural circulation and Bulletin 79-06B HPI b requirement s. Explain, in detail the uncontrolled reactor coolant activity release 16. directly to the auxiliary building. Include why it happened, how it was detected, its release path (s), a. how long it continued, the amount, type and form (liquid, gaseous, particulate) of activity released, personnel exposures at the site and potential dose rate at the site boundary and beyond. Given the inadvertent operator error, equipment failure, or I b. combination thereof, involved on September 25, 1979, state whether an uncontrolled release would have been prevented had the piping to the process vent from the high level waste drain tanks been installed as called for in the plant design rather than in the as-found conditions. If the-answer to b above is in the_ negative, propose a design c. nodification that will prevent a future uncontrolled release l of activity.outside containment. l & cv
b ,3, 17. The FUR indicates that a high level in the VCT will alarm in the control room and divert the letcown stream to the boron recovery systen. Did any part of this alarm and diversion system activate prior to or during the time that the VCT relief valve was open (assumed to be approximately 10 ninutes)? Would such activity affect the release in 16 above? O O 19-5/
APPENDIX VIII SEQUENCE OF EVENTS DURING SEPT 25, 1979 TRANSIENT AT NORTH ANNA 1 _ SEQUENCE OF REACTOR EVENTS MUCH OF THE FOLLOWING INFORMATION WAS TAKEN FR0h THE DATA ACQUISITION S THE TIMES GIVEN ARE APPROXIMATE.- 0544 REACTOR POWER WAS AT 78%. A TUBE RUPTURE IN THE DRAIN COOLER FOR LOW PRESSURE FEEDWATER HEATER SB CAUSED DRAIN COOLER DUMP VALVE TO CYCLE. THE VALVE APPARENTLY FAILED CLOSED CAUSING EXTRACTION STEAM CONDENSATE TO BACK UP INTO THE FEEDWATER HEATER. 0509 THE TURBINE TRIPPED ON HIGH HIGH LEVEL IN FEEDWATER HEATER 55. THE REACTOR TRIPPED BECAUSE OF THE TURBINE TRIP.THE MAIN STEAM DUMP VALVES OPENED AUTOMATICALLY TO REDUCE THE REACTOR COOLANT BELOW THE NO-LOAD STEPOINT WHICH IS 547'F. OSE OF THE MAIN STEAM DUMP + VALVES FAILED IN THE OPEN POSITION CAUSING THE REACTOR TO COOL DOW RAPIDLY BELOW 547'F. 0610 THE LOW PRESSURIZER PRESSURE ALARM ACTUATED AT 2022 PSIG. THE IN LOOP 2 0F THE REACTOR COOLING SYSTEM WAS 537*F,100* BELOW THE SATURATION TEMPERATURE. AT LEAST ONE BANK OF PRESSURIZER HEATERS ACTUATED. 0611 PRESSURIZER PRESSURE WAS 1901 : 16 PSIG. THE TEMPERATURE OF THR REACTOR COOLANT SYSTEM COLD LEG WAS E33*F, 94* BELOW THE SATURATION TEMPERATURE. CONDENSATE PUMP 1C TRIPPED. MAIN FEEDWATER FUMP C l TRIPPED BECAUSE OF LOW CONDENSATE PRESSURE. STEAM GENERATOR LOW LEVEL ALARMS ACTUATED. 0613 SAFETY' INJECTION INITIATED AUTOMATICALLY BECAUSE OF A LOW PRESSUR PRESSURE SIGNAL. THE LOW PRESSURIZER LEVEL SIGNAL HAD BEEN PREVIO 0; ADMIN 1S m Ti n LY 1ar m D BASED ON IEB 79 0eA. PRESSuarzER LEVEL WAS AT 0.2% AND THE ALARM SETPOINT IS AT 9%.
. 0 0615 HIGH PRESSURE SAFETY INJECTION PUMP B STARTED. THE LOW PRESSURE SAFETY INJECTION PUMPS STARTED. MAIN FEEDWATER PUMP A TRIPPED BECAUSE OF THE SAFETY INJECTION SIGNAL. 0617 PRESSURIZER LEVEL WAS AT 9.4%. 0618 PRESSURIZER PRESSURE WAS AT 2161 5 PSIG. 0619 MAIN STEAM TRIP VALVES HAD BEEN CLOSED TO STOP STEAM DUMP. STEAM PRESSURE WAS 596 PSIG. SAFETY INJECTION WAS RESET. CHARGING PUMP (HPI) B WAS TRIPPED. REACTOR COOLANT SYSTEM COLD LEG PRESSURE.WAS 2314 PSIG. 0620 0627 THE POWER OPERATED RELIEF VALVE CYCLED TO LIMIT THE PRESSURE RISE ~ IN THE REACTOR COOLANT SYSTEM. LETDOWN TO THE CHEMICAL AND VOLUME CONTROL SYSTEM WAS INITIATED. 0628 PRESSURIZER PRESSURE WAS 2340 PSIG. STEAM GENERATOR 3 PRESSURE WAS 612 PSIG. 0629 REACTOR COOLANT PUMP 1B WAS STARTED. AUXILIARY PRESSURIZER SPRAY WAS INITIATED. 0631 PRESSURIZER LEVEL WAS 63%. 0639 PRESSURIZER POWER OPERATED RELIEF VALVE WAS CLOSED.
