ML20126A439
| ML20126A439 | |
| Person / Time | |
|---|---|
| Issue date: | 09/08/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1672, NUDOCS 8002140636 | |
| Download: ML20126A439 (425) | |
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' v h r i h TABLE OF CONTENTS 233RD ACRS MEETING h g' i gIJ!. g@t.
SEPTEMBER 6-8, 1979 .- + / / 6' '? f,f !'I f:?3D" i :: !^ I. Chai rman 's Repo rt (0 pen to Publ i c ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 A. " viewers .................................................... 1 B. NRC/TMI Special Inquiry Group ............................... C. Kemeny Commission Request ................................... D. Assignment of Review of LERs to ACRS Fellows . . . . . . . . . . . . . . . . II. Meeting with NRC Staff on the Implications of the TMI-2 Accident (0 pen to Public) ................................................... A. S u b comi t tee Re po r t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B. As ses smen t o f TMI-2 Acci den t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . III. Meeting with Representatives of the Stone and Webster Engineering Co. on the Seismic Design of Nuclear Power Plants (0 pen to Public) . A. Introduction .................................................... 5 B. Survey of the Seismic Performance of Stone and Webt Designed Piping .................................. ........... 5 C. Mo d i f i c a t i o n s Ma d e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 D. Conservatisms of the Original Seismic Analysis . . . . . . . . . . . . . . . . . . 5 IV. Briefing on the Proposed Activities of an Institute of Nuclear Power Operations (0 pen to Public) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 V. Meeting with the NRC Staff on the Test Results from the Mark I Full-Scale Facility (0 pen to Public) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 VI. Meeting with Members of the NRC Staff on Systematic Evaluation P rog ram (0 pen to ? ubl i c ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
. VII. Executi ve Sess ions (0 pen to Publi c) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 j A. Fo re i gn Me e t i n g s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 B. AC RS Meeti ng Schedul e for 1980 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 C. S ubcomi ttee Repo rt . . . . . . . . . . . . . . . . . . . . . . . . . ................... 7 l 1. Plant Arrangements ...................... .................. 7 1
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233RD ACRS MEETING
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- 2. Reliability and P'robabilistic Assessment . . . . . . . . . . . . . . . . . . . . . . . 8 3 Anticipated Transients Without Scram (ATWS) .................... 9
- 0. El ectri c - Gri d S ta bi l i cy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 E. Revi ew of ACRS Gen ri c Items Li s t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
F. ACRS Report to Congress on NRC Research Program .................... 10 G. Review of Nuclear Regulatory Function and Progress . . . . . . . . . . . . . . . . .10 H. Future Agenda ...................................................... 11
- 1. Honor to ACRS Consultant ........................................... 11 J. ACRS Reports and Letters ........................................... 11
.l. Evaluation of Licensee Event Reports . . . . . . . . . . . . . . . . . . . . . . . . . . .11
- 2. Response to Questionnaire from the NRC/TMI Special Inquiry Group ........................................................ 11
- 3. Letters to Members of Congress Proposing Delay of the ACRS Annual Report on the NRC Safety Research Program . . . . . . . .11 VIII. Executive Sessions (Closed to Public) ................................. 12 A. Conduct of Members ................................................ 12
- 1. M r . Mo e l l e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2 l
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i . 1ABLE OF CONTENTS FOR APPENDIXES TO 233RD ACRS MEETING SEPTEMBER 6-8, 1979 APPENDIX I - Attendees APPENDIX II - Items for uiscussion at 234th ACR$ Meeting APPENDIX III - Schedule of ACRS Subcommittee Meetings and Tours APPENDIX IV - ACRS Meeting Schedule for Calendar Year 1980 APPENDIX V - Kemeny Commission Query APPENDIX VI - Member's Reply to Kemeny Commission Query APPENDIXVII-ProposedDefenseinDepthdancept APPENDIX VIII - Class 9 Question APPENDIX IX - Memo, Mattson to Cunningham, Board Question Concerning Class 9 Accidents APPENDIX X - Background Material for Discussions on the Seismic Design of Nuclear Power Plants APPENDIX XI - Seismic Pipe-Stress ;alculation Techniques and Reanalysis Effort APPENDIX XII ~ Seismic Performance of Welded Piping APPENDIX XIII - Modifications Made to Plants Shut Down for Seismic Reanalysis APPENDIX XIV - Conservatisms of the Original Seismic Analysis APPENDIX XV - Proposed Activities of an EPRI Institute of Nuclear Power Operations APPENDIX XVI - Memo, H. R. Denton to Commissioners, Mark 1 Containment Potential Problem APPENDIX XVII - Systematic Evaluation Program Review APPENDIX XVIII - Zion Station LER Study APPENDIX XIX - Electric Grid Stability APPENDIX XX - Current Status of ATWS Problem APPENDIX XXI - Generic Items Review Assignments iii
APPENDIX XXII - Review of Licensee Event Reports (1976 - 1978) APPENDIX XXIII - Response to Questionnaire from NRC/TMI Special Inquiry Group APPENDIX XXIV - Letters to Members of Congress Proposing Rescheduling of the ACRS Annual Report on the NRC Safety Research Program APPENDIX XXV - Additiona,1 Documents Provided for ACRS Use I l iv
l 3 Fedecal Registec / Vol. 44. No.173 / WehAay. Septnder 3.1779 / No"'" 518~5 1 j the enetter M whecher to hoki a nauccal evaluatica wiu help the staH maka Under 6is proposal. NCAQ would j coclerence. the Caateussa>o especaUy necessary changes and !ciprovements. estabush and pay expenses of a panel of encourages comments on the fouowtng in ceder to conduct a fair and thorou,h individuals who coUectively represent questions: , i . evaluation, the Commissian wilt the major parras Wuestad la atr i e Should NCAQ select parecipants Gf A. Der.=/op ew2/ac6cs cmenc. Based qua Py issues and involved in the week i so. acx:ctding to what mteria1. of should ca ec= neo s received on its sectou of of da Cocmusstart.Througn attendance be opent the Pubbc Pstticipatm Plan, the , independent interdews. and utcg tha f
- Should NCAQ reimburse some Cocumssoc wiu develop sewral evalusta enterf s. the panel would participants for travel? How? What enter'.a to measure process ic meeting assen Cer:umste eHor's to cre-t the f j selecton cntena should be ued to determine eligibility for retmbursement?
the puhhc partic:pation goal Those cntena d beam part of the M public partcpation objectre and would
,ausgest appropriata adficaticas to the What alternauve funding mehiems Part>cipetion Plan. NCAQ is specdically pubile partcipation procam.
2 are available7 seeking connent on appropriate . Ib.a NCAQ La especially laterested in
" eenacon cntena, comments on the feasibdity and value of IV. Evaluados B. Schedule periodrc em'uodoas of such an omeme a& pan evaluam.
a N NCAQ staff wiu coattnuaU MP' hoce in mung de @ buppm ces sM popm evaluate accocipHshme r s emi ac inties pamcipcuono W e b emitsd uledon cmes y pasts. ad these intemal evaluatme wid be appropriate evalusWa proudure. If an included in the NCAQ bbrary Ms. de e t bfe v T bc c : a an d{na a is meant to be dexible, evolymg to cwe4 needs as they anw. A funal cud-po.at , C. Cor.sidenha ra/aes andcasts of an evaluate mtena to be used by the independe:tt midpoint er:!ucuan. , cama, msuy, Sarewy el Nanai Meems Propee.d tv NcAo I Ts= w.
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NUCLEAR REGUL.ATORY na Commmn wc! herr and discurs a COMMtSS10N nport by the NRC Staf resang ta ne Commutu wiu hear and dimu a l implicatons of the acadect which occ.ttred report regardig anas of poteobauy tugh j f Advisory Committee on Raactor seewes m the Mark I con smment resultmg at the new me Island Nuclear P' ant Unit 2. km reccet tage scale eew data. l S,aloguards; Revised Notice of Meeting , tw PM-m P.M: Momius with stone and Regarding the previotts Federal ***MMWM a px_ya py, a,c.gn Smm y, ' C Register Nocce (published en August ::3. ne Comcunu wd haar ed Ams ne Cocunmee wul haar and discuu the 19~9 (44 FR 49527j) for the meetmg of the e muu h npnmaraum of de kne report of its Subcomcuttas an S review of l Advisory Committee on Reec"or and Webster Enginunna W Licensee Eveet Reporu a.bcatted by tha ! Safeguards to be held on September e-4. W 8 68 operators of cocunerciaCy Ucensed nacleat 19~9 in Washington. D.C changes in ' pant sWems.""" d**@ ' 8"CI**' M*'"' pone plants durma the penod of wa. s scheduling are indicating below: 88 ** 88 ne Comcattee wiU dbcuss the statu of i S*o"**'Y t8 "" Pramy h uon. senon bec; taken to realuate ryttems Tha'idey, S.apiernher s. trs ' toterschecs at 6 7.Jon Nacf mer Stabon and M PN-as P.M M*cas on Nacar ladlea Point Nuclear Ptaat Urut S.
' &M AR-stCO A.M: hecuan Seuian Cpercuoan inutiMe(Opnf (openj Repmentattm of ta Atomic Industnal Frids y, Sepsnaber 7, w?S
' ne Comnee 4 hear and dmaa da Pc. rum Poitor Commsttee on Followarp to be l
Three Mile Island Accident *nu benrf 6 w AR-r:x PM: nt=e Mae Island Unit
' 2 Accident Implicatons (Opmf us te re to Cy Comuee ng&g m proposed actnd" scuvitas. Opweens tasota ne Cam.~.ittee wdl discusa undadying causes which contntuted to the accident h21 AR-225 Noorr Morang wrd NRC im PR-wpH Meeting with NBCSk:;jr which occur ed at ce nree Mle Island S W (Open) (Open)
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31876' Federal Register / Vol. 4 No.173 7 Wednesday, September 9,1979 / Notices l Lx PhfA30 Pht.: Leevare Session (Open) of impediments to international trade la September 14-13.1979, the Comndsslon The Committee will hear and discuss the services. Public views and comments will meet in closed session to discuss its report of its subcommittee Chstrman conceming foreign barriers to U.S. Endings and recommendatiens. regarding a proposed plan of accon to review service industnes were solicited in These meetings will be held pendm, g and evaluate the status of genene itema connection with the Tokyo Round of cottflcation and approval by the CSA * , which were listed in ACRS Report No. 7 Multilateral Trade Negotiations. (Seet 42 Administratot. dated March 21.19*9 es applicable to light
- FR 3M45,42 FR 59784. 42 FR 38445.) inquiries should be addressed to ,
wster nudear reactors.
- For further infor: nation concerning Barbara Jorgenson (202/853-7677).
P 'd A rec m end con to e ar services and service industries. Dated: August 31.1979. Regulatory Comnusaton regarding reference shoul,d be made to:, U.S. hibars largensoo. tmplicauons of the accident which occurred Service Industries in World Marxels: gg gj, at the Three Mde island Nudear Station Unit Current problems and future Policy
- 2. Development, U.S. Dept. of Commerce, The Committee will discuss the proposed 1978, avedable for purchase by scope and organization of its ar.nual report to reference to publication No. PB282528 Congress oo the NRC Safery Research from the Department of Commerce, Program. National Technical Information Service, DEPARTMENT OF STATE Saturday, September s.1975 Washington, D.C. 20230* Office of the Secretary ex A.Af.4x p.3f.:Lecuure Session l0###I inven a of im e t (Pubuc Notice CM-4/221)
& Committee will discusa proposed intemational trade in services will be '
ACRS comments and recommendations ava lable for public inspection by National Committee of the U.S.
' Septemb/r 7,1979 at the Office of the Organization for the International
$ c at Nile sbad on March 28.1979,its proposed report to the Special Representative for Trade f Negotiations. Room 735,1800 C Street.
Telegraph and Telephone Consultative Committee (CCITT); Meeting ? NRC regarding evaluation of Licensee Event NW., Washington. D.C. The Departrnent of State announces Reports, and proposed ACRS posioon/ For further information contact: comments regarding other matters discussed that the National Committee of the U.S. dunns this mentag Nancy Adams or Dorothy Dweskin. Room Organization for the International 1 735.1800 G Street. NW, Washington, D.C. Telegraph and Telephone Consultative ! 2G506. (Telephone 202-395-5140). - ac ci t d to t e a lop et Committee (CCITT) will meet on-quanutsove nsk cnterta for nuclear facilities. C. Micha.1 Hathaway, September 24,1979 at 1:00 p.m. En Room The Committee will discus 4 proposed Acung Cenem/ Counsel, 1912. Department of State,2201 C Street resolution of Anticipated Transients Without % r,,. w w % m w .3 N.W. Washington. D.C. Scram for Boiling Water Reactors. est t.asa cooe siw The National Committee assists in the The Commuttee will discuss its schedule for resolution of admittistrative/ procedural 4 future acuvities and ce activities of individual members. problems pertaining to U.S. CCITT I Dated: September 4.19-9. PRESIDENT'S COMMISSION ON THE activities: provides advice on matters of ACCIDENT AT THREE MILE ISLAND policy and positions in the preparation John C. Hoyl*- for CCITT Plenary Assemblies and l Advisory Commir:ee Manegement Officer. Meetings meetings of the intemational Study W D'=JS-NM *' 2" "I In accordance with the Federal Groups: provides advice and emo coos is"-" recommendations in regard to the work Advisory Committee Act (Pub. L 92463), artnouncementis made of the following of the U.S. CCITT Study Groups: and OFFICE OF THE SpECIAL meetings: recommends the disposition of proposed Name: ide U.S. contributions to the intemational i REPRESENTATlVE FOR TRADE ogi sien on the NEGOTIATIONS CCITT which are submitted to the i place: Washington, D.C 2 co M Street. N.W. Committee for consideration. Impediments to intemational Trade in Time: Wednesday. September 12. 9 as-.-4 The purpose of the meeting on Services; Solicitation of Publio Views p.m. Thursday, September 13,9 as-a September 24 is to review the , and Comments ps Fnday. September 14. 9 as--a ps Organization's charter, which must be Sa turday. 5eptember 15. 9 as-6 ps renewed on October 15, and to discuss
- 1. Solicitation of Views: Preparation Proposed Agenda: prsparations for U.S. participation at the for discussion in the Organization for I. Discussion of staa reports. Vllth Plenary Assembly of the CCITT. to Economic Cooperation and u. Discussion of findings and be held November 10-21,1980, at Development (OECD) concerning recommendations. Geneva, Switzerland.
Impediments to intemational trade la tu. Discussion ofissuance of subpoenae ad Members of the general public may services is now underway..It is :estificandum and duces tee-un. attend the meeting and join in the contemplated that public sector IV. Oder bustness- discussion subject to instructions of the advisory committees will be formed to The Commission was established by Chairman. Admittance of public provide advice on this subject. Pending Executive Order 12120 on Apnl11.1979. members will be limited to the seating the avatlability of advice from such to conduct a comprehensive study and available. In that regard, entrance to the committees the Office of the Special investigation of the accident involving Department of State building is Representative for Trade Negotiations is _the nuclear power facility on Three Mile controlled and entry will be facilitated if particularly laterested in receiving Island in Pennsylvania. arrangements are made in advance of public views and comments and hereby On September 12-13,1979. the the meeting. It is therefore requested solicits public views and comments on Commission will meet in closed session that prior to September 24.1979. work undertakan to prepare an to discuss staff reports and issuance of membys of the general public who plan inventory to define and categorize types any additional subpoenas. On to attend the meeting inform Mr.
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Revised: Sectemoer 5, 1979 OETAILED SCHEDULE AND OUTLINE FOR DISCUSSION 233RD ACRS MEETLNG SEPTEMBER 6-8, 1979 WASHINGTON, DC Thursday. Sectember~ 6,1979, Room 1046,1717 H Street, NW, Washington, DC
- 1) 8:30 A.M. - 9:00 A.M. Executive Session (0 pen) 1.1) 8:30 A.M. - 9:00 A.M. - Chair-man's Report (MWC) 1.1-1) Request from NRC/TMI
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Special Inquiry Group for comments regarding NRC licensing and re-view process 1.1-2) Request from President's Commission on the TMI Accident for comments
; regarding review of the TMI-2 accident 1.1-3) Results of pressure sup-pression test program -
areas of high stress in Mark I containment
- 2) 9:00 A.M. - 12:00 N00N Meeting with NRC Staff (Ocen) l 2.1) Report on Lessons Learned and l other NRC task force activities regarding TMI-2 accident !
implications
- l 12:00 Noon - 1:00 P.M. LUNCH
- 3) 1:00 P.M. - 2:00 P.M. Meeting with Stone & Webster Engi- !
neering Corocration (Ocen) l 3.1) Comments on seismic design l of nuclear power plant systems I (Portions of this session will be closed as necessary to discuss , Proprietary Information related i to tnis matter)
9
- 4) 2:00 P.M. - 3:00 P.M. , Nuclear Ocerations Institute 4.1) Briefing on proposec activities of the Nuclear Operations Insti-tute
- 5) 3:00 P.M. - 4:00 P.M. Meeting with NRC Staff (Ocen)
Report regarding Mark I containment areas of potentially high stress
- 6) 4:00 P.M. - 7:00 P.M. Executive Session (Ocen) 6.1) 4:00 P.M. - 6:00 P.M. - Discuss proposed ACRS report to NRC on Evaluation of Licensee Event Re-ports (DWM/AB) 6.2) 6:00 P.M. - 7:00 P.M. - Report of ACRS Subcommittee on Plant
. Arrangements regarding evaluation of systems interactions at the Zion Nuclear Station and Indian Point Nuclear Plant Unit 3 (MB/RKM)
Friday, Sectember 7,1979, Room 1046,1717 H Street ~, NW, Washington, DC
- 7) 8:30 A.M. - 12:30 P.M. Executive Session (0 pen) 7.1) Discuss uncerlying causes contribut-ing to TMI-2 accident (MWC/RFF) 12:30 P.M. - 1 :30 P.M. LUNCH
- 8) 1:30 P.M. - 6:30 P.M. Executive Session (Ocen) 8.1) 1:30 P.M. - 4:30 P.M. - Discuss proposed plan to review and re-solve generic items applicable to light water reactors (MB/JCM) 8.2) 4:30 P.M. - 5:30 P.M. - Discuss proposed ACRS recommendations on TMI-2 accident implications (00/RKM) 8.3) 5:30 P.M. - 6:30 P.M. - Discuss proposed organization and scope of ACRS Annual Report to Congress on the NRC Safety Research Program (CPS /0Z)
B Saturday, Sectember 8,1979, Room 1046,1717 H Street, NW, Washincton, DC
- 9) 3:30 A.M. - 12:30 P.M. Executive Session (Ocen) 9.1) 8:30 A.M. - 10:30 A.M. - Discuss
- p. oposed ACRS comments regarding NRC licensing and review process and TMI-2 accicent implications 9.2) ~10:30 A.M. - 11:30 A.M. - Discuss proposed ACRS report to NRC on evaluation of LER's 9.3) 11 :30 A.M. - 12:30 P.M. - Develop-ment of quantitative risk criteria for nuclear facilities (00/GRQ) 12:30 P.M. - 1:30 P.M. LUNCH
- 10) 1:30 P.M. - 3 :3G P.M. Executive Session (Ocen/ Closed) 10.l) 1:30 P.M. - 2:30 P.M. - Dis-cuss proposed resolution of ATWS for Boiling Water Reactors (WK/TGM) 10.2) 2:30 P.M. - 3:00 P.M. - Future Senedule 10.2-1) Anticipated ACRS Sub-committee activity 10.2-2) Anticipated ACRS activity 10.2-3) Proposed ACRS review of grid-stability (JJR/GRQ )
10.2-4) Meeting dates for CY 1980 l 10.3) 3:00 P.M. - 3:30 P.M. - Miscel- I laneous 10.3-1) Activities of Members (Closed) - Discuss activities of Dr. Dade W. Moeller in support of EPRI research ; contract with the Harvard ; School of Public Health i 1 00 3
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Issue Date: November 21, 1979 (FOIA EXEMPTICN (b) 5) MINtRES OF THE Ig 233RD ACRS MEETING . J SEPTEMBER 6-8, 1979 f j WASHINGTCN, CC ' gg The 233rd meeting of the Advisory Comm'ittee on Reactor Safeguards, held at 1717 H Street N. W., Washing ton, CC, was convened at 8 : 30 a.m. , Thursday, September 6, 1979. (Note: For a list of attendees, see Appendix I. Mr. Mark was not present on Friday, September 7.] We Chairman noted the existence of the 'published agenda for this meeting, and the items to be discussed. He noted that the meeting was being held in confonn-ance with the Federal Advisory Committee Act (FACA) and the Government in the Sunshine Act (GISA), Public Laws 92-463 and 94-409, respectively. He noted that no requests had been received from members of the public to present oral state-ments. He also noted that copies of the transcript of some of the public portions of the meeting would be available in the NRC's Public Document Room at 1717 H Street, N. W., Washington, CC, within approximately 24 hours. (Note: Copies of the transcript taken at this meeting are also available for purchase from Ace-Federal Repo rte rs , Inc., 444 North Capital Street N. W., Washington, CC 20001.] I. Chairman's Reoort (Open to Pablic) (Note: Raymond F. Fraley was the Designated Federal Employee for this portion of the meeting.) A. Reviewers ; The Chairman named Messrs. Bender and Ray as reviewers for the 233rd ACRS Meeting. B. NRCMMI SPECIAL INQUIRY GROUP The Chairman noted that a request was received f rom' the NRC/IMI Special Inquiry Group to interview the Members of the Committee, preferably during the time that the Committee will be in Washington for its 234th Meeting, for a period of approximately 4 hours, he Members agreed to be interviewed, probably on Saturday, October 6,1979, to answer questions and provide personal views on matters touching on the inquiry. 1
MINUTES CF THE 233RD ACRS MEETING SEPTEMBER 6-8, 1979 C. Kemeny Cormission Recuest The Chairman noted that the Committee, as a tedy, has not received a query from the Kemeny Cocaission, but that letters were sent by the Kemeny Cocmission to Members as , individuals (see Appendix V) . He also took note of Mr. Lewis' reply to the Kemeny Commission's query (see Appendix V!). D. Assignment of Review of LERs to ACRS Fellows The Committee agreed that a routine review of all LERs should not be made by ACRS Fellows. Specific reviews in selected areas will be done on a case-by-case basis. II. Meeting with NRC Staff on the Implications of the TMI-2 Accident (open to Public) (Note: Richard K. Major was the Designated Federal Employee for this portion of the meeting.] A. Subcomittee Recort i Mr. Okrent, Subcommittee Chairman, briefed the Committee on the matters discussed at the September 5, 1979 Subcommittee Meeting in the order of that meeting 's agenda. He noted that members of the NRC Staf f reported on the status of the Systematic Evaluation Pro-gram (S EP) , and that, in his opinion, the rate of pecgress of this study is inadequate. (Note: At a later portion of this meeting, I a presentation will be made to the Full Committee of the status of I the SEP.] He said that the Subcocmittee also received a preliminary oral report from the Long-Term Lassons Learned Task Force, but that the task force has not yet completed its work, and there is no writ-ten report available at this time. He suggested that the NRC Staff l is having difficulty in reaching positions and/or conclusions on l some of the more complex issues raised by the accident. Itfe NRC Staff I has not made recommendations yet regarding improvements for future nuclear power plants, nor have they reached conclusions regarding backfitting requirements for operating plants and those plants still l under construction. The NRC Staff discussed a proposed defense-in- ! depth concept that is still in a formative stage. Questions raised ; here involve the adequacy of engineered safety features and the current l siting policies. He said that in considering protection beyond current i design bases, the NRC Staff has outlined only one of several options, i l Mr. Okrent suggested that the Committee should interact with the NRC j Staff to develop approaches that best protect the public health and I safety. 2
; l MINUTES OF THE 233RD ACRS MEE"'ING SEF" EMBER 6-8, 1979 In answer to a question, W. Minne rs , NRC Staff, said that the mat-ters which the NRC Staff has brought to the Committee and Subcommit-tee are currently in a preliminary stage of development. Currently, the NRC Staf f is thinking in terms of a five-year plan, considering operating reactors and those plants currently under construction.
Accidents are being redefined, and the engineered safeguards must mitigate those accidents. At Mr. Okrent's request, W. Minners reiterated the proposed defense-in-depth concept that he discussed with the Subcommittee on September 5 (see Appendix VII). Mr. Okrent noted that the NRC Staff is considering the development of _ an approach to supplement the Single Failure Criterion. Conside r-ation is being given to the application of fault tree and event tree 1 analysis methods, to assess the role of operator action. We recommen-dation is that with the incorporation of these probabilistic methodol-ogies, and with some review, the current Sirgle Failure Criterion could continue to be used. He said that with respect to water-metal reactions, the Long-Term Lessons Learned Task Force Group does not appear to be able to reach a recommendation before the task group will be dissolved. 4 It is a matter which will require further study by the Committee. Mr. Okrent suggested that one of the current problems facing both the Commission and the Committee is how safety questions can be placed on the table for discussion, without interfering with the regulatory process. He also noted that the NRC Staff is reversing a position taken in one of its early bulletins following the TMI-2 accident, namely that reactor circulation pumps should be lef t on following a , small LOCA. his may be a confusing matter, and should be carefully considered before any final decision is made. B. Assessment of OtI-2 Accident J. Voglewede, NRC Staff, said that an Atomic Safety and Licensing Board asked the NRC Staf f to determine whether the classification scheme in the proposed annex to Appendix D of 10CFR 50, applies to the Pu-2 accident. In effect, the question asked whether the accident was a Class 9 accident. He noted that although this annex has never been adopted, it was originally published in the Federal Register in 1971. De annex also appears as an appendix to Regulatory Guide 4.2, l which deals with the preparation of environmental reports. J. Voglewede said that the NRC Staf f has determined, assuming the I , strict interpretation of the words in the proposed annex, that the
'IMI-2 accident was a Class 9 accident. Bis conclusion is based on 3
l [ l l
l I 1 MINUTES OF THE 233RD ACRS >tEETING S EFl'mBER 6-8, 1979 the proposition that the sequence of f ailures at TMI-2 were more severe than those used in the design bases. Specifically the two successive failures at DtI-2, a small break LCCA and the failure of the ECCS, are not normally postulated in the design basis. On the other hand, with respect to consequences, which in fact were neg-ligible, the accident cannot be considered Class 9. But, based on the annex, it was concluded that the TMI-2 accident was Class 9. Mr. Kerr pointed out that we now have a situation in which consequences of Class 8 accidents can be more severe than those of Class 9. In answer to a question regarding dissents from this opinion by members of the NRC Staff, J. Voglewede listed the followirg: e a stuck-open PCRV should not be considered a small break LCCA, e the throttling back of high pressure injection should not be considered a failure of the ECCS, and e since the physical integrity of the containment structure was maintained throughout the accident, failure of the contain-ment should not be considered to have occurred. (For details of J. Voglewede's presentation, see Appendix VIII; for the memorandum from R. J. Mattson to G. H. Cunnirgham III, answering the ASLS question concerning Class 9 accidents, see Appendix IX). Several members of the Cocmittee concluded that Annex D is deficient and cannot be used for the classification of accidents. In answer to a question, P. Novak said that the Commissioners have not considered the issue of whether the StI-2 accident was a Class 9 accident. . With regard to the issue of whether reactor recirculating pumps should be stopped following a small break LCCA, Mr. Ebersole suggested that the operating instructions that are issued should state when these pumps should be restarted. With respect to the NRC Staff's proposed position regarding siting policy, a member suggested that not much can be done with existing plants beyond monitoring nearby population and/or installiry additional engineered safety features. He said that while attention to siting is necessary in the near future, there is a question of the current timing. He suggested that it is more important to ded with existing plants and those under construction, first, before being involved in future siting requirements. He also recommended that the Committee pay close atten-tion to, and review, the NRC Staff's proposals. 4
MINUTES CF THE 233RD ACRS MEETING SEPTDtBER 6-3, 1979 i III. Meeting with Representatives of the Stone and Webster Engineering Company on the Seismic Design of Nuclear Power Plants (Ocen to Public) i (Note: Elpidio G. Igne war the Designated Federal Employee fo r this portion of the meeting] _ 3
- (For background material of the following discussions, see Appendix X) j A. Introduction W. J. L. Kennedy, Stone and Webster Engineering Corporation, pro-vided the Committee with the background issues relating to the seismic design analysis used at five licensed nuclear power plants i for which the NRC Staf f ordered a shutdown and reanalysis (see Appendix XI) . He noted that the reanalysis showed that because of ;
other design conservatisms, there was no need for pipe replacement, )
- but that in some cases, modifications were made to pipe support i i structures because Stone and Webster determined that these modifica- l tions were cheaper than detailed stress analyses of the affected pipes. . He restated the Stone and Webster position that the affected plant piping was safe as initially designed. He said that S&W found )
no rules against using the algebraic method of stress summation. This algebraic summation was used in plants designed prior to 1970. He i further of fered the opinion that well laid out piping will meet l
- requirements in almost all cases.
l B. Survey of the Seismic Performance of Stone and Webster-Designed Piping J. R. Hall, Stone and Webster, discussed the seismic perfo rmance 4 of welded piping (see Appendix XII) . He said that welded piping j has never been known to fail f rom seismic events, even when no consideration had been given to such events during design. In his opinion, welded piping is very resistant to seismic effects, and there ! is no reason to expect f ailure resulting f rom seismic events. I C. Modifications Made E. J. Siskin, Stone and Webster, discussed the mcdifications made j , to the affected plants to expedite their return to service. (Fo r his verbatim remarks, see Appendix XIII) . D. Conservatisms of the Original Seismic Analysis K. F. Reinschmidt, Stone and Webster, discussed the conse rvatisms - employed in the original seismic analysis (see Appendix XIV), and concluded that the reanalysis was unnecessary.
, - _, , , _ a u,_ xa .a . . _ . - . -
l 1 l MINUTES CF THE 233RD ACRS MIETIm SEPTEMBER 6-8, 1979 l i IV. Briefing on the Proposed Activities of an Institute of Nuclear Power l Ocerations (Cpen to Public) Fraley was the Cesignated Federal Employee for this l (Note: Raymond F. j -! portion of the meeting.} j W. Lee, Duke Power Company, representing the Electric Power Research
- Institute (EPRI), briefed the Committee on activities planned for the proposed EPRI Institute of Nuclear Power Operations (see Appendix XV) .
He concluded that one of the purposes of the institute is to change attitudes in power plant control rooms from one of complacency to one of alertness in which it is recognized that at any time, in any shif t, an accident can occur. i In answer to a question regarding the sources of personnel for this institute, W. Lee said that EPRI plans to recruit from the nuclear in-dustry, electric utilities, NASA, and from other sources that have person-nel of the required expertise. ! W. Lee also noted that EPRI is planning to establish an Institute of Nuclear Systems and Equipnent (INSEQ) to consider mechanical problems with regard to nuclear power plants. V. Meeting with the NRC Staf f on the Test Results From the Mark I Full-Scale Facility (open to Public) (Note: Andrew L. Bates was the Designated Federal Employee for this portion of the meeting.] B. Grimes, NRC Staff, provided information on what had been considered a potential problem in the design of the General Electric Mark I contain-ment System (see Appendix XVI) . He described the Mark 1 Full-Scale Test Facility. Data and analyses obtained in recent tests in the Full-Scale Test Facility have shown that the potential problem does not develop, and that with tie-down straps on the downcomers, as currently provided in the operating plants, there is a safety factor greater than two in the f atigue usage during a design basis accident. Owners of plants utilizing the Mark I Containment System have been notified not to re-move the downcomer tie-downs. , VI. Meeting with Members of the NRC Staf f on Systematic Evaluation Program (Open to Public) (Note: Richard K. Major was the Designated Federal Employee for this
- portion of the meeting.)
6
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MINUTES OF THE 233RD ACRS MEETING SEPrEMBER 6-3, 1979 Mr. Okrent noted that the following presentation on the NRC Staf f's Sys-tematic Evaluation Program (SEP)' was presented at the T'.I-2 Implications Subcomittee Meeting, and that the Subcomittee believed that the status ard proposals for this program should be heard by the Full Comittee. R. Vollmer, NRC Staff, provided an overview of the SEP (see Appendix < , XVII). He said that the reviews have been delayed for up to three years because of manpower limitations within the NRC Staff. In answer to a question, R. Vollmer said that the SEP is writing the criteria for definition of deficiencies. He said, howewe r, that the determination of a deficiency is based upon the judgment of the reviewer. A member cueried the methodology used in the documentation of the pro-gram, and of fered the opinion that the program has not identified the
- serious problems that he believes probably exist in the older plants.
He was also of the opinion that the SEP was progressing too slowly. The Comittee agreed to consider further the SEP at its 234th ACRS Meeting (October). VII. Executive Sessions (Open to Public) (Note: James M. Jacots was the Designated Federal Employee for this portion of the meeting.] A. Foreign Meetings The Comittee requested the Executive Director to make preliminary arrangements for meetings in Europe with both the French Group Permanente and the German RSK. The desired dates for the sit-down ; meetings are during the week of May 18, 1980 for both, with facility visits to be scheduled for the weeks immediately preceding and follow- , ire the sit-down meetings. Members were requested to inform the l Executive Director of the topics they believe appropriate for discus- l sions and the facilities they wish to visit. j 1 B. ACRS Meeting Schedule for 1980 l l
'Ihe Comittee agreed on a schedule for its meetings during calendar l year 1980 (see Appendix IV) .
C. Subcomittee Recott
- 1. Plant Arrangements l l
Mr. Bender, Subcommittee Chairman, said that he was reporting , on a Subcommittee meeting that he had been unable to attend. He said that in connection with the Comittee's request that Indian j , 7 l l t l' i l
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i , 4
. 1 i l
- MINtfrES OF THE 233RD ACRS MEETIM SEMEMBER 6-3, 1979 l-I Point Unit 3, conduct a " systems interactions study", the licensee l has proposed that it be permitted to respond in the same manner as ,
1 Zion had done: 1.e., on the basis of a review of LERs. It was ' i the sense of the Committee that it recogni::e the Zion study as i being useful, but also that this study of LERs did not fully meet j the intent of the Committee's request concerning Indian Point 3. j ! The Committee agreed to consider some proposed amplification of l l Committee interest regarding this activity at its 234th Meeting l 1 (October). j (For background information sterial regarding the Zion study, j see Appendix ,WIII.) i
- 2. Reliability and Probabilistic Assessment j .M r . Okrent, Subcommittee Chairman, announced that the Subcom-i mittee would meet in Los Angel 9s within a week to plan activities for the development of quantitative risk criteria. He said that
- he had met with Commissioner Ahearne, who was interested in the l Committee's plans regarding this matter and in the available l literature on this subject. Preliminary plans include the following
l e participation of interested foreign groups, French and q German, d e sy:nposia with. people of diverse backgrounds, to be held in ! January or February, 1 I e solicit input from all known groups who might have an l interest in cooperation in a project of this type, i j e obtain input regarding . legal considerations, j ] e obtain input regarding economic considerations,
; e solicit cocznent from other government agencies and groups, I
i e obtain input frem mjor insurers to learn how they deal with risk, and l e obtain accident data from the National Safety Council, t One question that should be resolved early is ' whether the pro-ject should consider only individual risk, or whether societal j risk should also be considered. 8
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4 MDR7FES OF THE 233RD ACRS MEETING SEPTE9ER 6-3, 1979
- 3. Anticicated Transients Without Scram (ATrS)
(For background on the current status of ATriS, see Appendix XX.) Mr. Kerr, Subcommittee Chairman, said that one of the current ATds concerns regarding boiling water reactors relates to the potential for a pressure pulse generated by fast closure of the main steam line isolation valve. A further short time concern is the rate of the response to a recirculation pump trip. On a longer time scale, there was concern that a temperature increase in the pressure suppression pool could cause chuggiry that might structurally damage the torus. Mr. Kerr said that he plans to discuss recently received documents with Messrs. Thadani and Hanauer, both of the NRC Staff. D. Electric Grid Stability Mr. Ray discussed electric grid stability, a concern raised by ACRS Consultant E. P. Epler (see Appendix XIX) and disagreed with Mr. Epler's suggestion . that nuclear sources have contributed to grid instability. He explained that grid stability involves the ability of a system to ride out an electrical transient in the bulk power system caused by one or more of its feeding generating units without cascading into a general system failure, an ability which is not influenced by whether or not the units are nuclear. Mr. Ray identified E. P. Epler's ccmments as express-ing a concern that, in the future, a widespread industry shortage of electrical generating reserve capacity resulting f rom the current economic situation could lead to unreliability in electric grids because of the relative unreliability of nuclear units stemming from frequent adverse systems inte ractions. To offset this potential, E. P. Epler recommended that a comprehensive review be made of the adequacy of the interfaces between electrical systems within plants since these systems are generally designed by different engineering organizations in the industry (e.g. the NSSS vendor, the A/E for balance of plant, etc.), to establish the possibility of improved reliability. Mr. Ray rioted that, though the forced outage rate of nuclear plants is only slightly higher than that of all other plants, it is considerably lowr than that of coal-fired plants of 400 MWe capacity and larger. He disagreed that there would probably be a general industry shortage of generating reserves, since the Federal Power Commission and rest state commissions ; now require that adequate system recerves be installed. He suggested I the likelihood that, in the immediate future, new generating capacity will probably be coal-fired, and that these coal plants are relatively unreliable, but he agreed with Mr. Epler that plant systems interactions and resulting failures seem to occur too f requently. He suggested, i however, that the proposed review of interfaces to improve reliability l wuld be a very large task and that, before such a study is undertaken, l a review should be made of operating experience to determine the facts l 9 ) l 1 _----___-_m__.- -___.__. __.___- . _ _ _ . - - - - _ - -
i i MINtfrES OF THE 233RD ACRS MIETnJG SEPrD13ER 6-3,1979 on the reliability of nuclear plant electrical facilities and the scope of their influence on plant availability. With this in hand, it should be evident where one should best direct efforts to improve plant reli-ability. He suggested that the Committee follow the recommendation of the Subcommittee to Evaluate LERs, and use the information available in the LER files for such purposes. There is already,available a printout of all electric power related LERs for 1976 throtyh 1978 (approximately 1100 LIRs) . An ACRS Fellow could be assigned to review these LEPs and identify system deficiencies and failure causes. Printouts could be obtained on control systems also, should it become desirable to expand the sttrly. He suggested that these matters be referred to the Power and Electrical Systems Subcomittee. It was agreed that a Fellow would be assigned to review the power-related LEPs and prepare a report on their significance for consideration by the Power and Electrical Systems Subcommittee. Mr. Ray proposed that E. P. Epler's recommendation fo r a dedicated reactor heat removal system, a protected bus for engineered sa fety features that would not be affected by degraded in-plant voltage, and the susceptibility of boiling water reactors to rod crop accidents be referred to the appropriate subcommittees. E. Review of ACRS Generic Items List The Committee reviewed its lists of both resolved and unresolvec generic items, and agreed to assignments to subcommittees for re-view and evaluation of generic items (see Appendix XXI) . ':he Com-mittee also agreed that designated subcommittee chairmen should provide the Ccemittee with plans of action regarding the generic items within tw months, and report on their actions within six months. F. ACRS Reoort to Congress on NRC Research Procram The Committee agreed, subject to the approval of the appropriate Congressional oversight committees, to defer completion of its an-nual report to Congress on the NRC Research Program until Februa ry. The purpose of the delay is to allow inclusion ir the report of the actual NRC budget provided to Congress rather than the NRC's pro-posed budget. The Committee also agreed that a working group of ACRS Staff members and Fellows be assigned the task of providing a "first cut" on risk reduction potential for the research items. G. Review of Nuclear Reculatory Function and Progress The Committee provided its comments on a draf t report on a review of the nuclear regulatory function and process. Mr. Sender agreed to incorporate these comments into a further draf t of this report to be considered at the 234th Meeting (October) . l 10 l l
t b A s 4 l MDR7FES OF THE 233RD ACRS MEETING SEPTEFSER 6-8, 1979 i H. Future Agenda , he Committee agreed on a tentative agenda for the 234th ACRS Meeting (October) (see Appendix II), , The Committee endorsed a suggestion that the Wolf Creek Subcommittee l
- Chairman review interveno r submissions regarding the seismic design bases for this plant and report to the Committee regarding the need for i Full Committee action.
he Committee agreed to examine its position regarding applications
- of the MI-2 experience to those plants for 'ahich an ACRS Construction j report has been provided but a CP has not been issued (e.g., Black Fox) . ;
The ACRS Staff was asked to provide a list of such plants for considera- ; tion during the 234th ACRS Meeting (October) . , I. Honor to ACRS Consultant i The Chairman noted that H. H. E. Plaine, ACBS legal consultant, has been l elected Vice Chairman of the Judicial Administration Division of the I American Bar Association. J. ACRS Recorts and Letters ; I
- 1. Evaluation of Licensee Event Recorts 1
te Committee completed its sttx$y, and approved a final repo rt, Review of Licensee Event Reports (1976-1978) (NUREG-0572) (see Appendix XXII) . I
- 2. Response to Questionnaire from the NRC/TMI Scecial Incuiry 1
- Group i
The Committee approved a letter responding to a questionnaire
- received from the NRC/MI Special Inquiry Group, dated August 7, 1979 (see Appendix XXIII) .
- 3. Letters to Members of Congress Procesing Delay of the ACRS
- Annual Recort on the NRC Safety Research Program l
The Committee approved letters to Representative Morris K. Udall and Senator Gary Hart procosing a rescheduling of the annual ACRS l report on the NRC Safety Research Prcgram (see Appendix XXIV). ) l 1 11 I l
. 1
e MINifrES CF Thr.: 233RD ACRS 'ICTING SFSTEMBER 6-3, 1979 I l l VIII. Executive Sessiens (Closed to Public) l l l A. Conduct of Members l l
- 1. I u }
I
'l Y
The 233rd ACBS Meetire was adjourned at 2:40 p.m. , Saturday, September 8,1979. F h 12 l . - - . . -. . . - __
e i APPENDIXES TO MINUTES OF THE 233RD ACRS MEETING SEPTEMBER 6-8, 1979
~ 233rd ACRS Meeting Meeting Dates: September 6-8, 1979 APPENDIX I ATTENDEES ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Max W. Carbon, Chairman Milton S. Plesset, Vice-Chairman Myer Bender Jesse Ebersole William Kerr Stephen Lawroski William M. Mathis Dade W. Moeller ! David Okrent l ACRS STAFF Raymond F. Fraley, Executive Director Marvin C. Gaske, Assistant Executive Director James M. Jacobs, Technical Secretary Herman Alderman Andrew L. Bates David E. Eessette Q U 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 ACRS CONSULTANT W. Lipinski C. Michelson
O NRC ATTENDEES 233R0 ACRS MTG, Sept. 6, 1979 Div. of System _s Safety Div. of Research J. Voglewede L. S. Tong A. Cappucci 0AB Div. of Operating Reactors F. Allenspach M. Hartzman C. Grimes 5tds. Development E. G. Adensam M. Medeired Management and Program Analysis E. Boyle b- T. Cintola J. Crooks O l l
l o PUBLIC ATTENDEES AT THE 233rd FULL COMMITTEE MEETING ; 1 September 6,1979
)
Name Affiliation ! l Joe Hebert AP , Roger P. Smith McGraw-Hill (New Jersey) ' Michael Hurp J' K Atomic Energy Authority (Va) John B. Hoch Pacific Gas & Electric Co. H. Kimmins Washington Public Power Supply System T. Martin NUTECH (Va) R. Borsum B&W Minie Meltzer Ace Federal Osen H. Davis PG&E Jim Schermbeck American University Scott Norsworthy American University Marie Newman EPRI O September 6, 1979 - Afternoon
- 0. C. Aabye Offshore Power Systems, Jacksonville, FL T. Tramm Commonwealth Edison, Chicago, IL N. P. Mathur PASNY, West Paterson, NJ 07424 L. Xornblith EPRI, Bethesda, MD Leyse EPRI Barry W. Roberts Dept. of Energy '
S. Kraft EEI L. S. Gifford GE lO l l l
/O
'v' INVITED ATTENCEES 233RD ACRS MEETING Sect. 6,1979 '
Atomic Industrial Forum Duke Power Patrick Higgins K. Canady F. S. W. S. Lee J. M. Maffre Stone and Webster A. Giambusso W. Kennedy S. Jacobs K. F. Reinschmidt P. A. Wild , R. P. Klause G. Flynn < C. Grochmal M. Holley E. Siskin J. Hall O o. sheve L. V. Senn W. Relal EEI S. Kraft l l () v l l
1 1 O NRC ATTENDEES 233RD ACRS MTG. Sept. 7, 1979 l Robert Baer l
)
i O l I i i I I l l l l O l I l
L ' O PUBLIC ATTENDEES i 233RD MTG. l l September 7, 1979 J. Schernbeck Wash., DC R. Borsum 8&W, Derwood, M0 J. Omang Wash. Post, D. C. J. B. Hock PG&E, 77 Beck St., San Francisco, CA 1 R. Leyse EPRI l B. Montgomery Bechtel R. P. Smith McGraw Hill l P. Higgins Atomic Industrial Forum R. E. Schaffstall KMC (3 v R. G. Neve S tafco E. L. Cox Stafco J. Giannelli U. S. Senate John Austin U. S. Senate September 8, 1979 Paul V. Holton Bechtel, Falls Church, VA E. Entwisle Environmental Law Inst., D. C. Patricia A. Young Self Patrick Higgins AIF, Falls Church l v)
i i i
=
2 APPENDIX II ITEMS FOR DISCU53 ION AT 234th ACRS MEETING October 4 o, 1979 i Review of Regulatory F roce s ses and Procedures (MB) 8 hou r s Diablo Canyon (DO) 3 hours Zimme r (DO) 2 hours Bulletins and Orders Subcommittee Report (WMM) 1/2 hour ACRS Report on NRC Rese arch Program (CPS) 4 hours Abnormal Operation Repo. 3 ( AORs) (DWM) 1 Hour Systemmatic Evaluation P ogram (DO) 1/2 hour Ice Condense r Containmer .s (CPS /DO) 2 Hours System Interactions (7 an/IP-3) (DO/ JC E) 1/2 hour Future Schedule 1/2 hour ACRS Review of Black Fox Seismic Design (HE) O
+f o, UNITED STATES
! y ,-q/[ % NUCLEAR REGULATORY COMMISSION
,i w ASHING TON. O. C. 20555 REVISED: 9/7/79 MEMORANDUM FOR: Dr. Milton S. Plesset, Vice Chairman, ACRS FROM: Robert L. Baer, Chief, Light Water Reactors Branch No. 2, Division of Project Management l
SUBJECT:
ACR$ FUTURE SCHEDULE The project reviews currently scheduled for the next five ACRS full Concittee I meetings are listed on the enclosed table. It should be noted that these schedules l are subject to f orthcoming guidance from the Corrnission on the resumption of l licensing activities. i A 1, . ,~ . _
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Robert L. Baer, Chief Light Water Reactors Branch No. 2 Division of Project Management As Stated j ces w/ enclosure: E. Case J. Knight D. Ross R. Tedesco S. Varga 0. Eisenhut W. Garnill R. Vollmer D. Skovholt B. Grimes F. Williams J. Miller J. St ol z W. Russell
- 0. Parr R. DeYoung L. Rubenstein F. Hebdon T. Speis R. Fraley /
i C. Heltemes M.Libarkin/ W. Ha a s s A. Abell J. Peterson R. Hartfield R. Mattson R. Clark F. Schroeder W. Ross W. Minners. DPM LA's . O i
RWISED:9/7/79 ENCLOSURE - - ACRS FUTURE AGENDA TYPE OF REACTOR SER IS5UE ACRS PEETING DA,it PROJECT REVIEW yENDOR 1- , October
! Resta rt- af.ter GE - -- 94M/7-9 --
- GE.TR show cause order Geology &
Seismology only Diablo Canyon OL W NUREG-0578, dtd 7/79 SC ucylh UL w NUREG;0578, dtd 7/79 deM , s OL W NUREG-Ub /Cdtd 7/ 79
) Zimer OL GE NUREG-0578, dt'd 7/79 November
- O ""*
Decembe r Floati ; Nu: lear ML W 11/1/79 F! ant January LaSalle 1 & 2 OL GE 12/3/79 TMI 1 Restart After S&W 12/3/79 Re fueling February Shoreham OL GE 1/2/80 San Onofre '0L -CE 1/2/80
. Watts'Bar OL W 1/2/80 l
i O
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e
/ c, _
UNITE D ST ATES
/? .. g '; NUCLEAR REGULATORY COMMISSION
,,, ) ADVISORY COMMITTEE ON RE ACTOR SAFEGUARDS
( , j,7.g j,1
./q %,. p WASHINGTON, D C 20555 September 6, 1979 i
1 l ACRS Members POTENTIAL AGENDA ITEMS FOR 234TH ACRS MEETING, OCTOBER 4-6, 1979 1 The following matters should be considered as possible agenda items for the subject ACRS meeting: l
.NRC Staff report on cracking in Boron Injection System piping at TMI-1 and related Bulletin
. Report on Surry Steam Generator Replacement
. Annual RSR report to Congress O "eer-term os for o4eb1o cemro" e"e ice comeemser eiemts (seavore".
McGuire) (Tentative)
.NSAC status report
. Briefing by BWR Owners" group re pipe cracking work
.GETR site seismicity / seismic design
. Meeting with Commissioners:
.TMI-2 implications for new construction permits
. Decisions on RSR funding (e.g., improved safety systems)
. Lack of information to ACRS (e.g., Mark I Containment experiments, Bulletin 70-05C and 79-06C)
Consideration of NUREG-0600, the . Investigation into the TMI-2 accident by I&E, has been postponed until the 235th ACRS meeting, November 8-10, 1979. M. W.' Libarkin , Assistant Executive Director for Project Review t,O v- cc: R. F. - Fral ey ACRS Technical Staff t {
M u a b x _~ e - _ _ _ _ _ _ ____ --__ __.__._m__ - - _ _ _ _ _ _ _ _ ___. mm-
J' 'o, UNITED STATES
! . NUCLEAR REGULATORY COMMISSION
$.y$"f / l.. / 7 ,;
. AovlSORY COMMl1 TEE ON REACTOR SAFEGUAnOS
*Q% g W A$HINC10N. O C. 2cS$$ APPENDIX III s e..*
September 8, 1979 f e 1 d ACP.5 Me: bets
, SCHEDLLE OT ACP.5 St~oCO'XITTEE ME!.TISCS. A';D TOU?.5 The felle ing is a list ef tours and Subco:.ittee tectin;5 cur-rently srbeduled, subject to che approval of the Advisory Cer-eittec ".anc;c:ent Officer. If ycu are listed and cannet attend a =eeting, or if you are net listed but vould like to attend, please advise the ACES Of fice as soch as possible.
. Mos t hot els currently being used b . ACF.S Me:bers in t he de.n-te. n ',.*athin , ten and Ecthesda arecs require a gue. rant ecd reser-vatien if arrivc1 is scheduled after 6:00 p.:. Tailure te use a rer: under these conditiens invelves forfeiture ef the cest.
Tlease advise the ACT.5 Of fice as soon as possible if you cannet
.p attend a ecting for which you are scheduled se that reserve-v tie:s can be cancelled i . ti e to avoid this.
l r
. l j f m. ~
l 6 % - H. W. Libarhin' . Assistant Executive Directo: for Preject Review ec: ACF.S Technical Staf f H. E. Vanderholt
- B. Dundr R. F. Traley D"P T H. C. Cashe D D
J. Jacobs j}F o-J . J, , P ' ' V. i.
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September 13 V 11- 12 Reliability & Prob. Assess. (Los Angeles, CA),(GRQ) - 00, WK, MP , HL , CM 13-14 Fluid Dynamics (Los Angeles, CA) (A8) - MP, CM 18-19 Waste Management (RM/PT) - SL , WK, CM, WM, JR 29 LAC 8WR (Spent Fuel Storage) (JCMcK) - WK, PGS, CPS October 2 TMI-2 Accident Bulletins & Orders (TGM/PB) WM, SL, JE, MB 3(A.M.) TMI-2 Impli . Re Nuc. Power Plant Design (Diablo Canyon /BWRs) (RKM) - 00, MC , WK, WM, MP, CPS (tent. ) 3 (P.M. ) TMI-2 Impli Re Nuc. Power Plant Design,(W/UHI ice condensers) (RS) - CPS, MC, WK, CM, WM, 00, MP 4-6 234th ACRS Meeting 16 Rad. Eff & Site Eval . (RSR Rpt) (RM) - OWM, JCE, SL , JR 17-18 ECCS (AB) - MP, JCE, HE 30 TMI-2 (I&E Rpt NUREG-0600) (RM) - HE, MP, JE, WM No vember 6 RSR (TGM) - 00, CPS, MP, JCM, HE, DWri, SL 7 Reg. Act. (GRQ/SD) - CPS, HE, WK, JR 7 TMI-2 (Impli . Re Nuc. Power Plant Design) (RCl) - 00, MC, WK, MP 8-9 235th ACRS Meeting 14 GETR (San Francisco) (EI) - WK, 00, CPS, MB(tent. ) 15-16 Extreme External Phen.(Los Angeles, CA) (RS) - 00, MC, DM, CPS 29-30 Advanced Reactors (Albuquerque, NM)(RS) - WK, MC, CM, MP, PS ' l l 1 l
( O December 4 RSR (TGM) - 00, CPS, MP, HE, PGS, WK, SL, JCM, DWM 6-8 236th ACRS Meeting O i I 1 I 1 l l O l l l l l
i eiii e-- - -- -
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APPENDIX IV ACRS MEETING SCHEDULE FOR C ALEND AR YE AR 1980 l
)
I i 237th - January 10-12 243rd - July 10-12 238th - February 7-9 244th - August 7-9 239th - March 6-8 245th - September 4-6 240th - April 10-12 246th - October 9-11 241st - May 1-3 247th - November 6-8 242nd - June 5-7 248th - December 4-6
)
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President's Commission i on the Accident at Three Mlle Island v.,vs .. 2100 M Street. NW Wasnington. OC 20037 r i Augus: 8, 1979 ,-
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tu./,c:ux sacc:.c..m ua a.a.- APPENDIX V Titie: Kemeny Commission Query a i s. .,. ., : : J
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e> g Ter assistance in its investigatien, the President's Cormission en the Accident at Three l'.ile Island would like to have you identify those questions that you feel should be answered by this Ccemission. The Ccemission wculd also like to have any suggestions, leads, or infermational material you feel are of relevance to its i investigation. [ t I The Ccesissien's assignment is best defined by qucting as foll::ws the direct charge which the President has given us: I t t The . cure.cse"of the ccenission is to conduct a ! comprehensive study and investiga:icn of the } recent acciden involving the nuclear pcwer facility en Three Mile Island in Pennsylvania. The Cermissien's study and investigation shall include: (a) a technical assessment cf the events and their causes; this assessment shall include, i~ but shall net be limited to, an evaluation , of the actual and pctential impac: of the events en the public health and safety and en the health and safety of workers: (b) an analysis of the role of the managing utility; (c) an assessment of the emergency preparedness and response of the ::uclear Regula: cry Ccemissien and c her Federal, state and 1ccal authcrities: D,
e August 8, 1979
- Page Two (d) an evaluation of the Nuclear Regulatory ,
Commission's licensing, inspection, operation and enforcement procedures as applied to this facility; (e) an assessment of how the public's right to information concerning the event at Three Mile Island was served and of the steps which should be taken during similar emergencies to provide the public with
- ' accurate, comprehensible and timely infor-mation; and (f) appropriate recommendations based upon the Commission's findings.
We plan to use your letter to help us insure that we are not overlooking seme important facet or implication of this accident. Since we are under a very tight schedule with our fina' t report due in the President's office by October 25, 1979, L your reply would be most useful to us if we could rece?.ve it by September 7*. Let me thank you in advance for any assistance you can provide us. If you have any questions, please call Barbara Jorgenson of our staff (202/653-7677). Sincerely, jA i
.' 6 9/
(.M- l John G. Kemeny i Chairman l
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Ti tle: Member's Reply to Kemeny John G. Kemeny, Chairman Commission Query President,s Cocaission on the Accident at Three Mile Island 2100 M Street, NW Washington, D.C. 20037
Dear Dr. Kemeny:
I appreciate your invitation of August 8 to contributeI to your hope work, and regret that I have been so long in responding. that this is not too late to be useful. Though my primary For the record, let me identify myself. vocation is as Professor of Physics at the University of California in Santa Barbara, I chaired the American Physical Society study of Reactor Safety, chaired the Risk Assessment Review Group of the Naclear Regulatory Cormission, and am currently In what a member follows, of the though Advisory Committee on Reactor Safeguards. s I obviously agree with the contents of those first two reports, I speak only as an individual and don' t mean to reflect my views upon any of my colleagues in these other enterprises. O ther material relevant to my views may be found in the Physical Society and Review Group reports, testimony before Congressmen Udall and McCormack, an impromptu speech I gave in May to the Atomic Indus-trial Forum, and an early letter to the Chairman of NRC about the implications of TMI. In the following I will confine myself principally to your , item (a), though some material relevant to your other charges will ! be found interspersed throughout. j A. The Events i By now the sequence of events involved in the '*.ree Mile Island accident is reasonably clear to everyone, and there ; nothing of great importance that I can add to the picture, though I did spend a full day at Three Mile Island with the operators who were on There was a feedwater transient, i duty at the time of the accident. l exacerbated by the closed auxiliary feedwater block valves, followed J by the leak in the pressure-operated relief valve, and a long series l of events in which the situation was sometimes improved by human l in te rven tio n, but all too often made worse. I am perhaps more l charitable than most in declining to second-guess people operating l in a crisis, but it is nonetheless true, and widely understood, l l that this accident need not have been as bad as it was. . f l [ l l i
August 31, 1979 John G. Kemeny O B. Operator Training You will have elucidated the exact errors of intervention, but it is quite clear that the training of these operators was funda-mentally inadequate to the job they were called upon to do at four in the morning. I believe that this is less the fault of the indi-viduals concerned than it is of superimposing a high technology upon an industry which has grown and prospered in a low technology environment. Many have compared the training of reactor operators to that of pilote (and I am a pilot) , and there are indeed fundamental differences. Every pilot understands that it takes no great skill to fly an airplane, but that his function is to react properly in the event of the inevitable crisis. All his training is geared to that eventuality, and he is frequently requalified in terms of his ability to cope. Among other things he learns early is the require-ment to believe his instruments, and to study them all and correla te - them before taking action in an emergency. I think It is obvious to everyone now that far greater emphasis must be given to this aspect of operator performance, since TMI reinforced the view that extensive human intervention is inevitable during the course of a reactor accident. Reactor accidents have the advantage over aircraft accidents in that, except for a few sequences, they occur over a long period of time, in which thoughtful human intervention can be an important positive factor. The training and upgrading required to achieve these objectives must be well considered, as well as the restructuring of control rooms needed to support construc-O- role tive human intervention. As part of this whole issue. the proper of automation has been the subject of numerous studies over the years, and that work should be incorporated into any changes that are now recommended. There is one caveat, in that some people are speaking about monumental computers which can provide to an operator on the scene of an accident information about his proper course of action. One should be aware that such a real-time simulator is only as good as the codes which are used to devise it, and the present generation of computer codes designed to simulate the behavior of a reactor under upset conditions leaves a great dealI to be desired, even when mention this only because not asked to do their job in real time. some people tend to put too much faith in computers. C. Probabilistic Risk Assessment 4 The TMI accident was, perhaps perversely, a triumph for proba-bilistic risk assessment. One of the conclusions of the Risk Assessment Review Group was The achievements of WASH-1400 in identifying the relative imoortance of various accident classes have_been inadequately reflected in NRC's policies. For example, WASH-1400 con-cluded that transients, small LOCA, and human errors are imoortant contributors to overall risk, yet their study is not adequately re-
~ ' (). flected in _the priorities of either the research i- or regulatory groups.
John G. Kemeny August 31, 1979 i ()andofcoursethesethreeitems-atransient, followed by a small LOCA, exacerbated by human error - were the major features of the TMI accident. It is difficult to avoid the view that, had the NRC followed the dictates of rational risk assessment methodology in directing its own emphases, the accident at TMI might have been avoided. I believe that it is not too late to go through the top 50 or 100 sequences in the Rasmussen Report, flawed though it may be, to ask for each of these sequences whether we are truly prepared for it, and whether the operator training and instrumentation necessary to nip it in the bud are really in place. This is simply an argument for rationality. I might add something controversial. In interacting with the NRC one cbserves that, in forming its own priorities with regard to reactor safety, it is quite responsive to outside pressure, even to the detriment of rational analysis. It is true that NRC is a public agency, and it is also true that some of the serious problems of reactors have indeed been brought to light through outside inter-vention, but it is also true that the vast majority of outside inter-vention is both ill-informed and destructive, at the very least distracting. It is not entirely capricious to say that the constant outside harping on large LOCA has had more of a lasting ef fect on NRC's energies than is justified by the risk to a reactor posed by large LOCA, to the detriment of energy and budget which might well have been expended upon those accidents which are most likely to occur, such as that at TMI. I believe that the public's right to O' know, which is essential in a democratic society, does not include the right to amateur management of activities which do indeed require ; some expertise. None of us would have an appendix removed by a j surgeon whose every move was guided by a popular consensus. To the l extent that popular perceptions of safety issues call attention to ' them it.is healthy and constructive, and to the extent tha t those i entrusted with the public health and safety are answerable to the l public for their performance, that too is healthy and constructive, l indeed essential, but the next step, public micro-management of l technical safety programs, is not. I have many examples. l 1 Operating Experience l D. (On this there is some documentation, beginning with my letter to Congressman Udall about a year ago, and subsequent exchanges between the NRC and Congressman Udall.) Reactors are complex objects, and despite the best ef forts of analysis and of probabilistic risk assessment, I believe it essen-tial to have an institutionalized procedure for learning by exper-ience. For the aircraft industry there is a quasi-judicial institu-tion, the National Transp.ortation Safety Board, whose function it is to turn aircraft misfortunes into statements of probable cause
-ecommendations for remedial action. This organization is
;n'hpendent of.the FAA, which is the corresponding regulatory organ-12ation. It has seemed to me for some time that reactor operating and O experience tra t there is is no a rich bettersource way toofpractice information aboutmedicine preventive reactor safety, on 1 actors than'to ask them in advance where their little pains lie.
'a fact, there are now some 3,000 Licensee Event Reports each year, 2 r which minor malfunctions are reported to NRC. Some of these i 1
receive a great deal of attention (TMI is an example), but o ther s
i John G. Kemeny August 31, 1979 O fall through the cracks. It is no secret that the event at Davis-Besse could have been regarded as a precursor to the kind of accident which finally occurred at Three Mile Island, and could even have been used to prevent the latter. i NRC has now taken a small step in this direction, by setting up a new office for the analysis of operational data, but it remains to be seen whether that of fice will have any measurable ef fect upon the regulation of the industry. 4 i In conclusion, I feel that we have a long way to go in making the regulation of nuclear reactors more ef fective and more rational, but am pleased that there are now some positive forces at work. My belief is that safety is best achieved, as in aircraf t, by making l a well-trained and competent operator the center of the ac tion, while < mobilizing the remainder of the system to support him in his job. This requires extensive use of probabilistic risk assessment to determine just what job it is the operator is most likely to have to do, reconsideration of the role of automation and of the training of the operators to make it possible for them to function in an emergency, considerable attention to the human engineering of control rooms to facilitate the provision of useful information to the operator, etc. As a separate item, it may well be that the NRC structure itself It was clear at TMI that the issues which i are known O needs in the to be military as C I - Command, control, Communications, reexamined.. 3 l and Information - need considerable attention. NRC is organized to regulate the nuclear industry, and is cast in the role of a regulatory ; agency like ICC and CAB, but is nonetheless sometimes called upon to function like a mission agency, and this is incongruous. I have no l specific conclusion here, and would make the question of reorganiza- ; tion a late issue, after the substantive questions are fully understood, i I could list here a number of proposed recommendations, but it is clear f rom the above what they would be - greater attention to human engineering and the man-machine interface, far more extensive use of rational risk assessment and operating experience in deter-mining' safety priorities, review of operator training and the connected issues of simulation and requalification, and better preparation for rational emergency response, with all the 3C I issues that entails.
-I hope that this overly lengthy letter will be of some use to you, -and would be happy to follow it up with a conversation at any ,
time. Should there be any questions, or need for clarification, I I will-be in Washington at the ACRS meeting on September 6 through 8. l l I Sincerely yours,
/
k Harold W. Lewis
'( } ,
Professor HNL/mm cc: Mr.1R. Fraley, Advisory Committee on Reactor Saf eguards i.
APPENDIX
Title:
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~
Et!ECi5s[ 9 Question IRE class 9 QUESTION O.
"WAS THE OCCURRENCE AT THREE MILE ISLAND , , , A CLASS 9 ACCIDENT?"
ASLB SALEM SPENT FUEL POOL - 7/10/79 "THE OCCURRENCES IN CLASS 9 INVOLVE SEQUENCES OF POSTULATED SUCCES$1VE FAILURES MORE SEVERE THAN THOSE POSTULATED FOR THE DESIGN BASIS FOR PROTECTIVE SYSTEMS AND ENGINEERED SAFETY FEATURES." 36 F.R.11113 - E/8/71 "THE ACCIDENT AT THREE MILE ISLAi!D UNIT 2 INVOLVED A SEQUENCE OF O SUCCESSIVE FAILURES , , , MORE SEVERE THAN THOSE POSTULATED FOR THE DESIGN DASIS OF THE PLANT. THEREFORE, WE CONCLUDE THAT THE ACCIDENT AT THREE MILE ISLAND WAS A CLASS 9 EVENT " MATTSON TO CUNNINGHAM - 8/16/79 0
er .J', l O OTHER FINDINGS OF THE CLASS 9 RESPONSE l THE ACCIDENT AT TMI-2 CAN ONLY BE CONSIDERED A CLASS 9 EVENT FROM THE STANDPOINT OF POSSIBLE, RATHER THAN ACTUAL, RADIOLOGICAL l CONSEQUENCES OF THE GIVEN SEQUENCE OF FAILURES. y l THE ACCIDENT AT TMI-2 CANNOT BE CONSIDERED A CLASS 9 EVENT BY THE l COMMON DEFINITION - AN EVENT WITH SEVERE CONSEQUENCES, DISSENTING OPINIONS EXIST ON THE STAFF RESPONSE TO THE CLASS 9 WJESTION WHICH SUGGEST DIFFERENCES OF INTERPRETATION O LACK OF A CLEAR RELATIONSHIP BETWEEN TMI-2 AND THE MORE COMMONLY HELD CLASS 9 EVENT SUGGEST IMPROVEMENTS TO THE CLASSIFICATION SCHEME RATHER THAN IMMEDIATE APPLICATION IN LICENSING ACTIONS, l l
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APPENDIX IX
Title:
Memo, Mattson to Cunningham Board Question Concernino Class 9 Ac.
& # Rich UNITED STATE 5
' y'o NUCLEAR REGULATORY COMMISSION g
. e WASHINGTON, D. C. 20555 Q
l
- E AUG 161979
, ,..... /
NOTE TO: Guy H. Cunningham, III, Acting Deputy Chief Hearing Counsel FROM: Roger J. Mattson, Director, TMI-2 Lessons Learned Task Force
SUBJECT:
BOAR.0 QUESTION CONCERNING CLASS 9 ACCIDENTS In response to your request of July 27, 1979, I have attached our response to the Salem spent fuel pool Board question or Class 9 accidents. C . Roger J. ttson,'Birector TMI-2 Lessons Learned Task Force
Enclosure:
As Stated l [ O: cc Barry Smith Janice Moore Gary Zech - l ( Walt Pasedag l Robert Tedesco John voglewede John Guibert Fred Anderson l l l (
EVE'STION* 3 proposed Annex to Appendix 0,10 CFR Pa.rt 50, appears to define a Class 9
/
(
.ccident as a sequence of failures which are rore severe than those which the The sequences of failures safety features of the plant are designed to prevent.
at Three Mile Island produ'ad a breach of the containment and a release of Was the occur-radiation which could not be preventad by 'the safety features. Was the risk to the rence at Three Mile Island therefore a Class 9 accident? health and safety and the environment " remote in probabil f ty" or " extremely low" at Three Mile Island, as those tems are used in the Annex?
RESPONSE
The proposed Annex (1) to Appendix 0,10 CFR ' art 50, states "the occurrences in Class 9 involve sequences of postulated successive failures more severe than those Oostulated for the design basis for the protective systems and engineered safety J eatures of the plant." The accident at Three > le Island Unit 2 involved a sequence (3) of successive failures (i.e. small-1reak loss-of-coolant accident and failure of the emergency core cooling syt;em) inore severe than those postulated for the design basis of the plant. Therefore, we conclude that the accident at However, at no time during the THI-2 Three Mile Island was a Class 9 event. accident were the radiological consequences to the public more severe than those calculated for the design basis of the planc. i. The commen definition of Class 9 accidents are those events whose consequences are The accident at Three severe but whose probability of occurrence is extremely low. Mile Island did occur, yet the release of rldioactive material to the offsite i
- Atomic Safety and Licensing Board in the utter of Public Service Electric and Gas Ccepany Salem Huclear Generating Station Unit 1 Spent Fuel Pool Expansion l
Docket 50-272 July 10,1979 l
P pulation was very We small . that the accident at D4I-2 can only be conclude considered a Class 9 event from the standpoint of oossible, rather than actual, radiological consequences of the given sequence of failures.
'In response to the final question of the risk to health and safety and the environment beina " remote in probability" or " extremely low" at Three Mile Island, we conclude tnat the radioactive material released during the THI-2 accident "recresents minimal risks (that is, a very small number) of additional health effects to the offsite population." (4) l
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P' Proposed An ex te Appendix 0,10 CFR 50. Originally published in 1. Although the Federal Recister on December 1,1971 (36 FR 22851) . the Annex was neve- adopted as part of the Code, it is currently carried as Appendix I to Regulatory Guide 4.2 (listed below). 2 Regulatory Guide 4.2, Revision 2 " Preparation of Environmental Reports for Nuclear Power Stations ," NUREG-0099, July 1976. .
- 3. " Investigation in .o the March 28, 1979 Three Mile Island Accident,"
NRC Office of Ins;ection and Enforcement, NUREG-0600, August 1977. I i . 4 " Population Do v and Health Impact of the Accident at the Three Mile Island N : lear Station," Preliminary Estimate Prepared by the
- Ad Hoc Interagency Dose Assessment Group, NUREG-0558, May 1979.
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- March 30, 1979 II Bu21etin No. 79-0!. Fage 3 of 3 A copy of your report (s) sheuld be sent 2'555to the t'nited Division of Reseter Cperatiens inspectien, Washin the app 31:c.ble requirements te report as set and license. i i 7.
All holders of construction permits for power res:ter "ter.s 1 ft:il t es are requested te describe ycur acticns te apun that . . .
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Office, completien Of yen- review and describe e.nyhehle ~ d discrepe.n:ies recting Items 1 thr:. e.. ugh . s:A 5 and, :yyifof necessa.f, your repcrt y0ur plans should be sent to the f r rese'.uti n. ..'..sa...'-..
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V O l UNITED STATES ! NUCLEAR REGULATORY COMMISSION I 0FFICE OF INSPECTION AND ENFORCEMEST WASHINGTON, D.C. 20555 l l ! April 14, 1979 l IE Bulletin No. 79-07 l SEISMIC STRESS ANA1.YSIS OF SAFETY-RELATED PIPING Description of Circu= stances: In the course of evaluation of certain piping designs, significant discrepancies were observed between the original piping analysis computer code used to analyze earthquake loads and a currently acceptable cocputer code developed for this purpose. This proble= resulted in the Nuclear Regulatory Co==ission order to shutdown five power reactors whose design had involved the use of the suspect computer codes (IE Infor=ation Notice No. 79-06) . The difference in predicted piping stresses between the two ce=puter codes is attributable to the f act that the piping analysis code used for a nu=ber of piping systems uses an algebraic su==ation of the loads O- predicted separately by the computer code for both the horizontal components and for the vertical component of seismic events. This is an incorrect treat =ent of such loads and was not recognized as such at the time the original analyses were performed. Such codirectional loads should not be algebraically added (with predicted loads in the negative direction offsetting predicted loads in the positive direction) unless certain more complex ti=e-history analyses are performed. Rather, to properly account for the effects of earthquakes on systems important to safety, as required by " Design Bases for Protection Against Natural Pheno =ena," General Design Criterion 2 of Appendix A to 10 CFR Part 50, such loads should be cocbined absolutely or, as is the case in the never codes, using techniques such as the square root of the su= of the squares. These combinations of loads conform to current industry practice. The inappropriate analytical treatment of load combinations discussed above beco=es significant for piping runs in which the horizontal seismic excitation can have both horizontal and vertical components of response on piping syste=s, and the vertical seismic excitation also has both horizontal and vertical components of response. It is in these runs that the predicted earthquake leads may dif f er significantly. ,
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IE Bulletin No. 79-07 April 14, 1979 Page 2 of 3 C Although the greatest dif ferences in predicted loads would tend to be li=1ted to locali:ed stresses in pipe supports and restraints or in weld attachments to pipes, there could be a substantial nu=ber of areas of high stress in piping, as well as a number of areas in which there is potential for da= age to adjacent restraints or supports. Any of these situations could have significant adverse effects on the ability of the piping syste= to withstand seis=ic events. The h*RC staff has not yet deter =ined that all of the piping syste=s i=portant to safety that were designed using a piping analysis co=puter code which contains the algebraic suc=ation error, have been identified. Certain infor=ation is needed in order to =ake this deter =ination. Action To Be Taken By All Licensees and Per=it Holders: 1 For all power reactor facilities with an operating license or a 1 l construction per=it: (1) Identify which, if any, of the =ethods specified below were e= ployed or were used in co=puter codes for the seis=ic analysis of safety related piping in your plant and provide a list of safety syste=s (or portions thereof) affected: Ecsponse Spectru Model Analysis:
- a. Algebraic (considering signs) su==ation of the
- codirectional spatial co=ponents (i.e., algebraic st.=:=ation of the maximu= values of the codirectional responses caused by each of the co=ponents of earthquake =otion at a particular point in the
=athe:.atical =odel).
- b. Algebraic (considering signs) su==ation of the J codirectional inter =edel responses (i.e., for the [
nu=ber of = odes considered, the =aximu= values of response for each =ede su==ed al t ebraically). Time History Analysis:
- a. A.lgebraic so-* tion o f the codirec tional =axi=u=
responses or the ti=e dependent responses due to each of the ce=ponents of earthquake =otion acting s1=ultaneously when the earthquake directional
=otions are not statistically independent.
l April 14, 1979 IE Bulletin No. 79-07 Page 3 Of 3 (2) Provide co=plete cocputer progra= listings for the dynazic ' response analysis portions for the codes which employed the techniques identified in Item I above. (3) Verify that all piping computer programs were checked against either piping bench = ark problems or compared to other piping computer programs. You are requested to i identify the benchmark proble=s and/or the co=puter l programs that were used for such verifications or describe in detail how it was deter =ined that these programs yielded appropriate results (i.e., gave results which corresponded to the correct perfor=ance of their Intended methodology). (4) If any of the methods listed in ite: I are identified, submit a plan of action and an esti=ated schedule for the re-evaluation of the safety related piping, supports, and . equf;=ent affected by these analysis techniques. Also provide an esti= ate of the degree to which the capability in the of the plant to safely withstand a seis=ic event interim is i=pacted. , [) The responses for Ice =s 1, 2 and 3 above, should include all subsequent piping system additions and =cdifications. Any re-evaluation required, in conf or=ance with Item 4, should incorperate the "as built" conditions. Licensees of all operating pcwer reactor facilities sheuld subeit the infor:ation identified in Ice =s 1 through 4, above, within 10 days ef the date of this letter. Holders of construction per=its for power reactor facilities should sub=it this inf or=ation within 45 days of the date of this letter. ' Reports should be subcitted to the Director of the apprcpriate NRC
. Regional Of fice and a copy should be forwarded to the NRC Of fice of Inspection and Enforcement, Division of Reactor Operations Inspectien, k'ashington, D.C. , 20555.
i Approval Approved by GAO, B180225 (R0072); clearance expires 7-31-80. was given under a blanket clearance specifically fer identified generie problems. V !
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(, , UNITED STATES NUCLEAR REGULATORY COMMISSION 0FFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, O.C. 20555 March 8, 1979
- IE Bulletin No. 79-02 PIPE SUPPORT BASE PLATE DESIGNS USING CONCRETE EXPANSION ANCHOR BOLTS
' ~ Description of Circumstances: While performing inservice inspections during a March-April 1978 refueling outage at Millstone Unit 1, structural failures of piping supports for safety equipment were observed by the licensee. Subsequent licensee inspections of undamaged supports showed a large percentage of the concrete anchor bolts were not tightened properly. Deficiency reports, in accordance with 10 CFR 50.55(e), filed by Long Island Lighting Company on Shoreham Unit 1, indicate that design of base plates using rigid plate assumptions has resulted in underestima-tion of loads on some anchor bolts. Initial investigation indicated that nearly fifty percent of the base plates could not be assumed to deheve es r4944 P'etes- ta ede4tioa. '4ceasee 4 aspect 4oa or eachor bott
'O installations at Shoreham has shown over fifty percent of the bolt installations.to be deficient.
Vendor Inspection Audits by NRC at Architect Engineering finns have shewn a wide range of design practices and installation procedures which have been employed for the use of concrete expansion anchors. The current trends in the industry are toward more rigorous controls and verification of the installation of the bolts. The data available on dynamic testing of the concrete expansion anchors J. show fatigue failures can occur at loads substantially below the bolt
' static capacities due to material imperfections or notch type stress
.. risers. The data also show low cycle dynamic failures at loads below T, '
the bolt static capacities due to joint slippage. Action to be Taken by Licensees and Permit Holders: For pipe support base plates that use concrete expansion anchor bolts in Seismic Category I systems as defined by Regulatory Guide 1.29, " Seismic Oesign Classification" Revision 1, dated August 1973 or as defined in the applicable FSAR.
.:~ Verify that pipe support base plate flexibility was accounted for 1.
" in the calculation of anchor bolt loads. In lieu of supporting analysis justifying the assumption of rigidity, the base plates
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'G IE Bulletin No. 79-02 March 8, 1979 should be considered flexible if the unstiffened distance between the member welded to the plate and the edge of the base plate is greater than twice the thickness of the plate. If the base plate is determined to be flexible, then recalculate the bolt loads using an appropriate analysis which will account for the effects of shear. - tension interaction, minimum edge distance and proper bolt
, spacing. This is to be done prior to testing of anchor bolts. These calculated bolt loads are referred to hereafter as the bolt design
- leads.
. 2. Verify that the concrete expansion anchor bolts have the following minimum factor of safety between the bolt design load and the bolt ultimate capacity determined from static load tests (e.g. anchor
. bolt manufacturer's) which simulate the actual conditions of installation (i.e., type of concrete and its strength properties):
- a. Four - For wedge and sleeve type anchor bolts,
. b. Five - For shell type anchor bolts. 'A 3. Describe the design requirements if acplicable for anchor bolts to U withstand cyclic loads (e.g. seismic loads and high cycle operating loads). .
4 Verify from existing QC documentation that design requirements have been met for each ancho, bolt in the folicwing areas: (a) Cyclic loads have been considered (e.g. anchor bolt preload is equal to or greater than bolt design load). In the case of the shell type, assure that it is not in contact with the back of the support plate prior to prelcad testing. I (b) Specified design si;:e and type is correctly installed (e.g. procer embedment depth). [ If sufficient documentation does not exist, then initiate a testing 4.' program that will assure that minimum desi met with respect to sub-items (a) and (b)above. gn requirements A sampling have been technique is acceptable. One acceptable technique is to rand 0mly select and test one anchor bolt in each base plate (i.e. some supports may have more than one base plate). The test should pr: vide verification of sub-items i (a) and 'b) above. If the test fails, all other bolts on that base l _ plate should be similarly tested. In any event, the test program should i
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assure thac each Seismic Category 1 system will perform its intended l function. O V 2 of 3 l i 1 l
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O te sellet4n No. 79-02 xer=h 8, 1979
- 5. All holders of operating licenses for power reactor facilities are requested to complete items 1 through a within 120 days of date of issuance of this Sulletin. A reactor shutdown is not required to be initiated solely for purposes of this inspection above. Maintain documentation of any samoling inspection of anchor bolts required by item 4 on site and available for NRC inspection. Report in writing within 120 days of date of Sulletin issuance, to the Director of the appropriate NRC Regional Office, completion of your verifica-tien and describe any discrepancies in meeting items 1 through 4
. and, if necessary, your plans and schedule for resolution. For planned action, a final report is to be submitted upon completion
. of your action. A copy of your report (s) should be sent to the United States Nuclear Regulatory Ccmmission, Office of Inspection and Enforcement, Division of Reactor Ocerations Inspection, !
Washington, D.C. 20555. These reporting requirements do not preclude nor substitute for the applicable requirements to report i as set forth in the regulations and license. !
- 6. All holders of construction permits for power reactor facilities are recuested to complete items 1 though 4 for installed pipe support base plates witn concrete ancnor bolts witnin 120 days of date of issuance of tnis Bulletin. For pipe support base plates f
which have not yet been installed, document your actions to assure l that items 1 though 4 will be satisfied. Maintain documentation of these actions on site and available for NRC inspection. Report in i writing within 120 days of date of Bulletin issuance, to the Director l l of the appropriate NRC Regicnal Office, comoletion of your review and describe any discrepancies in meeting items 1 though 4 and, if necessary, your plans and schedule for resolution. A copy of your recort should be sent to the United States Nuclear Regulatory Comission, Office of Inspection and Enforcement, Division of Reactor Construction Inspection, Washingten, D.C. 20555.
. Apcroved by GAO 3180225 (R0072); clearance exoires 7/31/80. Aporoval was given under a blanket clearance specifically for identified generic problems.
Enclosure:
l List of IE Bulletins i Issued in Last I Twelve Months i i 1 Q 3 of 3 l l
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- a
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I . 1 - [p nsQ*o, UNITED STATES , !y)3 7[h NUCLEAR REGULATORY COMMISSION f c- '
-i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o, [ W ASHINGTON, D. C. 20555 l
' *...= July 2,1979 i I
)
l o 1 . M. Bender, Chaiman, Subcornittee en the Combination of Dynamic Loads , P. Shewmon, Chairman, Subcommittee on Metal Components '
SUBJECT:
PLAllT SHUT 00'liis 00E TO PIPIf;3 SYSTE.'l It!ADECUACIES FRCri SEIS!!IC EVEilTS 1 i The following attached letter from W. Kennedy, Vice President of Engineering at Stone and Webster, and a report by R. Cloud on " Seismic Capability of iuclear Piping" should provide some interesting reading in light of the current problem on piping seismic analysis. In the Kennedy letter, he provides a summary description of Stone and Webster's
- n reanalysis activities and procedures and a brief report of their findings with U respect to the evolution of the calculational techniques, i
The Cloud repo: t which was submitted as background infomation on the Beaver Valley, ' Jnit ' docket provides a general review of the methods applied to seismic analysis of safety class piping. He takes the position that piping systens designad ear'ier in the evaluation of dynamic analysis have an inherent seismic j resistance. ,, v, Jyu> . E. Igne I Staff Engineer Attachments: As stated cc: ACRS tiembers 1 I ACRS Technical Staff H. Corten A. Pense Z. Zudans T. Pickel R. Scavuz:o W. Gall 2 O
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STONE O WEBSTER ENGINEERING CORPORATION t' 245 SvMMcA S t a c t T. B o sT O N. M ASS ACHUSCTTs mooncss Ass commtsnosocsec To a o. een asas, sosron. wass. canor er s.s~
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. . ,.me Mr. Harold Denton, Director .
Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission i Washing to n , D. C. 20555 ,
Dear Sir:
Stone & Webster Engineering Corporation (S&W) and the utility, companies involved have been working diligently on the reanalysis required by the NRC Orders to Show cause of March 13, 1979. We have progressed to the stage where substantial amounts of data are flowing to the NRC and the process of confirming the adequacy of each reanalysis is well under way. The Maine Yankee reanalysis required by the Order to Show
, cause has been'ccmpleted. Work en the other plants is progressing. We have endeavored to do everything necessary to assure full compliance with the Orders, including develep-ment and use of special procedures. Concurrently, we have reviewed the evolution and history of our calculational techniques to help establish a clear understanding of them by all concerned.
Given the urgency of the situation, we believe that it would be helpful to both the ACRS and the NRC to have a summary description of cur reanalysis activities and procedures to the evolution and a brief report on cur findings with respect of our calculational techniques, particularly with respect to satisfaction of the requirements of General Design Criterion 2. REANALYSIS ACTIVITIIS AND PROCEOURES The reanalysis effort has centered around the fact that certain piping (SHOCK II) systems were designed by S&W using a computer programwi
. force combinatien in one of the steps. The NRC has contended that the use of algebraic summation in this case was incorrect (s~g / (we do not agree) and has therefore required that all systems designed with the SHCCK II subreutine be reanalyzed using currently acceptable programs.
l i
l I" In view of the seriousness of the impact of the shutdowns and j the extensive effort required to ameliorate the situation, immediate steps were,taken upon receipt of the Orders to establish the scope of the reanalysis effort, develop work plans, organize teams for each project, develop procedures to assure a high quality and timely response, and carry out the work expeditiously. Over 400 people, operating on a two-shift basis, were assembled to gather documentation, perform the analyses, confirm the l adequacy of the analyses, and verify current as well as formerly used computer programs. Project teams and review groups were established; administrative and, technical procedures were developed; orientations were conducted; and schedules of activities were prepared and tracked. . 4 The basic elements of the rernalysis effort are highlighted ! below: !
- 1. Establishment of Scoce of Effort Efforts were made to establish the scope of the reanalysis effort in conjunction with the NRC and the utilities '
shortly after issuance of the Orders. .
- 2. Development of Work Plans l A work plan was developed for each of the five units.
Each plan consisted of identification of af fected systems, determinaticn of reanalysis sequence logic, determination of project manpcwer recuirements and potential sources, and establishment of service priorities including computer use.
- 3. Retrieval of Documentation and Establishment of Desica a
Basis for Reanalysis 1 All available documentation relating to pipe stress and , pipe supports was retrieved frem archives and/or field records. This documentation included ccmputer and hand calculations, piping design drawings, flow diagrams, piping and support sketches, and project job books. Extensive field verificatien was performed to assure that the reanalysis effort would reflect the "as-built" config-uration of the system being evaluated.
- 4. Computer Procram Verification NRC required extensive verification of the calculational techniques to be used in the reanalysis. This consisted primarily.of applying the S&W programs PSTRESS/ SHOCK III
# and MUPIPE to standard benchmark problems defined by NRC.
Another program (E-?!?I) was verified by NRC itself. S&W l l
- l. 3.
supplied typical problems frcm each of the affected (l plants to NRC for confirmation using E-PIPE. The overall 'u/ verification effdrt was based on a stringent interpretation of Standard Review Plan 3.9.1 and the application of verification techniques and procedures beyond those generally required of the nuclear industry by the NRC.
- 5. Reanalysis of Pice Stresses Following verification of the "as-built" configuration, the computer model for the system was receded, if required, and a computer run was made using a verified computer program. If the pipe-stress results were within allowable limits, the evaluation sequence proceeded to review of supports, no::les, and penetrations (see 6.). If the results indicated a possible stress condition in excess of allowable, the system model was scrutinized in greater detail, revised utilizing current,more sophisticated techniques still consistent with the original licensing bases, and rerun. If the results still indicated a possible stress condition in excess of allowable, the system was analyzed using current methodology, such as amplified response spectra with soil-structure interaction, consistent with the original licensing bases and with current requirements.
If this latter rerun does not resolve the problem, additional g3 supports or snubbers are considered. %) 6. Reevaluation of Pice succorts, Nozzles, and Penetrations Following ccmpletion of a pipe-stress run where the recalculated loads were within allowable li,mits, the support and end-reaction loadings were tabulated and compared with the original loadings. If the loadings were equal to or less than the design limits, the supports or no::les or penetrations involved were considered j acceptable. If the loadings were abcve the original design limits, new calculations or supplements to'the original calculations were performed, using techniques licensed for the units. The item was then determined to be either acceptable or in need.cf modification or replacement. If no:cle reactions exceeded those initially approved, vendors were contacted for approval of the new 1 loadings. The vendor either approved the new reactions or requested reductions. In the latter case, other system modifications were identified and made. DEVELOPMENT OF SEISMIC PIPE-STEISS CALCULATIONAL TECHMICCES The use of alcebraic summation for ccmbination of intramodal seismic forces in cice-stress calculations has been characterized in various ways by government officials and the media. Many of
~
f) '- the characterizations have been incorrece or misleading. From the outset, as stated in public releases, S&W has agreed i that the SHOCK II subroutine may not always yield results as
4. j, uniformly conservative as procedures used in the nuclear industry today. We believe, however, that the use of algebraic j (-s) summation was an acceptabic, ccamon practice at the time it was used by us, and that the piping systems designed with the use of this method will perform their intended saf ety functions. A review of the history of calculational technique development by the nuclear industry confirms that continuous ef fort was made to better approximate seismic effects since the initial modal ccmbination " rule" was suggested in 1943. This " rule" established an upper bound to maximum seismic response by use of absolute value summation. Because this approach was considered to be unrealistically conservative, alternative techniques have evolved. These include the use of algebraic summation, square root of the sum of the squares, and various combinations of all three (absolute, algebraic, and SRSS). Through the years we have continually mcdified our calculational techniques to lead with advances in both the state of the art In 1972 and 1973, we started and regulatory requirements. using new ccmputer programs which introduced a number of changes, including substitution of a modified square-root-of-the-sum-of-the-squares-(SRSS) precedure for algebraic summation to ec=bine intramodal forces. In late 1974, NRC formally identified specific acceptable calculational techniques in Regulatory Guide (RG) 1.92. Our modified SRSS technique, then already in use, met
~h the requirements of RG 1.92.
(V Then, as now, we recognized the evolutionary nature of ecmputer programs and th'e nature of their use in the overall design process. Therefore, we began to employ the revised techniques on new projects while we continued to use the previous one en certain projects. In our view, the new techniques provided a means for obtaining more uniformly conservative results. That is why we adopted them. This does net imply, of course, that the previous technique was unacceptable. It merely indicates that, consistent with sound professional engineering judgment, we were employing the most advanced calculational tools. The computer programs developed by us and by others are 1 calculational ecols used to predict forces and resultant i stresses that may occur during an earthquake. Such predictions l are an integral, but not necessarily the most important, part of the total process of establishing design adequacy of piping systems. The present seismic analysis programs utilize more i refined calculational techniques, but they still provide only ; one stress componen among several that must be considered, viz., thermal, pressure, deadweight. Furthermcre, system modeling and the assumed characteristics of the earthquake used for design has greater impact en theWith solution than the respect to the 73 method of intramedal force cc=bination. (/ trend toward more uniform conservatism, it should be nc:ed that SRSS, which is deemed acceptable and adequately ccnservative, 1 l 1
3.
/ '
1 is based on statistical concepts and may, in seme circumstances,
~
be less conservative than earlier approaches to seismic analysis. ! ( ' I i (~ The evolutionary nature of calculational techniques for seismic analysis was similar to the development of other improvements I in analytical methodologies, e.g., the change frem simple hand calculations to complex computer codes and frem static analysis to dynamic analysis. This evolution was acccmpanied by a parallel development of seismic criteria by the AIC and the NRC sta: ting in the mid-1960s and continuing well into the 1970s. There was no specific ASMI code requirement for seismic design during that period. In response to the recent NRC I&E Bulletin 79-07, additional data have been submitted to the NRC which provide further historical information on the various techniques used in seismi: analysis of piping systems. It appears that a number of other responsible organizations have usod algebraic summation of intramodal forces in a seismic pipe-stress j I analv. sis for a number of nuclear power plants - c.e:hao.s 15 to 1 20 plants designed by firms other than S&W. CONCLUSIONS The efforts we have expended to date have provided results While from which some ceneral conclusions mav now be drawn. it is. recognized that the algebraic suhmation of intramodal
~
(~%
'/ forces may not always yield results as uniformly conservative l as the present-day techniques, it was a widely used technique l I
acceptable to th'e profession during the evolution of the art of seismic analysis. It is unlikely that major physical changes will be required in the piping systems .<ni which it i was used. It appears more likely that any changes that may l be required to satisfy current requirements will be confined , l to pipe supports. It is a well known and demonstrable fact that piping is not particularly sensitive to seismic events. Such simple, commonly-used design techniques as fastening equipment andto restricting the movements of'large masses Further, are the sufficient various assure a high level of integrity. analytical approaches generally show large dif ferences at low stress levels and small differences at high stress levels. In summary, virtually any reascnable design approach or ccmputer program will ensure a high degree of earthquake protection. During the late 1960s and early 1970s the use of dynamic analysis superseded static analysis and the utilization of ccmputers repiceed manual techniques. During the same
. period, respcnsible organi:ations used algecraic summation (sN) of intramcdal forces in calculational techniques empicyed to determine values for use in seismic design of piping systems.
- 8 l . ,
Most important is the fact that these changes did not replace design review by knowledgeable engineers. This resulted in designs no less rugged than past designs, which have withstood
- actual seismic events of great severity. It is noteworthy that a number of knowledgeable individuals have observed that the designs resulting from techniques employing algebraic summation do not appear any less substantial than designs resulting from other techniques.
Our present reanalysis effort has demonstrated that she , differences between the earlier technicues and cur, rent practice, resulting from gradual improvement typical of high technolcgy developments, do'not yield substantially different final ' designs. - 1 j We trust this summary of our reanalysis activity and conclusions are helpful to you.
' Yours very truly,
. /. #4
- J. L. Kh .nel
ice Presiden 6 O g 4 4 e e 4 O 9 5 e e O t
. l i
s e i s 9 8 9 h I 1
. _ _ _ _ _ _ _ _ _ _ _ _ ___ l
3EAVER VALLIY FC'a?.,R STATION, " NIT l A??INDIX K SIISM:0 CAPA3:LI!T CT NUOLIAA ?!?!NG b7 Robert L. Cicud Robert L. Cloud Associates Inc. Manic Park, Calif. May 1979 1 i i i l ts l l I l l
. 1 3IAVIR VALLIY PC%TR S*AT:CN, CN:7 L APPINDIX H
*A3LI CT CON INTS Seetien h h
- 1. INTRCDCCTION. . . ............... . . . . . . . . . . . X-l
- 2. SI:SMIC ANALYS:S CT NUCLIAR PLANTS. . . . . . . . . . . . . . . . . . M-2 1
- 3. PIPING ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . H-3 1
6 ANS: 331.1 CCDI . . . . ..... . . . . . . . . . . . . . . . . . . H-6
- 3. 3 31.1 AN') LATIR CODIS . . . . . . . . .. . . . . . . . . . . . . . . H-7
- 6. SI:StCC PIRTCRMANCE CT PCb'IR ?:7ING . . . . . . . . . . . . . . . . . M-9 6.1 Long 3each Steam Statica . . . . . . . . .. . . . . . . . . . . . . H-3 !
l 6.2 Re n Ccunty Steam Station. . . . . . . . . . . . . . . . . . . . . . H-9 6.3 Alaska Iarthquake of 1966 . . . . . . . . . . . . . . . . . . . . . H-LC 6.6 San Ternando, Califcznia, 197L . . . . . . . . . . . . . . . . . . . H-ll 6.5 Managua, Nicaragua, 1972 . . . . . . . . . . . . . . . . . . . . . . H-12 s 7. 3 ASIS TCR SIISM*C CAPA3:LITY CT PC%'IR FI? NG. . . . . . . . . . . . . K-13
- 8. CONCLUS:CNS ANL IMPLICA!!CNS TCR MODERN NUC'IAR PLANTS. . . . . . . . H-13
- 9. RITIRINCIS. . . . . . . .... . . . . . . . . . . . . . . . . . . . X-13 LTST CT TA3 TIS
*ia %,. .I n. _-3ble +
X-1 Seismic Analysis cf Nuclear Plants M-2 Seismic Analysis of Piping Systems , X-3 Damaged Iqui;=ent at the Ivaluf Pcver Plan: X-i ____._____.___.______________m_,_.
3EAV!2 VALLEY ?C7IR STATICN, 'JN!! 1 C
- l. INT 2CDCCT:CN 3efc a the development of the ANSI 331.7 Code for Nuclear Piping in the late 1960's and subsequent inclusion of piping under the ;; visions of ASMI Code, Section III, all nuclear safety class pi;ing was designed te meet the requirements of the ANSI (formerly USAS) 331.1 Code for Fever Piping. As a result, many of the operating nuclear pcVer plants in the United States today were designed and built to meet the p cvisicas of the 331.1 code.
A general review of the methods applied to the seismic analysis of 331.1 safety class piping is given including :sference to the historical evolutica of these methods. The 331.1 code itself is discussed and it is de= castrated that, contrary to the belief of many, this code :ssts en an advanced technical base; sufficiently advanced in fact that very few changes had to be made, other than notation, :: upgrade it to the 331.7 nuclear cede and then to ASMI Section :::. The older piping code, unlike that fe vessels, contained all the main features of current codes. Fever plant, chemical plan:, and refine:y piping designed to 331.1 is not cu: dated and behaves very well in earthquakes. The available da a en performance of piping in seismic events are reviewed, and it is shcwn that engineered piping systems have perfor:ed extremely well in ;cus: plants that have experienced substantial earthquake-induced g: und metien. This cu: standing peric mance has been exhibited even at plants in which the piping syste:s were designed for seismic icadings far less severe than current cri:eria veuld associate vi:h the 3:cund =c:icns actually ,( experienced. It appears that piping systs:s engineered fe the p: essure and
\ temperature conditiens typical of power plants are inherently resistant to the effec:s of seismic-induced =ctions of their su;;cets, whether :: not such effects are specifically addressed in the design ;;ccess.
Fever plant piping aivays has been designed to de=anding standards. 'Jith the int:cducti:n of nuclea pcVer, these standards have been =aintained and streng:hened to sc=e deg es. , Thus, it is reasonable to expect that piping systems in nuclear pcVer plants that may experience earthquake :::icas vill perform as well as have the piping systa=s in non-nuclea: pcVer plan:s. I4:17 in the int:cduction of nuclear power, a =aje devele; ent effer: began in methods of dynamic analysis of st:uctural :esponse to earthquake 3 :und me:1cn. This deveic;aent was focused, almost exclusively, en syste=s (cf piping and suppcrting s::uctures) eenserva-ivelv assu=ed to res;cnd in a linear elastic mode. On the other hand, criteria of allevable piping stress, vbich had been developed before seismic loadings were of major interest, remained relatively unchanged by the substan:tal evolution of =etheds of dynamic analysis. Stress c:iteria which had been rationally and conservatively developed fe; relatively vell-defined pressure-te=;erature-:1:e condi:icns were a;;1ied with mino: difi:a:1:ns to the less well-defined and ve:y di!!, en- earthquake condi:icas.
- n consequence, the inheren: seismic resistance of nuclear pcVer plan: ;i;ing systems designed earlier in the evolution of dyna =ic analysis may go unrec:gni:ed. Since systems fe: which earthquake effects vers not even considered in design cisa:17 have substantial resistance : such effects, it s
s-K-l
1 4 3IAVIR VALLEY PCVER STAT:CN, UNIT 1 J is unvarranted to caly judge the seismic safety of a particular piping syste= I by the particular 3:cund =ction specified and analytical =sthods used to predict response at the time of its design. Te illustrate, there was a period when algebraic ce=bination of intra =cdal seis=ic ef f ects (cf earthquake 3 cund
=ctions in differing directions) was ce==en practice throughout the industry.
This technique could either overesti= ate er underesti= ate earthquake leads en .
'a piping system. In the li=it, it might indicate essantially :ere earthquake leading; i.e., equivalent to c=itting earthquake f ce the design conditions.
Obviously this is in no way the same as designing with ze:c resistance to , earthquake. On the contrary, the syste=, designed to conservative criteria I for pressure-ta=perature-ti=e conditicas, veuld certainly be resistant to l substantial earthquake g:cund =ction as =ay be seen by ecmparison with the power plants discussed herein.
- 2. SI:SMIC ANALYSIS CT NUCLIAR PLANTS ,
1 Table M-1 shows a rough chrenelegical development ei se=e of the =41n features of seismic design and analysis mesheds for nuclear plan'es. The first plants were designed with static =ethods using lateral force coefficients as static icads in the =anner of various building cedes. These plants were, in the main, built in regions of lov seismicity. Dyna =ic considerations were int:cduced at about the ti=e plants were built in , regions of higher seis=icity. In reco gnition of the amplified response j
. possible when shaking =ctions have f:equencies at c near the natural
( frequencies of buildings and equip =ent, design 3:cund response spectra vere int cduced ic: design. Several papers that describe the derivation and I applicatien of response spectra methods are centained in the section en Seis=ic Analysis of Reference 1. This :sference was compiled to provide technical backg cund for the advances and changes of various codes !c: design and constructica of pressure vessels and piping, es;ecially for nuclear applications. As such, the key papers that influenced the develcp=ent of nuclear seis=ic technology by seismic specialists such as Newmark, Mali, j Cicugh, Cc:nell, and others a:e reprinted ccaveniently in one place. l
. 1 To obtain the seismic response of piping systa=s, it is necessary te study the passage of g:cund =etien th:cugh the scil, buildings, and equip =ent, all of which cause modifications of the =ction before it reaches the piping.
Originally, design response spectra were applied to piping in the simplest way considering the fi:st mede of each span and taking the :sspense directly f:cm the 3:cund spectrum. This app:cxi=ation was an improvement over purely static methods, but is quite si=plified c:=pa:ed to later matheds. ; 1 Subsequently in the 1960's, the effect of building =ction en piping systa=s was incorporated into the design 7:ccess en an industry-vide basis although I the concept had been deveicped =uch ea:11er.828 Ccnceptually, this is done by analy:ing the building fer.the effect of 3:cund =ccion and developing new spectra at the ficers and valls of the building where piping is suppc ted. In practice this was done at first using recc ds of actual earthquakes, Taft, Il i Cent:c, etc, normali:ed :: the design acceleration level chosen fe the site. (~ The accelerations were applied to lumped mass- building =edels in- a time-history fashien. At first, very few = asses veuld be used :: represent the H-I
3IAVIR VALLEY FCifIR STA!!CN, CN:7 1 building, say less than 10. Also app:: xi= ate matheds were devised te obtain the effect of building a=plification on the design spectra'88 direc Ly withou: a time-histery analysis of the building. Design fleer spectra vere developed by these =eans and used for several p'. ant designs.
- n the 1970's, several =ajor chaages in matheds cf nuclear plant seismic analysis were =ade. The key changes vere a standardi:ation of design 3: und spectra, a require =en: fe three-d'rectional analysis, and use of increased da ping values. The not effect was a : ore rational approach to seismic analysis, but in any given case, cceputed seismic s : esses tended to be ce=paracle to those obtained by the ==:e app;cximate methods since the higher da= ping ce=pensated for the additional i= posed =ction. In any event, this appendix is addressed more to 331.1 plants and subsequent developments vill not be discussed further.
- 3. p:P:NG ANALYSIS Seismic analysis of piping sys: ens in nuclear plants has also unde: gene an evolution, outlined in Table M-2, censistent with the g veh and develop =en:
of seis=ic =etheds for the plann as a whcie. Ia:1y =e:heds were based ce static analysis using a cons:an: lateral force coefficient that was a specified fraction of the total mass of that part of the piping under censideration. As =entioned prevleusly, when spectra vere ft s: used, , spectral accelerations censistent with span f:equency were applied to the piping systa=. These veuld be applied ner=al to the plane of the pipe, i.e., in the vers: ' direction' and ce=bined with a vertical ce=penent. Later, =edal response spectra analysis was applied to safety class pi;ing as a=plified fice s'pectra beca=a available and vere specified.8 The applica:ica of this app;cach varied between different c:gani:ations and with ! the times. Although the funda entai steps and basic =4the=atics were generally ec==en to all, certain chcices had to be made in ce=bining responses j for each directica and each =cde. These ce=binations are in a sense arbitrary since the =cdal response spect:1. analysis method relinquishes ti=e as a paramete and relationships with respect to time, including phasing, are icst. Sc=e analyses have been done by eva!.uating each of three directions separa:ely and ce=bining contributions f := ea::h directi:n. In many cases the heri: ental direction that causes the highe.s t stress is ::=bined with the vertical and this 2-0 plana response bece=es the basis fe evaluation. l The directional ce=binations have also been made in other ways. Since the l various :sspense quantities are signed, algebraic su==ation of :espenses f; = l each direction within each =ede has been done. Analyses have been ec=pleted using c hor Options, SRSS and absciute su=. The la::e is probably everly con's erva tiv e . After ce=bina:icns have been =ade se tha: the respense f:: each =cde is c =plete, the su: ef all the medal.:ss; cases =ust be ob:41ned. Analyses have been ec=pleted using several differen: =ethods of ce=bining these res;enses. The =stheds include a st:aightfe:vard " square-::c:-ci-the-su=-ef-the-squares"
\ c SRSS, the absolute value of the single =aximu =edal :sspense ;ius the SRS3 M-3
3EAVER VALLEY PC'JIR STA!:CN, CH:: 1 i C of the re=aining =cdes, and other ce=binations including absciute values of response of closely spaced =cdes plus the SRSS ci the re=aining =cdes. The general i=;etus for the advance of seismic analysis and evaluation =e:heds ca e fic= a videly felt need both within the industry and regulatory agencies to better understand seis=ic behavier of piping and equi;=ent. As :ssults beca=e available f::= develep=ent activities, they would be used for specific plant analysis. The i=petus for doing so would as often ce=e f:ce the utility or the manufacturer as f c= the :egulatory agency. It was a peried of rapid technical 3:cvth in which all groups concerned with the issue participated. 4 ANS: 331.1 CODI Prior to the appearance of the ANS: 331.7 and the ASMI Section II: Cede, all safety class piping was evaluated accc: ding to the ANS: (for=erly USAS) 331.1 Code for Power Piping. Tc the present discussion, the 1955 and 1967 versions of this code 4:e the issues of interest. There was little c; no basic change in 331.1 between the 1955 and 1967 ve:sicas. The 1955 version hcvever was a !
=aje departure fre= the previous issue of 1942 and su;ple=ents. In fact it I was in the 1955 version of 331.1 in which the basic rules and technical philosophy were established for the design of power piping that are essentially in existence today.
The advanced features and underlying technical sophistication of the 331.1 Code have gene relatively unnoticed in this era of rapid technical change and i innovation. The 331.1 app cach, first established in 1953, contained previsiens fe: limiting the ther=al strain range; recogni:ed the self-li=iting nature of thermal. stress; contained design rules for icv cycle fatigua; incer; orated the =4ximum shear stress theory; and contained other i=p :ve=ents. The ASMI 3cile and Pressure Vessel Code contained none of these features a: that ti=e. In fact it was not until the ASMI I:I Nuclear Vessel Code came out nine years later (1964) that these technical i=p;cve=ents ve:s applied to ;; essure vessels. The fundamental intent of piping design lies in developing a system that has sufficient flexibility but is sufficiently well cent:clied as discussed fur:her belev. The conce;: cf cen::clied flexibility is the key to successful piping design. The Code recogni:ss this with an entire see ica devoted to piping flexibility. The app;cach can be seen f::= the felieving, quoted frem paragraph 119.5 of the Code:
" Fever piping syste=s shall be designed to have sufficient flexibility to preven pipe seve=ents fre= causing failure f:c everstress of the pipe
=aterial c anchors, leakage a joints, c detri= ental diste::ica of connec:ed equi;=ent resulting f == excessive thrusts and n =ents.
Tiexibility shall be p;cvided by changes ci direc:icn in the ;iping th: ugh the use of bends, loops, c: cifsets; c ;;cvisions shall be made to absorb thermal =cvements by utili:ing expansion, svivel 0; ball joints,
- cc :ugated pipe."
Explici: guidance is given :: ebcain balanced syste=s and := avoid ;;:blems Of ( strain concen::a:ica caused by ----" d'- = flexibility. In this c nnectica M-4
k 3IAVIR VALLEY 70L'IR STATION, UNIT ! the concept of elastic fellev-up is discussed. Design configurations vulnerable to st:ain concentrati:n are explained and cautioned against. The phone:ena of low cycle fatigue are acecunted fe; in the design cf 331.1 piping systems also. The basic allevable value of expansion stress is multiplied by a facto: which is related to the nu=ber of stress cycles. The facter functions as an allevable stress reduction fae c due to fatigue service. The values of f are given belev, where N is the number of stress cycles. E L 7,000 and less 1.0 7,000 to 14,000 0.9 14,000 to 22,000 0.8 22,000 to 65,000 0.7 65,000 to 100,000 0.6 100,000 and ove 0.5 These s::ess range reduction f ae:c:s are based upen tests of full si:e pipes made by Marki.'82 Not caly is the basic fatigue p;ccess censidered, but also the deletericus e!!act en f atigue st:ength of varicus fittings, albevs, tees, etc. This is acecmplished by a requirement ec multiply the basic ce=penents of the expansion st:sss by " stress intensification fac: cts" doncted by 1. The numerical values of i vere also derived f:ce full scale tests snd are givsn in l I' the Cede. The stress intensificatien f acts; bears only a nc=inci relatien to the stress concentra icn facters of elasticity; rather, i fe: a gi,:n f ting is related to the rati: of the fatigue strength for the fitting to that of straight pipe. It is in fact a fatigue strength reduction facter. These various fatigue conside ratio ns have been condensed and codified in i apparently simple terms; but it is important :c keep in sind that the app:0ach ) has a basis in full scale testing and, where si=plifica:icns have been made, I
- hey are conserva:ive. It is also ::ue that even :cday with apparently l inexhaustible computer resources available, a single piping systa= is an extraordinarily cc plex s::uctu:e and in a single nuclear plan: the safety class piping sigh: :ssolve d.evn := as many as 90 to 100 piping p :ble=s. ::
can be seen the simplift:ations are not caly desirable, : hey are necessary. l Although an evidently s::aightfervard censiderati:n, the use of the maximu= shear s :ess instead of the maximum normal stress (as a limit of strength) is worth mentioning. The advanced technical nature of 331.1 can be better understeed when it is reali:ed that the videly accepted Seile and F:sssure i Vessel Code used the less accurate maxi =us principal stress thee:y up until i 1966 1 The Ccde has a brief parag;aph that states that earthquake leads, when l a;pli:able, tus: he ::nsidered; havever, ne explicit guidance is prevised. This =ates: vould c dinarily be left :o the designer. However, in nuclear practice, the =agnitude of design basis earthquakes is established as par: ef the licensing ;; cess. Tur: hor, the motheds used to seismically qualify a fs t l H-3 l
- - - - - - - - - _ - _ _ _ _ _ _ _ , _ ___,_,,___m___.__.
i l 1 3EAVIA VALLEY PC7IR STATION, UNIT 1 ! I plant a:e subject to regulatory body ap;; eval, so this ce=binatien of l requirements severned seismic design of 331.1 piping en nuclea plants. As discussed previcusly, in all except the very early plants, a seismic g:cund
=ction in the form of g cund spectra and ap;;e;;iate accelera: ion levels veuld I be specified. This motion would be applied to the buildings and i a plifica:icns of the ground me:ica a varicus levels threugheu: the buildings l vould be ce=;uted in the form cf ficer response spec':1. : :: is the latter i that were used as design bases fe nuclear piping. l The qualification of large piping systems of safety class categories is nearly ,
always dene by means of a ec=puter analysis. A dynamic analytical : del of l i the piping system is derived in which the mass of the system is concentrated at a finite number of = ass points and the flexibility of the system is j - re; esented by springs cennecting the masses. System damping is included as ) visceus damping, normally with highly censerva:ive numerical values of 0.5 c 1 percent of critical damping. The e mpleted medel is then analy:ed for the i a;;;c;;iate seismic spectral =ction en the cceputer. Usually, one a=plified ficer respense spect:um is used as an input acceleration at eve:y ;cin: ef sup;c:: or connection te the building. This simplificatien can be an 1 ;creant conserva:is: especially fer piping systems , travc: sing different vertical levels c: different buildings. The medel of the piping system is passed threugh the cc puter several times te account for all directions of =etion and both the 0;erating and design basis earthquakes, a r Inertia forces are devele;ed first fer all directions vi:hin each mode of vibration, then the ,centributiens of each =ede are ce=bined :o ebtain the tetal fe:ce. A current cen::eversy lies in the fact that feree ce=binattens vi:hin each mede vere in some cases combined algebraically se that se=e leads veuld subtract f:c the :::al. The alterna:ive veuld be te cc bine forces in such a vay that subtraction could net occur, which is the case if an SRSS approach is used. Effee:s of the inertial fe:ces 'are combined with effects f:cm relative building dis;1acements, gravi:y (veight) effects, and internal / external pressure leadings en :he pipe.