- O p)- sa
. 0 0648 THE RELIEF VALVE ON THE VOLUME CONTROL TANK LIFTED, THUS TRANSFERRING REACTOR COOLANT TO THE HIGH LEVEL WASTE DRAIN TANK AND RELEASING DISSOLVED NOBLE GASES TO THE AUXILIARY BUILDING VIA THE WASTE SYSTEM VENT. AN INCORRECTLY CONNECTED VENT LINE ALLOWED VENTING DIRECTLY TO THE AUXILIARY BUILDING. 0700 AUXILIARY BUILDING AIR INCREASED TO 1000 MONITORS INCREASE TIMES SACKGROUND - RETURNED TO BACKGROUND BY 0900 0700 EVACUATED AUXILIARY BUILDING TWO PEOPLE 0705-0710 AUXILIARY BUILDING AIR 100-150 TIMES MPC O SAMPLES 0830-1430 AUXILIARY BUILDING AIR NOBLE GAS ACTIVITY SAMPLES 10% MPC BY 1030 1700 SOUTH FENCE LINE TLDs CHANGED NO EXPOSURES ABOVE BACKGROUND O W
O O O SEQUENCE OF EVENTS AT 00aIU_aBN6_0N_SEPIEMBER_25 _1919 1 0544 REACTOR POWER AT 78% 0609 TURBINE TRIP / REACTOR TRIP FAILURE OF STEAM DUMP VALVE 0613 SAFETY INJECTION 0627 PORY CYCLED TO LIMI T PRESSURE LETDOWN TO CVCS INITIATED I 0648 RELIEF VALVE ON VCI LIFTED b 0700 AUX'lLIARY BUILDING EVACUATED 0900 AUXILIARY BUILDING' AIR MONITORS RETURNED TO BACKGROUND t v
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"/ .) ) ( d Yb$b^ W' LETTER FROM MI E CA OALITI0r' FOR M-g); 3d . ENERGY ALTERNATIVES TO NRC COMMISSIONERS REGxRDING SEISMIC DESIGN OF. WOLF CREEK {g gg-- g 4 yz//-))gus=,W: ' ~; ~m% N_ ['i(('Q::{q >7 q ,2 < M' a-M[&%g('[hd,,,kMshuh N [_ 6, if# d Mid-America Coalition For Energy Alternatives 9130 antl.S*CN ACAO Seelersest esissace, As 64204 19131282 8023 June 29, 1979 Joseph Hendrie, Chairman Peter Bradford, Cornissioner [i f l % [ Vic:cr Gilinskv, Commisstoner Q$id) @J"s iW j\\ f,ji_, Richard Kenned {e, Commissioner On9 %' John Aherne, Commissioner ~ ~ U.S. Nuclear Regula: cry Commission dashing:cn, D.C. 20555
Dear Commissicners :
I wrote you on January 19, 1979, on behalf of my clien: i for the asking that you suspend the construction permit Wolf Creek project in view of documented quality centrol of the reactor _croblems specifica11; related Oc the base ma containment building. You responded by publishing a notice - ~,,,.-c i of ou-request in the Federal Register. ,m /d _ This letter is to advise of certain new determinations to the seismic character of the area and to with respect for at least a partial suspension of the renew our request construction permit in view of the significance of those determina:icns in conjunc Lon with existing unrescived issues regarding base mat integrity. Ycur attention is directed to a report of the Kansas to State Geclogical Survey (KSGS) prepared under contrac: your Cicision cf Reac:cr Safety Research, Office cf Nuclear Recula :rv Research, entitled " A Revised and Auemented List Of' Earthquake Intensities for Kansas, 1367-1977 NCRI:3 /CR-3 2294, August, 1975. The report details the cenclusion of the KSGS tha: the largest histcrical earthquake in Kansas cccurred at a differen: Icca icn and was of a different This erthquake magnitude than had been previously believed. was used as the basis for the design of the non-standa-diced I (safety related) portions of the plant. Cc=monly Categer) Oc have kncwn as the 13c7 Manhattan eartnquake and thought been of the size Modified Mercalli VII. L:s epicenter was of assumed to have been approximately 22 miles northwest Manhattan, Kansas. The applicants argued that the earthquake was related to a presumed ":cne of weakness" associated with the contact of the Keweenawan mafic volcanic bel: n) and the Nemaha Ridge (Nemaha Uplift). The neares: ( 7 907 090M / $ 8 '00% Aecyc:ed [ [(/?/* O I Pa, er SIO k
~ b NRC Commissioners -- 2 according to the SNUPPS PSAR Wolf approach of the zone,is 75 miles from the Wcif Creek site. On Creek Addendum, basis the applicants urged the adoption of a safe that shutdown earthquake (SSE) with a.10g horizontal acceleration. basis for the applicants' assumption Finding insuf ficient the earthquake was related to such a zone of weakness, that the SSE be based on the your staff apparently insisted thatthe 1967 Manhattan earthquake could occur on assumption that approach to the Wolf Creek the Nemaha Ridge at its closest Such an assumption would, concluded the 50 miles. site. and the site staff, yield a safe shutdown earthquake of.12g, was licensed accordingly. In light of the new information developed by the KSGS concerning the size of the 1867 earthquake and the actual location of its epicenter, and recent microseismicity recorded along the long inactive Humbol Fault, the postulated.12g horizontal acceleration safe shutdcwn earthquake does not now appear to be conservative. The KSGS report concludes, on the that the 1867 basis of extensive review of historical records, " Manhattan" earthquake was at least a Modified Mercalli VII-VIII stronger than the >M VII that both applicants and staff had assumed. It concluded also that its epicenter was in the ( )) Wamego vicinity, and was, accordingly, associated with the Humbolt Fault. The Humbolt Fault defines the eastern boundary of tne Nemaha Ridge and passes within 50 miles of the Wolf Creek site. In addition, since January, 1973, numerous microearthquakes have occurred along the trace of the Humbolt Fault north of the Wolf Creek site and south in Oklahoma. this means stress While the KSCS has not yet concluded that is butiding in the vicinity of the nearest approach of the fault to the plant site, they site successful earthquake prediction experience elsewhere in the country which indicates that such is often the case. The size of the appropriate safe shutdcwn earthquake for the Wolf Creek site can be determined by reference tc ycur staff's Safety Evaluation Report for another of the SNCPPS Both Tyrone, in Wisconsin, and Wolf Creek are units, Tyrone. The located in the Central Stable Region Tectonic Prevince. following Tyrone SER discussion elucidates the reason for se::ing the Tyrone SSE at.28 hertrontal acceleration: " Based on historical accounts, the area of the Central Stable Region in which the Tyrene site is located is setsmically very quiet. No historical earthquakes have been reported within 100 mile ; of the site, and only ten earthquakes of intensity >M IV or greater have been reported wi:hin 200 miles of the site. The nearest historical earthquake in (") the vicinity of the Tyrone site, which occurred sometime between 1865 and 1370, had an estimated () / ~ p
$1 9 o y 0 rQ 7 I N i a eb j s N_) NRC Commissioners -- 3 intensity MM VI-VII and occurred slightly more than 100 miles west of the site. n "The Midcontinent Geophysical Anomaly is located approximately 45 miles northwest of the Tyrone site. to a region characterized This feature correspond: by gravity and magnetic anomalies, which over much coincide with mapped basement faulting. of its extent, The Midcontinent Geophysical Anomaly extends generally from the Lake Superior region south-west through Minnesota, across Iowa, and into Kansas where it trends into the Nemaha Uplift. The largest histcrical earthquakes which have been located along this feature have had reported epicentral intensities of MM VIII. However, as has been noted above, the characteristics associated with at least one of these intensity FD1 the Keewenaw Peninsula earthquake of VIII events, 1906, would indicate that the intensity level may have been influenced by local geology. If it is assumed that an intensity FBI VIII earthquake could occur on structures associated with the Midcontinent Geophysical Anomaly at its closes: approach to the site i.e. 45 miles, the intensity at the site due p)s (_ to attenuation would be reduced to intensity MM VII-VIII. "In 1954 Neumann developed an empirical relationship between earthquake intensity and p ound acceleration. Mere recently Trifunac and Brady (1975) have published a relation between intensity and accelera:Lon which was developed using many additional observations. Irifunac and Brady's data essentially cerroborate the relationship publisted by Neumann. C:ilt:ing either the Neumann or the Trifunac-Brady relation between in:enst:y and acceleration, the mean acceleration corresponding to intensity >Si VII-VIII is 0.2g. Based en this analysis we consider 0.2g to be the appropriate acceleration for the seismic design of the proposed plant at the Tyrone site." pp. 2-16, 17, 13 With respec: to the base ma of the Wolf Creek reac or building, the significance of-setting the safe shutdown earthquake at.2g horizontal acceleration is substantial. Your staff has been unable to dt.iclude that the 90-day concrete cylinder tests, which showed that the base mat concrete failed to meet the design specification of 5000 pounds per square - inch, were in error. Acccrdingly, it fs(") ordered the applicants, who carry the burden of proof on all /h*(o/
O 0 D D 3'k f ( EN elfb A k Tb _, NRC Co==issioners -- 4 such matters, to show that the concrete is of sufficient strength, on the basis that the 90-day tests are assumed to be accura:e. The Wolf Creek architect / engineer, the Bechter Power Corporation, performed the reanalysis by first determining that actual concrete strength as shown by the 90-day tes:s was 4460 pounds per square inch (by working backward from the acceptance criteria) and then by performing computer simulations to show tha: the base man was adequate at that strength to permit the safe shutdown of the plan even if it is subjected to a horizontal acceleration of.2g -- greater than the.12g earthquake for which, as ncted above, the Wolf Creek site is licensed. The standardized portion of all SNUPPS plants must be built to be shut down safely after a.2g earthquake. The Bechtel Report notes tha: this safe shutdown earthquake is " controlled by a site other than Wolf Creek", but does not specify which one. The Report states that the use in the reanalysis of the greater than required.2g assumption "is consistent wi:h the general methodology used for the project, is in accordance with the commitments made in PSAR Section 3.7 and provides additional conservatism." " Seismic loads were conservatively determined at the SNUPPS envelope "g" (~T level, which is considerably higher than that for which the s_) site is Itcensed", states the Report in its conclusion-We submi that the reanalysis was, for the reasons discussed above, n21 conservatisa -- that the Bechtel Report shows, if Lt is valid, only that the base ma is not expected to crack during the largest probable earthquake, if the concrete undercoes no deterioration. However, no allowance is made in the 3echtel Report for normal de:erioration of the base mat due to routine plan: operation. In addition, evidence exists tha: the base ma: concrete is presently undergoing spontanecus detericratien due to some as yet unknown cause. As you are aware, some of the 90-day test results were lower than the 2S-day test results. Unless he reason for
- his anomaly is explained, it constitutes evidence that de:erieration is taking place -- evidence which, under your agency's rules, it is the responsibility of the applicants to refute.