'Jhen lead ccebinations are cc ;1ste, bending =ccents and stresses in :he pi;ing system are cesputed accc: ding to 331.1 equations. 3asica117, :vice the max 1=um shearing stress in the pipe due to bending and tension is computed and limited te 1.2 Sh for the 03E and 1.8 5 3 for the D!E in a manner very ce=;4rable to ASP.I III today. S h is the tabulated value ci allevable st:ess as provided by the Cede, in the het condition. In 331.1, S n
is based en the lever cf 5/8 Yield Strength c 1/6 Ulti= ate Strength at c;erating tempera:ure, excep: certain austeni:ic =aterials are pe::itted S h va ues a: c; era:ing temperatures up to 90 percen: of :yteld s::eng:h because of the greater scughness and due:ility of these materials. These values of allevable stress are the. IcVes: in use fer any pi;ing in the Uni:ed States. ASP.I I:: Class i nucisar piping has higher allevables, as does 331.3 Refinery and Chemical n 71an: ?iping. 331.4 and 3 31~. 3 fer Gas and Oil Transmissica ;iping
~/ - respectively permi: allevable stresses u; :o 70 percent of the ul:ima:e X-6
N 3EAVER VALLEY FCVER STATION, UNIT 1
~h strength. When nuclea: plant piping was =cved under the aegis of ASME Section II, Safety Class 3 and 2 centinued to be designed by 331.1; hevever the allevable s::ess ic: the faul:ed plant conditien was raised to .6 Sb' Men:ica is made of certain of these fae:s as an observatien of :ne conservative nature of the 331.1 Code even when ec= pared to other codes :ha:
use the same calculational basis. The mothed of st:ess evaluatica just described is a simplified everview of the , acutal ;;ccess. One of the =cre :cublesome aspects of the vc k is ace:unting l for elbevs. tees, at:achments, and other stress raisers. This is acce plished i by a mandatory multiplication of the stress at ;cints of concentra:ica by I I tabulated " stress intensification factors" c; i factors.
- 5. 331.1 AND LATER CCDES The fi:st versien of the 331.1 Cede was published in 1925, and a revised second edition was published in 1942. Then a :hird editien was issued in 1951. This was a period of rapid development in pi;ing design me:heds and i:
vas found desirable to publish ancther revised edition of the Code in 1955. A brief history is given in :he forevord to the 1955 edition ci 331.1. What is not =en:icned there, however, is that the 1955 edition of the piping code had ; several far reaching engineering imprevemen:s, which have been men:icned earlier herein. The develeptent of the 1955 edition and seme of the changes therein are discussed in References 6, 7. Subsequently, a new edition was published in , 1967, and although there were a number of changes and rino: revisiens, ne new i concepts were int: duced. In 1969 the ANS: 331.7 Code ic nuclear piping was firs: published. The basi: philosophy of this cede was to have nuclea: ;;i=ary syste: ;iping designed to l simila ::iteria as nuclea ;;1:ary system vessels. This required 331.7 to ade; similar a;;: aches c the different possible types of failure and p 0 Vide :=pa:able margins with Section ::: ef the ASME Code. The : des of failure for which ;te:ection is ; evided explicitly by the stress evaluation
;:::edures of See:icn ::: are bursting, excessive plastic deformation,
; egressive diste::len, and :hermal and mechani:a1 fatigue failure. Of ecurse '
c:her possible types of failure are considered in c her areas of the Ccde, specifically in ma:erials selection and fabrica:ica guidelines. , 1 The obvious ap;;cach :: devele; a pipin;* code comparable :: Section !!! for vessels was to at:em;: to adapt the existing 331.1 Cede, which was the ap;::ach taken. McVever, as i: turned out, she 331.1 Cede already centained almes: every ;: vision of See:icn :::, in a differen: forma pe: haps, bu: all ) the basi eence;:s were in place. The devele;=ent of 331.7 then was a ma::e: ef recasting the iginal ;;cvisions of 331.1 in:0 Section ::: fe: mat. Only one technical additica was required tha: eculd be censidered a nev conce;:, 4 I and tha: vas :he addition of censideration fe: :adial temperatu:e gradients th: ugh pipe valis. :n cer:ain situa:icas 0; p:: cesses :his : uld be an importan: censidera:icn, be: in neelaar plants 1: rarely determines :he acce;: ability of piping sys: ems. The ne: resui: is that 331.7, even though l q differen in a;;earance and ;er:1::ing slightly thinner ;i;e vails due :: l 1 l l
.u. .1 ~
3IAVIR VALLIY F0WIR STATION, 7N:7 1
\
higher Section III S values, was not funda=entally di!!srent f:c= the 331.1 Code. This was especially t;ue in the : s: i=pertant aspects of piping design, the limitatica en the =ain expansien strain range and thermal f atigue censidera:icas. The stress indices. Ci and K2 of 331.7 (and Section III), are even generally related to the cid i indices of 331.1. Cs Ks = 2L l This :elationship and other backg:cund en the development of the cu::en: ASMI l Section ::: Piping Code are in a forthceting editica of the ASMI Criteria 3ackg:cund 3ceklet.<** The essential peint of the preceding discussion has been to =ake clear that safety class piping designed to meet the requirements of the cide: ASA 331.1 Code vould aimest vinhout exceptice also =eet the require =ents of the lates: 1 l versica of the ASMI Code. A little more needs t0 be said about seismic design hcVever. The 331.1 Code of 1967 and 1953 clearly spells cut that seis=ic l stresses are to be considered but dcas not say hov. To nuclear plants buil: to these cedes, hevever, this is not significant for present purposes since rigorous seismic analysis was ec=pisted ic these plants to satisfy licensing requirements.
- 6. SIISMIC pIRyCRMANCI CT pC7IR ?!?:NG Although there appear to be no cent:clied experi=ents of seismic perfor:ance of actual piping syste=s, there is, nevertheless, a sur;;ising a= cunt ci very
( interes:ing data on the respense of pcue; piping to actual earthquakes. In the icileving, pcVe plant behavier in several recent earthquakes, Managua 1972, San Ternande 1971, Alaska 1964, Xe n County 1952, and L ng 3each 1933, is discussed. No atte=p has been made te sert c: classify the observa:icns, rather all significan data that could be f und in a she:: ti=e are repersed. Possibly the =es interesting cf the observa:icns are these per:aining : the Kern Steam Statica in the Kern County earthquake, and the Inaluf Steam Plan: in :he Managua earthquake. 3c:h'these plants were designed by conventional p :cedures, both unde:ven severe g:cund shaking, and neither suffered any failures of the piping syste=s. The =aximum g:cund acceleraticas were es:imated :: he as high as possibly 0.6 g at Inaluf, which was adjacent to the
=ain fault causing the quake, and abcut 0.03 g for the Xe n Coun:7 S:ea Plant. Ti=e and again it is seen that piping systa=s cc::se:1y designed for normal service are relatively impervicus to earthquake damage. The basic concept of cent:clied flexibility built into pever piping : enders these syste=s scre esilient than the buildings f::s which they are su;;c::sd.
6.1 Long 3each Steam Station This statice was 'ccated
. en To: inal : stand in L:ng 3each, Califernia, abou:
& =iles f::= the faul: :ta: caused the Long 3each earthquake en March 10, 1933. This ea::hquake was ci magni::de 6.3 and caused accelerati:ns 4: the si:e of the s sas plan: estima:sd :: be adeu: 0.03 g. Ia= age in Long 3each i:self was ve ry e:::ensive , but thers were ne actual a::elerece:e rece:ds of gg
( ,/ the earthquake. 1 I i H-3 4
SIAVIR VALLIY PCWIR STAT:CN, UNIT 1 0 f (-) At the steam station site there were actually three independen: plants. Plant I consisted of one unit and vas built in 1911. It was either out et service c; in intermitten: service in 1933 and the butiding was severely damaged in the earthquake. Plan: 2 consisted of two units and was built in 1922. Plant 3 censisted of three units and was built in 1908. This and subsequent inic :ation was obtained f:cm W.T. Sviger of the Stone & Webste: Engineering Cc:peratica, designers and builders of the plant. To: c hor
- easons it was necessary te re-examine the design of the plan: 4: a later time and it was determined the, plant se:uctu:es were designed for lateral static forces of 0.2 3 Toundations of both plants were heavily reinic:ced cone:ete mats supported by veoden piles 50 tc 60 feet long driven to hard sands. No infor:ation is available on seis=ic design of the piping and equip =ent, but ,
considering the state of the art it is p:chable that either the 0.2 g static design was used, c: else seismic design was not censidered. Neither plant, that is te say, none of the five units, suffered any I significant damage. Sc=e sinc: damage such as to lighting fixtures was l repc :ed; hcVever, the steam plants either operated th:cugh the ea:thquake c vere shut devn due :: icss of Icad and vere back in cpe:a:ica the sa=e day. The i=por: ant peine is that five steam units designed with a: =cs: static
=etheds to a g level (0. ) p cbably lever than actually er.perienced (0.25) were unda= aged and in particular, no piping was damaged.
6.2 Karn Coun:y Steas Statica This cil-fired 60 MW steam plant was designed and built in 1947-8. :: is iccated on the Ke:n River near 3akersfield, Cali!cenia, abcu: 25 miles f:c= ((b the epicenter of the July 21, 1952 Kern County earthquake. This earthquake, sometimes referred to as the Taf:, the Tehachapi, c: the A vin-Iehachapi, was of magnitude 7.7. It was the =cs: severe earthquake recc:ded in the centinental United States since that of 1906 in San Trancisec. It eccurred alcar the 'Jhite Wolf fault scu h and east of 3akersfield. Da= age l vas extensive in 3akersfield and to cil production facilities in the area and 1
-to the Scu: horn Pacific Rail cad. The rail cad tunnel nea: Sealville crossed the fault and was dest:cyed.'
The structures of the plant vere designed for 0.2 lateral icad en a static basis vi:h s::sss limits increased by 0.33 for centined dead, live, and earthquake leadings. Teundations are soit bearing icetings at shallev depth. Anchcrage systems ci all maje equipment including switchgear vere carefully' i reviewed for resistance te lateral leads. This is one of the firs: electric pcVer plants te have piping designed by dynamic analysis. The 31ct'2' s:ce:hed respense spec::um was used ic: the ' design of the =ain steam and boiler feedva:e piping. The :espense spec:: = l vas ac::aliced :c 0.1 3 4: 3:cund level and 0.3 g at the sep ficer ci sne buildings, vi:h linear inte:pelatica 4: ether levels. In this vay an amplified res;ense spectra was available as ave:y ficer, even thcugh 1: vas cf na::cv band and heavily damped ce= pared to spec::a used fer nuclear plan:3. The spee::a vere applied fe :he s:eam and feed lines by calculating the firs: na: ural frequency of each span of ;ipe censidered as a simply suppo::ed bea=. I s X-9 s l
l i. ! SIAVIR VAL *IY FC'JZR STA7!CN, UNIT 1 ("~ . then applying the app;cpriate lateral g force. Based en the cynamic analysis of the =ain piping, psuede-static g 1: ads were developed ft other piping systa=s. These leads were also used to design guides and s: cps and te find 1 cads acting en the supporting structure. It is of interest to note that sc=e guides and steps en the =ain steam line had gaps c: rattle space cf as =uch as
. 2 inches.
An acceleratica recc d obtained at Taft, California, was farther f;0= the l epicsnter than the Kern County Plant. Maximu= acceleration : ace:ded 4: Taft was 0.17 g and it was esti=ated that g cund acceleratica a: the plant site was i a very substantial 0.25 g. The plant operated through the earthquake with no i , significant damage. It was shut devn after the earthquake due to icss of Icad i but was returned to service in a few hours. There was sc=e minor damage to I oil tank seals and a small house turbine thrust bearing, but no da= age at all to piping systems. This is a very clear and graphic exa=ple of the al=est l ce=piste seis=ic p:::ection that is p;cvided by even the = cst rudi=entary seis=ic design precedures (by today's standards). Of course, there was even greater inheren: reserve in the piping systems due to thei: natural centrolled flexibility. 1 l 1 6.3 Alaska Zarthquake of 1964 This earthquake of 8.4 =agnitude was the largest rece:ded earthquake of =edern ti=es. It was centered east of the cit 7 of Anchorage, near the sewn of Valde:. There was videspread destruction thecughout the area, net only f::=
' earth vibration, but f c= the tsuna=1, the failure of poc scils, and fire.
P some observaticas by kncviedgeable engineers of power piping are available, ; but the:e is =cre detailed inic =atica that is yet to be chtained. In a panel
-discussi:n en the Nuclear Piping Cede, observations ve:e noted of pcVer piping behavic; by an experienced piping engineer with a leading Architect /Ingineer). Mr. Tred Vinsen repo::ed that he reviewed the da= age at two power stations i==ediately folleving the earthquake. The pcVer s:ation at an air base in the earthquake ene had no da= aged pi;ing although there were some " bent hanger cds," damaged lighting fixtures, and an eve::urned cent:01 panel due to absence of ancher belts.
A second pcVer plant in the earthquake :ene incurred =cre da age :: the plant, although there was ne failure of pcVer piping. There were fatiures ci se=e equipment supports made of =alleable iren, and an ash handling line conne :ed with patented couplings is reported to have f ailed due to i=p =per support. The significant finding of the observations of Reference 11 is that two power plan:s : de cut the Alaska earthquake with ne failures of the power piping, even though the exae: g levels 4: the sites were net repersed and the design basis was ne: given.e:he :han :c say "very little was done in the way of seis=ic design for the pretection of anything." l i A brief mentien is =ade in Reference 10 cf the Chuga:h Iles:ric C =;4ny plan: in An:horage. This fessil-fueled plan: of abeu: 50 M'4 was built be:veen 1949 and 1957. The plant was designed :e 0.1 g by :he "nifer: 3uilding Code. The buildings were of s:sel f rame censt::cti:n vi:h cc :uga:ed ;anel vails. "here c, 9 H-10
3EAVIA VALLIY POWIR STATION, CN:7 L i was no damage in the turbine ;ces nor to piping and critical equipment. There was minor damage in the boiler room consisting of bending of sc=e bracing
= embers and appreciable damage to fra:ing supporting the coal bunkers. Many piping hangers en the main steam lines vere b cken, but the piping itself was undamaged. The plant was returned c service at full power in less than 10 days.
The consulting fi:s of Ayres and Mayakava of Les Angeles was asked to review all nonstructural damage to buildings due to the Alaska earthquake as part of the investigatien performed by the National Academy of Sciences at the reques of ? ssident Lynden Jchnson. In thei report88 power plants were no: , discussed separately, rather observations of piping systems of all types were discussed on a generic basis. The discussion is based on a study of large modern s::uctures located, with few exceptiens, in Anchorage. The reference report addresses general piping systems of all types, but mainly that required in modern buildings. With the exception of certain fire p;ctection ;iping, nene was seismically designed. Because of the b: cad basis of the repert, the ic11 ving paragraph is quoted directly frem the section entitled " Piping Systems."
"The overall da=43e to piping systa=s was surprisingly icv. Many instances were reper:ed where piping systems re=ained intact, despite the significan: structural and nons::uctural damage suffe:ed by the building.
To example, the plumbing pipes in the Inlisted Men's service Clue at Ter: Richardsen remained standing after the ea::hquake although the valls around the: cellapsed. Centracters also reported that =cs: systa=s were put back into service when pressure-testing :evealed nc isaks." The general conclusion was that piping systems are basically earthquake
- esistant. Tailures occur if at all at threaded fittings. Velded steel pipe dcas no fail. One instance of pcVer piping failure was noted. Small sesa:
pipe drain lines anchered te building valls were :::n f r:m the steam line as it responded :c the earthquake at the Fort Richardson pcVer plan:. This is the type of unbalanced design va:ned agains: in the piping code. ?:Operly-detailed systems had no ;;chie=s. 6.4 San To:nando, Calif :nia, 1971 The San Tornando Zarthquake of 1971 was centered in the n :thern par: cf the San Tornando Valley. C:cund accelerations of 0.1 to 0.19 3 vere recorced in Los Angeles at distances of 33 km and 0.37 g at Lake Mughes, 23 km f:cm the epicenter. Tigure M-l shews recorded g levels !: the 1971 earthquake at varicus 1ccations near Les Angeles. There was severe damage :o a numbe- a# ' s::uctu:es in the valley. The Valley Pcver Plan: is a fossil fuel plant vi:h three units en the si:e located ' d d' es f rem the line ci surf ace rupture (Lakeviev Segment) cf :he primary fault. break. Accelera:1cns at the si:s were es:ima:ed : be in en:ess of 0.05 g based upon the 1: cation of various rece: dings. The s:a:1:n was designed to 0. 'c: 0.05 g at:heugh actual details are ne: kn vn. i l X-!! l'~ l
BIAVIR VALLIY PCVIR STAT:CK, UNIT 1
) In any event there was no da=' age to the plant. It was tripped off the line by action of sudden pressure relays and icss of icad, but was back en the line inside of 2 hoursi'88 There vas significant metien of the pi;ing and seismic
- holddown bars came inte play' "$, but other than insu14
- icn, the piping itself was undamaged. This is a graphic exa:ple of the basic point that voll designed piping :: re gular ccamercial prac: ice is highly resistant to ea:thquake damage. Piping designed te nuclear standards is that much more resistant.
There were other power plants in the area at Playa del Rey, San Ped:c, and Seal 3each that ve:e no: as close to the epicenter as the Valley Plant and none of these were damaged. The San Tornande ?cver Plant is an cid hyd: plant built in 1921 and there was a structural failure of the building which led to a pensteck failure. There were cu= arcus failures of electric trans=ission facilities due to ::acking of pc celain bushings and =cve=ent of pec:17 anchered equi;=ent. There vere ne power piping failures in the San Tornando earthquake. 6.5 Managua, Nicaragua, 1972 ; An earthquake cf magnitude 7.5 struck Managua en December 5, 1970. There was
=uch damage and great icss of life. The icss cf life was largely unrelated to damage of industrial buildings and facilities since the earthquake cccurred nea- ~42-#ght. A repcrt en the damage was sponsored by the National Science Toundatica and several p;ciessional societies together with the Ministry of Public Wo:ks of Nicaragua <i'8 i 1
1 Tigure X- taken frem Reference 15 shevs the fault lines along which =cvement i l occurred running tb cugh the city of Managua. The 1cca:ica of tvc industria' facili:les, the ISSC refinery and the INAL"T Pcver Plan , are also nc:ed. The earthquake res; case of these two facilities will be discussed since they contain industrial piping systems of interest ic: present purposes. A cc=plete acceleregraph record was cbtained at the ISSC refinery. The peak seasured accelera:ica was 0.39 g I-W and .0.34 g N-S. The design of the refinery =e: p:: visions of the Unifc:s 3utiding Code for 0. g, including tall fractienating :: vers, seme of which enceeded several hundred feet. There was almes: ne damage at the refinery and none to the piping systa=s. Some pi;ing jumped cut of saddle supper:s and was pushed back in:o place. The facili:y was shut devn for an inspectica but was c;erating at full capacity within 24 hours even though there was a icss of cifsite,pcVer. The refinery p;cvides a clea; example of the seismic capacity of welded steel pipe that has been designed !c: seismic conditions, albeit statically. 3ased en the earthquake =agnitude, accele:ation reco:d at the :sfinery, and the loca:ica ci the INALUT ?lan 12:ediately adjacent to the causative fault, it is p;cbable this plan: en;erienced accelera:icas en the c: der cf 0.6 g. The ;cver plant censists of three cil-ft:ed uni:s, one of 30 MW and svc of 0 M%' . All th:ee uni:s vers :aken cif-line by ;;c-se:ive :alays. The p'an: suffered s::e damage but nene : the pi;ing systems. It was one of the firs: 1 industrial facilities rest::ed :c servi:e afts: :he earthquake. One uni: vas I~ r M-L: l l l . -
SIAVII VALLIY PC7IR STAT:CN, TNIT l f operating in two weeks, the second in three weeks. Operation of Unit 3 was delayed due to ru:bine p;chle=s. The specific da= age to the three units is listed in Table H-3. Note that no damage occurred t: the piping, and that =any of the pecble=s resulted f:c= absent c; inadequate anchers. To: exa=ple, tu:bine bearings vere ics: because emergency de oil pu=ps were inoperative due to the batteries ::=bling out of thei racks. The basic facts about the pcver piping are that, with unkncvn seis=ic design . applied, but certainly less rigorous than used for nuclear plants, the piping survived site accelerations en the c:da of 0.6 g with no failure. Modern velded steel piping with built-in cent:clied flexibility is inherently highly resistant to earthquake damage.
- 7. 3 ASIS TCR SIISMIC CA7A3:LITT CT PC7Il P:?!NG In the previcus section the perfor=ance of piping syste=s in ;cver plants and a refinery during actual earthquakes was reviewed. :: vas shevn that there vero no piping failu:es even though 3 cend acceleratiens up to 0.6 g vere ex;erien:ed and seis=ic design was usually based on static analysis to the lever value of 0.2 g. This app cach to seis=ic design veuld be considered rudi=entarybyn'jeles: standards.
In the ic11 ving paragraphs, the probable reasons for the ex:ellent perfe:=ance of piping systa=s in earthquakes is ex;1cred. The funda= ental seis=1: ca; ability of piping syste=s ap;arently derives f::= three sources:
- 1. The pover piping design and construction : de, ANS: 33L.1, is quite conservative.
I
- 2. lesigns that are successful for ther=al expansion also p:: vide seed seis=ic ca; ability.
- 3. The large damping factors that becc=e operative in severe shaking are neglected in nor=al design ;:actice.
Taking the above facters one at a time, it is first noted that in Secti:ns 4 and 5 cf this :spert, the 331.1 code for power piping was discussed and it was shcvn that the nuclear power piping codes derived f := 33L.1 have =uch in ce==ca vi h the parent cede. McVever the basic censervatis: vas not covered i in detail. There is substantial =argin provided by the design rules of 331.1. l The average stress in the pipe vall due to the design pressure is limited :o 1/4 of the tensile strength of the steel. Ther=al expansion of the pipe may cause s:: esses due to rest:aint of e:7ansion, but these are displace =ent : strain cent:clied. That is, the strains vill no: bec =e large: :han indica:ed by the assceia:ed te=;erature and vill always be stable, unlike a dead lead : l pressure stress. The s::ain range due :: the =al expansion is limi:sd := a very smali !;a::ica of the s :ain capability of the pipe, censidering :he repetitive nature of the t h e r= al' expansion leading. The code at:e: pts := consider all the ca:ege:ies of leading :ha: a piping syste= vill experience I s 0 l ud t 3-L3 r--, - - . ---mi- y -- g .-r
3EAVIR VALLI 1* PC7IR STATION, UN!! l O
%,s/ nd maintain the pipe in a state of s=411 strain, including the effects of
- ess intensificati n at elbevs and tees.
McVever, the significance of the rules fe; f abrication and construction given by the code tend to be ove:1ceked in discussiens of design capability. The p;cvisiens ic; sound veld design, veld qualifica:ica tests, heat :: eat =ent, inspection, and tests all cc bine to p;cduce piping syste=4 as sound in the field as they 4;; ear en the draving bea:d. The significance of the requirements ic: constructi:n beceses even =cre visible as one reads the references that describe the results ci field inspections ic11cving earthquakes. Occasional references to f ailures ei piping in plumbing systems are made, e.g., Reference 12. In these cases the preblems invariably occur at threaded joints and cecasionally at flanged joints. 7:cught i en and cast iron pipe aise perfer: pectly in ea:thquakes. McVever, p cpe:17 designed and hung velded steel pcVer piping did not fail in even very severe earthquakes. Ividently the cent:011ed flexibility built into voll designed piping systems i= parts substantial seis=ic ca; ability also. !!.. in the design, ;;cvisien is cade for pi;e displacement due te thermal g::v:h, the pipe is then late: un::cubled by forced seismic displace ents. The ;;cvision fc: flexibility may be the : s 1 ;c:: ant aspect of seismic design and is an especially i:;ct:an censideratica in selecting and sicing pipe hangers. : is significan that piping hangers vere reported en one occasica to have failed'2, but the piping 1:self did net. Scund piping material can underge cyclic strain of several percent fe :he limited number of cycles that veuld be 1:;csed by an
~Nearthquake.
_(Y he da ping associated vi:h severe shaking is one of the ecs: i=;ortant
.cnservatis:s in existing a;proaches to nuclea piping design. No : ally visecus da:;ing is assumed with damping facters of 1/2 c 1 percent of critical da=;ing. In a large earthquake hcVever, several energy dissipating mechanisms vill bece=e operative; ordinary material da: ping, i= pact damping, friction c: coule=b dam;ing, and plastic defer:ation when there are large ;ipe me:icas. Taken :0gether, it is clear that damping ratics much greater than design values can be expected.
Sch 'S has presented a reasonably ec prehensive survey of damping in reacter systems. Unfe: tuna:ely, :he data available vere 411 fe: :elatively small dsfiecti:ns. M:vever, there is a clear cc::ala:ica ci damping values vi:h amplitude of vibration. Tigu:e T-3, taken f:c: Reference 16, shews the increase in damping with deflection fe; the data eb:ained f:c tests of full scale nuclear plants. There are aise se e data f:ce the San Cnef:e Nuclea Plan: in the Il Cajon and San To:nando earthquakes.
.:t is interes:ing :o note that the San Ternando earthquake ;;cduced g :und ac:elerations of 0.013 g maximu= a: :he San Onci e site and damping of be:veen
- and a percen: for deflec:icas of abeu: 0.03 inches vere measu:ed by plant ins::::en:atica en the ;;i=ary equi; en:. 04 ping cf 3 :: 3 percen: vas re;c :ed :o have been ceasured in pluck :ss:s a: the Tsuruga Nuclea Plan:.
- n general, damping tha: is =uch highe :han the design value was measured in several tests 4: very s=ali deflections and it increases with amplitude Of x-ta
D**D D 'Tl 6
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1:09089-19 W18M9 041 , 3 L NlE A V Al.LI.*I 10'.'t K S T ATI C:: , t* NIT 1
- c. 6ection. Extrarclatinit the curve of ."1gura M-3 to 0.5 in:h deflection :3.11 j yi Ads 10 percr.nt famping. 23.12 ,;
As olartietty develers in the r i r t r'.c e ve a 2 n '- ^ 1 1 A to': n * * . da ptnr it-Los of 23.l? 10 rerernt and hir.he: Art d e f i".1 t c ly t o t t crT *" t e d. .! n !cft, therC ts 4 23.15 najor Freja:t unde:vay at t h e : t e s erir. ' ' to revel :- s eism e r i:: t : 41 r.e s b.'. s e d on eyelf e glas ticity of t ! c .e t t r e r t e . ,re etre.r..p l <: ett.y e the : e. l .i t ! n n s h i. p 03.17 , betvean d .= . s n ie . a e. c a l a v e s s e n savvi. ans 6 -i - f, em i t. . . - ta r> pins .4 c a a 23.18 nea t. i nc r e a s e p rsis c i t a o t.f.t e l v vi t ti 13.r t i e v e <. i r r t ' f ? ' Igvete e ..i t r.f. s is d .r ot in large part to dr.crn4ser in (avring leveir as (c f l.ce '. i cas t n'. t e t s c . 23.19
- g. C0!P"LUSICits At;D n!!LIC'NTIONS TOR McDr.H litT!.I/.R P Lti.flT S 23.21 1ho' evolutien of reis.ie d e = $. r;: r ett ce in r.uclear ecuet T1?!tf h .19 been 2 2 . ." .'
- avs n'.re d t e re t be r vi t h t i.e d t 'vr lon ent of t. L c ri r ! r.: ceden. .
vu rN:ur. tha? 23.25 n u e l. s , r yaanss t ha t me e t th. e L t.8. r E.it.1 c c. <' e v52' -n r e, t. : in 3ik*)y also . p.etisf7 the nov nuc lea r c oc'e s t.h at have bettet e.v an t i f t r 'J r e nt e l vr '. i t is. 23.*f ova.ilzble data ora the actvAl e e t .* - t e re:ferntice cf reve: c t r ! r.c rvrtere vere 03.!? revifved. It v.12 Pbtvn t!.4? c r a t at ir.g J ova.1 p l e nt :, <l p i rr'e e. d ic ec vety hi g h 23. 7. IcVelf cf trarmic c .n e.bi l i t y . Of the st. vet 0! r l a r.t s t he.t s > - t r i ne d e r vr' r e 23.29 sround rec ti ce f r era c. t o c . o it , the:a vare r. f e : A r : c r. of u- : s e.d escrj pcun; :..) P i r i r* y . f < nr $ dc e t r. :( t li es o c c e n t t v 't .' of t. b - .i s * . ) . et a . .- r. . . -. n ,- ehe . s i $ r 5 i .'. ) t y nf
. 24.I the .CF r !. g r. prratieas. , l:3
- 1* .t rl e n i r l l r r.t r .' a f t . ' sud ': A rt o r .1 v 4,a,9 e been t.64 cj e 2.9 t' Y the n a t.u r a l r e r 311.c re;y er ;tever p 3 p 2 3. s. , ; 4. 6 p eP).ble ,
O p r ve ab! st retrons ter this n .s t ur e. ) r es il: n.cy vo t e d i e c o s t r. d r.e x t . It is 24.6 Jeved that the erin reas.rns ,t i c : f a rn . t h e. vtstan:Jej eence votie, of the Cet t e for Fever T i r a r.tt . 331.1. m:lt dit.? !Se s t ovf ? : e r.- fe: c A t e r :.al s . *k." fabtfr.4 tion, anr4 ,c e r.a t. : e c t s e n t recend. :.5 : t
'**1rn of tiriq4 fer thr: mal 00.B e n r a u r, s.e n T r e vi d e.
- 1 r.he r ens t c e i e r.1 - c e r a te l li t y ; f.nd *trd, that dar;ing .
inert:at es very Tr.ridj y eith def lectier. I c v r l.e . T.S c litru f r:nc i n t !rctort :=.I' , provent t v il e'up ef r e i s.-i : d i - t 'i r t.r.r.e e t j u ' t r et:10t $ Y t t i v. t . 2.t i$ hallCVed 2E.11 those reacone explair. the t erm t sbl e r a s i c N ot e of ritirw systems in e r. r t hqv a ka r . N s e e' vren the f et e ge f r.t nB r e rec t i ftne . It f* vary 1.prebabir- that ;1 ring- 04.17 relat ed saf ety pa r . J an.2 vt'u l d e : e .' t n ::ut. l e n s t e n +. - n t !.c n a t t a. : n 'l r.i t e d . 24.!3 S t A t. fi r: due to suut t c' i .* t p r N n c e r . *bert
,. p l0! . a h a"+
. :r.r v iu 'vi r, t otin d me t t ol:a 24.14 e f O .15 gt t t he y ha ve b e e n d e tt a r. .c 4 ty dyni .:: Ir..!yric nr.d all saiety r 2 r 1:: s 04.)S systr.m nave bee n t r ee:fically 9::vtinJ:ed. C:ntrzed t lJ s = 2 t v .'t t i e n wi t h r a y 2 4 .1 '-
the nr.rn county piar. .:M r o 0.75 e vae a r.t v a fy c e r e! r i r ne tid a t.d .a./ p i t e i t 2 4 . l A r.a l y s i s. vas patfor ad en!y on the ste:am nd fred linc*; ce the F. ? * 'i t'.' T plant
- which vas ytebat.f.y M 31. . d ne.itically and es
- rtv:enced r e t h. :s s 11 . 1 P. ~.h e 24.19 l
c o n t r .= s t if t i ;rly tro r.rret: 7 : r t n r. f ailu rrt if nucitar safety t y r. t r m thould not rnault f.t t . e A r t h qu a l.c s in t he Ur.atad States. 24.20 2 P.T.F E MNC CS 24. P 1 C l ou ri . R.L. et A1. L.' i t o r s . T r u s e'ir r Ver t ri! and P' ring , lie s i te n a nd 26.:! , /' Ana l ys : s , Ar'e r , he , of Mech. L' ng e s . lt . T. , ::. Y . , 19*:. 2 4 . '.> ( e M-l$ e a.. ee n.e e >mem
l l 1 h-1:090f9-19 06/18/?? 041 3 E AV E R VI.L L L Y } 0k'L F. STA13ON. rJNI T 1 ,
\' i ;, 3iet, M.A., Analyticel a nd T.r p e ritre nt al !!c t huvo in f.ngineering Seirrulesy, 2 4 . .*l '
Iract ASCL '08 p 363-40e, 19 :. Os.' y, , stygs, J.M., Introductsen to Structural Liyn;.ni:s, ?!:0: 4v Mill Book Co.. 25.1 196(. 4 perkowitt, t. . . t o S tr.i c Analysia. ef T:t . r v P a r ! ta> Fy s *.-M f e r Uve le a r 05.2
~
Matl 05.3 , c e ne r at ins: S r e t *' t.3 . he ac t. o r ..nd f u e l f r e' et s t r.w : v c t. tic l o gy, Atgunne
- t. a b , fall, 1969.
A.K.C.. T a t ic'.'e T o f t ? of Piring CoNitr.cnte. Trano A,M! V 74 1952. 25.4 l J. Marki, 1 Brock, J.E . E n en'ie n a nd Ti e r.t b l i t y , h t. p . 4 ?:rit.y "andhcoh. Ith Ed., 25.5 ' Xitig and Crecker (Tds.) .W Grr Hil' re!!!?).ing C3.. 1 9 '. 7 . 25.6
], . Ma rkl, A.C.K. , Tiring TIFY1I*i!Liy Agaiysir. II4Mi AGT , 19 M , p 619. !).I A. Criteria of the AOME !' c i ' e r end Prce ure 'l e 9 5 < 1 cede for Denign by .5.9 Au41y?19, 6 er. s tie , of flech. rents.. !?*? (to be publithtd). 25.9 2 suiget, '4.T., pe r s ona l c ote.:.u r.ica t ion. ? ay l??? . *5.1
- 10. Feiter, W.T., b'e t e s on Plantr De ri r"* 4 tv S t. o ne f. ticb*trr Which Save 25.1 p E rr r r '. p nc e d I,r r S ee l'.a r t hq v a a st : , 19 7 9. 'lrt t Li s ha d . 25.1
- 11. Hev !fe:1 ear P i r i r.g Code T.v i c a *'i13
- frditance M etr.t: kaei;:n led ey and 25.1 Tomorrov, Ee o'. i ng , Pir a ng , a d A;r Cene:tien ra. Jene, 1970 p 69. 25.1 1
- 12. ihe Great Alarks C a r t hq t'a h e of 1 6 . J.r.y.ineet:nn, Nctional Acadtmy of 25.1 S c a rnc es , W^r ttinb ton. D.C. , 1772. 25.1
- 13. Fan T e rna rid o s CA!ifornia. ' at tbstrP e of Februa ry 9.
. 1971 tronard P.urphy. 25.)
Lei . Ceo rd, U. S . L'ert. of Ccam., N D /i A . a t ui :t r. t e n . l' . C . , 19)J. ~5.1 14 Snyder. Atthur I., Deen to Mechantu.1 t e c i r .- e n t er c P. r e. e l t of t he .5.1 Teb. 9, 14 71 I a rt hqttar e : n S en Te t n4nca , c c.l i f e t nia , ' ,c i r nes e tiesign and 2 5 . ." Analysis , Ame r . S oc . of Occh. Tngr?., l Y t . 25.0
- 15. !!a ns. cp a , lric e r estu s f*aithqu ee of Dec. 23. 1?*, La t thquake tr. git ee rin;; 25.0 X c ti c a r ch J ns t . , Nov.. !?73. 25.0
- 16. D r.ht , George 1
!'a wr i ns !er r.yn a.-o r Antuyvit n! ?.<!c e t e r Ce o l a nt toep 25.
FYot059 i O p t.1 c o l l'r " LDR en hifaetor. Safety. S a l t. 1.a k e C i t. y , V '. A h . liatch 25.? 197 3, conf- 7 ? O3 0.. Ava ilab i c N7: 5. I
- 17 nuch, Spencer. Dattellr Unrt?.ve
- t Laboratolias, P e r s cinal C oa.. uni c a t ion. '5.!
i i O TELECOPIED ggg wur - en a@ a k. gd d M LN l b W r 5 , >, . <. ../,./ l l't'A
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3EAVIA VALLIY PCIR STA!!0N, UN!! 1 IA3LI H-1 CHRONCLOGY TCR SI SMIC ANALYSIS CT NCOLIAR PLANTS l 1953 Static Methods 1960 Int:cductica of*C:cund Spectra; 3utidings considered Rigid 1963 3uilding Metics and Amplification of Spectra Considered Dyna:ic Analysis and Amplified Respease Spectra Tirst Applied to Piping G:cund Spectra Change 19'c Scil S :ecture Interaction Conside:ed; G:cund Spectra Change 3 Direction.tl Earthquakes Regulatory Guides 1.92, 1.61, 1.60 !amping Changed 1973 Higher Site g Levels Considered; Systa=atic Reevaluation P:cg:a=; Seismic Safety Research i l l l I i I l l l l l t
! cI 1
3EAVIR VALLIY pC'4IR STATICN, CN:7 1 s TA3LI M-2 CMRCNCLOGY TCR $! SMIC ANALYS:S CT P:?:NG SYSTIMS 1955 Static Methods 1960 Static Application of Spectral Accelerations 1965 Re s pons e Spectra Dynamic Analysis; Censideratica of 3rcadened Amplified Spectra; 331.7 Code - Evaluation Criteria 19[0 ASMI Code Section III Applied 3 Directional Earthquakes; Damping Changed; Regulatory Guides 1.92, 1.61, 1.60 1976 Cecasional Time Mistery Analysis; Cecasional Flastic Analysis b t 9 1 c31
-w .
3EAVIR VALLIY POWER STATION, 7N!! l . TA3LI H-3 l i DAMAGED IQU:7 MINT AT TXI ENALOT PCRIA PLANT l l t
' Unit L L. To:ced-draft fan was out of aligt.nent.
- 2. Induced-draft fan was cut of align =ent.
J. 3 earings of the condensate pump burned cut. 4 640 V ac panel fell.
- 5. Condensate pump intake valve was b:cken.
- 6. Sc=e tubing and refractory valls of the beile: vere b cken.
- 7. Deae:ater nu=ber i fell f:cm its base.
- 8. Stack suffered 5:cken splice belts at =id-elevation.
Unit 2
- 1. Tc ced-draft fan was out of alig=nent.
<[~
Induced-draft fan was cut of align =ent. J. Refractory valls of the beiter vare damaged. 4 Deaerate number 2 fell f;cs its base.
- 5. The condensate pu=p intake valve was b:: ken.
CnL 3 l 1
- 1. One 440 V ac cent:01 center fell. i
- 2. Main transformer bushings were broken. l 1
- 3. Starting transic::e bushings were b:: ken.
4 Some preboats: seals were damaged.
- 5. Tour turbine bearings burned cut when the de-;:vered emergency tube oil pu=; batteries b:cks.
- 6. A 69 kV switch bushing was 5 ken.
- 7. 30ile: su;;c:t tubes ever :he preheats: vere broken.
~ 1 l
,. c.e -
SEAVER VALLEY PCVER STAT:CN, UN:T 1
.( s 7A3LI H-3 (ccat)
- 8. Terced-draft-fan cent:ct linkage vas damaged.
- 9. Miscellanscus ali tubes and other tubing. vere b:cken.
- 10. Evaporator drip valve vas b cken.
- 11. Three :scirculating valve bodies were becken.
- 12. Batteries in the battery cca fell f:cm their sup;c :s and b cke.
Miscellaneous Damage
- 1. Turbine bay crane rails were bent and electrical supply conductcts were b cken. Crane remained in place.
- 2. One 138 kV substatien fell.
- 3. Several transic;ner bushings vere b:cken.
4 Tive lightning :cds (69 to 138 kV) were b;cken c: damaged.
- 3. Cns capacito: transformer was b:cken.
\
- 6. Miscellanseus insulate:s were b:cken.
~
- 7. Water sef tener units f ell f:cs their supports and were da= aged.
- 8. One end of the bridge crane in the building that hcused the diesel-electric generators fell f cm the crane girder.
- 9. Other =iscellanecus =iner damage.
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' l a - 1972 PRINCIPAL FAULTS ,g g MAtJ AGU A. tJICAll AGUA """ s u s s a. 1972 SECONDAllY f AULTS $p E AltlllOU AKE SC{M 23 DEC.1972 hh SURVEY StT E NO.
===
* ,'~5D kt its of Alnitanua. ilicaracita $1'"wisic 8ta' ide 888"u$ aiI8etti8se' a'$d lituidings 8 ele 8'e'l '" la 8hicP"'t. See Table i far a list of it.e nacilitics.- . . -
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/ S IPP SkAes F 6 5 Hs $s Gea stata, K (GER Shakes T 5 92 Hs & 5 82 Hs St Gen (W & NSas**
i fGER 8tais I 5 t Hs Se Gem NS*2s C tre SkAes I ~ 2 4811s $s Gen H4t'* M SONGS Shdes I 5 67 Hs 1318 Hs Pa.mp IIS & IWah 0 APP Shaes I 315 ses St Gea Rad'b N SONGS Shdes I. 2 87 Hs Psessusises NS** (- 9 i LPP ShAes I 4 5 lis Pa.mp lan 'a 0 SONGS Shdes I I 87 Hi St Gem f w'* I trP SLA es I 5 3 R a t'*
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yM r G tPP Sbdes I 2 6 Hs HsP.[en S: laa4* P SONGS teste (seek (4seh 2 O lds Se Gee (Wa** 0 SONGS S f essande f asth I $ Its & 2 $5 Hs St Gee Taa & Rar8-A IPP Pseep I 2 4 Hs & 3 25 las St Gen Rad il 4 st 18 ter Skdes I 2 4)Its I iPP Shoes I 315 las Se Gem lan.a is Gen I .a '.' $ gyp p,eep I. 2 4 Its & 3 25 H St Gen Ian II 4dr'* c.': s J IPP Shdes I 6 9 Hs Se Gem I a I IPP Paes, I 4 0 Hs & 4 9 Hs Pump 114dr'* N' en D I I I I I I I I I I I I I I I I I I I l IfI w-b MS Col 0t DEFLECTION (IN) - Fig. H-3 Percent of Critical Damping for_ Reactor Coolant Loop Components Deflection (in) i
OFFICE OF NUCLEAR REACTOR REGULATION WEiiX EN0ING AUGUST 24, 1979 "I. LSTONE 2
-n 21, 1979 meeting with the staff, N:rtheast N" ,ea r Ine ;y e sti ng In C ::g Augustan.y (NNECO) presented che results of radiographic ultraseni:a They of the Millstone 2 feedwater lines as recuired by su 'etin 79-13.
,our 1: cations, the re::rtedihat crack indications were discovered at!
no::le generater. safeqend-to-pipe weld and the pipe-to-eibf weld of each Engineering (CE) designed facility. '
,/
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\s NNECO cnaracteri:e'd s the worst crack indication, located at the pite ed;e of -he safe end-to-pipe weld of the A steam generator as two para cir:umferential indica't(ens about The :
ac;coxicately 12 inches and maxir m depth of 0.110 ' O.0 A fracture mechanic analysis, :erfer ed by l i depth of the worst indicatic . l Westingnouse for NNECO, ind cates that the thermal cycle ;roduced cracks sh0uld be self limited a depth' c.f approxi ately 0.1 in:h. s NNECO stated that rpair cf these four' weld areas would This re:uire weldabout is six weeks due to lack i access to the safe end-::-pipe weld. located six incb.s inside the reinforced concrete wall which provides seismic suppos for the steam generater. s NNECO wil be dccumenting the inspection results a'nd fracture themechanics worst crack analys' , a ec=itrent to install monitoring equipmen(a: ind,i,c tien weld aria before startup, a c:mmitment to' prqide a complete rapair program by Octcber 1,1979 and a ecmmitment Oc shutdown for reinsp f these weld areas before the end of October 1979. Surry Power Station, Unit No.1 On August 22, 1979, NpJs issued an Order permitting resumptien of operation l of Surry, Uni t 1. After an extensive reevaluation using currently ap;reved analytical techniques, it was determined that As theasafety-related result of the reanalyses, piping l systems, with modifications, are ac:eptable. approximately one-third (19 of 63) of the safety-related piping systems in the plant required modifications to correct everstress under pestulated earthquake conditiens. The modifications involved installatien of new or modified pipe supports (shock abscrbers and restraints) and repair of existing supports. Approximately 130 safety-related supports required modification. The majority of the modifications plant were to ake the "as-built" restart is ex:ected the system c nform with the intended design. l early part of September, i l Turkey point - i On Au:ust 3,1979, an a- nd Licensinc Ecard (ASL5) crdered that f a hearing be hejld M ...g M'eam generat:r ie: air pr: gram. In res;:nse ; lQ to the SpMurging that the ; t, st ties meet pr:m;tly recarding these l l
- nt= ..:ns not ~ vet ru el d en and a .hedule for disc:ver'v a meetine of the ~ '
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i APPENDIX XI
Title:
Seismi:. i> f oe-S tress Calculation Techniones and , O Reanalysis Ef fort SEISMIC PIPE-STRESS CALCULATION TECHNIQUES AND REANALYSIS EFFORT ACRS Meeting O Septemeer 8. ,979
/ .O siome a weester ensiaeerias corporeiiom Boston, Mass.
TABLE OF CONTENTS Section Title A Letter from Mr. W.J.L. Kennedy, Stone & Webster Engineering Corporation,to Mr. Harold Denton, U.S. Nuclear Regulatory Ccmmission B Figures:
. Seismic Design Regulatory Guides
. Seismic Design Regulatory Guides (Continued)
. Seismic Design AEC Reports
. Seismic Design Regulations
. Technical Data
. Program Comparisor.
. Seismic Design Const rvatisms es . Seismic Design Conse#vatisms (Continued)
. Seismic Review
. The Review Process
. Computer Program Imr.'.ementation Dynamic Analysis
. Seismic Analysis of Piping
. Computer Program:3 Used for Dynamic Analysis at S6W
. Comparison of Program Methods
. Comparison of Program Methods
. Comparison of Program Methods
. Comparison of Progra.n Methods
. Summary C SEISMIC CAPABILITY OF NGCLEAR PIPING n
v 1
STONE 6 WEBSTER ENCHNEERING CORPORATION 245 SUMMER S TR E ET, B O S TO N, M ASS ACHU SETTS J aconces ALL comarspowersce to n o. sox nas. uceros. wass onio7 25f4l" n'febo. Oftes ;;T.'!,,os. , n,o;.m.s.w o.no, . ~ ,
"avnab May 18, 1979 Mr. Harold Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Sir:
Stone & Webster Engineering Corporation (SSW) and the utility companies involved have been working diligently on the reanalysis required by the NRC Orders to Show Cause of March 13, 1979. We have progressed to the stage where substantial amounts of data are flowing to the NRC and the process of confirming the adequacy of each reanalysis is well under way. () The Maine Yankee reanalysis required by the Order to Show cause has been completed. Work on the other plants is progressing. We have endeavorea to do everything necessary to assure full compliance with he Orders, including develop-ment and use of special procedu_.es. Concurrently, we have rmetM ';b ? cvolution and hi cory of our calculational technanves to help establish a clear understanding of them by all concerned. Given the urgency of the situation, we believe that it would be helpful to both the ACRS and the NRC to have a summary description of our reanalysis activities and procedures and a brief report on our findings with respect to the evolution of our calculational techniques, particularly with respect to satisfaction of the requirements of General Design Criterion 2. REANALYSIS ACTIVITIES AND PROCEDURES The reanalysis effort has centered around the fact that certain piping systems were designed by S&W using a computer program (SHOCK II) with a subroute containing an algebraic intramodal force combination in one of the steps. The NRC has contanded that the use of algebraic summation in this case was incorrect (we do not agree) and has therefore required that all systems designed with the SHOCK II sabroutine be reanalyzed using currently acceptable programs. O
2. i 1 O In view of the seriousness of the impact of the shutdowns and the extensive effort required to ameliorate the situation, i immediate steps were taken upon receipt of the orders to establish the scope of the reanalysis effort, develop work
- plans, organize teams for each project, develop procedures to assure a high quality and timely response, and carry out the j work expeditiously.
Over 400 people, opettting on a two-shift basis, were assembled ! to gather documentation, perform the analyses, confirm the ! adequacy of the analyses, and verify current as well as formerly used computer programs. Project teams and review j groups were established: administrative and technical procedures 1 were developed; orienta. ions were conducted; and schedules of activities were prepared and tracked. The basic elements of the reanalysis ef fort are highlighted d
- below-i 1.
Establishment of Scope of Ef fort Efforts were made to establish the scope of the reanalysis e f fort in conjunction with the NRC and the utilities shortly after issuance of the Orders.
- 2. Development of Work Plans
- A work plan was developed for each of the five units, i
Each plan consisted of identification of affected systems, determination of re: nalysis sequence logic, determination of project manpower cequirements and potential sources, L and establishment of service priorities including computer use. i
- 3. Retrieval of Documentation and Establishment of Design Basis for Reanalysis All available documentation relating to pipe stress and pipe supports was retrieved from archives and/or field records. This documentation included computer and hand-calculations, piping design drawings, flow diagrams, piping and support sketches, and project job books.
Extensive field verification was performed to assure that the reanalysis effort would reflect the "as-built" config-uration of the system being evaluated.
- 4. Computer Program Jerification
'NRC required extunsive verification of the calculational techniques to be used in the reanalysis. This consisted primarily_of aptlying the S&W programs PSTRESS/ SHOCK III and NUPIPE to standard benchmark problems defined by NRC.
(} Another. program tE-PIPE) was verified by NRC itself. S&W
3. supplied cypical problems from each of the affected () plants to NRC for confirmation using E-PIPE. The overall verification effort was based on a stringent interpretation l of Standard Review Plan 3.9.1 and the application of verification techniques and procedures beyond those generally required of the nuclear industry by the NRC.
- 5. Reanalisis of Pipe Stresses Following. verification of the "as-built" configuration, the computer model for the system was recoded, if required, and a comauter run was made using a verified computer program. If the pipe-stress results were within allowable limits, .tbe evaluation sequence proceeded to review of supports, nozzles, and penetrations (see 6.). If the results indicated a possible stress condition in excess of allcwable, the system model was scrutinized in greater detail, revised utilizing current more sophisticated techniques still consistent with the original licensing bases, and rerun. If the results still indicated a possible stress condition in excess of allowable, the system was analyzed using current methodology, such as amplified response spectra with soil-structure interaction, consistent with the original licensing bases and with current requirements.
If this latter rerun does not resolve the problem, additional supports or snubbers are considered. () 6. Reevaluatian of Pipe Supports, Nozzles, and Penetrations Following ;ompletion of a pipe-stress run where the recalcularad loads were within allowable limits, the support e t.) end-reaction loadings were tabulated and compares with the original loadings. If the loadings were equal to or less than the design limits, the supports : or nozzles or penetrations involved were considered acceptable. .If the loadings were above the original design limits, new calculations or supplements to the 1 origir.al calcula tions were perf ormed, using techniques l licensed for the units. The item was then determined l to be either acceptable or in need of modification or replacement. If nozzle reactions exceeded those initially ' approved, vendors were contacted for approval of the new loadings. The vendor either approved the new reactions or requested reductions. In the latter case, other system j modifications were identified and made. ' DEVELOPMENT GF SEISMIC PIPE-STRESS CALCULATIONAL TECHNIQUES The use of algebraic summation for combination of intramodal seismic forces in pipe-stress calculations has been characterized-in~various ways by government officials and the media. Many of the characrarizations have been incorrect or misleading. (f From the outset, as stated in public releases, S&W has agreed that the SHOCK II subroutine may not always yield results as l
4 4. () uniformly conservative as procedures used in the nuclear industry today. We believe, however, that the use of algebraic l summation was an acceptable,ecommon practice at the time it was used by us, and that the piping systems designed with the use of this method will perform their intended safety functions. A review of the history of calculational technique development oy the nuclear industry confirms that continuous effort was made to better approximate seismic effects since the initial modal combination " rule" was suggested in 1943. This " rule" l established an upper bound to maximum seismic response by use of absolute value summation. Because this approach was j considered to be unrealistically conservative, alternative tecnniques have evolved. These include the use of c1gebraic l summation, square root of the sum of the squares, and various ) combinations of all three (absolute, algebraic, and SRSS). ) Through the years we have continually modified our calculational l techniques to lead with advances in both the state of the art and regulatory requirements. In 1972 and 1973, we started l using new computer programs which introduced a number of changes, ) including substitution of a nodified square-root-of-the-sum-of-the-squares (SRSS) procedure for algebraic summation to combine intramodal forces. In late 1974, NRC formally identified specific acceptable calculational techniques in Regulatory Guide (RG) 1.9'. Our modif+.ed SRSS technique, then already in use, met (n,) the requirements of RG 1.92.
.Thea, as now, we recognized the evolutionary nature of computer pretrams and the nature of their use in the overall design p re~ e s s . Therefore, we began to employ the revised techniques ;
ra new projects while we continued to use the previous one on
;ertain projects. In our view, the new techniques provided a means for obtaining more uniformly conservative results. That is why we adopted them. Thin does not imply, of course, that the previous technique was unacceptable. It merely indicates that, consistent with sound professional engineering judgment, we were employing the most advanced calculational tools.
The computer programs develooed by us and by others are calculational tools used to credict forces and resultant stresses that may occur during an earthquake. Such predictions are an integral, but not necessarily the most important, part ! of the total process of establishing design adequacy of piping systems. The present seismic analysis programs utilize more refined calculational techniques, but they still provide only ane stress component among several that must be considered, viz., thermal, pressure. deadweight. Furthermore, system modeling and the assumed cha racteristics of the earthquake ased for' design has greater impact on the solution than the method of.intramodal force combination. With respect to the trend toward more uniform ccnservatism, it should be noted that fO JRSS, which is deemed accept.able and adequately conservative, I' i
5. l (~N is based on statistical concepts and may, in some circumstances,
~ \- be less conservative than earlier aoproaches to seismic analysis.
The evolutionary nature of calculational techniques for seismic analysis was similar to the development of other improvements in analytical methodologies, e.g., the change from simple hand calculations to complex computer codes and from static analysis to dynamic analysis. This evolution was accompanied by a parallel development of seismic criteria by the AEC and the NRC starting in the mid-1960s and continuing well into the 1970s. There was no specific ASME code requirement for seismic design during that period. In response to the recent NRC I&E Bulletin 79-07, additional data have been submitted to the NRC which provide further historical information on the various techniques used in seismic analysis of piping systems. It appears that a number of other responsible organizations have used algebraic summation of intramodal forces in a seismic pipe-stress analysis for a number of nuclear power plants - perhaps 15 to 20 plants designed by firms other than S&W. CONCLUSIONS The efforts we have expended to date have provided results from which some general conclusions may now be drawn. While it is recognized that the algebraic summation of intramodal (\/') forces may not always yield results as uniformly conservative as the present-day techniques, it was a widely used technique acceptable to the profession during the evolution of the art of seismic analysis. It is unlikely that major physical changes will be required in the piping systems on which it was used. It appears more likely that any changes that may be required to satisfy current requirements will be confined to pipe supports. It is a well known and demonstrable fact that piping is not particularly sensitive to seismic events. Such simple, commonly-used design techniques as fastening equipment and restricting the movements of large masses are sufficient to assure a high level of integrity. Further, the various analytical approaches generally show large differences at low stress levels and small differences at high stress levels. In summary, virtually any reasonable design approach or computer program will ensure a high degree of earthquake protection. During the late 1960s and early 1970s the use of dynamic analysis superseded static analysis and the utilization of computers replaced manual techniques. During the same period, responsible organizations used algebraic summation of intramodal forces in calculational techniques employed to (} determine values for use in seismic design of piping systems.