Yet, on June 7, 1979, your staff issued a summary l c f the - pubite meeting held in Burlington, Kansas en May.15, 1979, to review with the appiteants the' Bechtel Reper and the base ma; problem generally, a principal conclusion of which wass l l '"1. 'There is no clear cut answer as to why some of (g the 90-day cylinder tes: results are icwer than 5000 s,/ pcuncs per square inch. Neither is there a clear cut an eer as to why some of the 90-day strencth results [ 2L
00 ?GC Commissicners -- 5 are lower than those obtained with the 23-day cylinders." Ue understand that your staff has now enlisted the technical services of the U.S. Army Corps of Engineers in an effort to illuminate the deterioration issue, and that several factors and combinations of factors are being investigated. We are aware of one such possibility, which we communicated to your staff two months ago. It involves the possible presence of opaline in the aggregate portion of the concrete mixture. Opaline has, after numerous investigations, been determined to be responsible for the unusual phenomenon attending concrete made utch river sand aggregate taken from nortnern Kansas rivers, including the Kaw, or Kansas, River s the concrete tends to expand and weaken over time, although this effect is seemingly somewhat unpredictable. It is our understanding that the source of the fine aggregate for the Wolf Creek base ma: was originally to have been a ilmestone quarry near Ottawa, Kansas, operated by the Haworth Company, but that Daniels, the Wolf Creek general contractor, with the assumed knowledge of the applicants, changed the source to Kaw River sand, to be supplied by Ho'liday Sand and Gravel of Bonner Springs, Kansas. The change precipitated a lawsuit by Holiday, which is pending ) in Coffey County. We do not know that your staff has addressed this. / Accordingly, we inquire whether the ultimate source of the aggregate was properly approved by your staff and whether the presence of opaline aggregate has been determined and' evaluated for its significance to the deterioration issue. In sum, (1) the largest historical earthquake in Kansas was bigger than your staff and the applicants were aware and took place on a fault which passes 50 miles from the plant site, which is only now known to be active, and whien may be developing a "seismi: gap" in the vicinity' of the neares: apprcach :o the plan; (_) no evidence exists tha: the base mat could survive such an earthquake after a period of wear and tear due to normal plan operations, or at anv Otme if spontaneous deterioration is taking place, and (3) evidence that such detericra:Lon is taking place exists. It is therefore imperative tha: those making decisions about :ne Wolf Cre~ek project know all that can possibly be known abou: the nature of the concrete in the base mat. We ask that you provide us a complete explanati:n of all the steps taken by you, other governmental agencies. -the applicants or their agents to - determine whether deterioration of the base ma: can be expected. Finally, 'we ask that you take action on our pettrion of f_~) January 19, 1979, concerning the Wolf Creek construction s
r { I o NRC Commissioners -- 6 V is your staf f's pos trion, expressed repeatedly, It permit.the applicants' decision, without staff authort ation, that to remove the voluntary " hold" placed on construction of the building, would cause the staff to seek an containment which they expect would be granted, -immediate order from you, such work be stopped. In fact, a vice-president requiring thatKGLE advised your staff in writing at the time bf applican: of the May 15, 1979 Burlington meeting that they indended to in the reactor containment building resume concrete placementIt is our understanding that " jawboning" within a few days, It remains our position that by your staff dissuaded them. a partial construction permit suspension is the only effective way for your agency to protect the public interest in this s ituation, and we hereby renew our request that you act accordingly,
- urs, sinc p/
V < r ' L. Wilitam H. Ward Attorney fcr MACEA W'rN bw () cci Domenic Vassallo,/NRC Roger Boyd, NRC s Olin Parr, NRC Carl Seyfrit, NRC H. D. Thornburg, NRC Stephen H. Lewis, Esq., NRC S. J. Chilk, Secretary, NRC Jay E. Stiberg, Esq. Kansas Congressional Delegation s b d ) /~3 V
s. a ~ sw ar... o.. c,,. 7..u w..D.. m ~a,ma.aa ;.a. .t -- ~ APPENDIX X ACRS CONSULTANTS' REPORT JOHN C. MAXWELL i mEDLe elst kh 53 3 3 WESTERN MlbLB ORIVE g AueTim vana rev3 e October 1, 1979 Mr. Harold Etherington Advisory Connittee on Reactor Safeguards Nuclear Regulatory ComT.ission
- Mail Stor H 1016 washingt'on, D. C. 20555
Subject:
I.etter of June 29, 1979, to the Commissioners, U. S.N.R. C., fro 5 William H. Ward, Attorney for Mid-A erica Coalition for Energy Alternatives
Dear Nr. Etherington:
- ".r. Ward's letter raises substantative questions on which Mr. Muller asked me to com.ent.
Undoubtedly, the Staff has considered these questions at length. My comments are based on material immediate1y available to me 1: Austin. 1. Validity of the choice of 0.12g for the Wolf Creek project. O c The selection of an SSE of 0.12g was controlled by the 1867 Manhattan earthquake, MM intensity VII, epi-center approximately 20 r.iles northwest of Manhattan, Kansas. Additional research carried for.:ard by the Kansas Geolocice2 Survey, under the sponsorship of the ::.R.C., iniicated thet the epicenter lay east of Manhattan, closer to the Wolf Creek site, and ap,troxi ately on the trace of the eastern boundary fault zone of the Nemaha buried uplift. The inten-sity ur.s elso reevaluated and raised to VII - VIII. Apparent 1:, the evidence for these changes is des-cribed in NUREG/CR - 0294, which was not immediate-ly available to me. The new position and intensity are, however, licted in NURIG/CR - 0666. which I do have in my file. Assu'11ng both the newly located 1867 earthquake epi-center and higher intensity are valid, then the SSE for the Wolf Creek Project should be redetercined on the basis of a sir.ilar earthquake occurrins 50 miles hT.' of the project site along the eastern margin of the Nemaha structure. I must agree with V e
0.h h. l tl I o '<r. Harold Etherington October 1, 1979 Mr. '<!ard's contention that the regional structural setting of the Wolf Creek project is similar to that of Tyrone, for which the SER recommended an SSE of 0.2g horizontal acceleration. This value would seem to be about ri ht for Wolf Creek also. 6 2.- The principle thrust of Mr. War.9's letter is to call attention to the possible deterioration of concrete in the base met of the reactor containment building, especially with regard to increased seismic risk related to the reevaluation of the 1867 Manhattan earthquake. The undesirable effects of opaline silica (usus.11y as chert or chalcedony sand grains and pebbles) on the strength of concrete are well known. I'm sure the staff is evaluating this situa-tion. In any case I have no basis for further com ent. Cordially, ~. John C. Maxwell () Consultant to A.C.R.S. l b f\\ W .. ~..,..