I l l 6. Most important is the fact that these changes did not replace ("]/ s- design review by knowledgeable engineers. This resulted in designs no less rugged than past designs, which have withstood actual seismic events of great severity. It is noteworthy that a number of knowledgeable individuals have observed that the designs resulting from techniques employing algebraic summation do not appear any less substantial than designs resulting from other techniques. Our present reanalysis effort has demonstrated that the differences between the earlier techniques and current practice, resulting from gradual improvement typical of high technology developments, do not yield substantially different final designs. We trust this summary of our reanalysis activity and conclusions are helpful to you. Yourp ver
/ /
[OM a.:/ [/y truly,.de,
- t. xeramer s
_, c) Vice President -
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O O O SEISMIC DESIGN REGULATORY GUIDES RG 1.12 Revision 0 3/71 Instrumentation for Revision 1 4/74 Earthquakes RG 1.29 Revision 0 6/72 Seismic Design Revision 1 8/73 Classification Revision 2 2/76 Revision 3 9/78 RG 1.48 Revision 0 5/73 Design Limits and Loading Combinations for Seismic Category i Fluid System Components RG 1.57 Revision 0 6/73 Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components RG 1.60 Revision 0 10/73 Design Response Spectra Revision 1 12/73 for Seismic Design of Nuclear Power Plants July, 1979
O O O i SEISMIC DESIGN REGULATORY GUIDES (Continued) RG 1.61 Revision 0 10/73 Damping Values for Seismic Design of Nuclear Power Plants RG 1.92 Revision 0 12/74 Combining Modal Response Revision 1 2/76 and Spatial Components in Seismic Response Analysis RG 1.100 Revision 0 3/76 Seismic Qualification of Revision 1 8/77 Electric Equipment for Nuclear Power Plants RG 1.122 ' Revision 0 9/76 Development of Floor Design Revision 1 2/78 Response Spectra for Seismic Design of Floor-Supported Equipment or Components i l RG 1.132 Revision 0 9/77 Site investi9ations for Foundations of Nuclear . i Power Plants l l July, 1979
O O O SEISMIC DESIGN - AEC REPORTS Published 8/63 " Nuclear Reactors and TID 7024 Earthquakes", Lockheed and Holmes and Narver, for AEC. HN-189 Published 3/67 "Some Considerations in Revised 6/68 Earthquake Resistant Design of Nuclear Power Plants", Holmes and Nanter, for AEC. Published 4/68 "Antiseismic Design of HN-192 a PWR Nuclear Power Station", Holmes and Narver, for AEC. July, 1979
O O O SEISMIC DESIGN REGULATIONS 10CFR50 Added 5/71 General Design Appendix A Amended 3/76 Criteria 10CFR100 Added 11/73 Seismic and Geologic Appendix A Amended 1/77 Siting Criteria July, 197
O O O TECHNICAL DATA UNIT SURRY MAINE BEAVER FITZPATRICK 1&2 YANKEE VALLEY 1 EARTHOUAKE: OBE .0 7 g .05 g .0 8 g .0 6 g DBE .15 g .I O g .15 g .12 5 g VERTICAL 2/3 HORIZ. 2/3 HORIZ. 2/3 HORIZ. 2/3 HORIZ. COMPONENTS 2 2 2 2 DAMPING: STRUCTURES CONCRETE A g CONCRETE 1 OBE 5% 2 *4 2% 2% 1% 2% 5% DBE 10 */. 5% 5% 3% 1% 2% 7% PIPING OBE O .5 */. 1.0 %
- O.5 % 0.5 %
DBE 1.0 %
- 2.0 %
- l.O % l.O %
COMPUTER l' PROGRAMS ALL > 6" AL L > 6 USED FOR SOME < 6" SOME < 6,
> 6" ALL > 6" PIPE DIAMETER:
- Verification done
- O.5/1.0 f or
- Steel frame
- Total soil-with O.5 % welded steel bolted / riveted containment
( ( Ref. : Seismic low-stress structure Design Review piping ** welded system Report ) between rigid supports l J t: ly, 1979 i 79-13,307
O O O PROGRAM COMPARISON PROGRAM INTERMODAL INTRAMODAL SHOCK I R=l Rim + s (I"R i j z )- R im _, Fx = Fxx ; Fxy = Fxz = 0 2 2 SHOCK H R= Rim + s (IDRjj )- Rim Fx = Fxx + Fxy + Fxz i=1 ( ALG EBR AICALLY) n 2 " MODIFIED SRSS" SHOCK llI s I Rjj 2 i=1 F=3 x Fxx + Fxz + Fxyl AVAILAB LE OPTIONS : AVAILABLE OPTIONS: 1.SRSS 1. Fx =(l Fxx l + l Fxyl)or(lFxzltlFxy!)
- 2. ABSOLUTE NUPIPE 3. "G ROU PING" 2.Fx =3 Fxx + Fxx 2 + Fxyl
- 4. 10 PERCENT,,
- 5. " DOUBLE SU M " z 2 2
- 3. FxSFu+Fy+Fu Example of nomenclature:
F = response in x direction due to y direction earthquake
.iuly, 1979
- 79-13,308
l O SEISMIC DESIGN CONSERVATISMS i
- 1. Selection of low-probability extreme event )
- 2. Wide band ground response spectra
- 3. Amplification factors
- 4. Enveloping synthetic time histories
- 5. Soil-structure interaction O 6. Three-component earthquake
- 7. Elastic dynamic analysis
- 8. Damping values
- 9. Multiple application of damping values
- 10. Load combinations
- 11. Assumptions of rigid boundaries
- 12. Analysis at peak value of response
. O July, 1979 I _.
l O SEISMIC DESIGN l CONSERVATISMS (Continued)
- 13. Peak widening of floor response system ,
- 14. Envelope response spectra for multiple-supported systems
- 15. System redundancy
- 16. Code minimum allowable stresses 1
- 17. Material specifications O 18. Designer's habits
- 19. Ductility to f ailure
- 20. Seismic stress not always significant fraction of total stress
- 21. Redundancy of structural elements lO July, 1979
O O O SEISMIC REVIEW Unit: Surry 1 & 2 Maine Yankee FitzPatrick Beaver Valley 1 Reviewer: NRC Stati NRC Staff NRC Statl NRC Statt Reviewed: FSAR to Amendment 32 FSAR to Amendment 34 FSAR
- 17 Amendments FSAR
- 11 Amendments Documented SER. 2-23-72 SER. 2-25-72 SER 11-20-72 SER.10-11-74 Reviewer: ACRS ACRS ACRS ACRS Reviewed: FSAR to Amendment 29 FSAR to Amendment 34 FS AR + 16 Amendments FS AR + 12 Amendments Documented: 12-17-71 Letter 1-13-72 Letter 12-15-72 Letter 11-20-74 Letter Reviewer: N. Newmark/Ha!! J. A. Blume Assoc. N. Newmark/ Hall Reviewed: FSAR to Amendment 25 FSAR to Amendment 25 FSAR + Amendments 1-5 and 8-12 Documented: 11-4-71 tSER Appendix) 6-18-71 (SER Appendixi 11-3-72 (SER Appendix)
Reviewer: ASLB ASLB ASLB ASLB Reviewed: (Later) (Later) (Later) (Later) Documented: Note: Note: Seismic Design Review Seismic Design Review Report Amendments 23 and 25 Report Amendment 35 i l { Jul., 19 h
O THE REVIEW PROCESS : i f DWG DATA ACQUISITION j CALCS f ARS ! VERIFICATION l CODING COMPUTER RUN
, PIPING STRESS ANALYSIS pens lSATIONS EQUIPMENT O VERIFICATION HANGER REVIEW HANGER ANALYSIS VERIFICATION ASSEMBLE REVIEW PACKAGE VERIFICATION S&W REVIEW CLIENT REVIEW NRC REVIEW
- Engineering Assurance
, T O p u o r b A e !!6 v, Janim'tN6 July, 1979 i
O O O COMPUTER PROGRAM IMPLEMENTATION DYNAMIC ANALYSIS
,1968 1969 1970 1971 1972 1973 1974 1975 1976 1977 1978 REGULATORY GUIDE 1.92
- JAN 68 JUNE 70 A SHOCK O ' :
MAY 70 OCT 71 , SHOCK 1 : : APRIL 71 AUG 74 SHOCK 2 . .-------------y FEB 72 SHOCK 3 - SEPT 72 MAR 74 NUPIPE - II (CDC) : : : - - - - -- - - - - - - - - - --i JULY 73 FEB 74 NUPIPE - SW - July, 1979
O O O SEISMIC AN ALYSIS OF PIPING INPUT APPROACH RESULTS USAGE o UPPER BOUND RESULTS
^ eM ST CONSERVATIVE e SMALL BORE PIPING SIMPLIFIED HAND CALCUL ATIONS FACTOR e MINIM AL DESIGN INFORMATION e MORE ACCURATE O LESS CRITICAL SMALL A&MRADON e CONSERVATIVE & MODERATE SIZE COMPUTER - STATIC ANALYSIS FACTOR e BETTER DESIGN PIPING INFORMATION A MP LIFIED e REASONABLY LARGE POWER RESPONSE COMPUTER - DYNAMIC PLUS SPECTRA & C O NSERV ATIVE PIPING STATIC DISPLACEMENT e GOOD DESIGN e MODER ATE & SEVERE PROFILE INFORM A TION EARTHOUAKES ACCELERATION e VERY ACCURATE
,p pp OR COMPUTER - DYNAMIC TIME e MINIMAL CONSERVATISM e SPECIAL CASES DIS PL A CE M N T HISTORY e EXCEL'FNT DESIGN S e SEVERE EARTHOUAKES INFORMATION July, 1979
O O O COMPUTER PROGRAMS USED l FOR DYNAMIC ANALYSIS AT S&W l SHOCK 0 SHOCK 1 SHOCK 2 SHOCK 3 , NUPIPE-IILCDC? UTILIZ6D UNDER ROYALTY AGREEMENT NUPIPE - SW PURCHASED FROM NUCLEAR SERVICES CORPORATION t July, 1979 -
O O O COMPARISON OF PROGRAM METHODS SHOCK 0/ ITEM SHOCK 1 SHOCK 2 SHOCK 3 NUPIPE SINGLE GROUND U NIO U E, U NIO U E, UNIQUE, , OR AMPLIFIED AMPLIFIED AMPLIFIED AMPLIFIED SPECTRUM SPECTRUM APPLIED SPECTR A APPLIED SPECTR A APPLIED SPECTR A APPLIED INPUT IN ALL THREE IN ALL THREE IN ALL THREE IN ALL THREE INPUT INPUT INPUT INPUT DIRECTIONS DIRECTIONS DIRECTIONS DIRECTIONS
- i July, 1979
O O O COMPArtISON OF PROGRAM METHODS SHOCK O/ ITEM SHOCK 1 SHOCK 2 SHOCK 3 NUPIPE INERTI A FORCES INERTIA FORCES CALCULATED CALCULATED IN THREE IN THREE INERTIA FORCES COOR DIN ATES COOR DIN ATES INERTIA FORCES CALCULATED FOR EACH INPUT FOR EACH INPUT CALCULATED IN THREE DIRECTION DIRECTION INTR A- IN COORDIN ATE COOR DIN ATES MODIFIED SRSS MULTIPLE OPTIONS MODAL ASSOCIATED FOR EACH INPUT SUM M ATION FOR SUMMATION WITH INPUT DIRECTION APPLIED APPLIED DIRECTION ALG EBR AIC STATICALLY STATICALLY SUMM ATION IN EACH MODE IN EACH MODE TO SOLVE TO SOLVE GENER AllZED GEN ER AllZED RESPONSES RESPONSES GENER AllZED MULTIPLE OPTIONS INERTI A FORCES INERTI A FORCES RESPONSES FOR SUMMING MODAL SUMMED BY G EN ER ALIZED "N AVY" METHOD "N AVY" METHOD p July, 1979
O O O COMPARISON OF PROGRAM METHODS SHOCK 0/ ITEM SHOCK 1 SHOCK 2 SHOCK 3 NUPIPE DISPLACEMENTS DISPLACEMENTS INPUT INPUT SEPARATELY IN SEPARATELY IN DISPLACEMENTS DISPLACEMENTS EACH DIRECTION- EACH DIRECTION-COMBINED WITH COMBINED WITH COMBINED COMBINED DISPLACEMENT INERTIA FORCES - INERTIA FORCES-EFFECT APPLIED TO APPLIED TO DISPLACEMENT DISPLACEMENT SYSTEM SYSTEM STRESSES PLUS STRESSES PLUS STATICALLY STATICALLY INERTIA INERTIA RESPONSES RESPONSES ADDED ADDED ABSOLUTELY ABSOLUTELY July, 1979
O O O COMPARISON OF PROGRAM METHODS SHOCK O/ ITEM SHOCK 1 SHOCK 2 SHOCK 3 NUPIPE SHOCKO SINGLE l PARTICIPATION FACTOR DIRECTIONAL DIRECTIONAL DIRECTIONAL OTHER PARTICIPATlON PARTlCIPATlON PARTICIPATION FACTORS FACTORS FACTORS DIRECTIONAL PARTICIPATION l FACTORS July, 1979
O O O
SUMMARY
~
- S&W PROGRAM DEVELOPMENT HAS BEEN RELATED TO A TIME FRAME
- THE RELATIONSHIP FOLLOWS THE
" STATE-OF-THE-ART" OF SEISMIC PIPING l ANALYSIS l
- AVAILABLE TECHNICAL LITERATURE AND
' USNRC/AEC DOCUMENTATION CONTRIBUTES TO AND HARMONIZES WITH S&W TECHNICAL DEVELOPMENT - July, 1979
l l l l O l l l l 1 1 1 SEIBMIC CAPABILITY OF MJCLEAR PIPHIG l l l l l l 1 ROBERT L. CLOUD MDfLO PARK, CALIFOR.V IA MAY, 1979
) REVISED AUGUST,1$r79 Review Performed For Stone & Webster Engineering Corporation Boston, Massachusetts rx
1 O SEISMIC CAPABILITY OF NUCLEAR PIPItIG Page INTRODUCTION 1 SEISMIC ANALYSIS OF NUCLEAR PIANTS 2 PIPING ANALYSIS 4 ANSI B31.1 CODE 6
.B31.1 AND IATER CODES 12 SEISMIC PERFORMANCE OF POWER PIPING 15 Long Beach Steam Station 16 Kern County Stes.m Station 17 Alaska Earthquake of 1964 20 San Fernando, California,1971 22 i Managua, Nicaragua, icf/2 24 Miya61-Ken-oki, Japan,1978 26 CONCLUSIO !S AND IMPLICATIONS FOR MODERtf NUCLEAR FIANTS 29 l
l 4 0 3 l l
)
SEISMIC CAPABILITY OF NUCLEAR PIPING f O l i INTRODUCTION There are several types and classes of piping in a modern nuclear j power plant and the seismic design requirements of the different classes 1 vary considerably. Further, if the historical evolution of seismic design f requirements is considered there is even more variability. Before the 1 development of the ANSI B31.7 code for Nuclear Piping in the late 1960's ! and subsequently the piping provisions of ASME, Section III, all nuclear safety class piping was designed to meet the requirements of the ANSI (formerly USAS) B31.1 code for Power Piping. As a result many of the operating nuclear power plants in the United States today were designed l and built to meet the provisions of the B31.1 code. A general review of the methods applied to the seismic analysis of B31.1 safety class piping is given including reference to the historical evolution of these methods. Then the B31.1 code itself is discussed and it is shown that contrary to the belief of many, the B31.1 code rests on an advanced technical base, sufficiently advanced, in fact, that very few changes had to be made, other than notation, to upgrade ! l it to the B31.7 nuclear code and then to ASME Section III. D e older I piping code, unlike those for vessels, contained all the main features of current codes. Piping designed to B31.1 is not outdated and performs I very well in earthquakes. S e available data on performance of piping l l in seismic events are reviewed, and it is shown that well designed piping is very difficult to damage as a result of its natural controlled i flexibility. O I j l i l. i.
2. O SEISMIC ANALYSIS OF NUCIEAR FIMPIS Table 1 shows the chronological development of some of the main features of seismic design and analysis methods for nuclear plants. The first plants were designed with static methods using lateral force coefficients as static loads in the manner of various building codes. These plants were in the main built in regions of low seismicity. Dynamic considerations were introduced at about the time plants were built in regions of higher seismicity. In recognition of the amplified response possible when shaking cotions have frequencies at or near the natural frequencies of buildings and equipment, design ground response spectra were introduced for design. Several papers that describe the derivation and application of response spectra methods are contained in the section on Seismic Analysis of Ref (1). This reference was compiled to provide technical bac%round for the advances and changes of various codes for design and construction of pressure n' vessels and piping, especially including nuclear. As such the key papers that influenced the development of nuclear seismic technology by seismic specialists such as Newmark, Hall, Clough, Cornell and others are reprinted conveniently in one place. To obtain the seismic response of piping systems, it is necessary to study the passage of ground motion through the soil, buildings and equipment, all of which cause modifications of the motion before it reaches the piping. Originally, design response spectra were applied to piping in the simplest way considering the first mode of each span O l l
3. O and taking the response directly from the ground spectrum. mis approximation was an improvement over purely static methods, but is quite simplified compared to later methods. Subsequently in the 1960's the effect of building motion on piping s'/ stems was incorporated into the design process on an industry wide basis although the concept had been developed much earlier (2). Con-ceptually, this is done by analyzing the building for the effect of ground motion and developing new spectra at the ficors and walls of the building where piping is supported. In practice this was done at first using records of actual earthquakes, Taft, El Centro, etc. , normalized to the design acceleration level chosen for the site. The accelerations i y l were applied to lumped mass building models in a time history fashion. l At first, very few masses would be used for the building, say less than 10. Also approximate methods were devised to obtain the effect of building amplification on the design spectra (3) directly without a time-history analysis of the building. Design floor spectra were developed by these means and used for several plant designs. In the 1970's several major changes in methods of nuclear plant r seismic analysis were made. S e key changes were a standardization of design ground spectra, a requirement for 3 directional analysis and use of increased damping values. The net effect was a more rational approach to seismic analysis, but in any given case, computed seismic stresses tended to be comparable to those obtained by the more approximate O
l l l 4 I l methods. In any event, this paper is addressed more to B31.1 plants and O1 subsequent developments vill not be discussed further. PIPDIG ANALYSIS Seismic Analysis of piping systems in nuclear plants has also l undergone an evolution, outlined in Table 2, which has been consistent with the growth and development of seismic methods for the plant as a whole. Early methods were based on static analysis using a constant lateral force coefficient that was a specified fraction of the total mass of that part of the piping under consideration. As mentioned previously, when spectra were first used, spectral accelerations con-sistent with span frequency were applied to the piping system. These would be applied normally to the plane of the pipe, i.e. , in the worst
" direction" and combined with a vertical component.
g Later modal response spectra analysis was applied to safety class piping as a=plified floor spectra became available and were specified (4). The application of this approach varied between different organizations and with the times. Although the fundamental steps and basic mathematics were generally common to all, certain choices had to be made in combining responses for each direction and each ecde. These combinations are in a sense arbitrary since the modal response spectra analysis method dispenses with time as a parameter and time relationships including phasing are lost, O
O Scnne analyses have been done by analyzing each of three directions separately and combining contributions from each direction. In many cases the horizontal. direction that causes the worst stress is combined with the vertical and a pinnar response is the basis for evaluation. D e directional combinations have also been made in different ways. Since the various response quantities are signed, algebraic sunnation of responses from each direction within each mode has been done. Analyses have been completed using the other options, SRSS and , absolute sum also. De latter is definitely conservative. - After combinations have been made so that the response for each 4 mode is complete, the sum of all the modal responses must be obtained. Analyses have been completed using several different ways of ceabining l these responses. De methods include a straightforward " square-root-of-the-sum-of-the-squares" of SRSS, the absolute value of the single maximum modal response plus the SRSS of the remainder, and other combinations of absolute values of response closely spaced modes plus the SRSS of the remainder. D e general impetus for the advance of seismic analysis and evaluation methods came from a widely felt need both within the industry and regula.ery agencies to understand seismic behavior of piping. As resultu became available from development activities they would be used for specific plant analysis. Be instigation for doing so vould come as often from the utility or the nanufacturer as O N m
6. from the regulatory agency. It had been a period of rapid technical growth in which all groups concerned with the issue participated. ANSI B31.1 CODE Prior to the implementation of the ANSI B31.7 and the ASME Section III Code, all safety class piping was evaluated according to the ANSI (formerly USAS) B31.1 Code for Power Piping. For the present discussion, the 1955 and 1967 varsions of this code s.re the issues of concern. Ihere was little or no basic change in B31.1 between the 1967 and 1955 versions. Ihe 1955 version however was a major departure
,, from the previous issue of 1942 and supplements. In fact it was in the 1955 version of B31.1 that the basic rules and technical philosophy were established for the design of power piping that are in the main and under different labels still in use today.
The advanced features and underlying technical sophistication of the B31.1 Code have gone relatively unnoticed in this era of rapid technica1 chs.nge and innovation. The B31.1 approach first established in 1955 contained provisions for limiting ther::n1 strain range; recogni::ed the self-limiting nature of thermal stress; contained design rules for low cycle fatigue; incorporated the maximum shear stress theory, and contained other improvements. The ASME Boiler and Pressure Vessel Code contained none of these features at that time. In fact it was not until the Nuclear Vessel Code came out nine years later in 1964 that these technical improvements were applied to pressure vessels. O
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i The fundamental basis of piping design lies in developing a system l that has the correct flexibility and, at the same time, is sufficiently j l well controued. The concept of controlled flexibility is the key to successful piping design. The Code recognizes this with an entire section devoted to piping flexibj.11ty. The approach can be seen from the following, quoted from Paragraph 119 5 of the Code: l 1 Power piping systems shall be designed to have sufficient flexibility to prevent pipe movements from causing failure from overstress of the pipe material or anchors, leakage at joints, or detrimental distortion of connected equip-ment resulting from excessive thrusts and coments. W Flexibility shall be provided by changes of direction in the piping through the use of bends, loops or off-sets; or provisions shall be made to absorb thermal movements by utilizing expansion, swivel or ball , I joints or corrugated pipe. Explicit guidance is given to cbtain balanced systems and to avoid problems of strain concentration caused by uneven flexibility. In this j connection the concept of elastic followp is discussed. Design con-figurations vulnerable to strain concentration are explained and cautioned against. l O
8. W e basic importance of the fact that piping operates in a strain range due to thermal expansion is recognized and explained. It is the strain range that is limited by the Code even though the limitation appears as a limit on calculated stress. Since piping in the ther=al expansion process is in a strain controlled loading situation, the magnitude of the strain range can be controlled by a pseudo-elastic stress calculation. Sir subtle concept was later adopted by ASME Section III. The phenomenon of low cycle fatigue is accounted for in the design of B31.1 piping systems also. he basic allowable value of expansion stress is multiplied by a factor f which is related to the number of stress cycles, he factor functions as an allevable stress reduction factor due to fatigue service. The values of f are given below, where N is the number of stress cycles. E ! 7,000 and less 1.0 7,000 to 14,000 0.9 14,000 to 22,000 0.8 22,000 to 45,000 0.7 45,000 to 100,000 0.6 100,000 and over 0.5 The stress range reduction factors are based upon tests of full size pipes =ade by Marke (5). Not only is the basic fatigue process considered, but also the deleterious effect on fatigue strength of various fittings, elbows, tees, etc. This is accomplished by a require-ment to multiply the basic ccmponents of the expansion stress by 0 l l
9 p V
" stress intensification factors" denoted by 1. W e numerical values of i were also derived from full scale tests and are given in the Code.
De stress intensification factor bears only a nominal relation to the stress concentration factors of elasticity, rather i for a given fitting is related to the ratio of the fatigue strength for the fitting to that of straight pipe. It is in fact a fatigue strength reduction factor. R ese various fatigue considerations have been condensed and codified in apparently simple terms; but it is important to keep in I i mind that the approach has a basis in full scale testing and where simplifications have been made they are conservative. It is also true j that even today with apparently inexhaustible computer resources 1 l available, a single piping system is an extraordinarily complex structure and in a single nuclear plant the safety class piping might resolve down to as much as 90 to 100 piping problems. It can be seen the simplifications are not only desirable, they are necessary. Although an evidently straightforward consideration, the use of the shear stress instead of the normal stress is worth mentioning. De advanced te:hnical nature of B31.1 can be better understood when it is realized that the widely accepted Boiler and Pressure Vessel Code used the less accurate maximum principle stress up until 1964 The Code has a brief paragraph that states earthquake loads, when applicable, must be considered. No explicit guidance is provided however. l v
10. mis a tter would ordinarily be left to the designer. However in nuclear O practice the magnitude of Design Basis Earthquakes are established as part of the licensing process. Further the methods used to seismically qualify a plant are subject to regulatory body approval, se this combination of requirements governed seismic design of B31.1 piping on nuclear plants. As discussed previously, in all except the very early plants, a seismic ground motion in the form of ground spectra and apprcpriate acceleration levels vould be specified. This motion would be applied to the buildings and emplifications of the ground motion at various levels throughout the buildings would be computed in the form of , amplified floor response spectra. It was the latter that are used as design bases for nuclear piping. g The qualificatica of individual piping systems of safety class , categories is nearly always done by means of a computer analysis. A dynamic analytical model of the piping system is derived in which the mass of the system is concentrated at a finite number of mass points and the flexibility of the system is represented by springs connecting the masses. System damping is included as viscous damping, norm 11y with conservative numerical values of .5 or 1 percent of critical da= ping. The completed model is then analyzed for the appropriate seismic spectral motion on the computer. O 1
11. Usually, one amplified floor response spectra is used as an input acceleration at each point of support or connection to the building. B is simplification can be an important conservatism especially for piping systems traversing different vertical levels or different buildings. W e model of the piping system is passed through the computer several times to account for all directions of motion and both the operating and design bases earthquakes. Se inertial forces in the piping system are combined with the gravity forces (weight) of the piping with contents and the pressure forces. B is is done first for all directions within each mode of vibration, then the contributions of each mode are combined to obtain the total force. A current controversy lies in the fact that force combinations within each mode were in some cases combined algebraically so that some loads would subtract from the total. 2e alternative would be to combine forces in such a way that subtraction could not occur which is the case if a SRSS approach is used. j When load combinations are complete, bending mcments and stresses ; in the piping system are computed according to B31.1 equations. l l Basically twice the maximum shearing stress in the pipe due to bending and tension is computed and limited 1.2 Sh for the OBE and 1.8 Sh for the DBE in a manner very comparable to ASME III today. Sh is the tabulated value of allevable stress in the hot condition. In B31.1, Sh is based on the lower of 5/8 Yield Strength or 1/4 Ultimate Strength l l l i
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12. O at operating temperature, except certain austenitic mterials are permitted Sh values at temperature up to 90 percent of yield strength because of the greater toughness and ductility of these materials. These values of allowable stress are the lowest in use for any piping b the United States. Nuclear piping has higher allowables, as does B31.3 Refinery and Chemical Plant Piping. B31.4 and B31.8 for cas and Oil Transmission piping respectively permit allowable stresses up to 72 percent of the ultimate strength. When nuclear plant piping was moved under the aegis of ASMS Section III, the Safety Class 3 and 2 continued to be designed by B31.1; however, the allowable stress for the faulted plant condition was raised to 2.4 Sh from 1.8 Sh . Mention is mde of certain of these facts as an observation of the conservative nature of the B31.1 Code even when compared to other codes that use g the same calculational basis. The method of stress evaluation just described is a simplified overview of the actual process. One of the more troublesome aspects of the work is accounting for cibows, ties, attachments, and other stress raisers. This is accomplished by a mandatory multiplication of the stress at points of concentration by tabulated " stress intensi-fication factors" or i factors. B31.1 AND IATER CODES The first version of the B31.1 code was published in 1935, and a revised second edition was published in 1942. Then a third edition O l
13. was issued in 1951. This was a period of rapid development in piping design methods and it was found desirable to publish another revised edition of the Code in 1955 A brief history is given in the foreword to the 1955 edition of B31.1. Miat is not mentioned there, however, is that the 1955 edition of the piping code had several far reaching engineering improvements, which have been mentioned c' 'ier herein. The development of the 1955 edition and some of the changes therein are discussed in (6,7). Subsequently, a new edition was ) 1 published in 1967, and although there were a number of changes and I l minor revisions, no new concepts were introduced. l In 1969 the ANSI B31.7 Code for nuclear piping was first pub-lished. The basic philosophy of this code was to have nuclear primary system piping designed to similar criteria as nuclear primary system vessels. B is required B31.7 to adopt similar approaches to the different possible types of failure and provide comparable margins l as Section III of the ASME Code. The modes of failure for which protection is provided explicitly by the stress analysis and evaluation procedures of Section III are bursting, excessive plastic deformation, progressive distortion, thermal and mechanical fatigue failure. Of course other possible types of failure are censidered in other areas of the Code, specifically in materials selection and fabrication guidelines. l i i
114 W e obvious approach to develop a piping code comparable to Section III for vessels was to attempt to adapt the exLsting B31.1 Code, which was the approach taken. However, as it turned out, the B31.1 Code already contained almost every provision of Section III, in a different format perhaps, but all the basic concepts were in place. The development of B31.7 then was a matter of recasting the original provisions of B31.1 into Section III format. Only one technical addition was required that could be considered a new concept, and that was the addition of consideration for radial temperature gradients throu6h pipe walls. In certain situations or processes this could be an important consideration, but in nuclear plants it rarely determines the acceptability of piping systems. Se net result is that B31.7, even though different in appearance and permitting slightly thinner pipe walls due to higher Section III S values, was not fundamentally different from the B31.1 Code. This was O especially true in the most important aspects of piping design, the limitation on the main expansion strain range and thermal fatigue con-siderations. The stress indices, C2 and K2 of B31.7 (and Section III), are even generally related to the old i indices of B31.1. C2K2 = 21 This relationship and other background on the development of the current ASMS Section III Piping Code is in a forthcoming edition of the ASME Criteria Background Booklet (8). O 1 1 1
15. O The essential point of the preceding discussion has been to make clear that safety class piping designed to meet the requirements of
- the older ASA B31.1 Code would almost without exception also meet the requirements of the latest version of the ASME Code. A little more needs to be said about seismic design however. 'Ihe B31.1 Code of 1967 and 1955 clearly spells out that seismic stresses are to be considered but does not say exactly how. For nuclear plants built to those codes, however, this is not significant for present purposes since rigorous seismic analysis was completed for these pIAnts to satisfy licensing requirements.
4 SEISMIC PERFORMANCE OF POWER PIPING Although there appear to be no controlled experiments of seismic performance of actual piping systems, there is, nevertheless, a surprising amount of very interesting data on the response of power i piping to actual earthquakes. In the following, power plant behavior 1 in several recent earthquakes, Managua 1972, San Fernando 1971, Alaska 1964, Kern County 1952, Long Beach 1933 is discussed. No attempt has been made to sort or classify the observations; rather l l all significant data that could be found in a short time are reported. Possibly the most interesting of the observations are those pertaining to the Kern Steam Station in the Kern County earthquake, and the Enaluf Steam 11 ant in the Managua earthquake. Both these O
16. O plants were designed by conventional procedures, both underwent severe ground shaking and neither suffered any failures of the piping systems. The maximum ground accelerations were estimated to be as high as possibly 0.6 g at Enaluf, which was right next to the main fault causing the quake, and about 0.25 g for the Kern County Steam Plant. Time and again it is seen that piping systems correctly designed for normal service are relatively impervious to earthquake da::nge. S e basic concept of controlled flexibility built into power piping renders these systems more resilient than the buildings from which they are supported. 1 LONG BEACH STEAM STATION D is station was located on Terminal Island in Long Beach, California, about four miles from the fault that caused the Long Beach earthquake on March 10, 1933. mis earthquake was of magnitude 6.3 and caused accelerations at the site of the steam plant estimated to be about 0.25 g. Damage in Long Beach itself was very extensive, but there were no actual accelerometer records of the earthquake. At the steam station site there were actually three independent plants. Plant 1 consisted of one unit and was built in 1911. It was either out of service or in intermittent service in 1933 and the building was destroyed in the earthquake. Plant 2 consisted of two units and was built in 1922 Plant 3 consisted of three units and was built in 1928. This and subsequent information was obtained from W. F. Sviger (9) of the Stone & Webster Engineering Corporation, O
17 O designers and builders of the plant. For other reasons it was necessary to reexamine the design of the plant at a later time and it was determined the plant structures were designed for lateral static forces of 0.2 g. Foundations of both plants were heavily reinforced concrete mats supported by wooden piles 50 to 60 feet long driven to hard sands. l No information is available on seismic design of the piping and equip- ' ment, but considering the state of the art it is probable that either the 0.2 g static design we a used, or else seismic design was not considered. Neither plant, that is to say, none of the five units, suffered any significant dama6e. Some minor damage such as to lighting 2 ' fixtures was reported; however, the steam plants either operated through the earthquake or were shut down due to loss of load and were back in operation the same day. The important point is that five steam units designed with at most static methods to a g level (0.2) probably lower than actually experienced (0.25) was undamaged and, in particular, no piping was damaged. KERN COUNTY STEAM STATION This oil fired 60 My steam plant -s 69 signed and built in 1947-8. l l It is located on the Kern River near Bakersfield, California, about j l 25 miles from the epicenter of the July 21, 1952 Kern County earth- ] quake. i 4 3 O I l l
18. + This earthquake, sometimes referred to as the Taft, the Tehachapi, 0 or the Arvin-Tehachapi, was of magnitude 7.7. It was the most severe earthquake recorded in the continental United States since that of 1906 in San Francisco. It occurred along the White Wolf fault south and east of Bakersfield. Damage was extensive in Bakersfield and to oil production facilities in the area and to the Southern Pacific Railroad. The railroad tunnel near Bealville crossed the fault and was destroyed (10). The structures of the plant were designed for 0.2 lateral load on a static basis with stress limits increased by 0.33 for combined dead, live, and earthquake loadings. Foundations are soil bearing x footings at shallow depth. Anchorage systems of all major equipment including switchgear were carefully reviewed for resistance to lateral loads. O This is one of the first electric power plants to have piping designed by dynamic analysis. The Biot (2) sa:oothed response spectrum was used for the design of the main steam and boiler feedvater piping. The response spectrum was normalized to 0.1 g at ground level and 0.3 g at the top floor of the buildings, with linear inter-polation at other levels. In this way an amplified response spectra was available at every floor, even though it was of narrow band and heavily damped compared to spectra used for nuclear plants. The O
l 19 O spectra was applied for the steam and feed lines by calculating the first natural frequency of each span of pipe considered as a simply supported beam, than applying the appropriate lateral g force. 1 Based on the dynamic analysis of the main piping, pseudo-static g i loads were developed for other piping systems. These loads were also used to design guides and stops and to find leads acting on the supporting structure. It is of interest to note that some guides l I and stops on the min steam line had gaps or rattle space of as much j as two inches (9). An acceleration record obtained at Taft, California was further from the epicenter than the Kern County Plant. Mavimum , 1 2 acceleration recorded at Taft was 0.17 g and it was estimated that ! ground acceleration at the plant site was a very substantial 0.259 g. The plant operated through the earthquake with no significant damage. It was shut down after the earthquake due to loss of lead but was returned to service in a few hours. There was some minor damage to oil tank seals and a small house turbine thrust bearing, but no damage at all to piping systems. This is a very clear and graph,1c example of the almost complete seismic protection that is provided by even the most rudimentary seismic design procedures (by today's standards). Of course, there was even greather inherent reserve in the piping systems due to their natural controlled flexibility.
20. 9 UIE AIASKA EARD{ QUAKE OF 1964 This earthquake of 8.4 magnitude was the largest recorded earthquake of modern tims. It was centered east of the city of Anchorage, near the town of Valdez. B ere was widespread destruction throu6hout the area, not only from carth vibration, but from the tsunand, the failure of poor soils and fire. l Some observations by knowledgeable engineers of power piping are l l available, but there is more detailed information that is yet to be l l obtained. In a panel discussion on the Nuclear Piping Code, some l l observations were noted of power piping behavior by an experienced i I tr piping engineer with a leading Architect / Engineer (11). Mr. Fred Vinson ) reported that he reviewed the dacage at two power stations immediately following the carthquake. The power station at an air base in the earthquake zone had no da= aged piping although there were some " bent gl l I hanger rods," damaged lighting fixtures and an overturned control Y l panel due to absence of anchor bolta, l A second power plant in the earthquake zone incurred more damage
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to the plant, althou6h there was no failure of power piping. There were failures of some equipment supports made of malleable iron, and ; l an ash handling line connected with patented couplings is reported to have failed due to improper support. O I i
1 21. The significant finding of the observations of reference (11) is l i that two power plants rode out the Alaska Earthquake with no failures I of the power piping, even though the exact g levels at the sites were not reported and the design basis was not given other than to say "very little was done in the way of seismic design for the protection of anything" (11) . A brief mention is made in reference (10) of the Chugach Electric Company plant in Anchorage. This fossil fueled plant of about 50 Mw was built between 1949 and 1957. The plant was designed to 0.1 g by the Undform Building Code. The buildings were of steel frame 1 construction with corrugated panel valls. There was no damage in the turbine room nor to piping and critical equipment. There was minor damage in the boiler room consisting of bending of some bracing l O members and appreciable damage to framing supporting the coal bunkers. Many piping hangers on the main steam lines were broken, but the piping it self was undamaged. The plant was returned to service at
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full power in less than 10 days.
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l The consulting firm of Ayres and Hayakawa of Los Angeles was l 1 asked to review all nonstructural damage to buildings due to the l Alaska Earthquake as part of the investigation performed by the i National Acader:y of Sciences at the request of President Lyndon Johnson. In their report (12) power plants were not discussed separately, rather observations of piping systems of all types were discussed on a gencric basis. The discussion is based on a study of large trdern structures located, with few exceptions, in Anchorage. O v
22. The reference report addresses general piping systems of all types, but mainly that required in modern buildings. With the exception of certain fire protection piping, none was seismically designed. Because of the broad basis of the report, the following paragraph is quoted directly from the section entitled " Piping Systems". The overall dama$e to piping systems was surprisingly low. Many instances were reported where piping systems remained intact, despite the significant structural and nonstructural damage suffered by the building. For example, the plumbing pipes in the Enlisted Men's Service Club at Fort Richardson remained standing after the earthquake although the walls around them collapsed. Contractors also reported that most systems were put back into service when pressure-testing revealed no leaks. The general conclusion was that piping systems are basically n. earthquake resistant. Failures occur if at all at threaded fittings. Welded steel pipe does not fail. One instance of power piping failure g was noted. Small steam pipe drain lines anchored to building walls were torn from the steam lines as it responded to the earthquake at the Fort Richardson power plant. This is the type of unbalanced design varned against in the piping code. Property detailed systems had no problems. SAN FERNANDO, CALIFORNIA,1971 The San Fernando Earthquake of 1971 was centered in the northern part of the San Fernando Valley, Ground accelerations of 0.1 to 0.19 g were recorded in Los Angeles at distances of 35 km and 0.37 g at 9
23. Lake Hughes, 25 km from the epicenter. Figure 1 shows recorded g levels for the 1971 earthquake at various locations near Los Angeles. There was severe damage to a number of structures in the valley. The Van ey Power Plant is a fossil fuel plant with three units on the site located about 5 to 9 miles from the epicenter. Accelerations at the site are estimated to be in excess of 0.25 g based upon the location of various recordings. The station was designed to 0.2 or 0.25 g although actual details are not known. In any event there was no damage to the plant. It was tripped off the line by action of sudden pressure relays and loss of load, but was back on the line inside two hours (13). Ibere was significant motion of the piping and seismic hold-down bars came into play (14), but other than insulation the piping itself was undamaged. This is a graphic example of the basic point that well designed piping to regular consnercial practice is highly resistant to earthquake damage. Piping designed to nuclear standards is that much more resistant. There were other power plants in the area at Playa del Rey, San Pedro and Seal Beach that were no as close to the epicenter as the Valley Plant and none of these were damaged. The San Fernando Power plant is an old hydro plant built in 1921 and there was a structural failure of the building which led to a penstock failure. There were numerous failures of electric transmission facilities due to cracking of porcelain bushings and movement of poorly anchored equip-ment. There were no power piping failures in the San Fernando Earthquake.
24 MANAGUA, NICARAGUA,1972 , An earthquake of magnitude 7.5 struck Managua on December 25, 1972 There was much dannge and great loss of life. The loss of 4 life was largely unrelated to damage of industrial buildings and facilities since the earthquake occurred near midnight. A report on the damage was sponsored by the National Science Foundation and several prafessional societies together with the Ministr/ of Public Works of Nicaragua (15). Figure 2 taken from (15) shows the fault lines along which movement occurred running through the City of Managua. The location of two industrial facilities, the ESSO refinery and the D{ALUF Power r Plant are also noted. The earthquake response of these two facilities will be discussed since they contain industrial piping systems of interest for present purposes. A complete accelerograph record was obtained at the ESO refinery. Se peak measured acceleration was 0.39 g E-W and 0.34 g N-S. The design of the refinery met provisions of the Uniform Building Code for 0.2 g, including tall fractionating towers, some of which exceed several hundred feet. Bere was almost no damage at the refiner / and none to the piping systems. Some piping jumped out of saddle supports and was pushed back into place. The facility was shut down for an inspection but was operating at full capacity within 24 hours even though there was a loss of offsite power. The refinery provides a clear example of the seismic capacity of welded steel pipe that has been designed for seismic conditions, albeit statically. 9
l i 25. O l Based on the earthquake magnitude, acceleration record at v.he refinery and the location of the ENALUF Plant immediately adSc. int l to the causative fault, it is probable this plant experienced ^ accelerations on the order of 0.6 g. D e power plant consists of three oil fired units, one of 50 Mw and two of 20 Mw. All three units were taken off-line by protective relays. The plant suffered some dama6e but none to the piping systems: It was one of the first industrial facilities restored to service after the earthquake. One unit was operating in two weeks, the second in three weeks. Operation of Unit 3 was delayed due to turbine problems. The specific damage to the three units is listed in Table 3, ) l Note that no damage occurred to the piping, and that many of the ! problems resulted from absent or inadequate anchors. For example, turbine bearings were lost because emergency D.C. oil pumps were inoperative due to the batteries tumbling out of their racks. The basic facts about the power piping however are that wi a unknown seismic design applied, but certainly less rigorous than used for nuclear plants, the piping sustained accelerations on the order of 0.6 g with no failure. Modern welded steel piping with built in controlled flexibility is inherently highly resistant to earthquake daw ge. O
26. O MIYAGI-KE:(-OKI, JAPAN,1978 nie Miya61-Ken-oki or Hiya61oki earthquake occurred on June 12, 1978 in the northeastern part of Honshu, the main island of Japan. It was of magnitude 7.h and the epicenter was located just offshore about 100 km, nearly due east of the modern city of Sendai and at a depth of 40 km. This earthquake was well charactericed because of the many strong motion accelero6raph stations in Japan. Fig. h, taken from reference (16) shows the epicenter, the locations of the city of Sendai, the Fukushima Nuclear Power Plant, and several accelero6raph stations. ,; This severe earthquake caused videspread damage in Japan. Approximately
., 28 Ice of earthen river dikes were damaged due to soil liquefaction and subsequent slumping, cracking, and settlement. Several thousand landslides and rockslides occurred both on natural slopes and artificial fill. In the modern city of Sendai, with a metropolitan area population of over one million, damage appeared to be confined to local areas, evidently related to soil cond-tions. Of particular importance was the fact that engineered high-rise buildings up to 10 floors that experienced 0.25 g at ground level (measured) suffered no serious damage. Several smaller, less well engineered buildings were badly dama6ed.
The Fukushima Nuclear Power Plant complex south of Sendai had five operating BWR plants and one under construction. The free field maximum acceleration at the site was 0.12 g. With the exception of one broken ceramic transmission line insulator, there was no dama6e to the site at all. Although not stated in Ref. (16), it is probable the design basis for the nuclear plants exceeded the 0.12 g so there should have been no damage. It is reassuring that there was none.
27. The new Sendai Thermal Plant is an oil fired facility with one unit of 350 W (1971) and one of 600 m (1973). The plant.is located about 15 km east of Sendai. No accelerogram was obtained because the instrument at the site was being inspected at the time of the earthqvake. However, in all probability the site experienced accelerations that were at least in the 0.25 g to 0.40 g range that were felt in Sendai,15 km further from the epicenter. A seismic alarm at the site was triggered at about 0.15 g. The seismic design applied to the plant was not reported. Minor damage was sustained inside the boilers; evidently some " spacer" tubes were sheared and suspended assemblies within the boiler pounded nearby structures. The details of this damage are not known, but it could not have been severe; repairs were made in six days. There was no damage to the power piping, although an additional reference on this topic, Ref. (17) confirmei there were no piping failures such as leaks or cracks, but that there vt.re some deformations and missing anchor bolts for the pipe hangers. Ref. (17) mkes no reference to the boiler tubing. As in the San Fernando earthquake, ceramic insulators in electric power I substations were shown to be vulnerable. A substation near Sendai experienced major damage to the insulators, lighting arrestors, etc. One additional facility deserves mention in this discucsion on power piping performance in earthquakes. Some failures did occt: when a large propane gas-holder, 38 meters in diameter, of the telesceping type collapsed and fell onto the piping systems in a gas plant. Cbvieucly this was not a failure of the piping itself, but of the tank. This tink was located near Sendai. n
28. Five other power plants ranging from 250 to 600 W, Ref. (17), were affected by the Miyagioki earthquake of June, 1978 These plants experienced intensities of 2 to 5 on the Japanese scale, corresponding to IV to VIII on
! the MMI scale; e.g. the intensity at the Pfew Sendai Plant discussed above was 5 on the Japanese scale. All of these additional plants were operating at the time of the earthquake and none were damaged. One of the five was shut down for one hour and inspected, but nothing was found.
a O mee I l 1 1
29. O CONCLUSIONS AND IMPLICATIONS FOR MODERN N'JCLEAR FIANTS The evolution of seismic design methods in nuclear power plants has been reviewed together with the developreent of the Piping Codes. It l was shown that nuclear plants that meet the older B31.1 Code will more l l than likely also satisfy the newer nuclear codes. It was made clear that the major specific feature of nedern piping design methods is to develop within tne piping system a high , i level of controlled flexibility and this feature that pemits successful ; piping design also imparts a high degree of seismic capability. ! 1 All available data on the actual seismic performance of power piping systems were reviewed. It was shown that operating powr plants do indeed have very high levels of seismic cepability. Of the several plants that sustained severe ground motion fmm 0.2 to 0.6 g there were O no failures of welded steel power piping. In one case a steam drain line was reported broken by differential n:ov ment, and this was the only instance of breached integrity found. Considering the magnitudes of the earthquakes and the variabi.11ty of the design practices, this is an excellent record and can only have been m de possible by the natural resiliency of power piping. Based upon the foregoing observations, it is improbable that problems would be expected in nuclear plants in the eastern United States. These plants have maximum ground cetions of 0.15 g; they have been designed by dynamic analysis; all safety piping systems have been O
30. specifically scrutinized. Contrast this situation with say the Kern Ccunty plant where 0,25 g was actually experienced and explicit analysis was performed only on the steam and feed lines; or the ERLUF plant which was probably designed statically and experienced perhaps 0.6 g. The contrast is simply too great; piping failures will not result from earthquakes in United States east coast nuclear plants, O l l l l l 4
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' ' ' ' ' I ' ' ' ' ' I 0.6 ooi o.oi o_ i D E FL ECTION - IN.
FIG.F-3 PERCENT OF CRITICAL DAMPlNG FOR REACTOR COOLANT LOOP COMPONENTS D E FLECTION - IN. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ - -_-_a- -__m_ _ _ _ _ _ - _ - . - - _ _ __-
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'44 136 14 2 X , SEISMIC ACCELEROGRAPH STATION
- FIGURE F4 LOCATIONS OF ACCELEROGRAPH SITES PROVIDING SIGNIFICANT RECORDINGS OF JUNE 12,1978 h
i TABLE 1 SEISMIC ANALYSIS OF ITUCLEAR PLAffrS 1955 Static Methods 1960 Introduction of Ground Spectra Buildings Considered Rigid 1965 Building Motion and Amplification of Spectra Considered Dynamic Analysis and Amp. Res. Spec. First Applied to Piping Ground Spectra Change 197 Soil Structure Interaction Considered Ground Spectra Change 3 Directional Earthquakes R.G. 1 92, 1.6., 1.60 Damping Changed l 1975 Higher Site G Levels Considered I ( -\-} Systematic Reevaluation Program Seismic Safety Roccarch l 1930
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1 1 i f TABLE 2 SEISMIC ANALYSIS PIPING SYSTENB
)
1
~1955 Static hkthods i
1960 Static Application of Spectral Accelerations 1 e 1965 Resp. Spec. Dynamic Analysis Consideration of Broadened Amplified Spectra i B31.7 Code - Evaluation Criteria l ASME Code Section III Applied 1970 3 Directional Earthquakes Damping Changed R.G. 1.92, 1.61, 1.6 , () 1975 Occasional Time History Analysis Occasional Plastic Analysis I i 1980 l l O' I l :.
, , , - - r -
TABLE 3 DAMAGED EQUIPMENT AT ' DIE ENALUF POWER PIANT Unit 1
- 1. The forced-draft fan was out of alignment.
2 The induced-draf t fan was out of alignment, i
- 3. The bearings of the condensate pump burned out.
4 A 440 v ac panel fell.
- 5. The condensate-pump intake valve was broken.
- 6. Some tubing and the refractory walls of ;he boiler were broken.
- 7. Deaerator No.1 fell from its base.
- 8. The stack suffered broken splice bolts at mid-elevation.
Unit 2
- 1. The forced-draft fan was out of alignment.
- 2. The induced-draft fan was out of alignment.
- 3. The refractory walls of the boiler were damaged.
- 4. Deaerator No. 2 fell from its base.
- 5. The condensate-pump intake valve was broken.
Unit 3 () 1. 2 One 440 v ac control center fell. The main-transformer bushings were broken.
- 3. The starting-transformer bushings were broken.
- 4. Some preheater seals were damaged.
- 5. Four turbine bearings burned out when the depowered emergency lube oil pump batteries broke.
- 6. A 69 kV switch bushing was broken.
- 7. The boiler support tubes over the preheater were broken.
- 8. The forced-draft fan control linkage was damaged.
9 Miscellaneous air tubes and other tubing were broken.
- 10. An evaporator drip valve was broken.
- 11. Three recirculating-valve bodies were broken.
- 12. The batteries in the tattery room fell from their supports and broke.
Miscellaneous Damage
- 1. The turbine-bay-crane rails were bent and the electrical supply conductors were broken. The crane remained in place.
- 2. One 138 kV substatien fell.
3.- Several transformer bushings were broken. 4 Five lightning rods (69 kV to 138 kv) were broken or damaged.
- 5. One capacitor transformer was broken.
- c. huscellaneous insulators were broken.
- 7. Water softener untts fell from their supports and were damaged.
- 8. One end of the bridge crane in the building that housed the
. diesel-electric generators fell from the crane girder.
9 Other miscellaneous minor damage m
t REFERENCES
, 1. Cloud, R. L. et.a1 Editors, " Pressure Vessels and Piping, Design and Analysis", Amer. Soc. of Mech. Engrs. N.Y. , N.Y . ,1972
- 2. ' Biot, M. A. , " Analytical and Experimental Methods in Engineering
' Seismology", Trans ASCE 108 P. 365-408, 1942 3.
Biggs, Book Co.,J. M.,4" 196 Introduction to Structural Dynamics", McGraw Hill j 4 Berkowitz, L., " Seismic Analysis of Primary Piping Systems for Nuclear Generating Systems", Reactor and Fuel Processing Technology, Argonne Natl. Lab, Fall, 1969
- 5. Markl, A. R. C., " Fatigue Tests of Piping Components", Trans ASME V 741952
- 6. Brock, J. E. , " Expansion and Flexibility" Chap. 4, Piping ,
Handbook, 5th Ed., King and Crocker (Eds h McGraw Hill i Publishing Co., 1967 ! l
- 7. Markl, A. C. R. , " Piping Flexibility Analysis", Trans ASME, l
, 1955, P. 419
- 8. " Criteria of the ASME Ecilor and Pressure Vessel Code for Design l by Analysis", Amer. Soc. of Mech. Engrs.,1979 (to be published) 9 Sviger, W. F. , personal corJmunication, May,1979
]
- 10. Sviger, W. F., " Notes on Plants Designed by Stone & Webster Which Have Experienced Large Earthquakes",1979, Unpublished
- 11. "How Nuclear Piping Code Rules Will Influence Piping Design Today and Tomorrow", Heating, Piping & Air Conditioning, June 1970, P. 69 4 12 "The Great Alaska Earthquake of 1964, Engineering", National Academy of Sciences, Washington, D. C. ; 1973 i
- 13. " San Fernando, California, Earthquake of February 9,1971",
Leonard Murphy, Sci. Coordi, U.- S. Dept. of Comm. , NOAA,
; Washington, D. C., 1973 1k. Synder, Arthur I. , " Damage to Mechanical Equipment as a Result of the February 9,1971 Earthquake in San Fernando, Califernia", " Seismic Design a.nd Analysis", Amer. Soc. of Mech. En;:rs., 1971
- 15. "Managua,. Nicaragua Earthquake of December 23, 1972" Earthquake Engineering Research Inst. , November,1973
- 16. _ Miyagi-Ken-oki, Japan Earthquake, June P. O., Editor Earthquake Engineering Eesearch Institute, 12,1978, Yanav, 8.
December,197
- 17. Letter, M. Hirata, San Francisco District Mgr., Toshiba ' International Corporation, to T. mci 1raith, Pacific Gas and Electric Co., San Francisco, CA.,. April 9, 1979, t
i I l 1 'O
- EARTHQUAKE: 1933 LONG BEACH M AG. 6.3 (EST.)
FACILITY: LONG BEACH STEAM l . STATION l l 5 UNITS: 2 Built in 1922 3 Built in 1928 DESIGN BASIS: 0.2G Static
- O SITE ACCELERATION
- 0.25G Estimated SITE LOCATION: About 4 Miles from Causative Fault DAMAGE: Lighting Fixtures, etc.
No Significant Damage; No i Failures of Structure or Equipment i O July, 1979 i l
l EARTHQUAKE: TEHACHAPI 1952 g
- (KERN COUNTY) MAG. 7.7 i
! FACILITY: KERN COUNTY STEAM STATION 160 MW OIL FIRED UNIT - 1948 DESIGN BASIS: 0.2G Static + Steam and Feedwater Lines Analyzed with Biot Spectrum; 0.1 to 0.3G SITE ACCELERATION: 0.25G (Est.)
#l Site was nearer epicenter than Taft record (.17G) location DAMAGE: Oil Tank Seals, Small i Turbine Thrust Bearing.
No Damage to Structure, Piping ; or Generation Equipment SITE l X g e xEasgreco l X X FT EPICENTER l O1 July, 1979
O , STANDARD SPECTRA ! PROPOSED FOR DESIGN ! M.A. BIOT 1941 i l l i !
)
l 1.0 - 4 1 0.8 -
- C s
9 06 - w O 0.4 l 0.2 - 1 i l I I I I 0.2 0.4 0.6 0.8 1.0 1.2 1.4 PERIOD, SEC. f O July, 1979
1 ? l i l 4 ! EARTHQUAKE: ALASKA 1964 l M AG. 8.4 l i l TWO POWER PLANTS REPORTED TO i HAVE BEEN SEVERELY SHAKEN WITH i NO PIPING FAILURES l The Chugach Power Plant at Anchorage g i experienced about 0.2G. It was designed l for 0.1G static. There was minor damage
- but no failure of piping, structures or generation equipment.
4 T O l July, 1979
4 O EARTHQUAKE: SAN FERNANDO : 1971 M AG. 6.1 l FACILITY: VALLEY POWER PLANT l 3 UNITS l This Plant was Located About 2.8 Mi.from Line of Rupture and About 5 Mi.from Epicenter. Based on the Locations of V ri us Recordings, the Site is O Estimated to Have Seen More Than 0.25G. DESIGN BASIS: .2 or .25G DAMAGE: No Damage to Plant - Operating at Full Power l 2 Hours After Event b O July, 1979
i ! l 9:' i l l EARTHQUAKE: M AN AG U A 1972 l M AG. 7.5 ! FACILITY: ESSO REFINERY i i l DESIGN BASIS: 0.2G UBC l ! SITE ACCELERATIONS: i .39G E-W' l 1 Measured l .34G N-S . O
- DAMAGE
- Essentially None; Some Piping Jumped from ,
l Saddle Supports and Was l l Lifted Back. Plant was Operating at Full Capacity 24 Hrs After Event l l l 9 July, 1979 l l
O l l EARTHQUAKE: M AN AG U A 1972 , M AG. 7.5 l FACILITY: ENALUF POWER PLANT 3 UNITS DESIGN BASIS: ? O SITE ACCELERATION: 0.6G Estimated Site is Right Next to One of Main Faults DAMAGE: Minor - No Piping or Structural Failures l l l l O July, 1979
EARTHQUAKE: MlYAGIOKI 1978 M AG. 7.4 FACILITY: SENDAl THERMAL PLANT 2 UNITS - OIL FIRED 350 MW (1971) AND 600 MW (1973) SITE ACCELERATION: NOT MEASURED BUT ESTIMATED BETWEEN 0.25G AND 0.40G. SEISMIC ALARM TRIGGERED AT 0.15G. DAMAGE: MINOR TUBING INSIDE BOILERS. NO DAMAGE TO POWER PIPING. SOME PIPE SUPPORT DAMAGE. O O O
O O O EARTHQUAKE: MlYAGIOKI 1978 M AG. 7.4 FACIL!TY: FUKUSHIMA NUCLEAR POWER PLANT COMPLEX 5 BWR UNITS - OPERATING 1 BWR UNIT - UNDER CONSTRUCTION 4 SITE ACCELERATION: 0.12G MEASURED DAMAGE: ONE BROKEN CERAMIC TRANSMISSION LINE INSULATOR. ESSENTIALLY NO OTHER DAMAGE. 4
ll l s l 7
l APPENDIX XII Titie: Seismic Performance of Welded Piping O O l l I SEISMIC PERFORMANCE OF WELDED PIPING COPIES OF FIGURES USED IN THE STONE AND WEBSTER PRESENTATION BY J. R. HALL SEPT. 6, 1979 O
O-1.40 1.20 - f DYNAMIC e, 1.00 - 2 y0.80 - s 0.60 - 0 0.40 - 0.20 -
/-STATIC ,
0.00 t i i i i i i i t i 1 0.00 1.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90 1.00 1.10 1.20 PERIOD - SECONDS DESIGN BASIS: 0.15g O
t ; O ! l EARTHQUAKE: 1933 LONG BEACH M AG. 6.3 (EST.) i FACILITY: LONG BEACH STEAM STATION 5 UNITS: 2 Built in 1922 3 Built in 1928 DESIGN BASIS: 0.2G Static O SITE ACCELERATION: 0.25G Estimated SITE LOCATION: About 4 Miles from , Causative Fault DAMAGE: Lighting Fixtures, etc. No Significant Damage; No Failures of Structure or Equipment O July, 1979 L
i l i l l lO EARTHQUAKE: TEH ACH API 1952 l (KERN COUNTY) MAG. 7.7 FACILITY: KERN COUNTY STEAM STATION : 1 160 MW OIL FIRED UNIT - 1948 DESIGN BASIS: 0.2G Static + Steam and Feedwater Lines Analyzed with Biot Spectrum; 0.1 to 0.3G SITE ACCELERATION: O 0.25G (Est.) Site was nearer epicenter than Taft record (.17G) location I DAMAGE: Oil Tank Seals, Small l Turbine Thrust Bearing. : No Damage to Structure, Piping or Generation Equipment I BAKERSFIELO X X EPICENTER TAFT July, 1979
i O EARTHQUAKE: ALASKA 1964 M AG. 8.4 i d TWO POWER PLANTS REPORTED TO
- HAVE BEEN SEVERELY SHAKEN WITH ;
NO PIPING FAILURES ' O The Chugach Power Plant at Anchorage ! experienced about 0.2G. It was designed for 0.1G static. There was minor damage but no failure of piping, structures or generation equipment. l l l O i July, 1979 l
l O EARTHQUAKE: M AN AG U A 1972 M AG. 7.5 l FACILITY: ESSO REFINERY DESIGN BASIS: 0.2G UBC 4 SITE ACCELERATIONS:
.39G E-W' O Measured
.34G N-S .
DAMAGE: Essentially None; ! Some Piping Jumped from Saddle Supports and Was i Lifted Back. Plant was Operating at Full Capacity 24 Hrs After Event O July, 1979
o 1 EARTHQUAKE: MANAGUA 1972 M AG. 7.5 . FACILITY: ENALUF POWER
- PLANT 3 UNITS DESIGN BASIS
- ?
O SITE ACCELERATION: 0.6G Estimated Site is Right Next to One of Main Faults . DAMAGE: Minor -
- No Piping or Structural 4
Failures O-July, 1979
O O O 'l EARTHQUAKE: MlYAGIOKI 1978 M AG. 7.4 FACILITY: FUKUSHIMA NUCLEAR POWER
- PLANT COMPLEX 5 BWR UNITS -
OPERATING 1 BWR UNIT - UNDER CONSTRUCTION SITE ACCELERATION: 0.12G MEASURED DAMAGE: ONE BROKEN CERAMIC TRANSMISSION LINE INSULATOR. ESSENTIALLY NO OTHER DAMAGE. l l i
i O O O 10 a l 4 l 0
~
l l-
~
E O F-Z n1 O 1 g - m CL a l 0 Z a. E Q 0.1 I I 0.001 0.01 0.1 DEFLECTION-IN. PERCENT OF CRITICAL DAMPING FOR REACTOR COOLANT LOOP COMPONENTS DEFLECTION - IN.
O s* i L ' _ A C T ,O I i I - R _ C _ F g - O I
/o O
g G i N . a l P @ i M A _ D 5 i i e OO O _ ' - ~ - - - _ - O O OO g 4- D m D C (D . O n <.- m -r O O ~
_ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ ____m__ __.___
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- 'N ,
PRESENTAil0N T TIE ACR3 ON SEPIEEER 6,1979 O ' ' " 1 APPENDIX XIII Ti tle : Modifications Made to Plants s' Shut Down for Seismic Re-Analysis AS HAS BEEN DISCUSSED, WE HAVE MADE SOME MODIFICATIONS l TO THESE PLANTS TO EXPEDITE THEIR RETURN TO SERVICE, l'D LIKE TO l EXPLAIN WHAT THESE MODIFICATIONS ARE ALL ABOUT. l l ALL MODIFICATIONS PRESENTLY PLANNED ARE TO PIPING SUPPORTS, l IHERE ARE NO PIPING CHANGES PRESENTLY PLANNED AS A RESULT OF OUR DESIGN BASIS EARTHOUAKE (DBE) ANALYSES, AS MR, KENNEDY MENTIONED EARLIER, EVERYTHING WE HAVE SEEN INDICATES THAT THE PLANTS WERE FULLY CAPABLE OF SAFELY WITHSTANDING A DESIGN BASIS EARTHCUAKE AS THEY EXISTED, AT THE TIME OF THE SHOW CAUSE ORDER, IHE SUPPORTS BEING MODIFIED CAN BE CATEGORIZED AS FOLLOWS: l IN FIRST CATEGORY ARE THOSE WHERE LOCAL STRESSES, USUALLY IN ATTACH-MENT WELDS, EXCEED VERY CONSERVATIVE LIMITS, $UCH LOCAL STRESSES ARE DISPLACEMENT-LIMITED, IHERE MIGHT BE SOME LOCAL YIELDING, BUT THE OVERALL BEHAVIOR OF THE SUPPORTS WOULD REMAIN ELASTIC, REMEMBER THAT WHEN THESE PLANTS WERE DESIGNED LOCAL STRESSES WERE NOT REQUIRED TO BE ANALYZED IN DETAIL BY CODE, REGLUATION OR CCMMON PRACTICE. IHE SECOND CATEGORY INCLUDES THOSE SUPPORTS WHERE STRESSES IN ONE OR MORE SUPPORT MEMBERS EXCEED ALLOWABLE VALUES, BUT ARE WELL BELOW CONSERVATIVELY DETERMINED ULTIMATE STRESS LEVELS, AT WOaST, SOME DEFORMATION MIGHT OCCUR, BUT THE SUPPORTS WOULD CONTINUE TO FUNCTICN, IHE THIRD CATEGORY INVOLVES A FEW CASES WHERE IT APPEARED THAT A SOPHISTICATED NONLINEAR ANALYSIS WCULD BE REQUIRED TO SHOW THAT THE VERY CONSERVATIVE LOADS BEING APPLIED COULD BE WITHSTOOD,
1
.,).
IN lilESE CASES WL GENERALLY llAVE REAtlALYZEU 'lHE SYSTEM ASSUMING O THE SUPPORT WAS COMPLETELY ABSENT. IN EACH CASE, THE PIPING WOULD HAVE C0f1TINUED TO PERFORM ITS FUtlCT10ft DURING AND AFTER ! A DBE. 1 IHE LAST CATEGORY INCLUDES THOSE FEW CASES WHERE SUPPORTS ARE BEING-ADDED TO REDUCE STRESSES IN PIPE THAT WOULD RESULT FROM THE VERY CONSERVATIVE LOADS HYPOTHESIZED DURING A DBE, THESE l MODIFIC,'.<.NS ARE TO RESTORE THE PLANT TO AS-DESIGNED MARGiflS. ' IN EACH .'s.E, WE BELIEVE THAT THE P!PE WOULD HAVE CONTIflUED TO . I PERFORM 75 FUNCTION DURING AND AFTER D8E, EVEN IF THE MODIFICATION l l HADN'T DEEN'MADE. ! l IT IS CRUCIAL TO REMEMBER THAT THE PERFCRMANCE OF THESE SUPPORTS ARE BElflG EVALUATED WITH SOME VERY CONSERVATIVE ASSUMPTIONS. THESE INCLUDE: FIRST, EXTREMELY LOW PIPE DAMPING VALUES HAVE BEEN ASSUMED. SECOND, THE SUPPORTS ARE BEING ANALYZED STATICALLY AND ELASTICALLY. BECAUSE ACTUAL DYNAMIC LOADS ARE DISPLACEMENT-LIMITED I- AND BECAUSE CONSIDERABLE DEFORMATION CAN OCCUR WITHOUT LOSS CF SUPPORT FUtlCTION, THESE ARE EXTREMELY CONSERVATIVE ASSUMPTIONS, IHIRD, THE' STRESS INTENSIFICATION FACTORS BEING USED INCLUDE: Aff ALLOWAf1CE FOR FATIGUE WHICH ISN'T APPLICABLE TO A DBE CASE. FOURTH, NO CREDIT IS BEING TAKEN FOR THE FACT THAT A SUPPORT i STARTING TO YIELD WOULD ABSORB CONSIDERABLE ENERGY, FURTHER INCREASING THE DAMP!NG INVOLVED AND THUS-FURTHER REDUCING THE SEISMIC LCADING. AN IMPORTANT ADDITIONAL CLARIFICATION MUST BE MADE. IHERE 4-HAS BEEN AN ATTEMPT TO CLASSIFY THE MODIFICATIONS BEING MADE AS s_ UNFORTUNATELY, h- -EITHER"AS-BUILT" PROBLEMS OR " REANALYSES PROBLEMS,
~ BASED 0ff THE WORK DONE TO DATE, IT SIMPLY IS NOT POSSIBLE TO MAK:
~
1 , l-i !
_3 SUCH A CATEGORIZATION ACCURATELY. I0 APPRECIATE WHY IT IS GENERALLY NOT POSSIBLE TO ACCURATELY SO CATEGORIZE THESE MODIFICATION, IT IS NECESSARY TO UNDERSTAND HOW THE ORIGINAL ANALYSES AND HOW THE REANALYSES WERE DONE, IHE ORIGINAL ANALYSES GENERALLY WERE COMPLETED BEFCRE ACTUAL CONSTRUCTION TOCK PLACE. DURING CONSTRUCTION, SOME VARIA-TI0flS FROM PLAN OCCURRED USUALLY BECAUSE OF CONSTRUCTION INTER-FERENCES. THESE WERE IDENTIFIED TO QUR ENGINEERS EITHER DURING THE ACTUAL FABRICATION PHASE OR DURING A " STRESS WALK" CONDUCTED AFTER CONSTRUCTION. WHEN THESE VARIATIONS WERE IDENTIFIED, AN l l EXPERIENCED ENGINEER REVIEWED THE VARIATION TO DETERMINE IF IT WOULD HAVE A "SIGNIFICANT" EFFECT ON THE ORIGINAL ANALYSIS. [F HE CONCLUDED THAT A REANALYSIS WAS NECESSARY, IT WAS PERFORMED. GENERALLY, HOWEVER, IT WAS APPROPRIATE TO ALL0d TnE ORIGINAL l i ANALYSES TO STAND. ! WHEN THE QUESTION OF ALGEBRAIC SUMMATION WAS RAISED ) THE ilRC STAFF DIRECTED REANALYSES OF THE SYSTEMS INVOLVED. WE l HAD CONSIDERABLE DISCUSSION WITH THE NRC STAFF CONCERNING THEIR REQUIREMENTS FOR CUR REANALYSIS. FIRST, THEY REQUIRED THAT THE REANALYSIS BE PERFORMED ON "AS-BU!LT" CONDITIONS. SECOND, THEY , REQUIRED A NUMBER OF ADDITIONAL INPUTS TO OUR ANALYSES WHICH 446' COMPLETELY SEPARATE FROM THE OUESTION OF ALGEBRAIC SUMMATION. SUBSEQUENT ANALYSIS REJECTED SOME OF THE SUPPORTS WHICH VARIED FROM THE ORIGINALLY DESIGNED CONDITION. WHETHER THE INCREASE p IN SUPPORT STRESS LEVELS CALCULATED WERE THE RESULT OF AN "AS-BUA T" PROBLEM, ALGEBRAIC SUMMATION OR ANALYTICAL INPUT VARIATIONS CANNOT
_q_ O DE DETERMINED EXCEPT ON A CASE BASIS AND AT THE COST OF EXTENSIVE AND UNNECESSARY AfiALYTICAL EFFORT, BASED ON THE fed CASES, WE DID REVIEW IN DEPTH AND ON MY PERSONAL INVOLVEMENT WITH MORE THAN HALF THE SUPPORTS THAT ARE BEING MODIFIED, [ BELIEVE THAT THE CHANGES IN ANALYTICAL INPUT RESULTED IN MORE MODIFICATIONS THAN ANY OTHER CAUSE. I WOULD LIKE TO SUMMARIZE BY MAKING ONE POINT. THOSE OF US WHO HAVE BEEN DEEPLY INVOLVED EVER $!NCE THE SHOW CAUSE ORDER WAS ISSUED ARE ABSOLUTELY CONVINCED THAT THESE PLANTS WOULD SAFELY HANDLE A 08E TODAY AND WOULD HAVE SAFELY HANDLED A D3E AT THE TIME THE SHOW CAUSE ORDER WAS ISSUED. O l l 1 l I e O
i
!2
APPENDIX XIV-
Title:
Conservatisms of the Original
, Seismic Analysis ACRS MEETING 6 SEPTEMBER 1979 STATEMENT BY KENNETH F. REINSCHMIDT AS HAS BEEN STATED, WE WERE CONVINCED ON 13 MARCH, THE DATE OF THE ORDERS TO SHOW CAUSE, THAT THE PLANTS IN CUESTION WERE SAFE AND THAT THE PIPING WOULD PERFORM ITS ESSENTIAL SAFETY FUilCTIONS IN THE EVENT OF THE DESIGN BASIS EARTHCUAKE. AT THIS TIME, HAVING COMPLETED THE REANALYSIS NECESSARY FCR STARTUP FOR FOUR OF THE FIVE UNITS, WE ARE STILL CCNVINCED THAT THE PLANTS WERE SAFE ON 13 MARCH, I WOULD LIKE TO DISCUSS A FEW CF THE REASONS FOR THESE CONVICTIONS.
AT THE TIME OF THE ORDERS TO SHOW CAUSE IN MARCH, WE BELIEVED THAT THE PIPE WOULD PERFORM WITHOUT LOSS OF FUNCTICM FIRST UF ALL BECAUSE WELDED STEEL POWER P!PIi1G HAS A HiSTCRY OF SURVIVING i.AJOR EARTHOUAKES EVEN WHEN DESIGNED FOR VERY LOW SEISMIC FORCES, AS CISCUSSED EARLIER SY OR. HALL, IHIS HISTORY CF COURSE INCLUDES THE KERN COUilTY STEAM STAT!ON DESIGNED BY STONE 3 WEBSTER WHICH SURV:VED THE IEHATCHAPI EARTHCUAKE UNSCATHED, IHIS CCNFIDENCE IN THE STRENGTH OF STEEL POWER PIPING STEMS IN PART FROM THE WAY IT IS DES!GNED. POWER PIPING IS LAID OUT TO MEET FUNCTIONAL AND THERMAL STRESS RECC:REMEilTS,
)
IT IS SIZED TO CARRY VERY LARGE PRESSURE - !NDUCED STRESSES, AND in SUPPORTS ARE LOCATED TO CARRY DEAD LOAD, THERMAL, AND TRANS!EhT FCRCES. IHEN, TH!S CONFIGURATION IS ANALYZED SEISMICALLY, [F THE PIPE AND , i SUPPORTS HAVE BEEN LA!D OUT ACCORDING TO SOUND ENGINEERIrlG PRACTICE, i I AS THEY WERE, THEN THE SEISMIC ANALYSIS SHOULD RARELY RESULT IN SUB- i STANTIAL MODIFICATIONS. IHAT IS, THE DESIGN C<F THE P!PE SHOULD BE l RATHER INSENSITIVE TO VARIATIONS IN THE METHOD OF SEISMIC ANALYSIS. D .Q i 1
l, WITH SPECIFIC REFERENCE TO THE MATHEMATICAL METHOD IN OUESTION, BOTH THE METHOD OF ALGEBRAIC SUMMATION AND THE METHOD USING THE SOUARE ROOT OF THE SUM OF THE SOUARES ARE APPROXIMATIONS, INTENDED TO FACILITATE THE USE OF THE RESPONSE SPECTRUM METHOD OF DYNAMIC ANALYSIS. IHE SRSS METHOD IS IN FACT A STAT!$TICAL APPRCACH, AND BY THE NATURE OF SUCH METHODS CAN ONLY BE STATISTICALLY CORRECT. THE JUSTIFICATION FOR THE USE OF THE SRSS METHOD IS NO GREATER THAN THE JUSTIFICATION FOR THE ALGEBRAIC SUMMATION METHOD. 30TH ARE APPROXIMATIONS, AND THE USE OF EITHER APPROXIMATION IS JUSTIFIED BY THE OVERALL CONSERVATISM OF THE DESIGN AND THE INSENSITIVITY CF THE RESULTS TO THESE DETAILS OF THE ANALYSIS. AT THE TIME OF THE PLANTS WERE SHUT DOWN, MUCH WAS MADE OF THE FACT THAT BY ALGEERA!C SUMMATION IT IS POSSIBLE THAT RESPONSE COMPONENTS WOULD CANCEL. Ih!S IS INDEED TRUE, BUT IT IS l () ALSO TRUE THAT IN SOME CASES IT IS CORRECT TO ALGEBRAICALLY CCMEIN: l THESE RESPONSES. MORE0VER, IT WOULD SEEM THAT THE STAFF SHOULD HAVE BEEN INTERESTED NOT ONLY IN THE WORST CONCEIVAELE CASE, IN WHICH THE RESPONSES MAY CANCEL, BUT ALSO IN THE LIKELIHOOD THAT THIS CONDITION OCCURS. ALSO, IT WOULD SEEM THAT, IF A CUESTION ARISES AS TO THE RELATIVE CONSERVATISMS OF TWO MATHEMATICAL FORMULAS, THE MOST REASONABLE APPROACH WOULD BE TO EXAMINE THE RESULTS OF THESE FORMULAS, BY MATHEMATICAL METHODS, RATHER THAN TO SHUT DOWN GPERATING POWER PLANTS. ACCORDINGLY, [ RAN SOME SIMPLE TESTS COMPARING THE RESULTS OF THE MATHEMATICAL ECUATIONS USED IN THE SHOCK 2 CCM= UTER
. PROGRAM TO THOSE USING THE ACCEPTABLE SRSS TECHNICUE. THE SHCCK 2 FR: GRAM DID COMBINE INTRAMODAL RESPONSES ALGEBRAICALLY, BUT IT ALSO COMBINED RESPONSES INTERMODALLY BY A FORMULA CALLED THE ElAVY METHOD 4 WHICH IS ALWAYS MORE CONSERVATIVE THAN THE SCUARE ROOT OF THE SUN OF THE SCUARES. IHESE RESULTS WERE COMPARED TO THOSE OBTAINED USING
o. inL SRSS ntinuu noin iNiuAMoDALLY AND iNIUUtuuALLY, lnL REGULiS Of THIS SIMPLE COMPARISION Sil0W TilAT lilr $ HOCK 2 METHOD CIVES RESULTS O SUBSTANTIALLY MORE CONSERVATIVE THAN THE $RSS METHOD, AS SH0hN IN THE F1RST FIGURE, IHIS F!GURE PLOTS THE PROBAD1L1TY OF THE COMEINED RESPONSE OBTAINED FROM THE TWO METHODS. FOR ANY GIVEN VALUE, R, OF THE RESPONSE, THE PROBABILITY THAT THE SHOCK 2 RESULT IS MORE CON-SERVATIVE THAN R IS ALWAYS GREATER THAN THE PROBABILITY THAT THE SRSS METHOD is LARGER THAN R, IN FACT, THE SIMULATION SHOWS THAT THE SHOCK 2 RESULT WILL BE MORE CONSERVATIVE THAN THE SRSS RESutT 92% OF THE TIME, LOOKING AT THE SAME RESULTS IN ANOTHER WAY, THE NEXT FIGURE SHOWS A PLOT OF THE RATIO OF THE SHOCK 2 RESULT TO THE SR$$ RESULT, ON THE AVERAGE, THE SHOCK 2 RESULTS ARE 37% MORE CONSERVATIVE THAN THE SRSS RESuLTS, CONSEQUENTLY, WHEN THE PLANTS WERE SHUT DOWN WE KNEW THAT POWER PIPING WHEN DESIGNED BY NORMAL ENGINEERING PRINCIPLES WOULD WITHSTAND THE POSTULATED EARTHOUAKES, AT THE PRESENT TIME, REANALYSIS HAS BEEN CCMPLETED FOR THE DESIGN 3 ASIS EARTHOUAKE FOR CRITICAL SYSTEMS FOR FOUR OF THE AFFECTED PLANTS, A NUMBER OF MODIFICATIONS HAS BEEN IDENTIFIED TO l RESTORE P!PE AND SUPPORT STRESSES TO BELOW ALLOWABLE VALUES. ONE l l THEN MAY ASK, DOES NOT THE FACT OF THESE MODIFICATIONS PROVE THAT THE PLANTS WERE UNSAFE ON 13 l'IARCH7 IHE ANSWER TO THIS IS MOST I CERTAINLY NO. WE ARE AS CONVINCED AS EVER THAT ALL THE PLANTS WERE SAFE ON 13 MARCH. , IHE PIPE WAS SAFE, REGARDLESS OF THE FACT THAT MODIFI-CATIONS HAVE BEEN MADE, BECAUSE OF THE NATURE OF THE ANALYSIS, IHE ANALYTIC METHODS USED ARE SATISFACTORY FOR DESIGN PURPOSES, IN WHICH THE OBJECTIVE IS TO ASSURE CONSERVATISM WITHOUT UNNECESSARY EFFORT,
- s. ,
IN MANY CASES IN THE DESIGN PROCESS AN EXACT ANALYSIS IS IMPOS$!BLE' l BECAUSE THE FINAL VALUES OF ALL THE VARIABLES HAVE NOT YET BEEN SELECTED. , I 1 l
fl, i IllEREFORE CONSERVATISM AND ECONOMY ARC MORE IMPORTANT THAN GREAT
#CCURACY. ONE SHOULD NOT ASSUME THAT THESE ANALYTICAL METHODS REPRESENT THE TRUE BEHAVIOR OF THE P!PE UNDER DYNAMIC LOADS; THEY D0 fl0T.
SEVERAL SOURCES OF CONSERVATISM WERE DISCUSSED BY l MR. $ISKIN! [ SHALL MENTION ONLY A FEW, PERHAPS THE MOST IMPORTANT BEING DAMPING. FOR THE PIPING REANALYSIS WE WERE CONSTRAINED TO ) l USE A VALUE FOR VISCOUS DAMPING OF 1% OF CRITICAL, [T IS WELL KNOWN THAT ACTUAL DAMPING IS MUCH LARGER, AND THE ACTUAL DAMPING WHEN THE i PIPE APPROACHES YIELD !$ LARGER SY AN GRDER OF MAGNITUDE, AS POINTED OUT BY DR. HALL, CURRENTLY THE NRC PERMITS THE USE OF 33 DAM?ING. WE HAVE EVALUATED THE EFFECT 0F THIS PARAMETER FOR SEVERAL OF THE UNITS, FOR ONE untT, REANALYSIS AT 15 DAMP!NG INDICATED THAT 19 CALCULATION PROBLEMS REQUIRED SUPPORT MODIFICATIONS TO REDUCE THE P!PE STRESS BELOW ALLOWASLE VALUES. WE RERAN FIVE OF THESE PROELEMS USING 3To DAMPING. WHEN THESE FIVE PROBLEMS WERE RUN AT 37. DAMPING, NONE WAS SIGNIFICANTLY ABCVE THE ALLOWABLE STRESS. AS THESE FIVE PROBLEMS REPRESENT 35To CF THE MODIFICATIONS RECUIRED FOR P!PE STRESS, WE CONSIDER IT AN ADEOUATE SAMPLE. MORE0VER, THE ALLOWABLE STRESS WE USED IS LOWER THAN THE CURRENT ALLCWABLE IN THE A3NE ((I CODE, AND IN SCME CASES WE HAD TO INCLUDE THE EFFECTS OF ANCHOR MOTIONS, WHICH ARE NOT REQUIRED AT ALL FOR THE UEE CASE BY THE ASME !!I CODE, IHUS, THE MODIFICATIONS ARE BASED ON ADDITIONAL LOADING CONDITIONS, LOWER DAMPING, AND LOWER ALLOWABLE STRESSES. IHE NET EFFECT 15 THAT THE SEISMIC STRESSES AS COMPUTED USING CURRENT ALLOWABLES AND 3% DAMPING, FOR THE SAME EARTHOUAKE. IHREE PERCENT DAMPING IS STILL VERY CONSERVATIVE. IHEREFORE, WITH MORE REALISTIC SUT STILL CONSERVAT!VE s_ DAMPING AND ALLOWABLE STRESSES, IT IS LIKELY THAT FEW IF ANY MCDIF12 CATIONS WOULD HAVE SEEN REQUIRED FOR SE!SMIC CCNDITIONS, l l E _ _ _ _._ _ __ . _ _____ _. _ __. _
l Jn IHERE IS ANOTHER FACTOR WHICH IS SIGN Fl CANT BUT DIFFICULT TO QUANTIFY USING CURRENTLY APPROVED METHODC. IH{$ IS A
\- THE EFFECT OF INELASTIC BEHAVIOR OF THE SUPPORTS. IN THE REANALYSlS WE WERE RECUIRED TO JUSTIFY OPERABILITY OF THE UNITS BY CCMPUTING THE ULTIMATE STRENGTH CF THE SUPPORTS, BUT WERE NOT PERMITTED TO USE ULTIMATE STRENGTH THEORY TO DO IT. ONLY LINEAR, ELASTIC METHODS ARE PERMITTED, NEVERTHELESS, INELASTIC BEHAVICR OR Y!ELDING OF THE SUPPCRTS IS VERY IMPORTANT TO THE SAFETY OF REAL P!PE, FOR THE FOLLOWING REASONS:
- 1. IF A SUPPORT YlELDS, THE NATURAL FRECUENCIES CF THE SYSTEM ARE REDUCED. IHIS WILL HAVE THE EFFECT OF SHIFTING THE SYSTEM RESPONSE AWAY FROM THE PEAK AREA 0F THE AMPLIFIED RESPONSE SPECTRUM, THEREBY REDUCING THE LOADS ON THE PIPE,
- 2. AS A SUPPORT YIELDS IT DISSIPATES ENERGY, IHIS IS EQUIVALENT TO GREATLY !NCREASING THE ECUiVALENT s
DAMPING OF THE SYSTEM, AND THEREEY SUBSTANTIALLY REDUCES ITS DYNAMIC RESPCNSE, AS EARThCUAKES ARE OF SHCRT DURATICN, THE EFFECT OF YIELDING ON THE FAT!GUE LIFE IS OF LITTLE CONSECUENCE.
- 3. IHE ONLY WAY IN WHICH EARTHCUAKE LOADS REACH THE P:PE IS THROUGH THE SUPPORTS, IHEREFORE Y: ELD!NG SUPPORTS LIMIT THE EFFECTS WHICH ARE TRANSM!TTED AND THERE3Y LIMIT THE LOADS ON THE PlPE, l
[N GENERAL IT CAN BE STATED THAT THE Pl?E IS PROTECTED \ BY THE YIELDING OF THE SUPPORTS, CONSECUENTL7, IT M!CHT EE THCUGhT THAT IT WOULD BE RATIONAL TO DESIGN THE P!PE TO FAIL SAFE EY DESIGN!NG THE 1 l SUPPORTS TO Y[ ELD FIRST IN A CONTROLLED MANNER,-THEREBY SACRIFICING THE s_
/~T SUPPORTS TO PRCTECT THE INTEGRITY AND FUNCTICN OF THE P!PE, Uf.FCRTudATELY,
, (_/ ! CURRENT NETHODS DO NOT PERMIT THIS PHILOSOPHY CF DESIGN AS LINEAR ELASTIC
fJ u I,1ETHODS DO NOT AND CANtt0T REPRESENi THE ACTUAu BEHAVIOR OF THE PIPING O svsTes u"oea ov"^stc 'o^os. [N
SUMMARY
, THEN, OR, HALL HAS PRESEfTED THE EV!DENCE SHOWING THAT POWER PIPING HAS NOT FAILED IN EARTHCUAKES. IHis EVIDENCE SUBSTANTIATES WHAT WE ALREADY KNEW, IWO OF THE MAJOR REASONS WnY IT HAS fl0T ARE DAMPING AND INELASTIC BEHAV!OR, FACTORS WHICH ARE NOT ADEQUATELY REPRESENTED IN THE DYNAMIC ANALYSIS, IHEREFORE, WE CONCLUDE THAT THE PIPING IN THESE PLANTS WOULD NOT HAVE FAILED TO PERFORM ITS SAFETY FUNCT10NS IN THE DESIGN DAS[$ EARTHOUAKE, O O
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i
i a. w X
APPENDIX XV
Title:
Proposed Activities of an EPRI Institute of Nuclear Power Discussion Paper on the Establishment of an Opera tions aucuc im Institute of Nuclear Power Operations Overview
- Help the utihtie3 to help them.
- Conduct seminars and generic sehes rather than preempt their trainmg for sarious utility emplo).
he electric utiht3 industr> s management re3ponsibilities eet including mstructors utiht3 T des elopmg plans for an Insti-
- Encourage escellence. esecutn es. and upper manage-tule of Nuclear Pow er Operations ment to ensure quality in the
'lN PO dedicated to ensuring th" operation of nuclear power In carning out this philosophy high quahts of operation in nuclear of operations. the institute will. programs inner plants its purposes. in hrtet. -
ar e to establish mdustn u ide hench-
- E3tablish industnwide hench.
- Perform studies and analyses to mark 3 lor escellent e in nuclear oper- marks for escellence in the man. support deselopment of critena ation and to conduct independent agement and operation of nuclear for operation for training and for esaluations to awist utilities in meet- the hmnan factors aspects at ing the benchmarks it will deter.
- Conduct independent esaluations dnign and operanon mine educational and training re- to determine that the henchmarks
- Prtnide emergency preparedness quirements for operalma personnel are being met coordination for the nuclear utihty and will accredit traimng orgam- " "
. Heuen nuclear power operatmg
#dh"""
esperiences for analysis and feed.
- Exchange information and esperi-Plans for the Institute are beinM hsk to the utilities incorporate ence with operators of nuclear deseloped under the leadership of lessons learned into training pro- Power plants in other countnes Dr Chaunce) 5tarr. Vice Chairman grams. Coordinate information re-of the Electric Power Research in3ri- porting and anal > sis with other tute +EPHI: Committees and adsiaol) orga nizations Institute Organization groups insched in the planning pro-ceo are listed at the end of this * . tablish educational and trainin.t rp he institute consists of a Board requirements for operations and 1 of Directors. an 3dsisen Coun-maintenance personnel and cil an Industrv Resien 5tructure. a Phdo soph.t of Operations dmelop screening and perfarnd President and' lis e disi3 ions The ance measurement 33 stems Board of Directors will set institute the philosophy of the Institute
- Accredit training programs and policies and direction and will he of Nuclear Power Operations is to eruf> instructors, composed of a chairman and elected
- Promote a lesel of professionalism in nuclear power operations com-mensurate with the importance to ihe public of safe reliable and '"**'**N"C'**'N"*'0**'*"*"'
$ tan Structwre economically efficient operation 3
- Imohe plant operating stafl3 in dmelopment of henchmarks and wg traming $3 stems m the con- oreciers duct of the operation evaluations
- Use the best asailable techniques and method 3 to deselop operating wsery **SW and training practices and the " "*'
s , human factors aspects of design and operation
- Ltilize independent professional o ,seem adsice and coun3cl tow ard3 accomph3hmg the Institute s ,
obiectis es l g l
- support and improse esisting O, practices and training wstems.
, ram &
ELea en cmca
' *ec u ta E,,arance
,,,,,,,c, s Ememen 3,c y,,c,,,
, as y "s,$nanc, w herm er possible rather than N'Sc" ,g7$,'
o b &c" bs ca supplant them. l l l
members trom the member unlities
- Relations n,th regulaton and gos- t he Dnision w di provide liaison fhe \dsison Councd will resiew ernment agencies on pubhc safets with architect +ngineers and sendor m organianons to ensure their partici-institute programs and prtntde ad- , Crtsi3 management and commune s h e to the Board of Directors it will cation with the public and Pdh""i"I"'U'"'" "'"'U"' I"
- be t omposed of di>tmguished mahon on relate temnwa rewad the regulaton agencies persons m areas related to the Insti- need3 such as mstrumentat on tuie s oh'ectnes and w ill include deselopment a di he 3ent to EPHI Operations F,rogram and other appropriate orgamzations pronunent educators 3cient:3ts enumeers mdustnabst s and health
- Upgradmg ot training and abthtte' To support entena deselopment s pecialist s The Industry Heuew of operators to handle off normal 3tudies and anah 3es nught be con-structu e w til perform resiew and es ent> ducted in such areas as direction funcuon3 assigned to it
- Increasing the understandme of . \dequacy of traimne simulator in the Board of Directors it will be plant systems and operatmns b) models and techniques of simu-composed of industn persons espe- supenisors and engmeenne lator use Deselopment of per-nenced m areas related to the Insti- support staff formance measures and esaluahon tute s actnities Its orgamzanon will
- Deseloping traimng programs for s) st ems parallel that of the statf structure su pen isors . Enectnene,s of aJs anced instrue.
T he President w ill manage the Insti-tional metnods such a3 computer-tute s operatmo and w til be chosen T'echnicians Program aided innrucnon and part task hs the Boani of Directors simulators it is estimated that the functions
- Des elopment of training specifi.
of the Instuute will require a staff of cations and performance stan-
- P3schological and phy siological about 30 The functions of each dards for skdled workers and for studie3 in area 3 such as shift rota-diuuon are de3cribed in the fol. technicians insohea in plant non. attentisenes> and 3tre33 low me wcuons maintenance and operation"
- M >s d su pport procedures Training and Education Disision The required staff of this Diu-
- Contial room alann presentation.
General fum nons of this Oni- 3 ion is esumated to be about 40 control concept 3 and information sion n ill he to processing and dt3 play
- Hes ien curricula lewon plan $ Criteria DeSelopment and
- Niethods for effectne use of oper
% and training malenals Ext 3tmg Analy sis Dhiston atmg and ca3ualty procedures approaches w di be upgraded
- Detailed human factors analy sis of The Dnision will deselop bench-il needed marks and guidehne3 for use in generic operating problems
- kcredit mstruction sy stems for nuclear pow er operanons. u di iden-
- Wen responw tesung for nuclear pow er operanons technol- tity training needs and will deselop 3rreemg and daebpem om (,ertily instructors as3ist in and evaluate training techmques their trainme and the deselopment
- Organiutional behamr and f The des elopment proces3 will '
ot tractung skilh a3 necessan I""I n d " "- include the following-l
- Mst m utiht3 planning to meet
- Informanon on present practices. The required 3taff of this Uni- 1 manpow er need' obtained from the Traieing and sion is esumated to be about to !
)
- Conduct u ncL3 hops 3emmar, and Education Dnision and the trainmg prngrams a3 needed Es aluation and Assistance Dnuion.
- Identification monitering and Evaluation and Msistance i
- Penmhcalh publish information of liis lu,o n interest to nuclear pow er opera- esaluation of operations problems tions and traimng staff, and collection of related infor. Tae obiectne of thi3 Oniston mation from Licensee Esent will be to identify and promulgate Imtial emphasis will be on Reports :LER3' and other infor- the best nubar plant operating (;
idenuhing and making generath mation sources practices and R inform and assisi asadable the best of esisting traming utilines when adopung 3uch l
- Loordinate data bases and esents matenals and resources and on till- practices might impros e their oper. l umans wah relmed work j mg mdustn needs u here no material done bs EPRI and other organi- aumn baluahon teann w W consist or approaches s et exist ong of from four to six member 3 of the '
t he actn uies of the Trainin4 In3titute staff. mostly members of and Educalmn Dntuon ndi include
- studies and anal)ses includme this Dnnion. who are expenenced the following area; human factors studies sponsored in plant operations management studie3 in direct 3upport of oper- some team members may be from p becurit e Program attons. and compansons with other INPO dni3 ions and the team
. \tanagement responsibihty and Practices in other countne' may be assisted bs persons from the phtlosophy on saten and reli-
- Res iew proposed guidehnes u ith stafh of member utthties b aluatmos abihty the Industn Reuew structure. will prohahh be annual and w di be
t.dlow ed h3 consultation with and Emergency Preparedness a good planning les el and w dl proude a report to the utihty Diniston sufficient scope to permit detailed This Onision will estabhsh and admument at In3ntute funWons as baluations will be plant ori- maintain a manpower h3tmg of may be needed during the initial > ears ented and uill emer the follnw mg espects in sarious Gelda w ho could f U * " P"'""" "'
""* be asailable to a utility esperiencing
* \tanaaement and organization an emergency The listing will in-
- Plant operating practice, dude the indiudual s qualifications Related Committees and and abilities Listed personnel could Advisory Groups
- t rainnte and quahlications be disided into regional emergency
- lechnological support technical support teams he following committees and
* \lamtenance practices and .
W on h a m n n aduson gmup3 are imolsed in material comhtion. ""*""'O emer8ene) equiP- the deselopment of plans for the ment. w here it is located, and w hom Institute
- Human factors aspect 3 of d'3igns- to contact concerning availability.
arrangemenh and practice > This imentory wi!! include equip-
- Radiological controls ment in sendorf shops. plants T\ll Ad Hoc Nuclear Osersight under construction. plants in oper. Committee
- Emergenc3 preparedness J Hew! industn committer ation national laboratories. and
- Prowdure3 documenlahon and federal agencies INPO s sersices admimstration- will be asailable to all plants at all Flo3d Lewiw Chairman
.\hddle South Utilities Inc.
- In-houw audit and qualits assur- MW5 ance & practice 3 In view of the broad utdit3 par- Thomas Ayers ticipation in emergency prepamdness Commonwealth Edison Co it is estimated that an institute staff of Lee Everett
\ bout 3ts esaluation teams uill abom 15 may be adequate for this Philadelphia Electne Co he required for mdu3tnwide em-U'"*I""'
er am staf f of the Training and Edu- ulh w cation Un t3mn and the Cntena Administration Ulvision Duke Power Co Dewtopment and Analpa Onision Frank Linder This Disision will prmide the will regularly participate in opera- niernal support. tras el arrange- Dairyland Power Cooperatiw tion 3 es aluation3 to ensure their ments per3onnel management, and Jack Pfister iontmume close familiarity with financial functions of the institute It Salt Rher Project operating conditions and practice' Mi be responsible for the contrac- John Selbv In3htute esaluations uill include - tual relationships with outside reuca 3 of custmg utihts audib. Consumers Power Co organizations consultants and Encouraging self audits will be an . " ."*" research subcontractors. All financial aim of the Institute. Problem 3 in Portland General Electric to management. osen new and auditing deternuning the effectiseness and udl be conducted in this Division adaptability of sarious practtee3 w di including the fundmg arrangements he r eferred to the Cnteria Des elop-with the supporting utilities INPO Steering Committee ment and Analpa. Onision for study It is estimated that a staff of .Os erseeing establishment of INPO: t he Institute statf u di need to work 50 will be needed for these aethities reports to Os ersight Committee-closel) w ith other organization > 3uch as NRC and m3urance com- Presidents Office William Lee. Chairman panie3 interested in the resiew in addition to the generally Duke Power Co D ' "'"
- recognized functions of the Pre 3ident Jack Pfister. Vice Chairman Estimated time per mm. .tmum as chief esecuthe officer. this office Salt Riser Protect esaluation is two weeks and based w di menee the relationships with .
on the proiected number of utilit) the Advisory Couned and the IndustO sites and plants. 213 team weeks per ac[fifCa a d Electnc Co. Resiew structure. A staff of 5. includ-3 car will he required for annual ing the President. is estimated to be Sol Burstein mdustmude esaluanon. Resulting isc ns n Electric Power C,o needed to proside centralized plan-nunimal statf is 4 5 teams for an ning and assistance. Richard Eckert ideal schedule. \s3ume the Dnision Public Sersice Electne and Gas Co has 6 teana of 5 members each. requirme 30 team members bume. All estimated staffing iesels are Cr on at Edison Co. 20 Deusion 3taf f for management for based on present conceptions of the direct support. and for assi>tance minion of the Insutute and the Frank Staszesky tunctions in area 3 indicated by the proposed actiuties of other related Bost n Edison Co ! es atuauons. This leads to an esti- organizations. both utility and got ern- James Taylor I mated Division statf of 30 ment The totalof 200 is belies ed to be Dairyland Power Cooperative
AIF Polley Committee on iloy H. Dunharn John II. slac\lillan Follou up to T\ll Accident Tennessee Valles Authorits Habcock & Wilcos Co
-Hecommended establishment of INPO reports tother >ight Committee' Illehard M. Eckert H arren II. Ouen P& se@e Uectric and Gas Co Duke Power Co Fly rr,n Lee, Jr., Chairman H alt Fee Ilomano Salt atori Cunimonw ealth t.dison Co Northeast Ctdities n estmghouse Electne Corp Hubert T. Sealay. Secretary D. Clark Gibbs Charles H. Sandford Atonut industnal torum Middle South sen ices. Inc. Ilechtel Power Corp Shepard Harinoff John H. Gore, Jr. James 11. Stoudt Jersey Central Power A Light Co Baltimore Gas and Electnc Co Gilbert Conunonwealth Companie3 Vlnrent S. Iloy er Stese Houell fluble A. Thomas Philadelphia Electric Co Consumers Pow er Co. Southern Company senices. Inc.
A. Phillip firay Hilliam J. L Kennedy Donald E. Vandenburgh
- General Liectric Co Stone & Webster Engineering Co Vankee Atomic Electne Co Itobert Cockrell Leonard J. Koch Milliam H allace. III u astungton Public Pow er sy stem lilinois Pow er Co Ebasco 5ersices Inc.
Joseph it. Dietrich Hilliam Lindblad John Mard I ( omhu3 tion Entmeenng Inc. Portland General Electne Co sargent & Lunds. Inc l l l\PO Lalabliahrnent hhisory Group To aduse on the f unctions and vigamzation of INPO. Dr. Starr has assembled a group of esperienced people from unious aetn itie3 u nh strong national safety programs or w ho has e participated in maior educational actisine> related to safet) Dr. S. A. liernsen and Frank \l. Staszesk) Jr. , Ilechtel Pow er Corp Dr. Joseph J. Hulmer President Hudson Valley Junior College i
-lormerh m charge of the Nasal Nuclear Power School at the Knolls Atomic Power Laboratory Leo Dutiey LGaG Idaho Inc 4 tormerly with flett:3 Atomic Power Laboratory' Jerome Lederer fletired tlonnerly. Director of safety. N A5A. and head of the Flight safety f oundationi
- Dr. \llies Les erett .
Retired ,lormerly with General Electne Co 1 Dr. Ilussell O'Nelli
- Dean School of Engineering and Applied science. Cnisersity of Cahfornia at Los Angeles H alker itapp Pacinc Gas and Electric Co Fred P. Hiley i southern California Edison Co i
i l l t i I
9 1 ? E_
APPENDIX XVI l
Title:
Memo, H. R. Denton to Commissioners, Mark l Containment Potential Problem 1 l DISTRIBUTION:
'O mam> Centret riies CGrimes l
Glainas l
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'ACR$ - A. Bates SHanauer MD'OPMDUP FOR: Chairman Hendrie Comissioner Gilinsky Comissioner Xennedy Comissioner Bradford Comissioner Ahearne FR0"- Harold R. Denton, Director Office of Nuclear Reactor Regulation . y
p ,., ,,,s.g r T T _,, , _- THRU: gi Lee V. Gossick, Executive of rector for Operations
SUBJECT:
[ Mf Dr. I PO';TAIPPiT PSTCNTI AL PRPLEu gs On Tuesday, August 28,194 the staff identified a potential significant t concern relating to the condensation loads on the downcomers in Mark I containnents. The Mark I design is used in 22 operating plants. This concern was discussed in the Daily Highlights for August 10 an? August 31. The phenomena under corsideration is but one of a number of various loads that have been evaluated under the Mark I containment Long Tern Program. The staff is currently developing acceptance criteria to begin implementation of the Mark I long Term Program. This particular loading relates to the possible overstressing of the downcomer - vent header system which coul d cause steam to byoass the suopression pool water. Details of this phenonenon are presented in the enclosure. Over the past few days, we have had several discussions with representative: of 8WR licensees, the Mark I Owners Group, and the General Electric Corcar.y. We have con #f rmed that all of the Mark I plants have dcwncomers that are tied together, which, when combined with the Owners Group's preliminary evaluation, leads us to conclude that the potential concern is not as severe as originally believed. We are continuing to evaluate this concern and will meet with renresentatives of the Mark I (Nners on Wednesday, Se; temher 5,1079, to reviev: their further assessment and detemine what action, if any, will be necessary. We will continue to keep you informed of our progress, cc: SECY 0@ul 54"! D ET OPE M.R.9M*0" LEossic's
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ENCLOSUP.E POTENTIAL CONCERN RELATING TO CONDENSATION LOADS ON 00WNCOMERS IN MMX I CONTA!NMENTS
Background:
Following a postulated loss-of-coolant accident (LOCA) in a Park I containment, steam will mix with the air in the drywell and flow through the vent system into the suppression pool where the steam will condense. The vent system consists of a main vent, a vent header and downcomers extending into the suppression pool, as shown in Figure 1. As the flow through the vent syst(.m continues the ' relative concentration of air in the vent flow will decrease as the inventory of air in the drywell is depleated. When the air concentration is low enough, " condensation oscillation" loads will occur on the suppression chamber walls and the vent system downcomers due to the rapid formation and collapse of the steam bubbles in the suppression pool. In order to quantify the condensation loads for the Mark I Containment Long Term Program, the Mark I Owners Group performed full-scale, prototypical tests. The Full-Scale Test Facility (FSTF) is a 221/2 sector of a Mark I
~
suporession chamber, containing four pairs of downcomers. The facility was instrumented to provide thermal-hydraulic and structural response data. Ten tests were conducted to cover small, intermediate and large break O accidents with both steam and liquid blowdowns. High amplitude condensation oscillations occur for large-break LOCAs and are characterized by a random motion of the steam-water interface outside the end of the downcomer. The rapid collapse of the irregular steam bubble in the pool causes lateral loading on the downcomers. The magnitudes and directions of the lateral loads have been inferred from strain measurements in the downcomers and vent header in FSTF. These measurements were taken for both
" tied" and " untied" downcomers. " Tied" downcomers have a strap which connecti, the ends of a pair of downcomers together.
Sta f f, Concern : The driving frequency observed in the test data for condensation oscillations was in the range of 4 - 8 h:, primarily at approximately 5.5 hz. The equivalent stresses (from measured strains) for " untied" downcomers during condensation oscillations were above the ASME allowable yield stress but below the material load bearing capacity. The staff had estimated that the natural frequency of the FSTF downcomer vent header system could be about 6.9 bz for
" untied" downcomers. Our concern was that in the event that the condensation i oscillation driving force was to occur at the natural frequency of the downcomer - vent header system, a dynamic amplication of the load would result.
The staff has estimated that, for the " untied" downcomers, a dynamic amplification of as much as a factor of six greater than the loads measured in FSTF could O l l
- -O i i i occur at resonant conditions. Such an increase in the inad would result in
}' fatigue failures of the downcomers. Since our previous discussions, the Mark I Owners Group has identificd data showing that the natural frequency of the i downcomer system of one operating plant was about 19 - 20 hz, hence well away ) from the driving force frequency. In addition, the Mark I Owners Group has reviewed the downcomer designs in all plant; and they have found that
. all plants have downcorer " ties" and, by comparisons of the relative downcomer-l vent header stiffnesses, they conclude that the natural frequency should remain well above the driving frequency.
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O SYSTEM DESIGN REQUIREfENTS 1 LESSONS LEARNED TASK FORCE CURRENTLY CONSIDERING THE i MAJOR AREAS:
- SAFETY SYSTEM Ut! AVAILABILITY
- CLASSIFICATION OF SYSTEMS IMPORTANT TO SAFETY
- OPERATOR INTERACTION (AS RELATED TO SYSTEM DESIGN REQUIREMENTS)
- POST ACCIDENT O
e 4 O
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O SAFETY SYSTEM UNAVAILABILITY
- 1) RECOMMEND APPLICATION OF FAULT TREE ANALYSIS METHODS TO
- FLUID SYSTEMS, I&C AND ELECTRIC POWER SYSTEMS IN SAFETY ANALYSIS REPORTS AND STAFF REVIEWS FOR SELECTED SYSTEMS.
- 2) RECOMMEND COMBINING THE FAULT TREE METHODS WITH THE sit!GL FAILURE CRITERIA TO DETEPJilNE SYSTEM UNAVAILADILITY OR RELA-TIVE RELIABILITY OF SYSTEMS USING FIXED INPUT FAILURE OR SUC-CESS ASSUMPTIONS. THESE EVALUATIONS WOULD BE USED TO IMPROVE THE CURRENT DETERMINISTIC LICENSING CRITERIA.
- 3) RECOMMEMD FACTORING INTO THE FAULT TREE METHODS OPERATOR ACTION, INACTION OR ERROR IN ADDITION TO THE SINGLE FAILURE,
- 4) RECOMMEND THAT THE REVIEW 0F PLANT OPERATING PROCEDURES BE O COUPLED TO THE SAFETY ANALYSIS TO ENSURE OPERATOR INTERAC-TIONS REQUIRED BY THE PROCEDURES ARE CONSISTENT WITH ASSUMP TIONS USED IN THE AtlALYSIS. THIS WOULD BE AN APPLICANTS RESPollSIBILITY WITH THE STAFF DOINGA'! AUDIT REVIEW 0F PRO- l CEDURE STliARIES CTSIS WOULD It'CLUDE TESTI?!G a MAINTENANCE P
- 5) RECOMMEND THAT WITH THE INCORPORATION OF THE AB0VE RECOMMEN TIONS INTO THE SAFETY ANALYSIS AND REC'JIRED STAFF REVIEW, THE CURRENTSINGLEFpILURECRITERIACONTINUETOBEUSED, j
i - 0 l
O CLASSIFICATION OF SYSTEMS IMPORTANT TO SAFETY
- 1) RECOMMEND THAT CURRENT REVIEW METHODS EMPLOY RELATIVE RELIABILITY ANALYSIS METHODS TO DETERMI!E THE IMPORTAfiCE OF SYSTEMS TO EACH OTHER,
- 2) RECOMMEND USING THESE METHODS TO DEVELOP A CLASSIFICATION OF SYSTEMS THAT ARE IMPORTANT TO SAFE PLANT OPERATION, BUT MAY NOT BE REQUIRED TO MEET ALL " SAFETY GRADE" CRITERIA,
- 3) RECOMMEMD THE STAFF INITIATE AN EWLl' A TIO" T TFE -
ADVISABILITY OF UPGRADING ADDITI0f!AL FLUID SYSTEMS, SUCH AS, THE ELECTROMATIC RELIEF SYSTEM, WASTE GAS SYSTEM, LETDOWN SYSTEM AND SUPPORTING SYSTEMS TO THE MAKEUP OR VOLUME CONTROL (g SYSTEMS.
- 4) RECOMMEND REQUIRING FAILURE MODE At!D EFFECTS ANALYSES BE C0f:-
DUCTED ON "NON-SAFETY" SYSTEMS TO IDENTIFY POTENTIAL INTER-ACTIONS WITH " SAFETY" SYSTEMS, RECOMMEND INCORPORATING THE REVIEW 0F EMERGENCY PROCEDURES, WITH RELATION TO SYSTEMt INTO THE LICENSING REVIEW. RECOMMEND THAT THE ENVIRONMENTAL QUALIFICATION OF SYSTEMS AND COMPONENTS IMPORTANT TO SAFETY BE INCLUDED IN THE LICENSING PROCESS. THIS WOULD INCLUDE PlP:t!G SYSTEMS, PUMPS, VALVES, BEARINGS, SEALS, ETC. TO WITHS'At!D ACCIDENT CONDITIONS, (IN-CLUDES SYSTEMS BOTH INSIDE AND OUTSIDE CONTAINMENT).
- 7) PECMEND A SEPAPATE SECTION OF THE SAR BE DEVEl.0FJ) TO SET FORTH TtE EVALUATION T THE INTEPACTION BEMBl " SAFETY" AND "tGF-SA=T(" SYSTEFE, E.G,' CONTPOL SYST95, C00 LITE SYSTEFS, WATER SlFPLY SYST9S.
d 7 l
O OPERATOR IKfERACTION (AS RELATED TO SYSTEM DESIGN REQUIREMEilTS)
- 1) RECOMMEND THAT AN EXPANSION OF AND EXPEDITIOUS EFFORT BE MADE BY NRC TO ESTABLISH A DATA BASE TO ASSESS OPERATOR PERFORMANCE IN ASSIMILATING, COLLATIllG AND REACTING TO THE COMPLEX INFORMATION PRESENTED BY THE REACTOR INSTRU-MENTATION, CONTROL AND DIAGNOSTIC SYSTEM UNDER BOTH NORMAL AND OFFNORMAL CONDITIONS. THIS COULD BE ACCOMPLISHED USING SIMULATORS.
- 2) RECOMMEND THAT, USING THE DATA BASE, A RELIABLE PROBABILIS-TIC ASSESSMENT OF PARTICULAR OPERATOR IETERACTIO!!S DURING 0FFNORMAL EVENTS BE DEVELOPED.
I
- 3) RECOMMEND THAT THE TECHNIQUES DEVELOPED IN (2) ABOVE BE USED AND COMBINED WITH THE FAULT TREE ANALYSIS TO ASCER-TAIN THE CONSEQUENCES OF OPERATOR INTERACTION DURING OFF O NORMAL EVENTS. THESE CONSEQUENCES COULD RESULT I'! CHAT!GES TO SYSTEM DESIGt! REQUIREMENTS.
- 4) RECOM'iEND THAT BASED ON (1) AND (3) ABOVE, REALISTIC TIMES FOR RELIABLE OPERATOR INTERACTION BE ESTABLISHED.