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O. M. Carson Seatomber 1A. 1979 g, .t 'a /t -q NOTE 5 ...J
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- 1) The two unit Allens Creek review was. suspended in December 1975 by the Applicant due to financial considerations. The project was-reactivated as a one unit plant in 1978. At the 217th ACRS Meeting in May 1978. the Corsnittee waived a re-review of the Project.
2) Interim Report 3) Interim Report
- 4) Early Site Review - Units cancelled
- 5) Review' suspended pending Applicant's decision on the site ACRS reviewed Cherokee /Perkins together. The Cherokee docket was 6)unco"ested; a CP was issued for the Cherokee units in December 1977.
Recently, the applicant has deferred final decision on construction of Cherokee Unit 3 and Perkins Units 1-3. O 8 I e k 4 c /f-c. Y (
s: 'g UNITED STATES if NUCLEAR REGULATORY COMMISSION h' I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4 D9 D 1D 7'Tl4 1
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WA$mCToN. D. C. 20555 $; j];j [ %,***,/ ed 1 October 10, 1979 APPENDIX XII LETTER TO REP. UDALL REGARDING PROPOSED HYBRID STANDARD NUCLEAR PLANT DESIGN The Honorable Morris K. Udall, Chairman Committee on Interior and Insular Affairs House of Representatives Washington, EC 20515
Dear Congressman Udall:
The Advisory Committee on Reactor Safeguards has reviewed the questions you asked concerning the development of a hybrid oower reactor design based on existing technology that would incorporate " plant features that maximize safety" in a single design to be developed by D3E and approved by the NRC. Answers to your questions are provided below. Question 1: "What would be the advantages and disadvantages of a regulatory system in which the only plants eligible for licensing would be those built in accordance with the D3E design?" Answer: There might be advantages from a single design incorporating the best combination of features to " maximize safety". he ACRS be-lieves a more desirable approach would be to establish a design team consisting of a Nuclear Steam Supplier and Architect-Engi-neer combination for each of the four water cooled reactor sys-tems and the HTGR system to develop a design incorporating the nest desirable features appropriate to that system. Rese de-signs should be carried out with the understanding that they would not be constrained by preestablished design concepts, cur-rent marketing restraints such as patent or proprietary limita-tions, or the preferences of specific utility customers. tey would consider capital and operating costs, and reliability as well as safety. A fif th IMR team composed of experts having a broad engineering background would be asked concurrently to arrive at an optimum conceptual design, drawing on all EMR technology. Each team would be requested to provide a recommended standard plant, optimized for safety. In addition, it is recommended that each design team be asked to provide two additional design variants, defining what additional safety measures they would include for say, 5 and 10% additional cost, and what incremental improvements in safety this would provide.
.(3 Honorable Morris F.. Udall October 10, 1979 hese results could then be evaluated to determine whether one or more should be required for use to the exclusion of all others in the licensing process. Such an action would be essential prior to establishing legislation restricting the licensing of nuclear power plants to a specific set of design requirements. he ACRS has reservations about the DOE's ability to establish a suitable integrated design. Its internal resources are limited and the design capabilities of its laboratories and contractors have become inactive in the reactor system design areas to a level where teams who have the overall breadth of knowledge for such a challenging assignment would be difficult to organize. If you are suggesting that DOE be the contracting agent for drawirg together teams of industrial participants, this might be a workable arrange-ment. ne ACRS would prefer that the study not be rigidly constrained by either a literal interpretation of the terr. " existing technology" or the avoidance of some research and development. Any system which departs significantly from the existing approaches will re-quire some experimental work to verify its performance character-istics before an explicit design can be reviewed and approved. Operational experience with an inUR is limited in the United States to Peach Bottom 1 and Fort St. Vrain whose power levels and compo-nents are not prototypic of a large commercial size plant. Be ACRS believes that considerable component research and development, as well as a large body of research in other areas might be needed to define a standard IMGR, and some years of successful operational experience with a commercial size plant would be appropriate before accepting the design as a standard plant mandated by act of Corgress. Further, the term " existing technology" can have several meanings, hat to mast it would mean using only features that are already in use on currently designed plants. W e ACRS believes that, regardless of current use, there should be some freedom to introduce safety features that measurably enhance safety in quantitative risk terms if the economics are tolerable and the time constraints associated with verification will permit. A basic design that could incorpo-rate such changes now or in the future would be an advantage if this capability were realistically established. v 4-76
o Honorable Morris K. Udall 0::tober 10,1979 v In proceeding with a study of this type the basis for comparison and selection of the best safety features would be difficult to define. An attempt should te made to preestablish a set of criteria for judging the safety features provided by design that would assist in selecting the most desirable plant for the intended purpose. If only one design were developed for licensed use its advantages might include: 1. A well established configuration that may be safer than current plants with identifiable features generally accepted as meeting public safety needs. 2. An op.mrtunity to concentrate attention on a single system with characteristics understood by a large number of plant owner / operators thus simplifying the educational problem and improving operational reliability. 3. Over a long period, some economic advantages if a number of reactor plants were to be constructed, since multiple units of the same design could spread the design costs over a broader base, realice the economics of quantity production and permit centralized inventory of parts and components by utility groups to reduce capital inventory while enhancing maintenance related reliability factors. 