THIS WOULD PROV!:~. A TIME BASED LEVEL FOR AUT0ATION OF SYSTEM DESIGNS. l
/
e o. I
O POST ACCIDENT DESIGN CONSIDERATIONS
- 1) RECOMMEND THAT ITEMS OF EQUIPMENT REQUIRED TO FUNCTION IN THE POST ACCIDEt!T PERIOD SHOULD BE DESIGNED TO SURVIVE THE ACCIDENT AND ALSO TO BE CAPABLE OF PERFORMING ITS DESIGN FUNCTION AFTER EXPOSURE TO THE ACCIDENT ENVIRONMENT.
- 2) RECOMMEND THAT PLANTS MA.KE DESIGN PROVISIONS FOR POST-ACCIDENT INSTALLATION OF EQUIPMENT AND PROCESS SYSTEMS THAT MAY BE NEEDED FOR ACCIDENT RECOVERY, PREPLANNING FOR PLACE-MENT LOCATIONS AND EQUIPMEllT AVAILABILITY IS ALSO REQUIRED.
- 3) OTHER POST-ACCIDENT FACTORS UNDER CONSIDERATION:
A. ALL IDENTIFIED ACCIDENT RELEASE PATHS FOR AIRBORNE / GASE0US EFFLUENTS SHOULD BE TREATED BY FILTRATION O AND ADSORPTION TO MINIMIZE OFFSITE RADIATION DOSES. B. RESEARCH AND DEVELOPMEIR ARE NEEDED FOR RADI0ACTIVTY MONITORING SYSTEMS TO IDEf!TIFY AND ASSESS RELEASES FROM PLANTS. SPECIFIC AREAS FOR STUDY INCLUDE: ACCIDENT RELEASES OF GASEOUS RADIOI0DIt!ES A "' PARTICULATES: MONI-TORING OF IN-PLANT LIQUID PROCESS s , STEMS, SUCH AS LETDOWil, UNDER ACCIDEflT C0f1DITIONS, x l l l l
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- APPENDIX XVIII [M
Title:
Zion Station LER Study , g ( . Q \ ,1 PLANT ARPANGEMENTS SUSCCIO!ITI'EE NASHINGICN, D. C. CCTCOER 25, 1978 ZICN SnTICN SYSTEMS IhTEFAC"'ICN STCDY . EIGRLIGIIIS - Ccemnwealth Edison Ccmpany presented a sumary of the Sys-tems Interaction Study performed by Fluor Pioneer Inc. dated June 16, 1978. 21s study was made in response to the Cemittee's recc=endations centai ed in its reports dated June 9, 1976 and June 17, 1977. The study was based on a review of Licensee Event Reports (LER's) frem all cperating reactors
- - through 1977 ( approximately 9,000 reports), tese reports involving sys-tems interactions and applicable to the Zicn design were studied in detail. l I
A copy of the Executive Sumary (Section I) and the complete Conclusions
, (Section V) are attached.
- The Subecmittee seemed to agree that the licensee had acecmplished what had been agreed upon. The subecmittee arrived at no conclusions and sug-gested that the ACES might want to hear an abbreviated version of this O presentation at a future meeting (date not specified) . ',
V l l . l E .
l
- y. I SECTION I EXECUTIVE
SUMMARY
1 j History I In a June 17, 1977 letter, the United States Nuclear Regulatory Comission's Advisory Comittee on Reactor Safeguards (ACRS) recommended that Comonwealth
)
Edison conduct a study of possible systems interaction related to the Zion ) Nuclear power plant's shutdown heat rernaval capability. The ACRS also re- l ' ferenced additional guidance contained in their letter of November 8,1974 After detailed review of these letters, Comnenwealth Edison personnel dis-cussed several possible approaches to a systems interaction study with a variety of consultants and with the Nuclear Regulatory Cc=ission staff, i Utili:ing this as background material, the plan upon which this report is based was devised and subsequently discussed witn an ACRS subcomittee on , i November 10, 1977. Methodolocy , The plan involved a study which was divided into three phases. Phase onc consisted of a review of over 9,000 Licensee Event Reports (LERs) which have been generated in the operation of U.S. comercial nuclear power plants be-tween 1969 and 1977. i The LERs were used to identify events which have occurred at operating pcwer O plants that involve systems interaction modes which had a potential for re-ducing the effectiveness of shutdown cooling systems under non-accident condi tions. The review covered not only 4 loop PWRs but all PWR, SWR, and gas cooled reactor LERs. The second phase of the study, which was conducted by Fluor Pioneer Inc.,
~
involved detailed analysis and investigations of each event to determine how and why the event occurred and its effect on the originating plant. Once the event was understood, an analysis of the Zion plant was implemented to determine whether an analogous type of event could o'ccur there. The analysis of the Zion plant was not limited to events identical to those listed in the LERs. The scope of the investigation was breadened to include review for similar interactions in other parts'of the plant. As a first step, this analysis involved a review of pertinent flow diagrams, ' piping layout drawings, equipment layout drawings, schematic diagrams, in-strumentation diagrams, electrical installation drawings, and equipment and contractor specifications. The second step consisted of a review of current and past operations history, procedures, data, legs, and special reports.
- The third step involved physical surveys of the Zion plant. Frequent plant visits were required for an adequate assessment of the physical relationship between systems' and thus, a determination of the field related effects. The
. fourth and final investigatory step involved detailed discussions with Zion e
. plant operating, maintenance, management, and engineering personnel. ,
- O. ,
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Oi
- e. A program should be developed to survey electrical boxes contain-ing open terminals which are used in safety or shutdown systems, and which are located in tne Auxiliary Building, Steam Generator Safety Valve Rooms, pipe tunnels, and crib house to determine if they could be subject to entry of water. For those boxes in this category, the existence (or lack) of box drain holes should be de-termined by inspection. If drain holes are not found in these boxes, they shculd be added, or some other technique should be used to pre-Yent potential shorting of the terminals by water accumulation (FPI !9943-1 and #g943-2).
- 3. Although the study did determine that some systens interaction could occur at the Zion plant, these occurrences would not significantly degrade the safety and shutdown systems in the plant and thus would have little or no effect on the public health and safety.
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- l SECTION V CONCl.USIONS Several conclusions were reached as a result of this study.
These are discussed below:
- 1. For the Zion plant, generic studies requested by the NRC and the implementation of their conclusions and recomendations involving such items as fibe protection, pipe break, low temperature primary system overpressure, etc. , have resulted in modifications which substantially reduce the possibility of the occurrence of a majority of the events studied.
- 2. The following investigations and/or plant mcdifications are recomended by this study.
.a. Following an evaluation of the benefits of J-tubes, which were -
installed in one of the steam generators on Unit #2, a determin-ation should be made as to the need for modification of the steam generators (FPI#9920-3).
. b. The containment spray pumo diesel fuel oil tank vent and fill lines' susceptibility to being blocked and covered after a significant gL snowfall should be investigated and/or corrected (FPI #9923-5),
(
- c. An investigation should be conducted to determine if ic'e can form on the Diesel Generator Room air inlet dampers to an extent that could be detrimental to the operation of the damper (FPI #9924-6).
- d. Before initiation of any steam generator maintenance that has the potential to affect the pressure retaining capability of the steam generator tubes, appropriate methods should be included in
,the procedures to check the integrity of the tubes prior to re-turning the steam generator to operation (FPI #9951-17).
- e. A program should be developed to survey electrical boxes containing open terminals which are used ic safety or shutdown systems, and Which are located in the Auxiliary Building, Safety Valve Rooms,, '
pipe tunnels, and crib house to detennine if they could be subject to entry of water. For those boxes in this' category, the existence (or lack) of box drain holes should be determined by inspection. If drain holes are not found in these boxes, they should be added, or some other technique should be used to prevent potential shorting of the tenninals by water accumulation (FPI #9943-1 and #9943-2). 0
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. . . . t
- 3. The large number of indicators and annunciators at the Zion Station serve effectively to infonn the operator of the presence 4
(dT of abnormal plant conditions including those associated with systems interaction. events.
- 4. The approach used in this study was found to be a satisfactory method for investigating systems interaction events. The method was successful because the key project staff members were senior personnel who had extensive experience.
- 5. Although the study did determine that some systems interaction could occur at the Zion plant, these occurrences would not significantly degrade the safety and shutdown systems in the plant.
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APPENDIX XIX Ti tle: Electric Grid Stability ELSERT P. EPLER NLC.tAA SYS?tMS CONSULTANT
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l l n u r e.asten [_} b/ Ny0 LEAR $v$*fst$ CCN$ULTANT l 712 si0trCA AvtNUi On RICGt, 135El37833 4 3 3.C9 94 Juin 23 197.9 Flerno ho lhe Si le - The biolus in new o<ders for nuclea/ pianis l5 now in ils lov/dh year. The prospeel cl a n impendiny bv3iness recessibn does nolhinc io oneovvoq e
}he placinc; in twelve lal ordeJo be cielO o h ideen shoda g e el generading capacil .yeon, Jhus insviinq lhal olililies will be opernding wik minimol reneve O '*Po'$ '"'u N'"1 "I 9']d insla biltl y n nd blo cic e v is.
Jocn Jhe ins]alleni nucleo >< copo cil will become on ilnpo</n nJ frac 4fon ei kJaI sysdem copor)/c nnd Jhe resuI4ing in/eraclions wiii becorne anpulons, auc A inle/ac b6ru w1ll inc /vc)e Jhe elfec/ cl nucleo / plands en c7 rid slobilily as well as 4he effec l cl 7 nd insin hihI on nuc len / plo n) a rca l lo bllil and sn /edy . Jyslom .sdobilily I s sdron9 q/ dependenl on
- spinn in c, reserve, t.e., planh in operadan had Wgn-i!n leaded,an<J inpn ble ad picl<ing up land insin nd1 9 an demand. A.s reserve capoet4 y dlrnInisher,3 o c.. it I J h e ,
spinning re. serve whhh alI of covise resoiJ ;<. a np Oo,) Q sin brli4 unnier pen k lond tenr}ilions wh en n. smo)] disluibame could resull in a recionn) blacle oud-
z Q' Il is well known lhal hej dro elecin*c un ris ennana e grid sla bil:l y A hydro unil con he brev;hl up 4 .hti pc ver almos4 insdanlly , and shov/d an2)h e/ yeneia lih4 unil be loyced olf-line, wov/d he able k pah up Jcod w;Ih mirurnn/ dislu/honce lo dhe stjslern. Fur!her, 4he hydro un;lis sh le p and rugg ed, hfile less ihon a. ditec] rircke of h' 7frdnh,y <cuid in use lhe unil k shd douvr) . 7i> do 9 ./he yh r)<o co ndei h u] Eon h a s h een d wn b.7. Jcss t l a nd nucleo / tapa ciI , Th e TVA so ce nw cl h elic g will be 0Ye/shaniceved h Jhe
't o, c c o en w on-bha and bv;Idihq, hn15 ol u hdh wil! h e nec leo r; 771e 3l0 bil(tinc, h yrito IT lheYe[cte beinq repiot enf b nut lea /
inn km 9 ii in,pcr/and Jhn ) Jhe e flec / cl nec leo / planJs O en sy5len, sin b;I:l h e e/o lun de,J., 5 t%Ieo / plan /s a re ,orclecded b nu,ne/cus sylern s whnh will shu4 Jhe plan l down on cau w enie of enn!Ivnc-lil>ns inlevnn) lo Jhe pinnd was n vesvil o f e:tleinal dis lv/ h a nt e.s The c hivpl opening o) a <> n u;} hien),e/ , ca us ent h a hundetsdctrr) could leic e n. Swa /c shv4 dcwn and reinnir1 ol[- lin e bo / a i2 hov/ ves!n v] ' reqirnein, /d is concoienhle Jh.o ] a muill-uni) pins) ecvid inbhi) n 1n r'n e / h e s},v<l dou. n, w hd h in l-urn u oviol o yO/ /ca d cl ho / plon).s ,. Ihv s (n us in q clhor ci/cul-l bi en )< e , s k Mp, wdh Jlle resull lha) nil nuclen/ n;ls no) ho vihq loc % /cor) te {ec-l ion capoh:lil , wovid i> ? kicent off-line 0 o nti bet owe vna /niin ble k' JP'em I hou 5 -
3 O Tnh suggesIs ,iha} c5 reserve cwpacily dilrunnhes, lhe nuc/eo/ ui.nl could hate a darnhianl e//ocI en syslem sda bili)y. we haie 3 een,in kva isn in mes,1 ha J ct .smo// disluyhance cov/d rn v.se n. Now Ycsk /ecj/eno l binc h. cul, and n;ain iha-l a -scrani o-l Tud e9 Peind cov/d blo c ). ot.-l 19hdni s' , we shovini 4 here (c /e be prepa <ed k lo e e p<chlenu. 7~here mo v, he nn add ll ten al nod more S evIcus consecj vent e od on in-le/a cl ion bc,lvveen -lhe plan l n nd Jhe udti;le, q vid . Al 17:llsdene o nac ic n v un.4 scra>r>nient or,d as a resvi l Ilie vuline, e cd Ihe endie regicn d<cpped. The Q veldac, e reduc}icn was vol svHiu, n4 le cnusc .Ine emergona; desel g en ern kn le adovi, h v l J h e terivi e,4 vcIla q e a os insufOcienl Ju poi) .n l h e n . r. n > c.Je < con-in c lavs. A.s a resull lhe con-lr=1 luse s hlew 'l<cn r oreiland nrrd <>ll hough a. c powe / wa s a rolla hje,. mc krs cculc) nc) yvn lc remo ye residun i hen], Thl1 pin; ed .Ihe pinn) in n- hiyhI ruinem ble tond l1bi1, Il 'i> ' lh erelcte f oss ) h ie I hn ) (L s/<17 e lren c l-c / c c v1,} , a.n n iren dq r}p.; rnoled scro m nnd Jnoseh 3 uvo /Jert C o 11 8 ,' I u n 0 l b h e V li ( l ((Ed, ov k -lh0 r?1vl].bho so/ern I nec len / un.4s could be sh) down and be
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H The condilibns len </i Ac, Jo a shorlnge ol 7enefadinej capaci] are tuell kno<ura nnd beyond on3 possihi-l$ of ton Irol. The t'eSuIking inIeracl ions bo Aw e en J h e nuc lea v' pin n ) n nd Jh e grid o <c h o w eve >/ su s j ec l ko rn odili<. a lib r, a nni should 4here,lo/e he co te Svil eynrn:rier] So m e exornple,s Jcilow .
- o. The BwR R a il D i o p , }nanleqciale pro]eclidn sy 54een response has led Je caudibus opern k 'an and cntessivel lano; ,slad up J inies, Infora cirbm heLan hv Indus)<1 a ~l 1he Reg uIn lu has y} ncw<nplished a reso),.Mn c(
lhe problen1, 9 6 Residun i Nea l Auncro). Dependenc e en gene <nt pu< pose lp anI syslems, as ofpased lo a dedita leni sy d ein det vendval hen) remo/a), hns made lhe plan) vulnein hie do r)egynded g rid cond,16ns ,luhn] is inore siynifican), x rnallunc libn al n Viln} serw'ce, u hich. Is essenlin [ Ic vesiduol henI venwcal, enn cnus e plan l Shv0down, a.hich in lwn, en r1 c a us e n. defranienl grid tondshion. C. As a res u ll c) TM I, n )) 64 w pinnls nee now recjvi/cd b NR C. k sc en ;n o n loss al lerdwn le / c/ Ju<h oh e l b rif , frev'ic u sI _61 w p la nbs we re cnfo ble o $ reyn nin /a o n. hol sInnd b on oc 'vueni e el J has e Jransien Is .Tht, na %hly vnihie nec,via laq alein chen a d ve<sel Q .aIlecliny sysdem sin blN . l l
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- >! salefu./ A ccmnion!,. encounIeren s;-lundici > 1 s innaeyvo h fecle:]icit response, w ich h coingens n ~ eI bs; M <]g m v Uc.'.1 n, m>M a o ,,,,1 cc, a e> vn In, e se i pei,> is , res u th n.3 , ,, ,,, cc/
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i l Ray 6/27/79 l O Implications Regardina Safety There were no Engineered Safety equipment malfunctions or danage due to I undervoltage. However, had the Emergency Core Cooling System or the containment Spray System been needed, it would not have performed as designed because of the premature RAS valve actuation. 4 Corrective Action The protective relays on the Bus Tie Auto-transformer were reset to permit supply to the auxiliaries of both Unit 1 and Unit 2. The settings of the inverter switches for DC trip voltage and time delay were corrected. Startup Transformer 42 is capable of supplying the emergency auxiliary loads O on one unit while simultaneously supplying the auxiliary loads necessary to orderly shutdown the other unit. The transformer can supply without damage the full auxiliary loads of both units long enough to parmit orderly shedd-ing of unneessary loads. This will be the mode of operation. To minimize the probability of recurrence of similar events in the future, a review of the industry experience with degradation in auxiliary power
' supply should be undertaken to establish a basis for review of typical current designs and the adequacy of design reviews.
40
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I Ray 6/27/79 O The Bus Tie Auto-transformer was now supplying the auxiliaries of both Unit 1 and Unit 2 thru their respective startup transformers. The Bus Tie auto-transformer has ample capacity to supply the auxiliaries of both units, but due to an engineering error the protective relays on this trans-former were still set for the operation of only one unit. Under the combined load of the auxiliaries for both units, the relays tripped and interrupted the supply to both startup transformers (il and 43) . The switching equipment automatically transferred the auxiliary loads of both units to the backup 42 Startup Transformer. However, the capacity of this , transformer is not adequate to supply the full auxiliary leads of both units. This was not realized prior to this incident and represents an engineering error that was missed in the design review. M a result of the overload on the l O stertee r<emere<me<, e8e vette9e euvelted te e8e etetice distriection evetem was degraded. During <nost of the incident, operating personnel for both units failed to recognize the degraded voltage condition, which could have been corrected by quickly shedding unessential loads. 21s constitutes an operating error. On Unit 2 undervoltage relays, installation of which the NRC Staff had required
, subsequent to a degraded voltage event in 1976 at Millstone 2, tripped the supply to the Engineered Safety Feature busses, as designed. (This fix had -
not yet been installed on Unit 1). Power supply to the .120 volt AC instrument busses should have been maintained from the D.C. busses (supplied from batteries) via inverters. However, during the event the inverters transferred to supply from alternate ESF busses which were isolated when the undervoltage relays operated. Loss of power to the instrument busses caused actuation of all Unit 2 l
1 Ray 6/27/79 O Engineered Safety Features, as designed, including startup of the emergency diesel generators, hhen power supply became available frem the diesels, the ESF equip:ent began to operate. l 1 Contaircent spray was initiated, dumping approximately 8000 gallons from the l l refueling water tank to contairment by way of the spray nozzles. Also the Recirculation Actuation System (RAS) cycled valves which momentarily opened ; a path to drain an additional 40,000 gallons of borated water from the tank to the contaircent sump. Subsequent investigation of the inverter operation revealed the fact that a combination of bad set points for transfer of the inverters to AC supply on low D.C. voltage (the set point was too high) and the time delay before such transfer (the time delay was too low) would explain maloperation of several of the inverters. It was established that the low time delay settings were missed in the preoperational tests. It is believed that several other inverters may have been lef t in the AC supply mode af ter maintenance during the previous month. Isolation of the ESF busses on Unit 2 and load shedding on #2 Startup transformer
, restored system voltage and permitted normal post trip recovery of Unit 1. The Unit 1 diesels were available for service, but were not required to start. ,
Frequency of Occurrence This incident represents an unusual coincidence of circumstances which would not be expected to be repeated. However industry experience includes many incidents of abnormal operating conditions and reactor scrams developing as the result of loss or degradation of of fsite power supply. O
0 h L ts E o b x X X
APPENDIX XX Titie: Current Status of p ATWS Problem
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O I M y,$* * * %q',a UNITED STATES ]I g
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..... !??? JJ!. 5 h 1159 Cr. Max Carben, Chairman u e..E~..e.. r.'".."."y.
'dEMCRANCUM FCR:
Advisory Cc=ittee en Reacter Safeguards > ^i;.,,e v p'.,R 9M e."* FRCM: S. H. Hanauer, Direc:cr Unresolved Safety issues Fregram, NRR SU5 JECT: STATUS AND FRCFCSED PLAN FCR RESCLUTICN CF " ANTICIPATED TRANS!!NTS WITHCUT SCRAM" SAFETY ISSUE ne 227tn The NRC staff las: met with the ACRS At tneteconclusien discuss theofATWS issue a:the that meeting ACES mes:ing ('darch S-10,1979) . C:mittee identified several matters to be discussec dy the staff at the As a resui: cf :ne need en the succeecing car: er 223:h meeting (A;ril, 1979).of :c:n the ACES anc the NRC staff Island 2 ac:icent, these discussicns have ne taken place.
- In ceder te facilitate the resum::icn of the dialogue on ATWS cetween :ne ACRS and the NRC staff, we are attaching to :his letter a discussion of n :
f' those ma?.ters :nat the Ccemittee has raised relative to BWR's. We are including any discussien, however, en the :WR rela:ac items until, as noted telcw, we have hac a tet.ter c;ccrtuni:y Oc assess :ne im:act of the Three Mile Island 2 accident en tre evaluation and resciutien of :ne ATWS issae for such :lants. As you are aware, as a art ci the NER Interim Crgani:ation, a technical ! review greue has been fer ed :: werk on a "dedicatec" basis en each of tne Unresc1ved Safety Issues recceted to C:ngres's in :ne 1973 NRC Annual Re:cr:. ATUS is, cf c:urse, ene cf tnese issues. As a result of the heavy expenditure of resources on Three Mile Island rela:ed activities, essentially no staff effort, and we think greatly reduced vender Fce effort, has been acclied to the ATWS issue for the last 3 enehs on the or sc. Sciling Water Reacters, the effect of Three Mile Island 2 General Electric . . evaluation of ATWS events is believed to be minimal. provided a submittal in May with answers to scme of the "early verifica.icn" cuestions whien were ransmitted to all of the venders by letter of February 15, 1979 from :. J, Mattsen. G.I. has c:mitted :: ::rovicing :ne balance of 7 in two additienal su:.-ittals, the early verifica:ica res;cnses for 3W,s one in early July anc :ne ctner in the Fall Of 1979. We intenc :c c meie:S the review of these submittais en an expedited basis with the ocjective cf , arriving a: a propcsed rule that will contain general NRC It recuirements f:r is cur inten: clan: =ccifications f:r varicus classes of BWR plants. that for the SWR's, cur revised schecule will be made as ax:editicus as e
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r l JUL 0 Or. Max Carbon 2-possible leading to a pec csal to the Ccemission f:r rulemaking in the next few months. For Pressurized Water Reacters, the staff's preliminary assessment is that the Three Mile Island '2 accident scenaric raises new questions with regard the a:precriateness and adecuacy of the pre;csed resolutien of :he ATWS issue for such plants. The reviewers assigned to ATWS witnin the '.'nresclvec Safety Issues Program are continuing their assessment of the imolications of This assessment includes Three Mile Islanc en cur evaluatien of ATWS. discussing the accident's possible implicatiens for ATWS with the NRR "Lesscns tearned" Task Group in July. l In addition, we ;ian Oc meet with re:resentatives of the ?WR vend 0rs and , u:ility re;:resentatives en July 25 te exchange views en :ne im:act cf Three l Mile Island en ATWS. Af:er this meeting, we ex ect te ceveic; a revisec l senecule fer ::::letinc cur evaluatien cf ATWS fer PWRs. I We will, cf c:urse, trevice the :CES with a complete revised sche ule for AT'.;S as seen as ene is available. ~ i kh :. s v..c. e .
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l S. H. Hanauer, Director )
' Unresclved Safety Issues ?regram ,
Office of Nuclear Rea::cr Regula:icn
Enclosure:
" AT'.:S - Outstancine .
ACRS Questions Fellowing ; 227 h esting - A :licable
- 5WE's Cnly" cc w/at*:
A. Thacani R. MattsOn ' l T. Su X. Kniel L. Ruth T. Speis S. Hosferd P. Check . K. Fare:ewski 0. Eisenhut - M. Srinivasan 3. Grimes H. Vander Mclen ?. "cenan M. Tckar J. Knight W. Regan R. 3csnak S. Coplan D. Muller F. Akstulewic: F. Schroeder F. Cherny J. Norberg l M. Ayccck {. jakel . T. Novak ACRS (21) H-1015 R. Tedesco FOR R. Cenise l
O \ ATWS - Outstandin; ACFS Questions Felicwinc 227th Meeting - 4:lica: 1e :: 5,,A>s aniy Question: a) What are the maximum temcerature and pressure transients one ::rus Or pressure su :ression ;ccis of GE centainments can ac:ect without eucture? What are the consecuences of tcrus or su pressicn Occi failure? Rescense: Our inter:retatien cf the firs : art cf this cuestien is: Our4ng the extended Saf ety/ Relief 'laive blewc wn resulting fecm an ATWS event, what is the maximum l c mbinatien of cci bulk tem:erature and escillating :cci :ressure that the terus or su:gressicn Occis cf G.E. c:ntainments could acce: withcut ru: ure? wnich failure woul:
- The 5:ecific ccc:inatien cf ressure anc tem:erature a l
be expected to c cur wculd be 01 ant s:ecific and has not been determined. We to knew, n: wever,fr ; testing that has been perfcr ed as a cart of the Mark I, II, and III Contain: ent Testing Fr: grams, that 0:ntinucus safety / relief valve steam discharge can result in a :hencrenen that has been termed
" steam c:ncensation instability". Ccndensatien oscillatiens occur as the steam is being discharged tc tne pool. The am:litudes of these vibrations are u cwever, c ntinued blowdown into relatively small at Icw pec1 tem:eratures, the occi, as would result fr:m an ATWS event, increases the :cci tem erature. '
If the tem erature rises far enough, a "thresheid temepra.ture" can be reachec f:r some ty es of cischarge devices, When the ; col tem:erature is abcve The resulting vibration and the threshold, steam c:ncensation is unstable. asscciated f0rces are very severe. In a foreign react:r, an event invciving ;r:10nged bl:wdown led to loss Of
.n . - . v . . .c w , , ..: . . w . n, . - . .
. . . g ? .. . . - . . - . .
O 2 i In testing, the integrity of the suopressi0n pool via this mechanism. condensation 1 cad has been c: served to be as great as ten times the normal os:i11ation Icad it pec) temperatures abeve the thresh:1d. Current p"a:tice fer operating BWE's is te restrict the allewa: Te c;erating tem;erature cf the su;;ressi0n 9001 via the Technical 5:e:ificatiens such tnat the thresneld tem:erature would not be rea:ned in a L:ss cf Ceclant A ::i 2rt. Ty:ical T.eennical See:ificati:n tem;eratures used for an ccerating Mark I plant are shcan in Figure 1. 2 Fer ATai events, G.E. has prc;0 sed higner maximum al10watie tem;erature limits thar. these u:cr. wnich the current Te:hnical Spe:ificatien valwes are (~ t based. 7ne staff agrees that hi:her tem:erature limits are a::re:riate for plants >'en cuencher type discharge devices are used. Present indications
# It is cur are tri; tem:eratares u: te a;;roximately 20C F are a::e: table.
un:t-standin: that til B'a' Clants, botn th0se i" c:erati0n and these under . c:rstructi:n. .ill be insta11in: cuencher dischar:e devices. An industry program has already generated a great deal of experimental and the:retical information en quenching phen 0mena. An intensive, detailed examinatien of all relevant data .;,,. related to sup;ressi:n ', - - Deci tem:erature limits is being conducted by the staff Is ca'rt of TAP.A29. All nree peci configurati:ns and varicus quencher devices are being reviewed. The ?cci temperature criteria to be established for ATW5 will be based en . thi. werk. Completier cf criteria is scheduled for Oe: ember, 1979, for T I and Mark II containments, and a few months later fer Mark III. s_) Mart
Tem:erature predictions for various event sequences and various boren injection capa:ities are given in Parts (b) and (c) of this res:ense. The sec:nd part of questien a) abov. was "What are the consecuences of t:rus er su::ressicn pcol failure?" Figure 2 illustrates, by means of an elementary event sequence ciagram, the relative safety significance cf maintaining su::ression p 01 integrity. This event secuence, anc tne related discussion ahich foli:ws, is basec u:en analyses
- erformed by the staff for a " typical" EWR/c plant with a Mark I containmenc and ecui::et with Recirculation Fum: Trip. However, we believe that.tne basic c:n:lusiccs are a::licable for all SWR c0ntainment designs.
/
( Referring to the bettem half of Figure 2 which summari:es the secuence of events (cr the case where.ccm:lete loss of su:;ression po01 integrity ec:urs, we nets the following:
- 1) Usin; corservative assumptiens the staff estimates that an ATWS event would result in f ailure of 10 percent tc 20 percent of the fuel by clad perferation early in the event i.e., a few sec:nds. Thus hignly centaminated steam would bicw down from the safety / relief valves t: the su:pression pool.
a
- 2) A major ru:ture cf the peci would result in a release ;cf a x 10" 15.
$ 0 to S x 10- lb. cf high tece;rature a:ce ximately 250 F , highty c:ntaminated water and steam to the auxiliary building. The exact quantity and tem:erature I is :lant specific. The water tem:erature was :alculated assuming l
- :lete mixing cc:urs in the pecl.
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3) The HPCI pump is located in the auxiliary building and thus wculd be exposed to a steam and radiation environment for which it has l not been qualified. If the HPCI pump fails because of exposure to the steam and radi5ticn environment, as was noted above, core melting would probably oc:ur. 4) A loss of suppression pool integrity could (decending en whe e the failure cc:urred) result in :ne loss of a major scurce of water fer reacter c:clant system invert:ry reciacement and core c: cling. With the less of the suscression scol (high temperature or less Of integrity) as a water source, the HF'I system can be continued to be su: lied frcm the alternate water scurce ',Cendensate Storage Tank). Althcugn the ; referred scurce of water su; ly ':r the HFCI system is frcm the Condensate 5t: :e 'mk, manual acticn is required t: re:lenish the C:ndensate Storage Tank in a:cui 'l minutes in ceder to kee: the c:re covered, additionally, if the initiating event is an CEE, the C:ndensate St: rage Even Tank may not be astilable en some plants to provide a water sue:ly. if the HPCI systam functicns using the Condensate Startge Tank and the I. core remains c0vered, it is likely that the steam pressure in the auxiliary building would cause tne building bicw out panels to ocen thus permitting a direct path for the contaminated steam to be released to the envircnment. . The staff estimate is that the resulting dose would e)ceed ICCFR100. General Electric, using different assumeti0ns, has reached a c:nclusion i different frem that cf the staff.
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- 5) In su= ary, su;;ression po01 integrity must be maintained to prevent ex:essive radioactive release to the envir0ncent and :: Orctect the HPCI pump and possibly other ' safe shutdcwn equipment in the auxiliary building from a steam and radiation environment fcr wnich they have net been cualified.
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4 Ouestion: b) Provide an evaluation and ccmcarisen of the effects of 43, 36, anc 400 gpm liquid boren injection rates en the predicted transients in the pressure suceression pool cr torus. Question: cl Provide a ecccarison of the effects of various time delays Of boren injection (to 10 minutes) en :he credicted transients in the pressure i su;cression cool or torus. Rescense: The staff has summari:ed in tabular form the results ci analyses performed by G.E. for the "werst case transient i.e., highest vessel cressure." 3 The informatien cresented in Fig. 3 demonstrates tne effect on the maximum
;cci tem;erature of varying both the licuic boren injection rate anc the SL:S actuation time.
Note that the 4C0 gpm SLCS injection rate (Alternate a;creach cer '!alume
-3 N'JREG-0460) provides for the following: l
- 1) Assurance that the sue:ression ;cci temperature core limit is not exceeded even if a single active failure is assumed.
- 2) Assurance that the core is not uncevered even if a singla active failure is assumed.
- 3) Assurance that the su;cression ecol temperature limit is not excesced even if the operater action (e.g., RHR in ecol cooling mode) is delayed beycnd 10 minutes.
(
7 () v 4) Additionally, im:lementatien of Alternate 4 will provide assurance that the suppressicn col tem;erature is icw enough t0 meet the HPCI system suction temperaturg recuirements and thus assure an available HPCI water source even if the Condensate Storage Tank fails. Scme plants do not have a Condensate Storage Tank designed for seismic loads, the CBE (an event with a substantial probability of occurrence over tne plant lifetime - ICCFR100, Accendix A) could be the initiating evert of an ATWS, al:ncugn of Icwer frecuency than other transients. O V 4 v
4 1 1 O
. BWR TECH SPECS I SUPPRESSION POOL TEfiPERATURE ( F)
ACTION 595 tiORtiAL OPERAT10tl-TESTIt!G LlillT BUT O START COOLING (24 HRS - 95 F)
>105
>110 SHUTDOWN
>120 SCRAli i
4 4 lO i i
. a SJFC, FCOL_ HPCI (1AKE-UP SOURCE cst
- Resucoliec ACCEPTABLE
. C0fiSEQUEilCES Functiens CST Not Resucolied ~
. CORE t1ELT LIKELY ltlTACT
'*"5 -
CORE MELT LIKELY cst Resue: tied
> PART 100 LIKELY Core Nc. Mel:
ATHS Func:1 ens O _ Csr Not aesue:1iee CORE tiELT LIKELY ~ FAILS
. Fails ~
CORE tiELT LIKELY I SUPPRESS 10ft POOL EVElli SEQUEllCE
' Condensate St: rage Tank .
Figure 2 1
PRELIMINARY BWR SUPPRESSION POOL MAXIMUM TEMPERATURE ATWS EVENT
- PLANT TYPE EVENT SLCS ACTUATICN PEAKPg0LTEMP. G.E. REFERENCE (GPM) TIME ( F)
(MIN) MARK I MSIV 36 2 200 March 9, 1979 9 :.' CLOSURE with NRC Staff 36 10 260 March 9, 1979 M ;. with NRC Staff 42 2 250 Tele hene Conver-sation w/ G.E. b\-- 43 5 290 Estimated fr:m NEDO-25016 anc
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- 11-10-75 Res:enses MARK II MSIV 36 2 180 March 9, 1979 M :. '
CLOSURE witn NRC Staff 86 10 210 March 9, 1979 Mt;. witn NRC Staff 400 1 150 February, 1975 Mtg. w/NRC Staff MARK III MSIV 26 2 165 March 9, 1979 M:9 CLO3URE , w/NRC Staff 26 10 190 March 9, 197; M:;. w/NRC Staff 400 1 150 February, 1973 Mtg. w/NRC Staf f O v
'C:nsicers Effe :slof Single Active Failures 1
Figure 3 1
t2 a b x X X
APPENDIX XXI
Title:
Generic Items Review Assignments APPENDIX X XI O Recomended ACRS Action Concerning Generic Items Agreed at 233rd ACRS Meeting Resolved Itees
- 1. NPSH for ECCS Pumps - Reactor Operations SC.
his is covered by Reg. Guide 1.1. We Reactor Operations Subcomittee could review this with the Division of operating Reactors to determine whether all plants are in compliance. Potential for vortex problems should be considered. , l
- 2. Dnergency Power - Joint Power and Electrical Systems and Reactor l 1
Cpetations SCs. Reg. Guide 1.6,1.9, and 1.32 in conjunction with portions of IEEE-308 l (1971) covers this matter. However, the question concerning loss of CC power or combined loss-of-of fsite- and -onsite-AC power are presently of concern from a risk standpoint. We Power and Electrical Systems subcommittee and the Reactor Operations Subcommittee should jointly review the status of emergency power requirements. W e question of grandfathering older plants should also be considered regarding emer-gency power.
- 3. Hydrogen Control Af ter Loss-of-cooling Accident - mI-2 Implications SC. ,
We present hydrogen control requirements are based primarily on the con- l cern for hydrogen build-up in containment following a IICA 4ere the fuel j temperature rises to the level at which zirconium-water reaction proceeds ! rapidly, leading to hydrogen generation suf ficient to cause burning or explosion. We Reg. Guide limits in 1.97 presume an oxidiation rate that is a function of surf ace area and a termination point related to ECCS capability. We tree Mile Island Accident displayed high hydrogen generation because the ECCS was not permitted to do its job. We WI-2 Implication Subcomittee should recomend actions for reevaluation of this generic item. l
- 4. Instrument Lines Penetrating Containment - No action required Reg. Guide 1.11 and its Supplement adequately cover this point and no further action is needed.
- 5. Strong Motion seismic Instrumentation - No action required 21s is covered in Reg. Guide 1.12 and there does not appear to be the need for further action.
- 6. Fuel Storage Pool Design Bases - Joint Plant Arrangements and Safeguards and Security SCs.
Wis is covered by Reg. Guide 1.13, however, the committee has frequently raised questions concerning the location of the fuel storage pool because of industrial sabotage questions. We Plant Arrangements and Safeguards and O v Security Subcomittee should review this matter and make recommendations to the full comittee concerning the need for further action, especially regarding the location of the fuel pool with respect to grade.
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- 7. Protection of Primary System and Engineered Safety Features Against Pump Flywheel Missiles - No action required
@is is covered by Reg. Guide 1.14 supported by knowledge developed in the Safety Research Program. Based on the staff evaluation of the R&D work, this matter appears to be adequately covered.
- 8. Protection Against Industrial Sabotage - Joint Plant Arrangements and Safeguards and Security SCs.
Reg. Guide 1.17 covers this matter, but since the issuance of Reg. Guide 1.17, comittee letters have continued to raise questions about the adequacy of industrial sabotage protection. h is matter should be addressed by joint effort of the Plant Arrangements Subecmittee and the Safeguards and Security Subcommittee.
- 9. Vibration Monitoring of Reactor Internals and Primary System -
No action required Reg. Guide 1.20 covers these matters and the recent review of the loose parts monitoring technology indicated that current interpretations of Reg. Guide 1.20 by the NRC Staff serve the situation adequately.
- 10. In-Service Inspection of Reactor Coolant Pressure Boundary -
Metal Components SC. Wis is covered by Section XI of the ASME Boiler and Pressure Vessel N Code and Reg. Guide 1.65 along with other modifications of the Code (d recently evaluated by the Reg. Guide Subcommittee. Questions remain as a result of Cuane Arnold piping problems and various PriR feedwater line problems. Wis matter is under active review by the Metal Componnts subcommittee and an update of recommendations concerning this matter should be provided from that Subcommittee.
- 11. Quality Assurance Curing Design, Construction, and Operation -
Reactor Opetations SC. Requirements of 10 CFR 50, Appendix B, ASME Boiler and Presure Vessel Code, Section III, ANSI-N45.2 (1971), Reg. Guides 1.28, 1.33, 1.64, 1.70.6, and proposed standard ANS-3.2, all address these matters. We NRC staff should be asked for a collective appraisal concerning the coverage in these documents. We Reactor Operations Subcommittee should then reassess the adequacy of this coverage. Recent experiences at t ree Mile Island and concerns about the seismic restraints justify a determi-nation concerning QA control adequacy.
- 12. Inspection of BWR Steam Lines Beyond Isolation Valves - No action requird his adequately covered by ASME Boiler and Pressure Vessel Code, Section XI.
- 13. Independent Check of Primary System Stress Analysis - No action required his is adequately covered by ASME Boiler and Pressure Vessel Code, Section III.
- 14. Cperational Stability of Jet Pumps - No action required We work on Dresden-2 and -3 installations and other operatirq experiences t]
b adequately satisfy the ACRS concern.
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- 15. Pressure vessel Surveillance of Fluence and NW Shif t - Metal Components SC (Review together with Item 16)
Wis is covered by 10 CFR 50, Apperdix A and ASDt Standard E-185. We NRC staff has recently recomended and the ACRS has approved the use of surveillance specimens from multiple reactor installations as satisfying the intent of the regulatory requirements. 10 CFR 50 will be modified accordingly under rulemaking proceedirgs.
- 16. Nil-ductility Properties of Pressure Vessel Materials -- Metal Components SC.
tis is covered by 10 CFR 50, Appendix A ard Appendix G, ASME Soller and Pressure Vessel Code, Section III and was addressed in the ACRS 1970 Report on Light Water Reactor Pressure Vessel Integrity, MSH-1285. We situation still appears. to be adequate from a safety stand-point, but the ACRS Metal Compnents Subcomittee should reexamine the nil-ductility problem as a function of temperature for some of the older vessels nearing the end of their specified life and any new questions that have arisen concerning the upper shelf properties of materials.
- 17. Operation of Reactor with Less han All loops in Service - No action required Standard Review Plan, Appendix 7A and Branch Technical Positit n EICSB-12 cover this matter adequately.
- 18. Criteria for Preoperaitonal Testing - Reactor Operations SC.
%is is covered by the most recent revision to Reg. Guide 1.66 but the uniformity of the preoperational testing program at variot.4 sites is unclear. We present concerns about plant operating skilli suggests a need to have the Reactor Operations Subcomittee examine _ne nature of preoperational test programs in order to determine whet er the require-ments of Reg. Guide 1.68 really satisfy regulatory needs.
- 19. Diesel Fuel Capacity - No action required Standard Review Plan 9.4 covers this matter adequately.
- 20. Capability of biological shield withstanding double-ended pipe break at It safe ends. Regulatory review practices cover this matter xlequately.
may be appropriate to have one of the ACRS consultants examine a few examples of the design treatment to ascertain whether the app;sach is based on correct safety criteria.
- 21. Operation of One Plant While Others are Under Construction - Have Fellows review
%e coverage under Reg. Guide 1.17; 1.70; Sections 13.62; 1.101; ANSI N-18, 1.7; and Standard Review Plan 13.3, Appendix A; and 13.6 are all relevant to this cpestion. Cne of the ACBS Fellows should be asked to review these dccuments to determine whether they-treat all of the ACPS questions that have been raised and whether any other matters desere attention, p te potential for a %ree Mile Island type of accident is peticularly i- relevant to this matter. IIRs should also be reviewed.
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- 22. Seismic Design of Steam Line - Combination of Dynamic Loads SC.
21s is covered by Reg. Guide 1.29 but the Combination of Dynamic Ioads Subcommittee is reexamining the design bases. Pocomnended changes to Reg. Guide 1,29 may evolve from the combination of dynamic loads review.
- 23. Quality Group Classification for Pressure Retaining Conponents -
Plant Arrangements SC. , Reg. Guide 1.26 covers this matter but questions arising from the i l interactive effect of non-safety grade equipnent as seen in the tree Mile Island-2 accident may lead to changes in these classifications. l We Plant Arrangement Subcommittee should review this matter. l l
- 24. Ultimate Heat Sink - No action required Reg. Guide 1.27 covers this matter satisfactorily.
1
- 25. Instrumentation to Detect Stresses in Containment Walls - No action required !
Reg. Guide 1.18 covers this matter but there are some controversial questions associated with grouted tendons. Current Staff interpreta- , tions provide adequate controls.
- 26. Use of Furnace Sensitized Stainless Steel - No ac tion required Reg. Guide 1.44 covers this matter satisfactorily,
- 27. Primary System Detection and I4 cation of Leaks - .'ower and Elec-trical Systems SC.
Reg. Guide 1.45 addresses this matter and experien :es at Duane Arnold and other plants indicate that the procedures are ;uitable. Exploring the use of TV cameras to find leaks could be explored.
- 28. Protection Against Pipewhip - Combination of Dynamic I/ cads SC.
> Wis is covered by Reg. Guide 1.46 but the Ceci>ination cf Dynamic [ cads SubcommPye will be reviewing these requirements as they are beire influencol oy combined load consid5 rations t e question of whether the rore elaborate requirements of combined loads introduce undesirable requirements should be examined.
- 29. Anticipated Transients Without Scram - MWS SC Although ttf.s matter was covered by WNSH-1270, iswed in September 1973, the NRC has not yet established an implementation plan nor are the technical bases fully established. We ACRS MWS Subcommit-tee should continue to review this matter and recc. mend actions to the full Committee.
- 30. EECS Capability of Current and Older Plants - Joint ECCS and Plant Arrangements SC 2e status should be updated through review by the ECCS Subcommittee, possibly with some support form the Plant Arrangements Subcommittee.
Concerns about the oldest installations, e.g.. Indian Point 1, have been s resolved by NRC licensing action over the past several years.
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- 31. Positive Moderator Coef ficient - No action required PWR's presently follow a practice that satisfies the concerns about mod- !
erator coefficients under normal conditions. We transient ques-tions associated with I4CA and the uncertainties associated with AWS effects are under review.
- 32. Fixed In-Core Detectors on High-Power PS's - No action required l In-core monitoring needs to be re-reviewed in the light of mI-2 exper-ier.ce, but it is unlikely that fixed in-core detector needs would change because of such a review. Bis item seems 0.K.
- 33. Performance of Critical Components (Pumps, Valves, etc.) in Post IOCA Environment - Power and Electrical Systems SC.
We qualification requirements in Rog. Guide 1.40, 1.63, 1.73, 1.89, and IEEE Standards 382 (1972), 383 (1974), 317 (1972), and 323 (1974), all address these matters. However, the experience at Three Mile Island-2 might alter some of these requirements. W e Power and Elec-trical Systems Subcommittee should axamine the need for new requirements.
- 34. Vacuum Relief Valves Controlling Bypss Paths on BWR Pressure Sup-pression Containment - ACRS Fellow te NRC staff requirements for Mark II and Mark III containments address these matters adequately. A review of 3ctual experience with Mark II de-O sign might be useful for updating our k,cwledge. One of the ACRS Fellows might be assigned to make such a review. LERs should also be considered.
- 35. Emergency Power for Wo or More Reacter ; at the Same Site - Power and Electrical Systems SC.
Reg. Guide 1.81 covers this matter. Shared diesels at oldet plants should be examined.
- 36. Effluents from Light Water cooled Nuclear Power Reactors -- No action required 21s environmental question is resolved by the requirements of Appendix I of 10 CFR 50.
- 37. Control Rod Ejection Accident - No accion required his is covered adequately by the requirements of Reg. Guide 1.77.
- 38. Main Steam Isolation Valve Leakage of EWR - No action required Reg. Guide 1.96 covers this adequately. l
- 39. Fuel Densif'ication - No action recuired Requirements of 10 CFR 50, Appendix K and case-by-case review of vendor fuel godels covers this matter sat!sfactorily.
- 40. Rod Sequence Control Systems - No action required te practices of the NRC staff, including those established by GE NEDO O 10527 cover this matter satisfactorily.
0 G 1 l
- 41. Seismic Category 1 Requirements for Auxilary Systems - Combination of Dynamic Loads SC.
Ris is covered by Reg. Guide 1.26 and 1.29, but may be reexamined l if new questions of interpretation arise out of a Combination of Dynamic l Ioads Subcommittee review. i
- 42. Instruments to Detect Limitsd Fuel Failures - Joint Power and Electrical Systems and Reactor Fuel SCs.
Although this has been addressed in an NRC document entitled " Fuel Failure Detection in Operat7ng Reactors" by Siegal and Hagan, June 1976, the experience of Three Mile Island warrants further review of this matter. We Power and Electrical Systems Subcommittee should evaluate this question in combination with the Reactor Fuel Subcommittee.
- 43. Instrumentation to Follcw the Course of an Accident -- Power and Electri-cal Systems SC.
Reg. Guide 1.97, Revision 1, addresses this matter but the requirements have never been recognized. We Power and Electrical Systems Subcommittee , should reexamine the requirements of 1.97 to determine whether they realistically define the need and whether a more definitive Reg. Guide should be provided based on MI-2 experience.
- 44. Pressure in Containment Fol.awirq LOCA's - WI-2 Implications SC.
t . WI-2 experience suggests the need to review this matter for low pres-sure containment.
- 45. Fire Protection - Fire Pro ection SC.
Branch Technical Position ?.5.1 provides a staisfactory review process. Reg. Guide 1.120 whose d !velopment has been suspended because of ACRS concerns should now be reinitiated with attention being addressed to the requirements found acceptable for current Standard Plant Designs.
- 46. Control Rod Drop Accidents (BWRs) - Core Performance SC.
R is had been adequtely covered by NRC review practices. However, LERs have raised questions, short period scram concern raised by E. Epler.
- 47. Rupture of High Pressure 'ines Outside Containment - No action required Standard Review Plan Sectiors 3.6.1 and 3.6.2 cover this matter ade-quately.
- 48. Isolation of Low Pressure from High Pressure Systems - Reactor Operations SC.
Standard Review Plan 5.4.7 addresses this matter. A few LERs have been identified which may have reopened concern for this question.
- 49. Nnitoring for Icose Parts Inside the Reactor Pressure Vessel - No action required Reg. Guide 1.133 covers this matter.
O d 50. Qualification of New Fucl Geometry - No action required Standard Review Plan 4.2, Revision 1, satisfles ACRS interest. i l l
l 1 l
- 51. Maintenance and Inspection of Plants - Reactor Operations SC.
Be ACRS originally accepted the postion that recent attention of the staff to these matters was adequate. Se experience at MI-2 reopens the question. We Reactor Operations Subcommittee should determine ! whether this matter needs additional effort.
- 52. Safety Related Interfaces Between Reactor Island and Balance of Plant --
Plant Arrangements SC. Standard Review Plan 1.8 covers the matter in an administrative sense, but systems inttraction questions from the 'IMI-2 accident experience warrent reexamination by the Plant Arrangements Subcomittee. Resolution of Pending Items
- 53. Turbine Missiles -- Get update from S. H. Bush.
Particular attention given to older plants.
- 54. Effective Operation of containment Sprays in a LOCA - Generic Items SC -
will follow at an appropriate time. This matter should be reexamined by the Generic Items Subcommittee. We selection c f chemical additives is still under review by the NRC Staff.
- 55. Possible Failur4 of Pressure Vessel Post-LOCA by Thermal Shock -
Meta 1 Component- SC. Reg. Guide 1.2 covers current practice satisfactorily. B e situation with respect te old plants is still unclear and the LERS display some events where .nermal shocks have exceeded Tech. Spec, limits. te implication:. of the LERs need m re attention. W e Metal Components . Subcommittee should address this.
- 56. Instruments to Detect (Severe) Fuel Failures - Power and Electrical Systems SC.
Be tree Mile Island experience justifies reexamination of this question. 1
- 57. Manitoring for Excess Vibration Inside the Reactor Pressure Vessel -
Power and Electrical Systems SC. : Methodology exi;ts to address this matter in the pressure vessel, but l the quality of its sensitivity has been related to actual safety needs. l The capability seems to be adequate but the matter should be kept under I surveillance by the Power and Electrical Systems Subcommittee, j l
- 58. Non-Random Multiple Failures - Single Failure Criterion SC.
Items 58.a, Reactor Scram Systems; 58.b, Current Sources; and 58.c, EC Sources, are matters of concern. We systems interaction work is now under active review by the Plant Arrangements Subcommittee and it should continue to assess this question. We single-failure cri-terion is relevant. I 1 l l l l
# *4 a4*.
- em A Q
- 59. Behas for of Reactor Fuel Under Abnormal Conditions - Reactor Fuel SC.
Recent experience at tree Mile Island-2 should be evaluated to detemine 1 what is needed in this area. %e ACRS Research Report has suggested that the PBF program be reoriented to address the question of intemediate level fuel degradation where fuel claddire has been significantly damaged and some fuel melting may have occurred.
- 60. BWR and PWR Primary Coolant Pump Overspeed During LCCA - Joint ECCS and I Plant Arrangements SC.
Ragaires review by ECCS and/or Plant Arrangements Subcommittees. l
- 61. Mvisability of Seismic Scram - Extreme External Phenomena SC. '
Information is available from the Japanese and from the Canadians with respect to seismic scram . We Extreme External Phenomena Subcommittee should evaluate whether this new information provides sufficient back-ground to make a judgment about when seismic scrams may be desirable in nuclear plants.
- 62. Dnergency Core Cooling System Capability for Future Plants --Joint ECCS and Plant Arrangements SC.
We requirements of 10 CFR 50, Section 50.3.4 (a) (4), 50.3.4 (b) (4) , 50.4.6, and Appendix K, establish fuel performance requirements that hav ' enhanced the emergency core cooling system capability of plants O, sin:e this generic item was identified. All of the LCCA evaluation mc31s have now been completed. W e need for other cooling approaches to improved ECCS capability needs to be reviewed by the ACRS. %e FCC ; and Plant Arrangements Subcommittees should jointly attempt to
.etermine whether this generic matter is adequately resolved, and if not, what actions are needed. l l
- 63. Ice Condenser Containment - ECCS SC.
Be ECCS Subcommittee should determine whether adequate design margin l exists during LCCA for ice condenser contairrnents. If design margins are of importance, the action required to establish design margins should be identified.
- 64. Steam Generator Tube Leakage - Metal Components SC.
Regulatory Guide 1.83 establishes a safe operating rrode, but the leakage frequency is still of concern. We Metal Components Subcommittee should review this matter and establish the path of action for generic resolution.
- 65. CRS/NRC Periodic Ten-Year Review of All Power Reactors - Reactor Operations SC.
We tree Mile Island accident reemphasizes the need to establish a policy concerning this matter, he NRC Staff presently has a program to review the older licensed reactor systems as a basis for defining periodic review policy. We ACRS Reactor Operations Subcommittee should evaluate this activity on a continuing basis until the NRC has established n fy an acceptable policy, i
t . l g%J i
- 66. Computer Reactor Protection System - Power and Electrical Systems SC.
Bis system continues to be reviewed by the Power and Electrical Systems Subcommittee and a periodic status report on the progress represents ade-quate action for the present.
- 67. Behavior of BWR Mark III Containments - Fluid Dynamics SC.
We experimental programs to verify Mark III contairment behavior are in progress and the Fluid Dynamics Subcomittee is maintaining an over-view of this work and reporting regularly to the full Comittee. Rese actions seem appropriate.
- 68. Stress Corrosion Crackirq in BWR Piping - Metal Components SC.
Wis matter is under active review by the ACRS Subcommittee on Metal Components. R&D work is underway under Industry sponsorship as wil as by DOE and NRC. We problem is still of concern bat the actions underway meet the present need.
- 69. Locking Out of ECCS Power Operated Valves - Reactor Operations SC.
%is matter should be examined by the Reactor Operations Subcommittee and appropriate action suggested.
- 70. Design Features to Control Sabotage - Joint Safeguards and Security and Plant Arrangements SCs.
(qj tis applies only to newly designed plants. W e Committee's intent is unclear. We Safeguards and Security Subcommittee should reexamine this question in conjunction with the Plant Arrangements Subcommittee for the purpose of establishing a direction for resolution.
- 71. Decontamination of Reactors - Joint Metal Components and Reactor Operations SCs.
% e t ree Mile Island accident shows the importance of this question but the original intent was primarily to address the decontamination of reactors to reduce operator exposure during in-service inspection and other circumstances. We status of the experimental work sponsored '
by Industry needs to be reviewed by either the Reactor Operations Sub-committee or the Metals Components Subcomittee. NCTTE: Radiological Effects and Sito Evaluation Subcomittee will consider occupational exposure aspects, and Waste Management Subcommittee will consider waste disposal.
- 72. Decommissioning of Reactors - Waste Management SC.
h is is an active NRC program of long duration and the status should tm reported periodically by the Waste Management Subcommittee.
- 73. Vessel Support Structures - Combination of Dynamic Mads SC.
We problem here is primarily asymetric load questions and load combinations. Bis matter should probably be addressed on a probaba-listic basis and should t:e reviewed by the Combination of Dynamic Mads Subcommittee.
- 74. Water Hammer - Fluid Dynamics SC.
%e mC staff is actively studying this matter tut the problem should be addressed on a case-by-case basis. An ACRS Subcomittee with compe-tent personnel to address the fluid mechanics questions should be assigned to review the status.
- 75. Behavior of BWR Mark I Containment - Fluid Dynamics SC.
%is matter is beirq addressed through R&D programs by the Mark I owners group and all of the open questions are nearirq resolution. Se ACRS needs an update of the status of this work. The Fluid Dynamics Subcom-mittee should be requested to sumarize current status and establish the actions ultimately needed to resolve open questions.
- 76. Assurance of Continuous Long Term Capability of Hermetic Seals on Instrumentation and Electrical Equipnent - Power and Electrical Systems SC.
We MI-2 accident reemphasizes the importance of this type of question and perhaps related ones. The Power and Electrical Systems Subcommittee should review this matter with the Regulatory Staff and Industry repre-sentatives to establish whether current practice is satisfactory, and if not, what actions might be appropriate to improve current practice.