4. Quality Assurance improvements derived from Standard inspection ethodology and equipment plus broader statistical knowledge of system and equi; ment problems af ter a period of use. 5. Regulatory simplification because only one set of documents would be needed, and an improved review. process could result. Disadvantages of a single design might include: 1. toss of operating experience. Although a hybrid design might use many components and parts which have established quality and performance as the result of previous use experience, the new arrangements resulting from hybridization could have be-havioral properties which were not adequately understood be-cause of lack of system operating experience. 'Ihe existing systems have about 20 years of use experience to draw upon, much of which might be sacrificed by departing from the established system designs. (v) n-,i
d Honorable Morris K. Udall October 10, 1979 2. Commercial advantage to the manufacturer with the most knowledge and mst extensive f acilities corresponding to the selected design since each of the existing manufacturing facilities in the U. S. is currently oriented to only one nuclear steam supply system. Eis might cause competitor organizations to withdraw from industrial participation. We loss of a large portion of existire expertise would be damaging to public safety, not only because of limited participation in the new design, but because of the loss of continuity in keeping track of those designs already in use. 3. Design inflexibility. One design would not maximice safety for every utility in every location thus reducing the opportunities for economic and safety tradeoffs. 4. Narrowing of design perspective. We multiviewed attention to ssfety areas provided by using several design teams to develop design variations would be lost. 5. Potential for a common flaw disabling the entire nuclear power complex, representing a serious economic risk. b) Question 2: "To what extent would such a design reduce the number of items on the NRC's list of high priority, unresolved generic safety issues?" Answer: Improved design could eliminate some of the generic safety issues not resolved for existing plants, but not all. In any case the ACRS believes that the current licensire and regulatory program must continue to address generic problems. A new design effort should not be pemitted to-defer their resolution. Question 3: "How much time would be required to specify an optimum design con-sisting of a hybrid of current designs?" Answer: We depth of study and the level of design detail needed would determine the lapsed time. We ACRS does not have a well founded basis for estimating the time requirements but can offer the fol-n lowing as a rough time estimate: V // ~7 :2_
O nenorab1e *><ris x. ude11 -s-october 10, 1929 Preparation of an initia1 comparison of existing systems Approx. 2 years Selection of engineering features for incorporation in a single design Approx. 1 year Development of a design in sufficient detal1 for cost estimating to estab-lish an economic basis Approx. 2 years Review and approval of safety features by tac personnel Approx. 3 years Construction of the first plant could begin immediately thereafter. The proposed design approach could have merit but, until a substantial effort has been applied, its net value cannot be adequately measured. We proposed requirements should not be established on a mandatory basis without the col-1ective willirgness of the af fected industry. We hope this response serves your need. . O si =ereiv. I Max W. Carbon Chairman i O f 3 0 /f-7 3
a,e ** cg UNITED STATES 9, NUCLEAR REGULATORY COMMISSION 1,, k:.,.,g' ( s ADVISORY COMMITTEE ON REACTOR SAFEGUARDS C,, ? h / ef WASHINGTON. o. C. 20555 APPENDIX XIII October 11, 1979 SYSIEMATIC EVALUATION PROGRAM Honorable Joseph M. Hendrie Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555 SUMECT: SYSTD4ATIC EVALUATION PR03 RAM
Dear Dr. Hendrie:
During its 213th meeting, January 5-7, 1978, the Advisory Committee on Reactor Safeguards (ACRS) received a presentation from the Nuclear Regulatory Commis-sion (NRC) Staff concerning the Systematic Evaluation Program (SEP) as planned. This program was intended to examine many safety-related aspects of eleven of the older light water reactors (LWR). We purposes of the program were to as-certain the degree to which these reactors complied with current LWR safety criteria and standards, and to enable evaluation in a systematic way of the possible need for backfitting, after the review of each reactor was completed.- The program also included the potential for identification of significant de-ficiencies which might warrant separate, earlier action prior to completion of the review. O V The SEP appeared to be generally responsive to the ACRS recommendation for a periodic, comprehensive (10-year) review of older reactors, first made by the Committee in 1966. An important difference was that the ACRS had recom-mended that the licensee perform the detailed safety analysis of his plant and report his results and conclusions to the NRC Staff for their review and evaluation, while in the SEP the NRC Staff performs the detailed review. In January 1978, the NRC Staff estimated that the SEP, as they planned it, would take about three years. During its 233rd meeting, September 6-8, 1979, the ACRS was again briefed on the status of the SEP by the NRC Staff. The Staff reported that progress had been far slower than expected and that the earliest completion date was now three to three and one-half years in the future even if the currently available manpower resources were not diverted to ether jobs. h e NRC Staff stated that, thus far, they had identified only a few potentially j significant deficiencies and stated that no criteria existed for identifi-l cation of such deficiencies by the Staff. l The ACRS believes that the pace of the SEP has been too slow and that the currently expected completion date is later than desirable, in view of the fact that most of the plants beirg reviewed in this prcgram were designed prior to the development of the first draft General Design Criteria and otherwise reflect an early era in the evolution of safety criteria. 3(a 7 L/ A w FT /
- = - lionorable Joseph M. Elendrie october 11, 1979 The ACRS still believes that the SEP chould be carried out in a manner simi-lar to the safety reviews at the OL stage; that is, the licensee should pre-pare a Safety Analysis Report for those portions of the plant being reviewed, this analysis should be reviewed and evaluated by the NRC Staff, and appro-priate actions should be required to remedy any significant deficiencies. The Committee believes also that criteria appropriate to the nature and intent of the SEP be developed on which to base the judgment of potentially significant safety deficiencies. Tne Committee recognizes that the SEP is in an intermediate stage wherein a reformulation of the responsibility for the safety reevaluation is not straightforward. However, in view of the patential importance of the safety reevaluation of the reactors under review, and in view of the importance of developing a suitable process for other reactors, the ACR3 recommends that the NRC undertake an early reevaluation of the current structure of the SEP. Sincerely, i Max W. Carbon i Chairman O f r L 9 l d O i 6
ps escg ,8 UNITED ST ATES c, s.7f % NUCLEAR REGULATORY COMMISSION ~/ E ADVISORY COMMITTEE ON RE ACTOR SAFEGUARDS "a f nasectoN,0. c. rosss October 9,1979 APPENDIX XIV XENON EMISSION CONTROL The Honorable Victor Gilinsky Commissioner U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Dr. Gilinsky:
In response to your letter of September 12, 1979, we offer the following comments relative to the emission of xenon from the MI-2 nuclear plant and the possibility of reducing such releases in similar accidents which may occur in the future. So far as we know, the coments on the various approaches for estimating the total quantity of xenon released following the accident appear reason-able. In terms of controlling xenon releases in accidents which may occur O in the future, we believe that chilled or cryogenic charcoal adsorption beds of adequate capacity would be helpful. Adsorption systems using this principle are commercially available and are in successful operation for handling routine releases at several nuclear power plants. The major problems would be possible decreases in the effectiveness of such beds under conditions which involve the larger volume and moisture content of releases accompanying an accident, and the difficulties in essuring that all the xenon emissions are collected so they can be effectively directed to and treated by the adsorption system. Other methods, such as low temperature liquefication and fractional distillation may also be usable for removal of radioactive noble gases. As noted above, the technology exists and has been applied in BWRs and PWRS to remove and retain for decay purposes certain radioactive noble gases before discharge as gaseous effluent. The systems have not, however, been applied under accident conditions where multimegacurie releases may occur essentially in bursts or over relatively short time intervals, such as was experienced during the 1MI-2 accident. The Comittee believes that a study should be undertaken to determine the applicability and desirability of available technology to minimize the release of radioactive noble gases during various postulated accident scenarios. The study should include assessment of the various potential pathways for radioactive gaseous re-leases as well as considerations of accelerated rates of treatment of large gas volumes such as those existing in large contalments. O &? G t
O rhe aenerab1 victor c111nsxv octoser 9. i979 Anticipating the need for answers to the types of questions raised in your letter, the Committee requested about a year ago that an ACRS Fel1ow develop a report sumarizing experimental data on the performance of charcoal beds under a variety'of parameters. It is anticipated that this report will be comp 1eted within the next two to three months, and we will plan to provide a copy to you at that time. Sincerely, x W. Ca bon Chairman ec: Joseph M. Hendrie, OCM Richard T. Kennedy, OCM Peter A. Bradford, OCM John F. Ahearne, OCM Samuel Chilk, SECY O IAe V. Gossick, EDO 4 4 O I 1
g s* ** %9* j q UNITED STATES !? g' t NUCLEAR REGULATORY COMMISSION 5 OJ i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS k// ( e. wAsmNotoN. o c 20555 October 12, 1979 APPENDIX XV SYSTEMS INTERACTIONS STUDY FOR INDIAN Mr. Leu v. Gossick POINT 3 Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
SYSTEMS INTERACTIONS STUDY FOR INDIAN POINT f00 LEAR GENERATItG UNIT NO. 3
Dear Mr. Gossick:
In a report dated July 13, 1978 concerning operation of the Indian Point Unit No. 3 at its full power level of 3025 MWt, the ACRS made several recommendations, including one that requested, " Review of the Station for systems interactions that might lead to significant degradation of safety." In its earlier report of June 9,1976 concerning full power operation of Zion Units 1 and 2, the ACRS had made a similar recommendation for that plant. In response to the recommendation for Zion, Commnwealth Edison arranged to have a study performed of Licensee Event Reports (LERs) covering the period between p 1969 and 1977 to determine which indicated a potential systems interaction question. 'Ibe results of this study were then applied to the Zion station to see if the potential for any of the same systems interactions were present and needed correction. The ACRS has recently been asked by Consolidated Edison and the NRC Staff whether an LER systems interactions study similar to that performed for Zion would be an adequate respanse to its recommendation for a systems interac-tions study for Indian Point Unit No. 3, which, like Zion, was designed and constructed prior to ACRS identification of the generic need to examine the matter of systems interactions (letter to L. M. Muntzing dated November 8, 1974). The ACRS believes that some types of systems interactions can be identified by an LER study such as that performed for Zion. However, the Committee believes that such an effort can only be considered to represent a treatment of psrt of the problem and does not recommend that type of study for Indian Point Unit No. 3. As the Committee has stated in NUREG-0572 (September 1979), " Review of Licensee Event Reports (1976-1978)," a detailed review of LERs cannot be expected to identify all systems interactions. By far, the bulk of the LERs deal with failure of individual components and equipnent, with relatively few cascades of failures resulting from an initiating event. It is not to be ex-pected that LERs will include a relatively comprehensive set of examples of low probability events involving the coupled failures of systems where the Q initiating event itself is unlikely. v A ~7 ?