- 77. Soil Structure Interaction - Extreme External Phenomena SC.
t e technology for evaluating soil structure interactions is developing O
\/ rapidly. %e ACRS should request one or nere of its consultants who are not actively pursuing personal interest in this question to sumnurize the current status of technclogy in order to determine whether the current situation satisfies the generic concerns. W e Extreme External Phenomena Subcommittee could undertake to sponsor such a review.
,_ . __ ~. _ _ _ . _ _. _
SUBCOMMITTEE ASSIGNMENTS GENERIC ITEMS
- REACTOR OPERATIONS SUBCOMMITTEE: (RKM)--ETHERINGTON,Ebersole,Mathis,Moeller Okrent, Ray
- 1. NPSH for ECCS Pumps
- 2. Emergency Power * (Power and Electrical Systems)
- 11. Quality Assurance during Design, Construction, and Operation
- 18. Criteria for Pre-Operational Testing
- 48. Isolation of Low Pressure from High Pressure Systems 51 . Maintenance and Inspection of Plants
- 65. ACRS/NRC Periodic Ten-Year Review of All Power Reactors
- 69. Locking Out of ECCS Power-0gerated Valves (Metal Components and Environmental) 71 . Decontamination of Reactors
- POWER & ELECTRICAL SYSTEMS SUBCOMMITTEE: (GRQ) -- KERR, Ebersole, Mark, Mathis, Okrent
- 2. Emergency Power * (Reactor Operations Subcommittee) i 27. Primary System Detection and Location of Leaks - use TVs
- 33. Performance of Critical Components in Post-LOCA Environment
- 35. Emergency Power for Two or More Reactors at Same Site (older plants with swing diesels)
- 42. Instruments to Detect (Limited) Fuel Failures * (Reactor Fuel)
- 43. Instrumentation to Follow the Course of an Accident
- 56. Instruments to Detect (Severe) Fuel Failures O 57. Monitoring for Excess Vibration Inside the Reactor Pressure Vessel
- 66. Computer Reactor Protection System
- 76. Assurance of Continuous Long-Term Capacility of Hermetic Seals in Instrumentation and Electrical Equipment
- METAL COMPONENTS SUBCOMMITTEE: (EI) -- SHEWMON, Bender, Etherington, Okrent
- 10. In-Service Inspection of Reactor Coolant Pressure Boundary
- 15. Pressure vessel Surveillance' of Fluence and NOT Shift
.16 . Nil Ductility Properties of PV Materials
- 55. Possible_ Failure of Pressure Vessel Post-LOCA by Thermal Shock 64 Steam Generator Tube Leakage
- 68. Stress Corrosion Cracking in BWR Piping 71 . Decontamination of Reactors * (Envirornental and Reactor Operations)
- THI-2 IMPLICATIONS: (RXM) .- OKRENT, Carbon, Kerr, Mathis, Plesset, Siess
- 3. Post-LOCA Hydrogen Control
- 44. Pressure in Containment Following LOCAs
- PLANT ARRANGEMENTS: (RKM) -- BENDER, Ebersole, Lawroski, Mark, Plesset, Ray
- 6. Fuel Storage Pool Design * (Safeguards & Security)
- 8. Protection Against Industrial Sabotage * (Safeguards & Security) L
- 23. Quality Group Classification for Pressure Retaining Components O.
- 3o. eccs caPee414tx or c#rreat and older Plaets* (ECCS)
Indicates a ' joint meeting of several Subcommittees.
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/- - PLANT ARRANGEMENTS (CONT'0)
- 52. Safety-Related Interfaces Between Reactor Island and Balance of Plant
- 60. BWR-& PWR Primary Coolant Pump Overspeed Ouring LOCA* (ECCS)
- 62. Emergency Core Cooling Capability for Future Plants * (ECCS)
- 70. Design Features to Control Sabotage * (Safeguards & Security)
- COMBINATION OF DYNAMIC LOADS: (EI) -- BENDER, Okrent, Plesset, Shewmon, Siess
- 22. Seismic Design of Steam Line
- 28. Protection Against Pipe Whip 41 . Seismic Category I Requirements for Auxiliary Systems
- 73. Vessel Support Structures
- EXTREME EXTERNAL PHENOMENA: (RS) -- OKRENT, Carbon, Lewis , Mark , Moeller, Siess
' 1 61 . Advisability of Seismic Scram
- 77. Soil-Structure Interaction
- SAFEGUARDS & SECURITY: (JCM/RKM) -- MARK, Bender, Carbon, Etherington, Lawroski l atMs , Shemon, Mess l
- 6. Fuel Storage Pool Design Bases * (Plant Arrangements)
- 8. Protection Against Industrial Sabotage * (Plant Arrangements)
- 70. Design Features to Control Sabotage * (Plant Arrangements)
- FLUID DYNAMICS: (AB/50) -- PLESSET, Ebersole, Etherington, Siess
- 67. Behavior of BWR Mark III Containments
- 74. Water Hammer
- 75. Behavior of BWR Mark I Containments
- WASTE MANAGEMENT SUBCOMMITTEE: (PT) -- MOELLER, Carbon, Kerr, Lawroski, Mark, Mathis Plesset, Ray
- 72. Decommissioning of Reactors
- RADIOLOGICAL EFFECTS AND SITE EVALUATION: (RM) -- MOELLER, Ebersole , Lawroski , Okrent .
- 71. Decontamination of Reactors
- relating to exposures (Metal Components and Reactor l Operations) l
- GENERIC ITEMS:
- 54. Effective Operation of Containment Sprays in a LOCA l
- ACRS FELLOWS:
1 f 21 . Operation of One Plant While Others are Under Construction
- 34. Vacuum Relief Valves Controlling By-Pass Paths on BWR Pressure Suppression O co"t 4"=e"t
- Indicates a joint meeting of several Subcommittees. .
l l
O - CORE PERFORMANCE SUBCOMMITTEE: (TGM/PB) -- KERR, Carbon, Mark, Okrent
- 46. Control Rod Orop Accident in BWRs
- ATWS SUBCOMMITTEE: (TGM/PB) -- XERR, Bender, Ebersole, Okrent, Ray
- 29. Anticipated Transients Without Scram
- SINGLE FAILURE CRITERION: (RXM) -- BENDER, Ebersole, Etherington, Okrent, Ray, Kerr, Lewis
- 58. Non-Random Multiple Failures
- - ECCS SUBCOMMITTEE: (AB/50) -- PLESSET, Ebersole, Etherington, Okrent
- 30. ECCS Capability of Current and Older Plants * (Plant Arrangements)
- 60. BWR & PWR Primary Coolant Pump Overspeed During LOCA* (Plant Arrangements)
, 62. Emergency Core Cooling Capability for Future Plants * (Plant Arrangements)
- 63. Ice Condenser Containments
)' - REACTOR FUEL SUBCOMMITTEE: (TGM/PB) -- SHEWMON, Etherington, Lawroski , Mark , Mathis , ; Okrent !
- 59. Behavior of , Reactor Fuel Under Abnormal Conditions
- 42. Instruments to Detect (Limited) Fuel Failures * (Power and Electrical Systems) 1
- ACRS CONSULTANTS:
- 20. Capability of Biological Shield to Withstand Double-Ended Pipe Break
, at Safe-Ends
- 53. Turbine Missiles - (Dr. S. H. Bush to update for older plants)
- FIRE PROTECTION: (JCM) -- BENDER, Ebersole, Etherington, Ray, Siess
- 45. Fire Protection I l
~
- Indicates a joint meeting of several Subcommittees.
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L APPEND [X X XII NUREG-0572 P O l l ( _ _ _ l' Review of Licensee Event Reports ? 11.976 - 1978:' L I
- O E3 Mlla :
' Advisory Committee _on Reactor Safeguards i
U.S. Nuclear Regulatory- l Commission i '.
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Available from GP0 Sales Program Division of Technical Information and Document Control U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l and National Technical Information Service Springfield, Virginia 22161
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#o 'g UNITED STATES
!? NUCLEAR REGULATORY COMMISSION 2 . h. , !f k3 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS SMI f WASWNG TON. O C. 20555 s
....o September 19, 1979 lionorable Joseph M. Hendrie l l
Chairman ' O.S. Nuclear Regulatory Commissico Washington, O.C. 20'555
Dear Dr. Hendrle:
In a let tar to rne Chairman, Advisory Con nittee on Reactor Safeguards l (ACRS), datad December 23, 1978, you recuosted that the ACRS conduct a review of Licensee Event Raoorts (LERs) covering tho three year oorlod, 1976 rnrough 1973. Tne goal af rnis stucy was to identify tnose ovants wnicn have implications far l'.10 roved reactor sa f e h . The Ot hched reocr r i has been preacred in response to this request. l l Although studies of LE:ls allI not identify alI sa fe ty problems in nuclear l noser plants, the report con firms that they da represent a source of , j
/aluable data and Information, For this reason, the ACa$ nod 3rses your !
recont es tad i l sh ren t af an Ooerational Oata Analysis and Evaluation Crou3 within toe 47C. Members of tna ACRS are avallaole to consult witn this Grouo o'n a periodic ba319, If you desire. As you know, a numDer of .)roup3, i ncl ud i n j tae AC'l$, are eval ua t ing ydrlous aspOCtS Of th3 InrCo Mile Island, Unit 2 aOCld0nt. Alth00]n h3r ape:lfically addressed as part of this report, the Ccomi t tee bel iev n it is aopropr iate to note uovaral LERs that relate to this avont. Then inciu le incidents of ?laanled Su<illary red aater sys tems (% ). , at th0 bncho Soco Nuc l ear Generat ing S tat ion, reoruary la, 1977) and powar anerated roller v,1ves (e.g., at rno Oconoo Nuclear Station, June 13, lg'/5, and 3r rno 03vis-des a Nuclear Powor Station, Wp Meoer N , N77) anich, througn pro;3er follow-up, coul d n ave al er ted tna industry to me posslaillry of tnis type of eccident. l l The C:mt t tee understJods indt the Jiv i slon n f Nuc I ear Power Cove i cat,eq+ of t n o d . '. . ilop ar tmen t or norm It conducting an anq aing rwiew o f jn-usual Uccurrenca boor ts suD~ti t ted b y the ir con tractor s . We urae that I the 'GC $ fart kaea d'arp as h Of this work and dDOly ino results, where G9 0 F C,Ur I O t e .
Honorable Joseph M. Hendrie September 19, 1979 The Ccrnmittee is pleased to submit this report for your use and hopes that It will prove to be helpful. Sincerely, Max . Carbon Chairman O l 9 . n*
TABLE OF CONTENTS Chapter Page Executive Summary S-l Findings and Recommendations S-2 I Origin and Purposes of the Study l-l 2 Methodology of the Study 2-1 3 Results of the Review of LERs 3-1 4 Observations and Recommendations 4-1 Acknowledgment References ! l l \ Appendix A Correspondence Establishing the Study A-l C Appendix B Subcommittee Members and Consultants 8-l Appendix C Meetings Held in the Preparation of the Report C-l Appendix 0 Reviews of SpectfIc Classes of LERs D-l Appendix E Statistical Analysis of LERs: A Trial Study E-l Appendix F ACRS Charter and Membership F-1 I l l l J l l-J L- .
n [
'N EXECUTIVE
SUMMARY
This report was prepared in response to a request by the Chairman of the U.S. Nuclear Regulatory Commission for the Advisory Committee on Reactor Safeguards to review, with specific objectives, the Licensee Event Reports (LERs) issued from 1976 through 1978. This request was encouraged by a letter from the Honorable Morris K. Udall, Chairman of the Ccmmittee on Interior and Insular Af f airs, U.S. House of Representatives, following a suggestton from Dr. Harold W. Lewis of the University of California, Santa Barbara. Approximately 8,700 LERs were filed during the three-year period under review. These described events ranging from the trivial to those of major safety significance. From a regulatory standpoint, LERs are generated in response to reporting requirements contained in the Technical SpecifIca-tions for a given plant. Many are based on violations of these Technical Speci fications while others may simply reflect an event having potential public interest. For purposes of review these reports were classified by components (e.g., valves), systems (e.g., the reactor coolant system), or general categories (e.g., !nstrumentation and controls), and each class was analyzed separately. The detailed analyses were conducted largely by ACRS consultants expert in their respective fields, f Several related studies have been made or are being made by others. A principal objective of the present study was to identify those events which have implications for improved reactor safety. As a result of this review, a number of classes of events were identified. These elasses were subsequentiy categorized into three groups: (a) those having potentially serious safety implications (e.g., failure of a vital Instrument bus), (b) events that must be regarded as matters of concern (e.g., failure of protection devices), and (c) events of lower safety i signi ficance but with excessive frequency (e.g., set point dri f t in l Instrumentation). The Committee concluded that, although many of the reported events are not particularly significant Individually, most should continue to be reported and evaluated because, in the aggregate, they may be indicators of more serious events. Moreover, the Ccmmittee found that plant operating logs contain records of many events that do not have to be reported; some of these are important as precursors to reportable events. The Committee also concluded that, too frequently, corrective measures consist of expedient remedies rather than permanent solutions to a basic problem. S-l J
s The numbers of LERs (both totals and those for particular classes) vary from plant to plant. An analysis of these variations shows that some variations in reporting rates are consistent wIth the assumption that I events occur randomly; other variations may be caused by differences in l plant design, operation, and Technical Specifications as well as real j differences in performance. Such analyses represent one of several tools which may be worthy of further exploration as a means of identifying events of significance. I This evaluation by the Committee demonstrated that, although LERs contain l Information important to safety, much of this information is not now being used ef fectively to prevent recurrence of an event at the same plant or at other plants. Appendix D contains detailed reviews of twenty-four specific classes of LERs. The information is presented in a uniform format, giving a general description of the class of event, the frequency of occurrence, implications regarding safety, and suggested corrective actions. It should be recognized that this study was conducted on an exploratory basis. The twenty-four classes are selected as examples and do not constitute a complete coverage or assessment of LERs. In conducting such a review, the results obtained and the conclusions reached are heavily dependent on the methodology used and the personnel involved. As would be expected, the study confirmed some previously held convictions as well as provided additional insight into certain problem areas. Findings and Recommendations
- 1. This study has confirmed that LERs represent a source of valuable data and information. This is particularly true if the people conducting the reviews go beyond the immediate LER to seek out the causes and ramifications of each event. Although this review did not identify any new safety problems, it did improve our understanding of the significance of certain classes of events. It also provided an improved Indication of the frequency of certain events and confirmed the importance of adequate corrective measures.
- 2. It must be recognized that a detailed review of LERs will not identify all safety problems likely to be encountered in the operation of nuclear power plants. Nonetheless, the analysis and evaluation of LERs is an important tool which should be more fully utilized. Although it wIII be necessary to continue to reIy heaviIy on systems analysis by knowledgeable specialists as an important mechanism for identifying safety problems, a more perceptive examination of operational data by systems oriented personnel will enhance the value of LER reviews.
S-2
- 3. The Committee believes that the objectives of the LER system need to be more clearly defined. The recent decision by the NRC Commissioners to create a headquarters croup dedicated to operational data analysis and evaluation has the oorential to be an appropriate mechanism to organize and coordinate the views of affected parties in improving the system. The Committee urges that the NRC review Group include input from industry and professional associations. The Commi ttee al so en-courages the NRC to assure that this Group has independence and that in olanning its operations it seek the advice of a wide range of potential users of the resulting data and findings. To the extent desired, the ACRS, through its LER Subcommittee and consultants, would be available to review periodically the planning, operations, and accomplishments of this Group.
- 4. Although the NRC has certain responsibilities with respect to LER data, these responsibilities are shared to a large extent by the nuclear industry. The ACRS believes that the industry must increase both the breadth and the depth of its ef forts in the evaluation, interpretation and aoplication of LER data.
- 5. This study confirmed that some LERs are Indicative of problems that require attention by the NRC. These include BWR control rod withdrawal patterns and methods, the unavailability of vital services following scram, and certain classes of systems interactions, particu-larly those that involve degraded performance of systems required for p vital functions. The study also showed that there are other types of
( LERs that are indicative of conditions that could lead to situations more serious than the original event. These include water hammer, i excessive vibration, leakage between interconnected fluid systems, and loss of AC power. These types of problems, which have the potential of being precursors of more serious events, deserve f urther investigation and evaluation.
- 6. Instrumentation failures represent another category of LERs to which ]
I more attention should be addressed. The more significant of these are I failures of post-accident monitors under ambient environmental condi-tions, less significant, but nonetheless important, are problems such as instrument set point drift. Although individually of little signifi-cance to safety, instrument set point drift constitutes approximately 10% of alI LERs.
- 7. Human errors in the design, operation and maintenance of reactor components and systems represent a signi ficant source of LERs.
Up to 50% of all failures appear to be attributable to this source. Contributing factors include the use of inadequate instructions, 1 S-3 o !
procedures and/or guides; a lack of adequate attention to man / machine Interfaces so as to reduce the probability of error; and the use of improperly trained personnel. The Committee recommends that these and related problems be given increased attention by the NRC.
- 8. The exploratory use of statistical analyses has shown that such techniques can be Instrumental in the identification of LER reporv! rig rates for speci fic power plants that lie outside those anticipated on the assumed basis of random variations. Such analyses can also provide insight into the performance of individual power plant systems and in selecting those systems potentially in need of further evaluation and/or investigation. The Committee believes it would be desirable to computerize such analyses for automatic processing of LER data as they are logged into the computer bank. Such a mechanism would make it possible to detect significant deviations from normal reporting rates and to follow up with further evaluation, as appropriate.
- 9. To assure increased effectiveness in the future use of the LER system, the Canmittee recommends:
- a. That the adequacy and completeness of the format for submitting LERs be thoroughly reviewed. Specific attention should be di-rected to the establishment of a more uniform system for identi-fication of the component and system involved, as well as its function. Attention should also be directed to the incorporation into the LER of a clear statement as to exactly what happened, the number of times a similar event has occurred in the past, its potential for recurrence, its significance relative to safety, and its Implications relative to any design, maintenance and operational changes that may be required or are suggested by the given event.
- b. That attention be directed to the development of procedures for Improving the quality of the reported data, including the classi-fication of LERs according to system and component, the correction of errors in LERs as submitted, and the attainment of greater uniformity and consistency in the resulting data. Accomplishment of l these goals might include the development of training programs for i personnel who initiate and review LERs.
- c. That a systematic procedure be developed so that the benefits or lessons to be learned from LERs submitted by individual licensees are fed back into the training programs for personnel involved in the design, construction, operation, and maintenance of all l nuclear power plants. J S-4 O
/~N h I. ORIGIN AND PURPOSES OF THE STUDY 1.1 Charge to the Committee in a letter to the Chairman, Advisory Committee on Reactor Safeguards, dated December 28, 1978, the Chairman of the U.S. Nuclear Regulatory Commission asked that the ACRS conduct a study of the Licensee Event Reports (LERs) submitted to the NRC by operators of ccmmercially licensed nuclear power plants. This request was encouraged by a letter from the Honorable Morris K. Udall, Chairman of the Committee on Interior and Insular Affairs, U.S. House of Representatives, following a suggestion from Dr. Harold W. Lewis of the University of California, Santa Barbara.
The charge to the Committee included the following direct and Indirect aspects:
- a. That the Committee should review LERs for the 1976 through 1978 period "to identify those events which have implications for improved reactor safety"; (Appendix A-1)
- b. That "In analyses of relatively minor incidents it is not ,
the investigation of facts, but the determination of cause l and the opportunity to make recommendations for Improvements"
} that is central to the review; (Appendix A-II) j
- c. That the review should be " directed toward the enhancement of reactor safety through the analysis of the precursors to major accidents", and should be conducted "with a view toward select- j ing incidents which may not have threatened the public health I and safety, but whose analysis might enhance the public health and safety". (Appendix A-Ill)
I.2 Depth and Extent of the Study The request to the Committee specified that the initial effort was to be a trial review and should cover LERs submitted during a period of three years, January I, 1976, through December 31, 1978. In conducting the study, the LERs from this period were augmented by the Committee as appropriate with some information from other years. The Commission stated further that they would appreciate a report within 6 to 12 months. I l-i i o I V l
1.3 Sources of LERs From a regulatory standroint LERs are generated in response to reporting requirements contained in the Technical Specifications for a given plant. Many are based on violations of these Technical Specifications while others may simply reflect an event having potential public interest. Regulatory Guide 1.16, "Rgporting of Operating Information - Appendix A Technical Specifications"I, serves as the primary reference concerning the reporting requirements of events at commercial nuclear power plants. Two types of reportable occurrences are defined: those requiring prompt notification with a written 14 day follow-up and those requiring a written report within thirty days of the event. The distinction between the two groups was meant to provide more prompt identification of thosa events of higher safety significance. Examples of the former type of reportable occurrence include f ailure to restore a safety system to operability following test or maintenance, failure of a safety / relief valve to close after pressure has been reduced below that required for closing, and welding or material defects greater than those allowable by applicable codesl. I Examples of the class of events requiring a written report within thirty days include the failure of one out of four undervoltage relays to actuate a reactor trip breaker during a test, one out of two centrifugal charging pumps inoperable because of a faulty bearing (redundant pump operability confirmed),l and failure to perform surveillance tests at the required frequency . The events requiring prompt notification are reported to the Director of the appropriate NRC Regional Office within 24 hours and confirmed no l later than the first working day following the event. A written follow-up l report is required within two weeks. The thirty-day written reports are also sent to the Director of the Regional Office. The Director of tho Office of Management and Program Analysis receives a copy of all written reports and is responsible for their entry into the NRC LER computer file. The Nuclear Safety information Center (NSIC) at the Oak Ridge National Laboratory also maintains a computer file of abstracts of LERS. l.4 Related Studies The application of the study of LERs to the field of reactor safety is not a now concept. Such studies have been conducted by several research groups. For example the Nuclear Safety information Center reviews safety-related l l l l-2 j O
LERs both on a bimonthly basis for publication in Nuclear Safety and on an
% annual basis in reports to the NRC2-5 and also publishes other reports directed to specific problem areas 6-8 In addition, a bimonthly column on operatlonal ex is pubIlshed in Nuclear News 9.periences, many of Other examples of which studiesrelate to LERs, in which the data contained in LERs have been utilized are discussed below.
In 1977, Crellin, Jacobs and Smith 10 of the General Electric Canpany presented results of the initial phase of the application of a defect analysis program in the design, construction and operation of nuclear power plants. In this program, LERs were used to identify " lessons learned" from existing operational experience. The model used in the analysis of the LERs was patterned after one which had been med success-f ully in the aerospace industry. 1 Moeller reviewed reported f ailures in air-cleaning systems from 1966-74 at nuclear facilitiesll. A second study by Moeller ld which utilized LERs l from the period January l, 1975 to June 30, 1978, was expanded in scope to ' include f ailures in air-monitoring and ventilation systems. In both cases l the investigator used the failure data to identify problem areas in the l systems studied. ! In 1978 Fluor Pioneer, Incorporated, published a reportl3 of a study performed for Commonwealth Edison Company. This study was based on a l review of approximately 9000 LERs from all operating reactors. LERs f.s involving systems interactions potentially applicable to the Zion Station (' design were studied in detall. The methodology used in the study was ; reported by Fluor Pioneer to be satisfactory for investigating those I systems interactions which are revealed by LERs. l The Electric Power Research Institute has sponsored several projects in which LERs have been utilized to examine operational experience. One such study involved an in-depth review of operating and maintenance experience wIth pump seals and seal auxii1ary systemsI4 Another stud covered the characteristics of instrument control systems failuresl . In addition to the above, the XYZYX Information Corporation recently conducted an exploratory study of the role of human errors as a contributor to LERs l6, The NRC itself has conducted a number of studies which utilized the information in LERs. Some of these have resulted in changes in plant designs and safety criterla. In 1979 the NRC published the results of a 1 1-3 ! l l l O '
l l study performed by the Universit of Dayton Research institute on the reliability of diesel generators 7 A review of the relevant LERs j furnished some of the information on diesel generator operation. A l similar study of diesel generator operating experience was conducted in i 1974 by the Office of Operations Evaluation of the Atomic Energy Com- i missionl8 During the same year, that agency also applied operational data to the study of set point drift in ins trumentat ion l9 l l A recent study by the NRC Staf f20 has demonstrated the use of LERs to identify a problem and to provide guidance in prescribing action to correct the situation. One observation of this study, which was directed to water hammer events in nuclear power plants, was that such an maluation required consideration of a number of factors besides LERs. Such actors included piping code requirements, NRC licensing procedures, and vendor and architect and engineer reports as well as review of fluid systems for ' water hammer potential. Thirty fluid systems were reviewed, of which ten were found to be important due elther to the frequency of occurrence of events within the systems or because of their safety function. This study resulted in a number of staff acticns and recommendations. These included revisions of review procedures for construction permits and operating licenses, and the preparation of new Regulatory Guides. O l l l-4 O
- 2. METHODOLOGY OF THE STUDY 2.1 Organization for the Study The ACRS assigned responsibility for the study to an Ad Hoc Subccmmittee on Licensee Event Reports. Although initially it was assumed that portions of the review could be conducted entirely through computer analyses of the existing LER data banks, this soon proved inadequate. Accordingly, a team of fifteen consultants, with expertise covering the full range of nuclear power plant operations, was selected to support the Subcommittee in the conduct of its detailed analyses (see Appendix B).
2.2 Sources of information The primary sources of information for the study were approximately 8,700 Licensee Event Reports submitted to the NRC during the assigned three year period. The abstracts of these LERs had been logged into a computer data bank both at NRC Headquarters (Bothesda, MD) and at the Nuclear Safety information Center (Oak Ridge National Laboratory, Oak Ridge, TN) . Both computer data banks were utilized. The ACRS obtained valuable information from a variety of other reports prepared under the, auspices of the flRC as welI as other organizations (see references following Chapter 4). During a series of meetings (see p Appendix C), the Subcommittee heard rcoorts of ongoing reviews of LERs
\ being conducted by the Headquarters NE Staf f, as well as by contractors, and interacted with groups such as the National Transportation Safety Board that are involved in related activities. The LER System was also discussed with the owners of several nucl ear power plants and the Sub-committee met with Dr. Harold W. Lewis of the University of California, Santa Barbara, whose suggestion served as the original stimulus for the study. Lastly, a variety of informal reports on evaluations of safety-related occurrences at nuclear power plants in foreign countries was made available to the Committee for its consideration, 2.3 Detailed Approach Computer printouts were requested of abstracts of all LERs pertaining to a given component or nuclear power plant system (Table 2-1) and these, in turn, were assigned to one or more ACRS members or consultants for in-depth study, in many cases, the original LER, as submitted by the licensee, was provided to the Subcommittee members and consultants for purposes of more in-depth revlow. This approach proved very productive and was followed throughout the study.
2-1 O 4
The Committee also examined the quarterly NRC Reports to Congress on Abnormal Occurrences 2I as a possible means of identifying the more significant LERs that should be considered. Although these reports were helpful in some respects, the f act that they contained only a smalI number of events relating to commercial nuclear power plants limited their usefulness in this study. O l 1 l 2-2
)
l Table 2-1 Comoonents And Systems Selected For Review Instrumentation and Controls Reactor Coolant System Electric Power Systems Engineered Safety Features Steam and Power Conversion Systems Olesel Generators 1 Boron Control Systems ! l Valves, Pumps, and Seals AC Power Supplies DC Power Supplies Containment Systems Radioactive Waste Management Systems 1 2-3 l [ _____ _
V 3. RESULTS OF REVIEWS OF LERs 3.1 Conduct of Reviews Ord Presentation of Data in the course of its study, the ACRS called upon members of the LER Subcommittee and its consultants to conduct in-depth reviews of selected LERs within the assigned three-year period. For the purpose of these reviews, the LERs were subdivided according to the components and systems previously described (Table 2-l). Appropriate groups of LERs were assigned to those Subcommittee members and consultants best quallfled to evaluate the given subject area. This effort resulted in reports pertain-ing to specific classes of LERs, each of which was prepared in accordance with a standard format developed by the Subcommittee. These reports are presented in Appendix 0. Because the study was conducted on an exploratory basis, these reports serve only as examples; further investigation would reveal additional and more complete Information, in conducting such a review, the results ob-tained and the conclusions reached are heavily dependent on the methodology used and the personnel involved. As would be expected, the study confirmed some previously held convictions as well as provided additional insight into certain problem areas. O 3.2 Interpretation of the Observations N.] The study clearly showed th'at the LERs are a source of valuable data and information. This is particularly true if the people conducting the reviews go beyond the immediate LER text to seek out the causes and ramifications of the event. The study also showed that there are de-ficiencies in the current reporting system, as well as in the procedures under which the data are recorded in the computer banks; these observations coupled with suggestions for improving the existing system are described in Chapter 4. As may be noted in reviewing the individual reports in Appendix 0, the implications of each class of events in terms of possible improvements i in reactor safety vary in importance. On the basis of this study, the Committee believes that the events can be divided into three basic cate-gories of significance. 3.2.1 Events Having Potentially Serlous Safety Implications included in this category are the following specific classes of LERs: 3-l O I U A
i 3.2.1.1 Separation of Control Rod from its Drive and BWR High Rod Worth Events Separation of a BWR control rod, coupled with the rigidly controlled sequence in which such rods must be withdrawn, could have very serious safety implications. Thirteen rod separation events and six rod-worth-multiplication events (resulting in periods shorter than 5 seconds) have been reported since 1976. Previous studies 22 have indicated that the probability of fuel damage from a rod drop accident is small. This study sugaests that the withdrawal of a prescribed rod whose worth has been magni fied by high core xenon concentrations may be more probable than the erroneous withdrawal of a high worth rod unoer normal start-up conditions. A determination needs to be made as to whether the high rod notch worth produced by high xenon concentrations extends over a large enough region of rod movement to cause fuel damage from a dropped rod. This study indicates that additional attention needs to be given to the manner in which control rods are withdrawn, particularly during conditions where rod worths deviate from the normal . (See Appendix D-1) 3.2.1.2 Unavailability gf,yltal Services Residual heat removal is dependent on general-purpose plant systems ano serv ices wh ich are redur4 int. Failure of these vital services may cause a scram which, in turn, causes these same services to be needed. Loss of a vital instrument bus is one e;; ample of this type of problem. Events of this class occur frequentl i; for example, when an inverter or DC bus failure causes a scram, the reactor is forced into the residual heat removal mode of operation with limited c/allability of vital services. Because of the , associated implications, the Committee believes that further consideration should be given to upgrading the reliability of such vital services. As one alternative, consideration might be given to requiring that commercial nuclear power plants be equipped with a residual heat removal system and supporting services that are dedicated to that single purpose and which would not be required to serve other less essential purposes. (See Appendices D-Il and D-Ill) 3.2.l.3 Failures Due to Water Hammer and Flow Induced Vibration Data show that there are 10 to 15 LERs originating each year due to water hammer. Additional events of this kind may not be recognized or reported. This problem and the problem of excessive vibration are important because they can often lead to damage of multiple components. These events are frequently precursors to more serious events Inasmuch as l l l 3-2 l l O 1
\ continued occurrences can result in pipe cracks, failures of valves and snubbers, and damage to electrical and mechanical equipment. Other aspects of this problem include the potential effects of water hammer and vibration on engineered safety features during severe transients. This latter type of problem is closely related to the class of events described below under the cat 6 gory of " Inadequate Design Criteria" (See Section 3.2.1.9). Although the NRC Staff is addressing those problems, the Committee believes that vibration problems may deserve more attention.
(See Appendices 0-IV and D-V) 3.2.1.4 Syster'. saieraction A number of LERs reviewed in this study reveal unusual and often unpre-dicted interactions among various plant systems. It is not surprising that interactions exist, since a nuclear power plan + is an extensive and complex facility; however, the nature of these interactions is often quite unexpected. When interactions involve degraded performance of systems required for vital functions, such as shutdown heat removal, there can be significant safety implications. The NRC Staff is currently studying systems interactions through Generic Task No. A-17 . Redundancy and defense in depth are widely used in essential reactor systems to assure their availability, implicit in such usage is the assumption that a high degree of independence exists between the redundant elements (or the various echelons of def ense in depth). Occasionally an (
\
LER discloses an unintentional or previously unrecognized interdependence between such elements. In such cases, interdependence reflects one type of systems interaction problem. Although there are few LERs that directly reveal such problems, there are many that hint at deficiencies of this natu o. Because of the potentially serious implications of such situatlans, more attention needs to be directed to seeking them out. Caref ul review of LERs can uncover such design errors, if they are consciously sought out. (See Appendices 0-111, D-VI, and D-Vil) 3.2.1.5 Valve Failures There have been many reported failures of valves to operate on demand, in some cases the cause was f ailure of the valve control system or the power supply, but in other cases the failures were associated more directly with the valve or its operator. Mechanical failures include breakage of the valve stem, detachment of the valve disc from the stem, excessive friction preventing valve motion, failure of a torque limiting device, breakage of an operator linkage, degradation of valve seats, and failure due to basic design inadequacies. This study suggests that better procedures should be developed for certifying and testing valves under both dynamic and static loads. (See Appendix D-Vill) 3-3 O V
3.2.l.6 Leakage Between interconnected Flul_d Systems Proper functioning of interconnected fluid systems (e.g., water / water, water / steam and air / water) depends on various isolation devices to separate the systems and to provide a neutral zone between them. The LERs i Indicato a high probability of valve leakage or inadvertent opening at some time In their lives; hence, there is a high probability of Invasion of the neutral zone by higher pressure fluid. If such Invasion is unnoticed by the operator, the inadvertent overpressurization of a lower pressure system or the mixing of two diverse fluid systems may cause other problems. This study Indicates the need for further review of existing design criteria for interconnected fluid systems. (See Appendix D-lX) 3.2.l.7 Problems in Containment isolation and Monitoring isolation of the contalnment is essential to prevent escape to the environ-ment of radioactive materials released inside the containment. Originally, purging of the containment was a practice 1imited prImariiy to perlods during reactor refuelIng; but today, purging is conducted even dur ing normal operations. The more signi ficant LERs within this category have involved purging of the containment while the particulate monitor isolation signal to the purge valves was bypassed. Others have involved reductions in the frequency of purging to reduce airborne releases, thus causing a decrease in the frequency with which the containment could be entered for visual , inspection of safety-related equipment. There was one instance in which a reduction in the frequency of purging led to a build-up of airborne radioactive materials to the point where t-+h the gaseous and particulate monitors inside the containment were at or near full-scale indication. As a result, the monitors were Incapable of detecting f urther increases that might have resulted f rom coolant leakage. The NRC Staff has already acted to correct the problem related to isolation of containment during purging. It is anticipated that the monitoring problem will be corrected on a generic basis through the application of Regulatory Guide 1.97. (See Appendices D-X and D-XI)
- ) .l.8 Failure of Containment Monitoring Systems Due to Environmental l Conditions Several instances were reviewed in which monitors within the containment, including those designed to monitor post-accident conditions, failed because of environmental conditions. The performance of an instrument that falls due to high ambient temperature or other environmental conditions would certainly be suspect in a post-accident environment. Without 3-4 O
[ adequate post-accident monitoring, the required actuation of isolation N systems may not occur. This is a matter which should be addressed on a prlority basis. (See Appendlx D-Xli) 3.2.1.9 Inadequate Design Criteria Although the reporting rate for events in this category is relatively small, it is one deserving more attention. In particular, consideration should be given to revising the LER system so that events of this kind, now reported as being due to some other cause, are properly attributed to inadequate design. In addition, failures which occur during normal operations should be evaluated in terms of their significance under accident conditions. A recent example in this category was the reported f ailure of a feedwater flow straightener. The immediate question that should be asked is whether this same component would have been able to withstand the blowdown forces associated with a pipe break. Similar questions apply to other Internal appurtenances such as flow orlfices, flow elements, thermocouple welis, flow scoops, flow tubes, diffusers and thermal sleeves. The detachment of these items under dynamic forces and their subsequent movement may interfere with the operation of equipment required for mitigation of an accident. Greater attention should be given to improving design criteria to prevent failures in non-safety systems from affecting safety systems. Many times it appears that the full Impilcations of such events are not appreciated. A case in point is the Browns Ferry condensate pipe failure which had the potential for serious flooding effects on critical equipment. This study Indicates a need for the NRC and licensees to evaluate such information in greater detail and to share the results of these evaluations. (See Appendix D-Xill) 3.2.l.10 Engineered Safety'Featurcs Degraded by Human Errors Although there has been only a limited number of reported events involving human error that have resulted in degraded performance or capability of engineered safety features, the contribution of human errors to LERs, in general, is high, Since events that might result in the unavailability of systems such as those related to emerger.cy core cooling and containment sprays could be serious, this study suggests that the NRC and the nuclear industry should continue to address the challenge of improving human porformance at nuclear power plants. Included within the scope of this effort, which is a responsibility both of the licensees and the NRC, should be problems such as the lack of adequate attention to human engineering in the design of key operating and emergency control systems, Instrumentation that provides inadequate or erroneous ! Indications of plant performance, and misinterpretation by plant personnel of oral and written procedures and guides, in some cases, the Committee ; 3-5 l l l t i I
. _ _ _ _ __ _ _ _ _ _ _ _- __-______-___ ___ A
noted that engineered safeguards may tend to mask human performance problems. Without more attention to improving the performance of the personnel responsible for maintaining and operating nuclear power plants, problems will continue to arise. (See Appendix D-Vll) 3.2.l.II Loss of HIgh-Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems v There is a varlety of problems associated with the operation of HPCI and RCIC systems. These include failures to isolate when called upon, as well as inadvertent isolation. The former are due to a range of events including valve leakage, improper valve lineup, and inadequacies in electrical support-ing systems. The latter include set point drift and relay malfunctions. A review of LERs also shows that failures of the cooling fans, in the areas through which the steam Iines to the HPCI and RCIC systems pass, have on a number of occasions caused the systems to be isolated. The reason for this sequence of events, which is an example of systems interaction, is , that the temperature sensors interpreted the temperature rise caused by ; failure of the cooling fans as a leak in the steam lines. A solution to the problem will probably require the development of a new design for the leak sensors. (See Appendix 0-XIV) 3.2.2 Events of Concern included in this category are events having somewhat less serious safety l Implications. 3.2.2.1 Failures in Air-Monitoring, Air-Cleaning and Ventilating Systems For the three-year period covered by this study, over half of the reported air related events for BWRs, and over one-third of those reported for PWRs, involved failures in the equipment designed to monitor these systems rather than in the air-cleaning and ventilating systems, themselves. The fact that a large number of LERs is being submitted reflects more the l Inadequacies in current monitoring equipment than in air-cleaning systems, i Such failures raise questions concerning the reliability of the data when such monitors report within acceptable ranges. This study suggests that more attention should be directed to improving the performance of monitoring equipment. (See Appendices 0-XV, 0-XVI, and 0-XVil) 3.2.2.2 Failure to Recognize and Correct the Cause of an Event Frequent 1y those submitting an LER wiII falI to recognize the root cause of an event. Sometimes, the solution to one problem will introduce another. Abnormal ambient conditions are often blamed for malfunctions 3-6 I O l
when closer examination would have revealed that designs had not been adequate to withstand condItlons that might have been expected. In stilI other cases, " operator error" is accepted as the cause when the design itself invites error. Although many of these events are trivial frcm the standpoint of safety, improper corrective action leads to repeated failures and hence unnecessary diversion of resources. This study Indicates that LERs should be searched for examples of this type and that the NRC and the nuclear industry should be encouraged to identify the root causes of such events, to evaluate their associated implications, and to specify proper corrective measures. (See Appendix 0-XVill) 3.2.2.3 Failures of Protective Devices for Essential Equipment A large number of LERs contain reports of incapacliation of essential equipment as a result of failures of fuses or other devices installed to protect the equipment or its services. Such devices are associated with nuclear power plant control systems, plant protection systems, and engi-neered safety features. It often appears that limits set for protective devices, such as fuses, are too restrictive. in addition, concern arises from potential common failures of fuses and circuit breakers due, for example, to ambient temperature increases. Safety Impiications -arise because such systems may not be available when needed. This study suggests that problems of this nature should receive more attention. (See Appendix 0-XIX) 3.2.2.4 Failures of Olesel Generators Operational experiences with diesel generators have been studied on several occasions by the NRC Staff l7 I8 and others and are continuing to receive attention. These studies have shown that the generators can fall through many modes. The review of LERs associated with such f ailures has revealed that operators, aware of the damage potential to components long idle, of ten preadjust automatic systems needed for rapid response of diesel units just prior to start-up, or make similar adjustments to improve the performance statistics. As a result, it appears that much of the reported test data may not accurately reflect emergency operating conditions. Too often, plant operators cope with inadequate performance of systems instead of seeking basic cures. In the case of die d generators, it should be recognized that perfect response of such uo t is not anticipated. The objective is to meet the necessary criteria so that, with redundant units, the availability of emergency power can be assured when needed. This study suggests that the NRC and the nuclear Industry need to address all aspects of this problem including the matter of possibly misleading statistics generated by unrealistic tests of emergency equipment. (See Appendix 0-XX) 3-7 A
i 3.2.3 Events of Lower Safety Significance but Whone Frequency of Occurrence Is UnnecessariIy High included in this category are the following speci fic classes of LERs: 2 3.2.3.1 Set Point Orlft in instrumentation Because safety Instrumentation channels are redundant, the significance ) of individual events of this type is generally small. Data show, however, that approximately 10% of all LERs fall into this category. This study indicates that the margins between the selected set points and the associated Technical Specification limits may be insufficient to allow for normal varia-tions in instrument accuracy. An associated problem is the large backlog of Facility Change Requests (requested changes in Technical Speci f ications) pending within NRC. Also to be considered is the possible need for improved instrument designs to meet system requirements. (See Appendix D-XXI) 3.2.3.2 End-of-Life and Maintenance Criteria A review of events related to leakage of valves, pumps and seals show that the causes of failure include the following: end-of-life, normal wear, and maintenance activity. The frequency of such events suggests , that the criteria used to establish maintenance periods and replacement i intervals regiilre further refinements. Actual plant experience should be factored into the determination of end-of-life statistics of plant components and maintenance schedules. Exchange of such information and i exporlence among plants wIth sImiIar components shout d be encouraged. j Certain programs, such as those at the Electric Power Research Institute i and the Southwest Research Institute, are directed toward this goal. (See ) Appendix 0-XXII) 3.2.3.3 Inadvertent Actuation of Safety injectior in PWRs This study Indicated that there had been at least forty cases of inadvertent actuation of safety injection during the three-year period covered. Safety injection systems are required to operate during loss-of-coolant accidents and other severe transients that require borated water addition to the pri-mary system. Operator response to an inadvertent safety injection involves l termination of the injection and resetting of the injection signal. Re- , peated operator exposure to inadvertent saf ety injection and its termination I may produce an unacceptable response to cases where the injection is required to provide core cooling water. Operator response to inadvertent injection should be reviewed to assure that sufficient information is available in an easily understood manner so as to allow an ea"ly decision on the termination of the event. (See Appendix D-XXill) 3-8 1
3.3 Commentary As previously pointed out, the exploratory nature of this study precluded the Identification of all safety related problems that complete analyses of LERs could reveal. For example, time did not permit an evaluation of pipe crack problems (e.g., Duane Arnold Energy Center *), grid stability problems (e.g., Turkey Point Station), and extended periods of loss of off-site power (e.g., Pilgrim Station), nor was it determined whether a study of LERs would have predicted these problems. The review of t.ERs in order to identif y precursors to major accidents is a difficult task requiring some insight into scenarios that may occur as a result of ar. Initiating event. Clearly some of the events identified here (i.e., water hammer, vlbration, leakage between interconnected fluid systems) have the potential to directly produce serious accidents; other events if couplea together could produce serious consequences. This study has revealed a number of problem areas which, if properly addressed, should lead to enhanced safety, in terms of additional " lessons learned", several aspects appear to be worthy of mantion. Licensees appear-Ing before the Subcommittee reported that at least twice as many even.a are judged worthy of noting in their plant files as are reported to the NRC in LERs. Since many of these unreported events are precursors to situations that wIII necessitate the fiIing of an LER, it would appear advantageous ! for the NRC Staff (particularly the Resident inspectors) to beccme thoroughly famtiiar wIth ihese fIIes. One category of events that should receive atten-tion concerns failures in non-safety systems that have the potential of d affecting safety systems. j in the opinion of the Committee, no ef fort should be made to change the reporting requirements simply to reduce the number of reported events; If there are problems in recording and processing them, then that phase of the operation should be expanded rather than seeking to reduce the flow of incoming data. Too little effort is being directed to finding the root cause and possible broader implications of each reported failure and correcting that cause to assure that repetition of the event wiIi be unlikely. The Committee noted in this study that, far too frequently, corrective measures did not address the true problem, or that expedient remedies were accepted. With the increasing cost of down-time at nuclear ; power plants, it would appear advantageous from the standpoint of economics, l as well as safety, that a better system for assuring proper and adequate corrective actions be Instigated.
"However, the ACRS did submit a separate report on this matter to the Chairman of the NRC on August 16, 1979.
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- 4. 06SERVATIONS AND RECOMMENDATIONS 4.1 Introduction This review Indicates that LERs collectively contain a wealth of data important to the improvement of the safety of commercial nuclear power plants. The review also shows, however, that some parts of the nuclear power industry are not using the experience available in such data. As a result, failures that occur at a single plant are repeated elsewhere as well as at that same plant. Obviously, much work needs to be done and some means should be developed to encourage the NRC and industry to utilize serlousiy the experience avaiIable in the LER data to effect improvements in equipment and system performance, rollability, and safety.
During its review tha ACRS also found a number of areas where improvements in the system for acquiring and managing LER data would enhance their usefulness. Observations and recommendations for improvements in the system for preparing, submitting, recording and analyzing LERs, as well as for using the resulting data to improve reactor safety, are presented in this Chapter. 4.2 Utility of the LER System The LER system has the potential to serve a number of important purposes. These include: O
- a. The collection of information on the safety significance of specific events, on the frequencies of f ailures and f ailure modes in particular safety components and systems, and on the importance of human errors as a contributor to such i events; l
l
- b. The provision to the NRC Staff and the nuclear industry of a means for monitoring the ability of licensees to operate nuclear power plants and their associated safety features and safeguards within the bounds of their Technical Specifications. Alternatively, the system can provide information on the adequacy and appropriateness of given Technical SpectfIcations. I
- c. The provision to the NRC Staff and the nuclear industry of a source of detailed data on safety problems being experienced industry-wide. The information can then be disseminated to plant operators for their guidance in effecting improvements in equipment 4-1 O
and system design and in performance, reliability, and saf ety. This information would be available to the public as a matter of record.
- d. The provision to the NRC of a guide for administrative and technical changes that should be required on the part of licensees to effect improvements in the safety of piant operations.
The Committee does not believe that the NRC has clearly identified the purposes of the LER system. Various NRC offices are using the system with different objectives in mind. it does not appear that the data collected completely meet the needs of any of the NRC groups using them. This situation needs to be corrected. 4.3 NRC Utilization of LER Data The ACRS believes that one of the major fallings of the present LER system has been the lack of coordinated feedback of present operating data into the licensing process, operator training programs, maintenance programs, and improved plant and equipment designs. For example, the NRC Operator Licensing Branch (OL8) has indicated that it uses LERs in reviewing the 1 operating experience at given facilities; however, there does not appear to be any systematic procedure whereby the NRC assures that the benefits or lessons to be learned frcm LERs are fed back into the training programs for plant personnel, it is disconcerting to note that, although the NRC Research Staff is currently conducting studies on human factors as related to reactor safety, none of the members of the OLB is involved in these studies, it is not sufficient to collect information; operational experi-ence must also be utilized to improve the present operation of nuclear power plants. The ACRS has observed that utilities hold a varlety of different opinions and attitudes toward reporting an event as an LER. It would seem that some time and of fort should be expended toward develooing a more uni formly positive attitude towards submitting LERs. While the NRC should avoid rating utilities on the basis of the number of LERs submitted, emphasis should be placed on the fact that repetitive occurrences of similar events, even if not serious when viewed in isolation, is representative of unaccept-able performance. TheACRShasnotedtgttheCommissionhasrecentlyapprovedaplanproposed to establish a headquarters group to coordinate the by an NRC Task Force analysis and evaluation of operational data (including LERs). The ACRS 4-2 O
endorses this move and encourages the NRC to assure that the group has adequate authority and independence, in developing procedures for data collection, analysis, and use, by such a group, the NRC should seek the advice of a wide range of potential users of the resulting data from within the NRC as well as from the utilities, reactor vendors, equipment designers, architect engineers, appropriate federal, state and local agencies, and professional groups involved in the development of safety-related standards and codes. Such an approach can be beneficial both in assuring maximum effectiveness of the system ultimately developed and in assuring support for such activities from the various groups affected, in each such contact, of forts should be made to involve the upper levels of management in the planning phases, since it is only through their coopera-tion that the success of the final system can be assured. Once under way, it should be the function of this group to coordinate all efforts related to the collection, evaluation, and dissemination of LER data. The ACRS offers the foilowing comments:
- a. The Committee urges that the NRC group include a person well l quall fled in human engineering and one well quall fled in data analyses. The group should also include people possessing experience in the design and operation of nuclear power plants.
The ACRS recommends that consideration be given to rotating NRC field personnel, and those from other headquarters units through the NRC Operational Data Analysis and Evaluation Group on a i systematic basis. An assignment with the Analysis Group might be I V considered a normal part of the career development program for I selected Junior and mid-level personnel,
- b. To assure proper dissemination of the results of the evaluations, it would appear that the NRC group should issue both topical and statistical reports. They should also report how they have used the results of their evaluations to interact with the utilities, vendors, architects / engineers, codes and standards groups, etc.,
to enhance the safety of nuclear power plants,
- c. To the extent desired, the ACRS, through its LER Subcommittee and consultants, would be available to review periodically the planning, operations, and accomplishments of the Analysis Group.
4.4 Licensee Utilization of LER Data Licensees to a najor extent, along with the remainder of the nuclear community, share with the NRC the responsiblity noted in the previous 4-3 O
l section of maintaining the coordinated feedback into the Industry of the lessons learned frcrn operational experlence for the purpose of improving reactor safety. At the present time, some licensees have requested that they be provided copies of all LERs from other plants, appilcable to their plants, or to given systems and components in their plants. The occurrence of a given event of potential significance at one facility after it has previously i I occurred at other f acili ties, even af ter the NRC Staf f has called atten-tion to it, Indicates that the present level of utilization is not accept-able. This study suggests that the licensees should develop and implement f ar better plans for formalizing requirements for the of fective feedback of such data into the design, construction, maintenance and operation of nuclear power plants. A spect fic topic that should be addressed is the large number of LERs that ' indicate the cause of the event to be operator error, procedural error, or maintenance error. This suggests that a formalized program of LER review should be Instituted within training and requali fication programs for plant l operations personnel. Consideration should also be given to the use of l LER data in the training and quali fication of maintenance personnel . I During the course of this investigation, the possible effects that the utili ty organizational structure has on the utilization and ef fective feedback of LER data were discussed. It was observed in at least one case that, while the utility engineering staff reviewed the lessons learned from LERs, as applicable to their plant, organizational barriers prevented the offective communication of these lessons to the plant operations staff. Although other utilities did not appear to have this problem, this observation does suggest that Individuel utilities review their organizational structure with regard to this potential def ect and, where applicable, institute appropriate corrective measures. 4.5 Analysis of Reported Events Resulting from Human Error in its study of LERs, the ACRS was impressed with the extent to which human errors play a role, in this regard, the Canmittee urges that the now NRC Analysis Group direct specific attention to this problem area. Subjects worthy of exploration include:
- a. How many events occur because of procedural errors, how many occur because of inaccurate communications, how many occur because of inadequate training, and how many are associated with boredcm? It would be usef ul to associate the frequency of operator errors with 4-4 0
time of day, with test scores on licensing examinations, with Senior Reactor Operators versus Reactor Operators, and with the estimated degree of stress at the time the errors occurred.
- b. This examination of LERs has disclosed operator errors which should not be dismissed solely as an Indication of the need for more operator training but rather the need for modl fication or redesign of plant features, such as plant systems or the control rooms, to reduce the Iikellhood and the consequences of such errors. it would be useful to determine (and to implement design changes where suggested) the extent to which operator er rors are caused by the lack of adequate information available to the operator and by the lack of appropriate consideration given to the man / machine interf ace.
- c. Information provided to the Committee revealed that plant personnel, who were not licensed operators, were frequently assigned duties which carry significant responsibilities for which they may not be fully quallfled, it would be useful to know to what extent people are put on the job before being properly trained. Data are also needed on the degree to which the use of such personnel influences the number of LERs associated with various power plant systems, in addition, it would be useful to know the frequency of human error in maintenance work performed by " imported" workers, such as those provided on a short-term basis by contractors.
- d. This review showed that a significant number of LERs are attributed to inadequate procedures. The Committee recommends that a study be made by a human-engineering group to determine whether procedures could be reduced in number and whether they might be organized in a manner that would be more easily understood and used by operations as well as maintenance personnel . This is particularly impor tant in view of the large number of LERs that arise curing and following maintenance activities.
4.6 Quality Control of LER Information The Committee has observed a range of problems in the quality of the Information reported in LERs:
- a. Coding of LER data by licensees leads to anbigultles, especially with respect to the classi fication of LERs according to system and component. The NRC Staff estimates that 25% of the LERs are inadequately coded or miscoded; the ACRS has observed similar 4-5 i
l
l l problems. This is partially because plant components often serve multiple functions and can consequently be classified under a number of different systems or subsystems. Another source of confusion is the system coding scheme used with the standard LER form. The existing scheme was derived, with modifications, from NSS vendor system codes. System tiiles are j not functionally organized or Internally consistent; the proper code for a system performing a particular function may differ according to the vendor. Mlscoding can result when a licensee, perhaps familiar with the terminology of only one vendor, must select a code word from a listing whose elements are not func-tionally unique. Miscoded information, when entered into the LER computer data banks, decreases the effectiveness of subsequent automatic searches of the data files. A related concern resulting from this problem involves the quantative accuracy of the results from such computer searches. The information retrieval system presently used by the NRC ls Insufficient and a better system should be developed. To this end the scheme used by NSIC should be considered for incorporation into the present NRC system. The NSIC system is more easily understood by the user and appears to be more comprehensive. The next generation of information retrieval schemes might be a hybrid of the two present methods,
- b. The Committee has noted a number of problems relative to the follow-up to correct what appear to be errors in LERs and to improve the means for retrieval of Information recorded in the data bank. To assist in resolving these problems, the Conmittee recommends that the NRC Headquarters Group be given the responst-bility to require licensees to explain or correct what appear to be errors in LERs, particularly in the assignment of causes of failures,
- c. The Committee observed several instances in which an LER, as abstracted for the computer data bank, canpletely missed the significance of the event. One example was the report of an j occurrence at Three Mlle Island Unit I, on May 17, 1977. In this case, the abstract stated that the nature of the failure was a l "significant difference between generated megawatts and the l 4-6 0
1 O expected value of computer-calculated core thermal power" and that the latter error was "due to a failed steam pressure trans-mitter". Reading of the complete LER, as submi tted, showed that there was a safety impilcation that was not apparent from the abstract, in that all four power range nuclear channels had been j calibrated earlier to agree with the erroneous calculated core thermal power and were therefore Indicating too low by 8% to 10%. The nuclear power channels af fected were not even mentioned in the abstract although they presumably serve to provide high power level scram.
- d. In many Instances, the necessity for submitting an LER ls dependent on interpretation of the Technical Specifications. In some cases, such Interpretation is made by the licensee; in some cases it is made by the NRC; in some cases it is a joint ef fort. Because this practice and di f ferences in the Technical Speci fications, thanselves,
! lead to differences from region to region in the numbers of LERs reported, the ACRS recommends that training programs be established ' 4 to seek uniformity and consistency and to elevate the quality of the LERs submitted. Such a program should include both the NRC l Staf f and utility personnel and should emphasize procedures for the follow-up on LERs which, as originally submitted, are found to be deficient. The ACRS recommends that consideration be given to the development or extension of a Regulatory Guide to cover not only this aspect of the LER reporting system but also the manner in which licensees should be expected to utilize LER data in the () training programs for their operating plant personnel . The Regulatory Guide should include procedures to be followed by utility personnel and the NRC to assure that the corrective action indicated by an LER has been implemented. This is especially pertinent in those cases indicating a required design change which may be generic in nature or in those cases that show that specific equipment manufactured by a given company is defective. 4.7 Revisions in the Form for Submitting LERs Although it is assumed that the new Group established by the NRC will direct considerable attention to the adequacy and completeness of the form , for submitting LERs, the ACRS offers the following suggestions: l
- a. A more uniform procedure should be established for the identification of the component'and system involved. Where applicable, the LER might also cite the number of the reference drawing showing the component involved.
l l i 4-7 d 1
--e >s -s e m e -, -m:p - - w e- er -
- b. The LER should include an identification of the function of the component or system in which the failure occurred; the time of day when the f ailure occurred; and the position In the licenseo*s organization of the Individual filling out the LER form,
- c. The LER should include a clear statement as to exactly what happened so that readers not familiar with the detalls of a particular plant can understand the failure, it would also be helpful If reports of f ailures of ccmponents with multiple functions included a list of systems of secondary function that were also affected,
- d. The LER should include a clear statement as to both the actual and the potential safety significance of the event,
- e. The LER should contain information, where appropriate, on the number of times a similar event has occurred in the past, including related incidents that may not have mot the threshold requirements for LER submittal,
- f. The LER should summarize any design changes that may be required or are suggested by the given failure as well as provide an estimate of the potential for recurrence of a similar event.