Mr. Lee V. Gossick October 12, 1979 tus, there will be important aspects of systems interactions wnich are un-likely to be exposed by a study of ERs. t e important question is how to uncover vulnerabilities which may have potentially serious ef fects the first time they occur. In its letter of November 8,1974 to Mr. Muntcing, the ACRS gave several examples of possible systems interactions to illustrate the matter. Since a question has arisen concerning what constitutes a reasonably appropriate study of systems interactions at Indian Point Unit No. 3, the ACRS has the following additional comments. There are at least two general areas of investigation of systems interactions which are unlikely to be covered by a review of ERs. 1. Were is a passibility of systems interactions within an interconnected electrical or mechanical complex. In such a sttxly, it is necessary to consider failures which may be outside the usual context of failure analysis. For example, a component may run away or it may partly fail and hang up somewhere between its normal and its " failed" state, in either case leading to some excess in whatever service (voltage, frequency, flow, pressure, temperature, etc.) is provided or controlled by the system com-lex under consideration. tis kind of failure, which usually is less likely than total functional failure of a sub-system, is unlikely to be revealed by MRs. Investigation of such failures generally will require an appropriate application of failure modes and effects analysis with the use of the systems diagrams. 2. %ere is a possibility of interactions between nonconnected systems due to the physical arrangement or disposition of equipment and to possibili-ties of transporting damaging influences, such as heat or water, within a given plant or site. Such interactions are likely to be unique to each plant and are unlikely to be revealed by ERs since the probability for such interaction to occur may be modest. Were are exceptions to this, of course, and mny reductions in the potential for systems interactions resulted from evaluation of the Quad Cities. event of June 9, 1972 in which a rupture in the circulating water system flooded the turbine building basement and some safety-related equipnent. Generally speaking, however, neither ERs nor a study of plant diagrams and other drawings will con-sistently reveal the potential for such interactions between nonconnected systems, because such drawings generally show single features or systems; composite drawings which include all systems are difficult to make without their becoming unmanageably complicated. Sus, uncovering the potential for interaction of nonconnected systems will usually require careful, in-situ examination of the physical plant. tis examination must consider all features having the potential to damage safety systems, including the safety systems themselves. We physical inspection of the plant could be approached by dividing the plant into "compartanents" following discernable structures - such as p) walls, ceilings, and floors with appraisable strengths and weaknesses. t Doors, stairs, ventilation ducts, piping, and other penetrations would be fA
Mr. Lc2 v. Go: sick October 12. 1979 ,3 evaluated for patential influence transpart (fire, steam, hot air, etc.). (J Structures, which act as barriers to the flew of a danaging influence, wauld be assessed for the adequacy of their resistance to such influences. In each compartment the elements of the safety systems, including such extensions as instrument lines and power or control wiring should be identified on a " train" tasis. We physical vulnerability of the safety system elements to nonstandard conditions (temperature, pressure, water, spray, etc.) should be identified. We characteristics of such systems as influence generators under faulted conditions would have to be assessed if such system elements exist as redundant elements within the identified " compartment" baundarles. We influence patential of all non-safety elements including such items as sewer and drain lines, combustible gas transport and storage, compres-sors, and heavy-power-circuits and transformers, within the given comp t-ment should be assessed with respect to potential for damaging or disrupting (as with induced electrical noise) critical system (s) within the "compart-ment" and the " compartment" boundary itself. %e invasion of damaging influences through the barriers or boundaries into the identified compartment would also have to be assessed. Eis wauld include consideration of entry of personnel carrying influence generators such as welding equipnent. Special consideration would have to be given to the identification of r'3 convergence of safety functions into single compartments and the degree (,) of convergence within the given space. he study of interactions between nonconnected systems would also have to include the possibility of non-visible interactions, such as the possibly adverse effect of failure of one buried pipe on a neighbor due to scouring. A study of plant drawings would be required in connection with this aspect. The A"RS believes that one practical method to pursue such a systems inter-actions investigation is by formation of a small but competent interdisciplinary team, perhaps four to six individuals, who would pursue the two areas of inves-tigation described above. We report of the team should identify the detailed approach mployed and tabulate the results in a r'eviewable form. The Comittee believes that the two areas of investigation described above can be used in defining a suitable approach to a systems interactions study for Indian Point Nuclear Generating Unit No. 3 and are generally applicable to such studies on other IRRs. Sincerely, Max W. Carbon Chairman v 1
APPENDIX XVI Additional Documents Provided for ACRS' Use gd 1. Wesen and Geschichte der Technischen Uberwachungs-Vereine, by Guinter Wiesenack (in German and English) 2. Draft, SAND 79-1088, Program Plan for the Investigation of Vent-Filtered Containment Conceptual Designs for Light Water Reactors, Allan S. Benjamin, Sandia Labs, Albuquerque, NM, Aug. 1979 3. Handouts from NRC Staff Meeting on Potential Unreviewed Safety Questions on Interaction Between Non-Safety Grade Systems and Safety Grade Systems, Sept. 18, 1979 4 Letter, Subcommittee on Nuclear Regulation, Senate Committee on Environ-ment and Public Works to Chm. J. M. Hendrie, regarding disposition of radioactive water at TMI-2, dtd Sept. 27, 1979 5. Listing of NRC Research Review Groups 6. Letter, Arkansas Power and Light Co. to J. F. Stolz, NRC Staff, Pre-Operational Test, Loss of Offsite Power, ANO-2, dtd Feb. 7,1978 7. Memorandum, S. Levine, NRC Staff, to Commissioner Ahearne, FY 1981 RES Budget Request, dtd Aug 6, 1979 8. Memorandum, T. E. Murley, NRC Staff, to R. J. Mattson, F. Schroeder, D. Ross, and D. Vassallo, Reactor Safety Research Significant Events Over the Next k Month, Oct. 3, 1979 9. Memorandul, H. R. Denton, NRC Staff to Commissioners, Resumption of Licensing Reviews for Nuclear Power Plants, Aug. 20, 1979
- 10. Memorandum, L. Beckwit, Jr. to Commissioners, Commission Participation in License Issuance, Aug. 3,1979
- 11. Design Criteria, Concepts, and Features Important to Safety and Licensing, G. H. Kinchin, UKAEA, Safety and Reliability Directorate
- 12. Memorandum, Commissioner Kennedy to L. V. Gossick, NRC Staff, 01A Investi-gation Report, "Michelson Report -- Events and Levels of Review, Aug. 24, 1979
- 13. NUREG-0591, Environmental Assessment, Use of EPICOR-II at Three Mile Island Unit 2, Aug.14, 1979
- 14. Preliminary Notification of Event or Unusual Occurrence, PNO-79-339, Trip on High Reactor Pressure, at Crystal River, Aug. 14, 1979 i
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