4.8 Problems in the Reporting of Data Although improvements in the form for reporting LERs will assist in the collection of useful data, other deficiencies in the data collection process need to be addressed. These include:
- a. Many LERs contain a promise of follow-up action; but it is difficult, under the present system, to determine whether the promised follow-up has been accomplished. There should be a j formal system to assure that every LER is monitored until it is .
completed and has been properly recorded in the data bank. The l' data-recording group should be given the responsibility to check back on inccmplete LERs and to have volds filled or inadequacles clarifled by the licensee,
- b. A better definition of the cause of each event submitted as an LER is needed. For example, " motor failure" leaves doubt whether the j cause was the f ailure of the motor or of the relay controlling the motor. Particularly troublesome are the LERs for which the cause 4-8 9
(A,) is listed as " unknown". In such cases, a conscientious effort should be made to assure that the events are subsequently analyzed and a complete report submitted. In addition, when human error is specified, efforts should be made to properly attribute the error to a spectfic subcategory (i.e., inadequate design or procedures, improper maintenance, or operator error).
- c. LERs as currently submitted provide information only on the frequency of f ailures which reach threshold reporting requirements.
They provide no information on the number of the given components or systems that are in operation or on the frequency with which they are called upon to respond--data that are needed it failure rates are to be computed. While it is acknowledged that the NRC supported Nuclear Plant Reliability Data system is gathering information on failure rates, attention should be directed to determining whether simple modifications or additions to the LER reporting system might provide additional useful data along these lines,
- d. Care should be taken to develop better means for assuring proper Identification of events resulting from deficiencies concerning component installation or storage during plant construction, in a number of cases LERs, filed several years after plant start-up, identifled improper installation of equipment or components during construction. Equipment and material are commonly del ivered to plant sites before construction has reached a stage which would
()T (, permit installation. Damage to equipment can occur due to adverse environmental conditions during storage, the consequence of which can show up later during service. 4.9 Commentary , One of the more elusive of the potential benefits of the LER system is the i use of LERs, that Individually have little or no direct safety signi ficance, { as a source of information for enhancing the safety of nuclear power plants. I This study suggests it would be worthwhile to make a continuing search of LERs to see if they could serve as a source of Information for revealing: { j
- a. Inaccurate concepts which have led to events that were misunderstood I at the time and therefore resulted in Inappropriate responses;
- b. Unusually high failure rates of common components when used in a particular duty as compared to similar conponents in general; l
l 1 l 4-9 l (qs_-) l l l l l i ! l l l
- c. Unusual coupling of events, failures, and human errors (This might very well be something that could be done by computer provided the data bank is properly organized; In fact, selected events might be grouped to identify potentially serious combinations);
- d. Types of failures in components of little safety significance ,
which coold have been sertous If they had occurred in sImlIar l components in systems of greater sensitivity or under difforent , operating conditions. While this review did not explicitly address the synthesis of probable accident sequences from LER Information, this is possibly one of the more usef ul and illuminative applications of LER data. For example, a number of LERs indicated the failure of diesel generators to respond to loss of off-site power signals during testing. Others identified events in which loss of DC power from batterles would have prevented operatlon of electrI-cal swItchgear to actuate elther alternate or mergency AC power supplles. Several LERs identifled events in which weather induced power interrup-tions led to loss of off-site AC power. In some cases, LER data may be sufficient to enable a reasonable estimate of the probability of such events. More generally, the information contained in LERs may be utilized to identi fy events which should be made a focal point for failure mode and offects analysis. As previously mentioned, approximately 8700 LERs were submitted by the licensees of U.S. ccmmercial nuclear power plants during the years'l976, 1977, and 1978. For several reasons, the number of LERs varies from unit to unit. WhlIe such variatlons may be Indicative of actual difforences in safety among nuclear power units, they may have other explanations. Certain dif ferences in the frequency of submission of LERs from unit to unit will occur as a result of the apparent random nature of the events being reported. Dif ferences greater than those expected on the basis of "randcmness" can be identi fled by statistical analyses. A trial study, using this approach, was , performed and is included as Appendix E. Deviations such as those revealed in this study can then be used as a moans for the identi fication of areas for possible further investigations. While such deviations do not necessarily imply safety-related problems, they should nonetheless be pursued in order i to determine their true implications. l l l 4-10 0
i i ACKNOWLEDGMENT The Advisory Committee on. Reactor Safeguards gratefully acknowledge the significant contributions provided by Its consultants (See Appendix B) in the preparation of this report. Their enthusiasm and dedication were major factors in helping to bring the study to a successful conclusion and in preparing this report in a timely manner. i i
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l l l O ' 4-11 1
REFERENCES
- 1. " Reporting of Operating information -- Appendix A, Technical Specifi-cations", NRC Regulatory Guide 1.16 (Revision 4) (August 1975).
- 2. Scott, R.L., and Gallaher, R.B., " Annotated Bibliography of Safety-Related OccPrences in Bolling-Water Nuclear Power Plants as Reported in 1976", MC Peport ORNL/NUREG/NSIC-137, Oak Ridge National Labora-tory (1977),
l 3. Scott, R.L., and Gallaher, R.8., " Annotated Bibliography of Safety-Related Occurrences in Pressurized-Water Nuclear Power Plants as . Reported in 1976", NRC Report ORNL/NUREG/NSIC-138, Oak Ridge National Laboratory (1977). ]
- 4. Scott, R.L., and Gallaher, R.B., " Annotated Bibliography of Safety-Related Events in Bolling-Water Nuclear Power Plants as Reported in 1977", NRC Report ORNL/NUREG/NSIC-149, Oak Ridge National Laboratory (1978). I
- 5. Scott, R.L., and Gallaher, R.8., " Annotated Bibliography of Safety-Related Events in Pressurized-Water Nuclear Power Plants as Reported in 1977", NRC Report ORNL/NUREG/NSIC-ISO, Oak Ridge National Labora-tory (1978).
- 6. " Reactor Operating Experience, 1975-1977", NRC Report ORNL/NUREG/NSIC-144, Oak Ridge National Laboratory (1978).
1
- 7. Hagen, E.W., " Common-Mode / Common-Cause Failure: A Review and a Sibliography", NRC Report ORNL/NUREG/NSIC-148, Oak Ridge National ]
i Laboratory (May 1979).
- 8. Scott, R.L., and Gallaher, R.B., "Sunmary and Sibilography of Operating Experience with Valves in Light-Water-Reactor Nuclear Power Plants for the Period 1965-1978", NRC Report ORNL/NUREG/NSIC-171, Oak Ridge National Laboratory (July 1979).
- 9. Verna, 8.J., "On Line With Verna", bimonthly columns in Nuclear News. l l 10. Crel l i n, G.L. , Jacobs, l .M. , and Smith, A.M. , "A Def ect Analysis Program Applied to Nuclear Plant Experience Data", CONF-770625, l pages 61l-635 (1977).
fl. teeller, D.W., " Problems in Nuclear Alr-Cleaning Systems", Nuclear Safety, Vol. 16, No. 4, pages 469-481 (July-August 1975). l R-l l lO . i
- 12. Moeller, D.W., " Failures in Air-Monitoring, Alr-Cleaning, and Venti-lation Systems in Commercial Nuclear Power Plants (January 1, 1975-June 30, 1978)", Nuclear Safety, Vol . 20, No. 2, pages 176-188
, (March-April 1979). , 13. Fluor Pioneer, Inc., "Conmonwealth Edison Co./ Zion Station Systems in teraction Study" ( 1978) .
- 14. Grant, W.S., " Recirculating Pump Seal Investigation", Electric Power Research Inst i tute, Repor t EPRI-NP-351 (Voi. 1) (1977).
- 15. Basin, S.L., Burns, E.T., Civi, V.I., and Loell, W. S., " Character-Istics of lastrument and Control System Failures in LWRs", Electric Power Research Institute, 9eport EPRI-NP-443 (August 1977).
- 16. Fuchs, F.H., " Anal ys is of Nuclear Reactor Events Reported in March 1979,"
XYZYX Information Corporation, Conoga Park, CA (May 1979).
- 17. Boner, G.L. , and Hanners, H.W. , ' Enhancement of On-Si te Emergency Diesel Generator Rellability", NRC Report NUREG/CR-0660 (1979).
- 18. Crooks, J.L., and Vissing, G.S., " Diesel Generator Operating Experience at Nuclear Power Plants", AEC Report 00E-ES-002 (1974) .
i
- 19. Hartfleid, R.A., "Setpoint Drift in Nuclear Power Plant Safety-Related instrumentation", AEC Report 00E-ES-003 (1974).
- 20. " Water Hammer in Nuclear Power Plants" NRC Report NUREG-0582 (July 1979).
- 21. " Report to the Congress on Abnormal Occurrences", NRC Reports NUREG- l
," 0090-3 through NUREG-0090-10, and NUREG-0090, Vol. I, Nos. I through 4 l (issued quarterly, 1976 through 1978) .
- 22. Richings, H.J., "A Statistical Examination of the RDA In Sone BWRs",
NRC Staf f Report (June 17, 1975).
- 23. "NRC Program for the Resolution of Generic issues Related to Nuclear Power Plants", N9C Report NUREG-0410 (January 1, 1978).
1
- 24. " Recommendations on Operational Data Analysis and Evaluation for l Nuclear Power Plants (Task Force Report)", with enclosures, NRC Staff, SECY-79-371 (June 4, 1979).
l
- 25. Seminara, J. L., Gonzalez, W.R., and Parsons, S.O., " Human Factors Review of Nuclear Power Plant Control Room Design", Electric Power Research Insti tu te, Report EPRI NP-309 (November 1976).
R-2 l l
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l
O k_-) 26. " Reactor Safety Study" U.S. Nuclear Regulatory Commission, WASH-1400, (NUREG-75/014) (October 1975).
- 27. "NRC Policy Statement on Abnormal Occurrences," U.S. Nuclear Regulatory Commission (February 25, 1977).
- 28. " Current Events - Power Reactors"; " Power Reactor Events"; and "Operat-Ing Experience - Bulletin Informatlan Report", NRC Reports (1976 through 1978).
- 29. " Instructions for Preparation of Data Entry Sheets for Licensee Event Repor t (LER) File", NRC Report NUREG-0161 (1977).
30 . " Nuclear Plant Reliability Data System", Electric Power Research Insti tute, Palo Alto, CA (1978) .
- 31. Koppe, R.H., " Power Plant Early Alert Reporting System", Electric Power Research Institute, Report EPRI NP-988 (1979).
- 32. Chakof f, H.E. , Speaker, D. 4. , Thompson, S.R. , and Cohen, S .C. ,
" Licensee Performance Evaluation", NRC Report NUREG/CR-Oll0 (1978).
- 33. Scott, R. L., " Outages at Light-Water-Reactor Power Plants: A Re-view of 1973-1977 Experience", Nuclear Safety, Vol. 20, No. 2, pages 210-215 (March-April 1979).
s_/ 34. Takeda, S. , and Uj ita, H. , " Human Error Analysis in Nuclear Power Plants", prepublication copy of report, Hitacht, Ltd., Japan (1979).
- 35. Hubble, W. H. , and Poloski, J .P. , " Rod Drive Mechani sm f ailure Rate Analysis Derived from LER's", Idaho National Engineering Laboratory (1979).
l l R-3
,y i
1 l
1 1 i ,1 l 1 1 l
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l APPENDIX A i 1 CORRESPON0ENCE ESTASL IStiING THE SFUDY i l l i
APPENDIX A-I j# %, UNITED STATES
# % NUCLEAR REGULATORY COMMISSION
[ , j WASHINGTON, D. C. 20535
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December 28, 1978
%....f CH AIRM AN Dr. Stephen Lawroski Chairman Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Dear Dr. Lu roski #>
In a recent letter to the Commission, Congressman Morris Udall has recomended that a review be conducted on a trial basis of Licensee Event Reports for a period of three years (January 1,1976--December 31, 1978) to identify those events which have implications for improved reactor safety. Congressman Udall has suggested that this review might be conducted by an NRC contractor or a subgroup of the ACRS. The Commission believes such a review should be undertaken. The Commission has considered the alternatives and has concluded that this review could best be accomplished by the ACRS through a Committee subgroup such as the Subcommittee on Operating Reactors. It is requested that the ACRS undertake this task and provide a report to the Commission. In organizing ACRS activity to conduct this review, the Committee may find it necessary to obtain additional support from its consultants and technical assistance type contracts. The Commission would support such increased effort to the degree considered necessary by the Committee. In scheduling the conduct of this review the Commission believes that a report within 6-12 months would be appropriate. We would be interested in Committee response as to the length of time the l Committee anticipates will be required to complete this assignment. In connection with the scope of the Committee's review, the Commission l feels that if the Committee finds it appropriate to examine LER's beyond i the time frame indicated in order to provide a more meaningful evaluation of the significance of reported events and related corrective actions, such an extension should be made. Also, other sources of information I pertinent to the task should be used. as appropriate. ! O v
Dr. Stephen Lawroski December 28, 1978 The Comission requests that it be kept informed regarding the progress of this activity. The periodic meetings of the Comittee with the Comission offer useful opportunities for discussion of Conmittee pro-gress in this area. fgcerely,
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CUf'O%'t.u Honorable Joseph Hendrie Chairman, Nuclear Regulatory Commission Washington, D.C. 20555
Dear Chairman Hendrie:
This is in regard to Dr. Harold Lewis' suggestions for the creation of a nuclear accident review board. You ' responded on August 7, 1978 to our initial inquiry and since them Dr. Lewis has had a chance to review both your reSly and that of the Advisory Committee on Reactor Sa f egual -Is . Dr. Lewis made some comments on your reactions, a copy c f which is enclosed. I believe the merits of a review group in the nuclear field similar to the National Transportation Safet'y Board (NTSB) are likely to be more substantial than your letter implies. In particular, I do not think adequate consideration has been given to the benefits that might result from analysis of " abnormal occurrences" -- or small accidents and events which could be precursors to more serious accidents, by whatever definition -- by a review group independent of the NRC regulatory staff. l I would like to recommend to the Commission that it institute a trial review of 1.his nature to determine whether information might be obtained which could lead to increased reactor safety. A subgroup of the Advisory Committee on Reactor Safeguards, or some other group under separate contract.with the Commission,-could review licensee event reports submitted between January 1, 1976 and December 31, 1978 for incidents whose analysis might enhance reactor safety. Incidents could be selected g using the " abnormal occurrence" criteria, or some other . criteria which may be more suitable. These incidents could be reviewed by the independent group to determine whether they have implications for improved reactor safety.
2 Hcnorable Joseph Hendrie Page 2 Dr. Lewis has noted, and it seems reasonable to me, that in analyses of relatively minor incidents it is not the l investigation of facts, but the determination of cause and the opportunity to make recommendations for improvements, that is central to the role of the independent reviewer. It seems, then, that with respect to fact-finding, the . techinical assistance of NRC staff, or use of data previously collected by them, would not damage the independence or productivity of the review group. Thank you for your attention. Sincerely, fsskazs/g( x..uoxts Chairman - Enclosure l l l l l l l l l h l l
l O APPENDIX A-III UNIVERSITY OF CALIFORNIA. SANTA BARBARA DLMELEY
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SANTA BAR8AM, CAur0RNIA 93l06 October 4, 1978 The Honorable Morris K. Udall, Chairman Committee on Interior and Insular Affairs United States House of Representatives Washington, D.C. 20515
Dear Congressman Udall,
I now have copies of the replies of the Nuclear Regulatory Commission and the Advisory Committee on Reactor Safeguards to your letter of January 27, 1978, about my suggestion for an "NTSB" for reactor safety. May I make a few comments on these replies? In the first place, both NRC and ACRS are concerned that any new agency would intrude upon their turf, NRC because it already investigates abnormal occurrences as part of its regulatory job and ACRS because it is statutorily independent, and has the power to do so in the appropriate circumstances. NRC does, however, recognize the enhancement of credibility-that might flow from the activities of an independent agency, and would consider establishing an independent review group for a major accident. (In fact, Browns Ferry was well studied by an internal review group, and was also the subject of extensive hearings by the Joint Committee on Atomic Energy.) Both replies reflect a concern about having outsiders poking around in their business, and NRC specifically raises the question of whether there are any knowledgeable cutsiders. There is more than a small hint of parochialism here. A much more important question raised is that of whether such a new board would have anything to do, since there have never been any major nuclear accidents leading to injury to the public. This, it seems to me, reflects a misunderstanding of the point of the proposal, which was directed toward the enhancement of reactor safety through the analysis of the precursors to major accidents. NRC reports a total of nineteen " abnormal occurrences" i in FY 1977, culled from a total of many hundreds of licensee i event reports, and asserts that each of these is appropriately investigated by NRC itself. I cannot quarrel with the selection
l The Honorable Morris K. Udall 2 of those nineteen, chosen as "significant from the standpoint of ! public health or safety", nor can I validate the selection. In fact, the whole point of the proposal revolves around the meaning of the word "significant" in the statement above. There ere ; certainly p~1enty of scare stories around. . My overall reaction is that;' leaving aside questions of turf, I would still like to have someone look at the last couple of years of licensee event reports with a view toward selecting incidents which may not have threatened the public health and safety, but whose analysis might enhance the public health and safety. I suspect that different criteria might produce a richer harvest than is projected by NRC and ACRS, but remai7 open-minded until it is tried. I I I hope that this is helpful to you.
\
Eincerely ours, H. W. lM ewis (Il HWL/etm f 0
i !. l l -l ~ l i l 1 l9 i l l 1 f 9 I i ) l ^9PENDIX B ! l SUBCOMMITTEE MEM3ERS i AND i I
- CONSULTANTS t
i u I
SUBCOMMITTEE MEMBERS D. Moeller, Chairman M. Sender H. Etherington W. Kerr S. Lawroski W. Matnis D. Okrent J. Ray C. Mark A. Bates, Staff D Johnson, ACRS Fellow CONSULTANTS J. Arnold Retired - Formerly Vice President for Operations, Air Products and Chemicals R. Burns, ill Staf f Member, Los Alamos Scientific Laboratory I. Catton Professor, School of Engineering and Applied Scienco, University of California, Los Angeles R. G. Colclaser Associate Dean, School of Engineering, University of Pittsburg s_ / S. Cromer Retired - Formerly Director of Engineering, Oak Ridge National Laboratory S. Ditto Development Engineer, Oak Ridge National Laboratory E. Epler Retired - Formerly Associate Division Director, Oak Ridge National Laboratory M. First Professor, Environmental Health Engineering, Harvard School of Public Health A. Grendon Retired - Formerly Blophysicist, Donner Laboratory, University of California, Berkeley W. Lipinski Tenior Electrical Engineer, Argonne National Laboratory C. Michelson Principal Nuclear Engineer (Nuclear Systems Analysis), Division of Engineering Design, Tennessee Valley Authority H. Parker President, HMP Associates R. Seale. Professor and Head, Department of Nuclear Engineering, University of Arizona
-s J. H. Warren Retired - Formerly Vice President, Atlantic Richfield Hanford Ccmpany g ,,/
Z. Zudans Senior Vice President, Franklin Institute Research Laboratories B-l
f f-i 1: e ! ) l 1 l i I l i t i t l l 1
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o i i i i t APPENDIX 0 i j' , o + MEETINGS HELD IN THE PREPARATION OF THE REPORT l i.______
i l 1 m s,) TOPICS OlSCUSSED OR PARTICIPATING MEETING DATE WORK PERFORMED ORGANIZATIONS LER Subcommittee February 7, 1979 (l) Organization of Study (I) NRC Staff (MPA) (2) Review of LER System (2) NRC Staff (RES, NRR) March I-2, 1979 (1) Use of LER Data and (I) National Transpor-Related Activities tation Safety Board (2) Reviews of Specific LERs March 22-24, 1979 (l) Review of LER Management (1) NRC (Region ll and Systems Headquarters I&E) (2) Review of LER Data Systen (2) Nuclear Safety Information (3) Reviews of Specific LERs Center (ORNL) April 26-27, 1979 (1) Background on the Study (l) Dr. Harold Lewis (2) NRC Evaluations of LERs (2) NRC (RES, 00R, OSE) (3) Reviews of Specific LERs (4) Initial Preparation of Draft Report May 24-25, l979 (1) Reviews of Specific LERs
\ (2) Continued Preparation of Draft Report June 28-29,1979 (1) Review of Licensee LER (1) Duke Power Company Reporting Systems (2) Canmonwealth Edison (2) Reviews of Specific LERs Corporation (3) Review of Draft Report Jul y 18-19,1979 (I) Final Preparation and Review of Oraft Report Full Committee August 9-11, 1979 (I) Review of Report September 6-8, 1979 (1) Review and Approval of ;
Report I 1 l C-1 j 1
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1
l l I I, i l l l l l- i l l I I I l I l I l 1 l APPENDIX 0 ! l
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REVIEWS OF SPECIFIC l CLASSES OF LERs i i i l I I k i I O
) in the course of its study, the ACRS called upon its members and consultents i i for in-depth reviews of all LERs which were issued during the assigned Three- l year period and pertained to given components and systems. Presented in J this Appendix are summaries of the key observations made as a result of these reviews. The su nmaries are grouped according to the categories shown below.
l Pagg i
- 1. Separation of Control Rod from its Drive and BWR High Rod l Worth Events......................................................D-3 )
II. Unavailability of Vital Services.............................. .. 0-5 j l Ill. S y s t em s I n t e r ac t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0-8 IV . Failures Due to Flow induced Vibrations...........................D-13 V. W a t e r H am me r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D - 1 6 VI. Failures that Indicate Interdependence of Redundant Elements. . . . . 0-19 s O
'"j Vll. Adverse Interaction of Safety Systems and the influence of Human Errors......................................................D-21 Vill. Degradation of Valve Seat Quality and Other Valve Failures........D-23 IX. Leakaga Between Interconnected Fluid Systems..................... 0-24 X. Prcb l ems in Conta i nment i so l a t ion and Mon i tor i ng . . . . . . . . . . . . . . . . . .D-2 5 XI. Unauthorized Bypassing of Interlocks............................. 0-27 Xil. Failures of Containment Monitoring Systems Due to Adverse Env i ro nme n t a l Con d i t i on s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-2 8 Xill, inadequate Design Criteria..... ..................................D-30 XIV. Loss of High Pressure Coolant Injection (HPCI) and Reartor Core isolation Cooling (RCIC) Systems..................... ...... 0-32 D-l \v]
. l l
XV. Failures in Air-Monitoring, Air-Cleaning and Ventilating Systems............................. .........................D-34 XVI. Sample Line Stockage........ ......... ...................... 0-35 XVil. Ai r f l ow Control Damper Fa i l ure. . . . ........ .................D-36 XVill. Failure to Recognize and Correct the Cause of an Event. . . . . . . .D-38 XIX. Failure of Protective Devices on Essential Equipment..........D-40 XX . Failures of Diesel Generators................................ 0-42 XXI. Set Point Drift in Instrumentation...... .....................D-45 XXil. End-o f-L I f e and Ma i ntenance Cr i ter i a . . . . . . . . . . . . . . . . . . . . . . . . . .D-4 7
- XXill. Inadvarrent Actuotion of Safety injection in PWRs.............D-48 XXIV. uightning Related Malfunctions.......... .................... 0-50 O
D-2 I
A D-l SEPARATION OF CONTROL ROD FROM ITS DRIVE AND BWR HIGH ROD WORTH EVENTS General Description BAR rods are withdrawn downward to increase reactivity and are Inserted upward for scram. Should a rod become separated from its drive, it could later fall from a fully inserted position, completely out of the core, at an uncontrolled rate. Multiplication of rod reactivity worth can be caused, in a large core, by lack of symmetry, if the central rod of a freshly loaded core were to be fully withdrawn, leaving all other rods fully inserted, its worth could be multipiled to 20 times its normal worth. To maintain core symmetry, rods are wtPhdrawn manually in a rigidly controlled sequence, supervised by a dedicated computer designated as the rod worth minimizer, or by hard wired sequencing circuits. A few hours after a scram the growth of xenon-135 will alter the reactivity pattern of the core thereby changing the worth of Individual control rods. The operator is then required to use an unsatisfactory withdrawal sequence with the result that one notch of withdrawal (48 notches per rod) has resulted in a number of reactor periods shorter than 5 seconds. Frequency of Occurrence At least thirteen rod separation events and six rod worth multiplication events resulting in periods shorter than 5 seconds have been reported since January 1976. All restarts shortly following a scram at high power encounter reactivity distortions and rod worth multipilcation, although periods shorter than 5 seconds do not necessaril y result. ( At Dresden-2 on a single day, two rod separation events and one 5 second period were recorded.) Implications Regarding Safety The defense against a dropped rod depends on operating procedures that minimize the reactivity worth of Individual rods. The recurrence of events with 5 second periods, when a single rod is withdrawn a single notch, is evidence that the required controlled conditions have not been maintained. A rod capable of causing a 5 second period on being withdrawn one notch, If uncoupled and stuck In the core, could In falling a number of notches cause a sign!ficantly shorter period leading to fuel damage. 0-3 0-
Corrective Action Previous studies 22" have indicated that the probability of fuel damage from a rod drop accident is small. This study suggests that the withdrawal of a prescribed rod whose worth has boon magnified by high core xenon i concentrations may be more probable than the erroneous wi thdrawal of a l high worth rod under normal start-up conditions. A determination needs to I be made as to whether the high rod notch worth produced by high xenon concentrations extends over a large enough region of rod movement to cause fueI damage f ron a dropped rod. If damage can occur, consideration should be given to the use of a di f f erent control rod withdrawal pattern during periods of high core xenon concentration. Consideration should also be f given to the use of grouped rod withdrawal. By withdrawing rods one at a timo to the f ully wi thdrawn position, the exposure to a short period event resulting fron a rod drop is maximized and the defense minimized. By with-drawing rods as a group, the exposure would be minimized and the defense max i n. l ze d . l l 9
*See reference list following Chapter 4.
D-4 0
l i [ D-Il
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UNAVAILABILITY OF VITAL SERVICES Gcneral Description Residual heat removal makes use of general-purpose plant systems and services that are intended to provide defense in depth. On failure, however, these systems and services, may cause a scram which, in turn, could cause the vital services to be needed. The vital Instrument bus which is served by an inverter is an example. Failure of this inverter, usually caused by tho f ailure of a diode or other minor component, causes the loss of power to vital Instruments, valves and to logic systems which, in turn, cause scram to occur. This, in turn, causes these devices and services to be needed for residual heat removal. (This same effect results from the loss of a DC bus). The interactions which result from an Inverter failure, or DC bus failure, will vary depending on the reactor vendor or the Balance of Plant designer. The interactions often lead to a chain of unexpected events, resulting from the use of general-purpose plant services for residual heat removal rather than having a self-contained dedicated system for that purpose. (O) v Frequency of Occurrence From 1976 through 1978 there were 51 instances of inverter failure. Nine of these caused the reactor to scram. These nine are listed in the accompanying table. Implications Regarding Safety When an Inverter or DC bus f ailure causes scram, the reactor is forced into the residual heat removal mode of operation with reduced availability of vital services and of ten with attendent or major equipment damage or spurlous actua-tion of unneeded engineered safety features. Corrective Action Ten or more times por year it is required that residual heat be removed over a long term, with no recourse to an alternative mode of operation. Because of the associated implications, this study indicates that further consideration should be given to upgrading the reliability of vital D-5 A
1 services which support residual heat removal systems. Another solution would be to provide a system, dedicated to a single purpose, which would f not be required to serve other less essential purposes and, above all, would not be called upon to perform as a result of its own failure. Such a system would be self contained with its own services and coolant supply, would be regulred to be fully available on demand and to be immune to large fires and plant equipment f ailure, and would be designed to protect ! against most acts of sabotage. The most frequently encountered Interactions could be minimized by an Interim program that would identi fy all features which, when deonergized, would cause the reactor to scram. Those features would be removed from the general-purposo DC and vi tal AC buses, essential to residual heat romoval, and be placed on separate AC or DC supplies. O l 1 l 1 j
\
l l D-6 O
a '. i Inverter Failures Causing Scram l Nuclear Power Cause of Failure or Plant Component involved Remarks
- Salem i Output transformer RHR pump, diesel generator and charging pump failed to start
!' Crystal River 3 Electric component Decay heat removal valve was , l caused to close ; i Farley i Unknown None f Crystal River 3 Unknown Lost feedwater, scrammed on high ; pressure j i
- D.C. Cook i Blown fuse Lost power to scram logic. Scram caused steam dump and safety In-jection s
Prairie Island l Capacitor Steam generator level control failure caused scram Crystal River 3 Diode Excessive cooldown rate resulted fran stean dump D.C. Cook 1 Unknown Safety injection and Isolation caused by high steam flow Fort St. Vrain Switching error Isolation and two coolant loops dumped l D-7 v 4 I
D-1I1 SYSTEMS INTfRACTIONS Generaf Descrlp_tfon One of the .nore promising routes to the improvement of nuclear power plant safety is through improved understanding of the control of designed interaction between plant systens and the limitation and mitigation of the consoquences of undostrable interactions. The ovents behind a substantial number of LERs represent unique sources of systems interactions experlonce. Of particular signi f icanco are experiences gained during abnormal plant operations under conditions which were not anticipated in design. Such abhbrmal interactions can cause less than the normal conplement of hafety equipment to be available when nooded. They frequently occur durinj the rapid operational transients which follow abrupt changes in operating modo such as loss of of f-sito power, turbine trip, reactor scram, loss of primary coolant circulation, ECCS actuation, and transfer of vital services from one source of power to another. The ovent sequences of three LERs are described briefly at the end of this review. Each of the three was chosen not only because of i ts own unique event sequence, but also because at least a part of its event sequence is similar to parts of each of the other two event sequences. Exanination of the throo LERs together Indicates the potential for increased understanding of systems interactions, and their frequency of occurrence, by combined appraisals. The NGC Staf f has used LERs as a source of Information on occurrence of systems interactions on a case-by-caso basis. However, the nood for prompt filing of LERs often limits the scope and quality of the systems interactions Information which can be developed in time to bo included in the reports. Unioss such information is included in a follow-up report, much of the exporlenco potentially availablo from this type of reportable ovont will not be available in the LER files. Impilcations Rogarding Safety 4 Nono of the ovent sequences of systems interactions described in tno 1976-1978 LERs appear to have af fected the health and safety of the public. Numorous reportablo events, however, did involvo undesirable systems inter-actions, including many which occurred during rapid oporational transients D-8 O
(O) of types not generally considered during the course of development of plant designs. It would appear that increased use of the information available from systems interactions experience could lead to improvements in nuclear power plant safety. Corrective Actlon Procedures should be established which require nuclear power plant operators to report to the NRC Staf f details of all undesirable systems interactions. Planning for such procedures will require early identification of those systems most likely to be involved in undesirable interactions and tne courses the interactions are most likely to take; any Instrumentation needed to identify and follow the Interactions; and criteria which will define, in advance, the scope and detail of tne information appropriate for systems interactions analyses and evaluation. Systems interactions Experience
- 1. The first example of systems interactions experience involves the events which occurred at the Joseph M. Farley Nuclear Plant Uni r I on June 6, 1978. The event descriptlon starts with a Iightning strike near one of the station water circulating pumps at' a time when the reactor was running at full power. The strike resulted in a trip of the circulating pump; an upward surge of back pressure on the main steam turbine condenser; and an operator initiated rapid reduction in Q reactor power to avoid a turbine trip. The resultant transient on V steam generator water level and feedwater flow caused a trip of both of the steam generator main feedwater pumps. As the transient proceeded, a reactor scram occurred in response to coincidental low steam generator water level and steam flow greater than feedwater flow.
Approximately eight minutes after the reactor tripped, a second lightning strike resulted in the loss of of f-site power to one train of the 4 Kv emergency buses and one vital Instrument bus, and also resulted in the automatic start up of the emergency diesel generators and the reenergizing of the vital buses. At the time a second vital Instrument bus was also deenergized because of a blown fuse in its DC power supply system. The coincidental existence of two deenergized vital Instrument buses in the brief period between the loss of power to and the reenergizing of the first vital instrument bus, provided the logic for safety injection. The injection was promptly evaluated as inadvertent and the established procedure for its termination was implemented. All safety equipment was reported to have performed according to design, except for the failure of one borated water charging pump and one residual heat removal pump to start when required. The failure of this pump was unexplained at the time, n N
The events recited in the LER show that many systems Interactions had occurred, but the information required for their analyses and evaluations are lacking. As a result, much of the information desirable frcm these interactions cannot be obtained from the LER files. Anong questions that might have been answered are the appropriateness of: (1) the Instrumentation systems to provide reliable measurements of the behavior of fluids (including the locations and magnitudes of liquid inventories, during rapid operarlonal transients); (2) the control and protection systems to initiate actions during such transients in a manner that preserves the margins of safety intended by the designs; and (3) the capabi!ity of the control, instrumentation and protection systems to prevent undesirable safety injections when all needed safety equipment performs according to design.
- 2. The second example of systems interactions occurred at the Beaver Valley Power Station Unit No I on July 28-29, 1978. The LER event descrIptlon begins with a short circuit in the main transformer which caused a prompt trip of the main turbine-generator and all tnree reactor coolant pumps, and the scramming of the reactor. Oli spray from the pressure relief valve of the transformer Ignited and the fire spread to the transformer and its environs. During the transient that followed tripping of the turbine-generator, the excess stean dump valves of the steam generators opened on a signal of high average reactor temperature. The indicated temperatures in two of the three coolant loops dropped more rapidly than in the third and af ter about one minute into the transient this temperature mismatch provided the logic for high pressure safety injection. The LER concluded that al l needed safety systems had functioned as designed.
For more than three minutes into the transient, adequate of f-site power was available to satisfy the needs of the emergency procedures being used to bring the reactor toward a hot standby condition. Then all off-site power was lost, immediately, one of the emergency diesel generators loaded its associated 4 kV emergency buses thus providing power for one of the redundant trains of safety equipment, i The second emergency diesel also started and came up to spead but the I generator field-flash contactor failed to close and so the second I emergency train remained unavailable. Approximately eight minutes into the +r:nslent, safety injection was terminated and seven minutes Iater the second emergency diesel l generator was running and providing power to the second train of I emergency equipment. A few minutes later, off-site power was restored and forced circulation of coolant was reestablished. Compared to tne eariler example, the file for this LER contains substantially more information on the transient behavior of individual operating variables, particularly for the period following reactor D-10 O, l l l
l---__-_-.
,s
( ,) scram. This information is not adequate for comprehensive analyses and evaluations of systems interactions, but it does provide some insights into the character of the operational transients which occurred. Following the reactor scram, the steam generator pressure surged from 830 to 950 psi, and then fell back to 870 psi while the steam dump valves were open. During tne first four minutes of safety injection, steam pressure increased rapidly to 940 psi and then increased to a peak of more than 1000 psi over the noxt ten minutes (seven minutes beyond the termination of safety injection) when steam was released to the atmosphere to maintain pressure, and to provide the temperature gradient needed for thermal circulation. The water level in the pressurizer fell rapidly from 52% to 21.5% following the reactor scram and then gradually rose to 40% during the course of - safety injection. Pressurizer pressure dropped rapidly fran 2235 psi following reactor trip and rapidly increased to a peak of 2330 psi, whereupon the operator opened a relief valve to terminate the pressure rise. Pressure remained at the peak for about ten minutes and then gradually decreased. The information in this LER failed to provide answers to the types of questions raised in connection with the first LER. Other questions left unanswered include: (l) the cause of the short circuit in the main transformer, and the likelihood of similar short circuits developing elsewhere in the plant which could lead to equivalent or more serious results; (2) the likelihood of releases of hot oil to the atmosphere in the vicinity of other vital equipment, and (3) the appropriateness of the control and protection instrumenta-N tion systems to maintain safety margins intended by plant designs s_sl during transients, such as the one that followed the failure of the main transformer.
- 3. The third example of systems interactions involves the events which commenced on November 27, I978 at the Salem Generating Station Unit 1.
At the time the reactor had been operating at f ull power. The LER event description begins with the deenergizing of one of the vital instrument buses, followed by a reactor scram. The loss of the bus was traced to a failure of the output transformer of the DC/AC power ~ inverter which normally provides power to the bus. Deenergizing of the bus resulted in a spurious signal to the reactor control and protection system and provided the logic for the reactor scram. The
+wo steam generator auxillary feedwater pumps failed to start automatl-cally. One pump was started manually and provided the needed feedwater flow.
Tho loss of the vital instrument bus resulted in spurious indications of high steam flow to the logic for safety injection. These Indications alone did not call for safety injection. However, during the recovery per iod following reactor screm the average reactor coolant temperature D-li (%
l i I l 0 dropped below 543 F, conpleting the logic for safety injection. The safeguards equipment control systen was activated following safety ; injection. However, the emergency diesel generator, borated water i charging pump and residual heat removal pump of one train of redundant safety equipment failed to start when needed. The other train provided the services needed to eqtinue the shutdown of the reactor. The operable residual heat removal pump ran until the end of safety injection, at which time it was shut down. An attempt to start this pump ninety minutes later was unsuccessful, when its breaker failed to close. The second diesel was declared operable seven and a half hours after the loss of the first instrument bus, and residual heat removal was started some four hours later. The LER contains little information useful for substantial systems Interactions analyses. Tne information available in the LER does not provide answers to the types of questions raised in regard to the first two LERs. Also, whereas it seems clear from the earlier LERs that at least one redundant train of safety equipment was available at all times, it is not clear that this situation prevailed at Salem. O i l l l l l l D-12 0
l 0-IV FAILURES DUE TO FLOW INDUCEO VlBRATIONS l 1 General Description Flow induced vibrations occur in equipment and piping carrying single and two phase fluid. Flow induced vibrations are caused by vortex shedding resulting from rapid area change, buffeting due to random flow turbulence, fluid structures interaction Instability, leakage excitation, steady ; operation of positive displacement pumps and cavitating valves. The vibrations frequently cause f ailure of equipment, electrical wiring or components, pumps, valves and piping systems. The three major f ailure ' i mechanisms are high cycle fatigue, impact and fretting (wear). Vibration problems inside the reactor vessel manifest thanselves as worn guide tubes, loose guide thimbles, cracked shrouds, cracked nozzles and spargers. Charging pumps have been damaged by cavitation as well as turbulent buf feting vibrations which show up as cracked casings and welds. Vibrating valve Internals result (in closed and open positions) in cracked and worn valve seats as well as cracked welds. Other f ailures resulting from vlbration include loosened bolts, broken fittings, leaking snubbers, 4 (s /') damaged pipe hangers, broken wires, thrown switches, loosened relays, damaged printed circuit boards, loosened instrument terminals, radiation monitor- f ailures, f alse Instrumentation activation, and open breakers. Frequency of Occurrence There were about five reports of vibration caused events per month during the 1976 through 1978 period. Of the 171 events ' reported, only 20 were noted by the licensee as caused by flow. However, closer examination of the LER abstracts Indicates that most of the remainder should be included in this category. No system or component seems to be immune and the frequency of occurrence seems to be relatively constant. The largest number of events resulted because of cracked welds In small diameter i piping (< l Inch) and cracks in pump casings. The specific components or systems involved in each of the 17 LERs reported during this period are Indicated in the table below. I i s D-13 l \ l p
-+ +_- _____u--
l Systems and Components involved in Related Vibration Events Charging Pumps 23 Pump Vibration 3 Damaged Valves 10 Wolds 36 l Electrical Components 16 Miscellaneous Pump 7 Problems Steam Gonorator 5 Mechanical Devices 12 Reactor Vessel 13 Valves Oponed & Closed 4 Internals Air Systems 4 Jet Pumps 2 Piping Cracks 12 Instruments 10 Snubbers and 10 Hiscellaneous 4 Pipe Hangers The number of LERs resul ting f ran steam generator vibrat ion was f ar lower than expected. A scarch of steam generator events was not made to f urther clarify this situation. ImpiIcatlons Regarding Safety The problem of excessive vibration is important because it can often lead to damage of multiple components. Those events are frequently procursors to more serious events Inasmuch as contiruod occurrences can result in pipe cracks, failures of valves and snubbers, and damage to electrical and mechanical equip-monts. Other aspects of this problem include the of fects 01 vibration (as well as water hammer) on engincored safety features following severo transients. CorrectIvo Action
- a. Correctivo action taken falls into one of the following categories:
, repair and add vibration dampers (9), rodosign (49), strengthen (27),
return to servico (54), modi fy flow system (9), study (14), or preventive maintenance (9). The number of " repair and return to service" scoms entirely too largo. In those cases there is no specific action Indicated to assure that the events will not recur. 1 l D-14 O
1 l 4 l A ('] b . ThisDesigners study suggests that the following considerations should be reviewed. need more information on flow induced forcing f unctions so - that proper margins can be incorporated into the initial design. To describe flow induced forcing functions, more studies cf operating condt-tions should be conducted. These studies should be designed to determine the ef fects of turbulence levels, damping and danping ch0nges with service, fluid structures and equipment interactions, failure mechanisms and trade-of f s between sti f f ening and sof tening of systems. Electrical equipment and systems environmental qualification programs should be reviewed to ensure that high cycle vibrations do not jeopardize service. l 1 l LJ 1 l l 1 l 1 i
)
D-15 ( 1 m : o) 1 i l l
D-V WATER HAMMER General Description Tne term, water hammer, referred to in LERs as hydraulic transients, defines fluid transients resulting in pressure pulsations and includes rapid valve closing, pump start-up with par tially empty lines, flow into empty pipes, as well as stean bubble collapse and water entrainment in steam Iines. Water hammer has occurred in steam generator feedrings and
- piping, ECC systems, RHR systems, containment spray systems, feednator and auxillary feedwater systems, service water lines and steam lines.
Frequency of Occurrence Ouring 1976 through 1978, approximately one event per four reactor years was reported. This is fewer than observed by the NRC Staff 20 for the 1971 through 1978 period. A summary of the 39 reported events is given in the following table. This table does not include cavitation and two phase flow events. As may be noted, data in the table Indicate that most safety systems are subject te water hammer events. O Summary of Water Hammer Events System 1976 1977 1978 Residual Heat Removal ! 5 0 , Core Spray i I I l High Pressure Coolant injection i 2 I I Service Water 0 1 0 Main Feedwater 5 2 1 ! Main Steam i O I l l D-16 l l Ol i
I l l 1 I I Summary of Water Hammer Events (continued) ; System 1976 1977 1978 Isolation Condenser 0 0 2 Steam Generator (feedrIng) 5 3 i Containment Spray 0 0 l Reactor Aux!11ary Cooling 0 I I Liquid Poison 0 0 i Total 14 15 10 it is interesting to noia that no events were reported for the auxillary feodwater system. Implications Regarding Safety Water hammer has resulted in damage In almost every incident re >orted. O Many such occurrences are not reported because no damage occurs or is observed, or because the event occurred in a system not requiring reporting. In reported events, most of the damage has been to piping supports, snubbers and restraints. Feedwater system water hammer events have resulted in pipe cracks and valve fallures. Several Inadvertent safety injections _have occurred as a result of steam generator feedring water , hammer events. Such events i.e also important in connection with the actuation of engineered safety features following severe transients. The NRC Staf f has concluded that the frequency of water hammer events is unnecessarily large and that they involve systems of sufficient safety importance that changes in design and plant operation should be sought. Corrective Action
- a. Corrective actton has involved prImarlly the modtfIcation of operating procedures. Recommendations developed by the NRC Include the preparation or modi fication of Regulatory Guides, revision of Technical Specifications for operating reactors, and initiation of technical studies *# ***"
is a generic Item and is a part of the Task Action Plan A-1 23 D-17
( ) [
- b. The technical studies to date involve for the most part the development of computer models for piping systems, insuf ficient ef fort is being devoted to determining the pressure pulse amplitude needed as input to computer model analysis, in particular, additional studies should be made of the anplitude of the pressure pulse resulting f rcr1 column separation and steam bubble collapse, j t ,
.l \ i I l t ) i I l l I i 0' D-18 9
D-VI FAILURES THAT INDICATE INTERDEPENDENCE OF REOUNDANT ELEMENTS General Descript!on Occasional LERs have dlsclosed situations where there is an uninten-tional or unrecognized interdependence between supposedly redundant and independent components or subsystems. This disclosure may be more Important than the event, if correction of the dependency is not a part of the follow-up action, the value of the lesson may be lost. At Oresden-l on July 29, 1977, a diesel fire pump failed to start because both batterles were dead. Investigation revealed that there were two chargers for these batterlos, but both were Inoperative because of a single fuse failure. On Augus t 29, 1977, at Davis-Besse-I a common fuse disabled both makeup pumps. A slightly di f ferent aspect of the same general problem Involves disabling of a warning or safety o7vice by the event being guarded against. On March 24, 1978, Surry-2 had a borun dilution event because of a flow controller f ailure that also disabled the flow deviation alarm circuit designed to inform the operator of such !oss of control. Some de-pendencies are more subtle as, for example, the miscalibration of all four s power range flux monitors as a result of a f aulty pressure transmitter at Three Mlle Island-l on May 17, l977. The practice of callbrating all such channels agalnst a single reference is not rare, although it clearly com-promises the Independence of the measurements. Frequency of Occurrence There are not many occurrences that clearly point out problems of this type. However, this may only Indicate a low visibility in the reports, impilcations Regarding Safety A great deal of credit is taken in reactor risk analyses for the use of redundant measurement channels as well as defense In depth. Whenever the assumptions of these analyses are not valid (as, for e<Pople, Interde- l pondencies are not recognized), conclusions can be overly at timistic. The l Impact on safety varles with the type of error. A small callbration error 0-l9 n. A
l I l l may have mlnimal effect while use of a common power source for redundani , vital engineered safety features actuation could have serlous safety Impil-cations. Corrective Action Any evaluation of LERs provides an opportunity for, and should include, a careful and deliberate search for any indications that channel Independence or defense in depth principles are being violated. This is an area in need of further exploration. I i l l lllh , i 4 l i , 0-20 till I 1
i O D-VIi ADVERSE INTERACTIONS OF SAFETY SYSTEMS AND THE INFLUENCE OF HUMAN ERRORS General Description An Improbable or unexpected coincidence of events can cause systems interactions which lead to degradation of engineered safety features. The effects of such Interactions can be compounded oy the contributions of human errors which occurred prior to the initiating event. On September 16, 1978, a sequence of events occurred at Arkansas Nuclear One - Units I and 2, which revealed a highly improbable conbination of events and human errors which led to systems Interactions causing degradation in the possible performance of the emergency core cooling system and the containment spray system. The interactions involved inadequacy of relay settings (both protective and undervoltage relays), overload of auxillary power supplies, degradation of auxillary power voltage, failure of redundant control power voltage, and failure of redundant control power supplies -- all compounded by the existence of a series of prior human errors In engineering, design f-~g review, testing, and maintenance procedures. The e/ent das initiated by ( j the f ailure of an air operator 3clenold on a main steam line isolation valve of Unit 1. This incident was reported in detall in Information Notice No. 79-04 issued by the NRC Office of Inspection and Enforcement on February 16, 1979. Frequency of Occurrence This unusual coincidenc6 of circumstances would not be expected to be explicitly repeated; however, the event demonstrates the vulner' ability of the highly complicated auxillary power supply and control systems to adverse interactions. Many incidents of abnormal operating conditions and reactor scrams have developed as a result of loss or degradation of off-site power supply, impilcations Regarding Safety in the incident reported above, there were no engineered safety equipment mal f unctions or damage due to undervoltage. However, had the emergency D-21
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\s,/ ' >
core cooling system or the containment spray system been needed for safe shutdown, it would not have performed as designed because of the premature recircelation actuarton system valve actuation. Loss or deocadatlon of the off-site power supply may cause serlous malfunc-tion or unav ' ability of vltal safety systems in the presence of systems interactioi human errors. Corrective Action The settings of the protective and undervoltage relays were corrected 1, consistent with design load and voltage conditions Operating procedure
- established to ensure prmer loacing of the auxiliary power supplies.
To minimize the probabl!1ty of recurrence of similar events, this study su that industry experience with loss or degradation of off-site power suppiv should be reviewed to datermine the adequacy of typical power supply syst designs. O D-22 ( O 1 (
. . . . -- . - . . ~ . . . . _. -. . - . - . . -. .y p
f ,, D-Vill 3 4 DEGR/JATION OF VALVE SEAT QUALITY AND OTHER VALVE FAILURES General Description
.'here T have been many' reported failures of valves to operate on demand, in
.some cases the cause was failure of the valve control system or the power supply, but in.other cases'the failures were associated more directly with the valve or its operator, hkenanical failures include breakage of the valve stem, detachment of the valve disc from the stem, excess!ve friction preventing valve motion, failure of a torque limiting device, breakage of an operator linkage, and f ailure due to basic design inadequacles.
All kinds of valves-can be expected to suffer a degradation In the quality of their seating surfaces at'some time in their lives. Specific causes reported
~ Included erosion and pitting of surfaces; deposition of solids, including
-crystalline boric acid on the valve seats; the scoring of seats by material
, entrained in the flow; and the degradation of inflatable gaskets on large
- valves.
r Frequency of Occurrence i
. N.
Examination of 570 LERs In-the categorles of containment integrity and valves revealed that over 125 pertained to degradatico of valve seat quality. Implications Regarding Safety The specific significance of a given event depends on the particular valves involved. Since most safety related fluid systems depend on correct func-tioning of valves, the consequences of such failures can be extremely serious. Corrective Action Where seats had deteriorated in quality, they were reground and restored to
. spec i f i ed . per formance ' cr i ter i a . In other cases the appropriate corrective
. action, including replacement of the valves, was taken. Some examination and improvement of malntenance schedule ~ rates and of valve selection and design criteria seem appropriate in view of the large data base that has now been accumulated. :This is one area where routine system inspection and maintenance appear to be an effective and much needed activity.
t 4 D-23 ,
D IX 1 LEAKAGE BETWEEN INTERCONNECTED FLUID SYSTEM 3 GeneraI Descr_IptIon Proper functioning of Interconnected fluid systems (e.g., water / water, , water / steam and alr/ water) depends on various isolation devices to separate the systems and provide a neutral zone betwoon them. This review of LERs indicates a high probability that valves will leak at some time in their lives, hence, a high probability of Invasion of the neutral zone by higher pressure fluid. If such Invasion is unnoticed by the operator, the 1 inadvertent overpressurization of the lower pressure system, or the mixing of two diverse fluid systems, may cause other problems. ) Frequency of Occurrence Connecticut Yankee reported vapor binding of two auxiliary feedwater
- pumps due to steam leaks through check valves into pump suction lines.
Davis-Desse reported overpressur izati n of the RHR system due to leakage 4 of high pressure wa ter f rom a core flud tank through a stop valve. Vermont Yankee reported overpressurization of the RHR system due to primary coolant leakage through a check valve and an injection valve. Other plants indicated problems with valve operators due to water or oil in air supply systems. Implications Regarding Safety 4 Several cases led to some degradation of shutdown heat removal systems. Speci fically, two cases led to substantial overpressurization and damage of the RHR system, and one led to degradation of auxillary feodwa ter pump performance. Neitner of the two systems would have been available at that time if called upon to operate. Corrective Action At present the most common corrective action is to repair the damage and clean up the system. This study suggests that the adequacy of instru-mentation to monitor the neutral zone should be reevaluated. This study also suggests that existing design criteria for Interconnected fluid , systems should be reviewed and, where necessary, upgraded. Further methodology development and analysis of the consequences of a breach of interconnected fluid system isolation devices should be encouraged. D-24 t 0
h
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D-X PROBLEMS IN CONTAINMENT ISOLATION AND MONITORING General Description isolation of containment is essential to prevent escape to the environment of radioactive materials released inside containment. Originally, purging of containment was a practice 1Imited prImariiy to periods during reactor refueling, but today purging is conducted even during normal operations. This practice has a definite relationship to the maintenance of containment integrity and the potential for radionuclide releases. Problems of this kind, which are primarily limited to PWR installations, include the following:
- a. Several Instances in which the containment was purged while the containment particulate monitor isolation signal to the purge valves was bypassed,
- b. One Instance in which excessive alrborne releases under routine gerations led to a decision to reduce the frequency of con-
"ainment
.- purging. A factor entering into this decision was the pr esence in this plant of 920 mm diameter purge lines; the N9C objects to continuous purging unless tne lines are smaller than 200 mm diameter. Reduction in the frequency of purging reduced airborne releases but necessitated a reduction in the frequency with which the containment could be entered for visual inspection of safety-related equipment,
- c. In another instance of reduced frequency of purging, the bulldup of airborne radioactive materials within the containment kept both the gaseous and particulate monitors at or near full-scale Indica-tion. As a result, the monitors were incapable of detecting further increases in airborne activity that would have occurred from a significant increase in roactor coolant leakage.
Frequency of Occurrence Although only a few LERs were reported annually in these categories, these problems appear to be generic in nature and the number of events may be greater than those revealed by the LERs. D-25 (a
im2lications Regarding Safety in case "a", the purge valves would not have closed in the event of a loss-of-coolant accident. In case "b", the reduction in the aDility and frequency with which inspections could be conducted within containment may have led to a reduction in the overall safety of the plant. In case "c", the fact that personnel had to purge containment simply to prevent gaseous and particulate monitoring devices from going of f scale appears to be a design defIclency in the range of the instruments. Corrective Action These problems have not been completely solved,
- a. In the first example, the NRC Staf f recognized the importanca of such events and classifled them together as an Abnormal Occurrence.
Althougn licensees have modi fied their procedures to avoid ver, ting containment with the particulate monitor isolated, the chosen 'lx appears to be administrative rather than technical. The ACR$ believes that the events leading to these LERs warrant a dettsiled review to determine how such a potentially unsafe situation could have been overlooked. This Is parrIcularIy urgent in Iight cf the f act that some of the events do not appear to have violated any Technical Specifications, even though they resulted in the degra-dation of the containment s leakage retention integrity and appear to have involved violations of basic safety precautions,
- b. Problems of excessive airborne releases from a nuclear power plant appear to be better addressed by directing attention to the source of the problem (or its control) rather than to approach it on the basis of reduced purging. A risk / benefit assessment should be conducted to evaluate the full extent of the potential impact of reduced Inspection frequency.
- c. It is anticipated that problems related to inadequacies in the operating ranges of radiation monitors inside containment will ultimately be corrected on a generic basis through the application of Regulatory Guide 1.97. As of this date, however, this Guide has not been implemented by the NRC Staff. The ACR$ has con-sistently urged that more attention be directed to providing instruments with ranges sufficient to assess a full range of conditions, including those accompanying a major accident. This study indicates that this matter needs further attention.
D-26 O
4 D-XI UNAUTHORIZED BYPASSING OF INTERLDCKS General Description The deliberate bypassing of containment building access door Interlocks in violation of Technical Speci fications was the cause for numerous LERs. These actions took place during ref ueling operations or at other times dur-Ing reactor shutdown. Cases cited Included the routing of lines to provide breathing alr for personnel In restricted ventilation areas, the transfer of large objects into the building, etc. Numerous LERs also resulted from Interlocking linkages being out of adjustment or having failed. 4 Frequency of Occurrence Cases occur during reactor shutdown without any special pattern in time. Twenty-seven failures in interlock Integrity were included in 164 LERs in the category covering containment Integrity, leak rate, containment valves, and violation of containment. Implications Regarding Safety These activities generally violated the Integrity of the containment. Corrective Action in general, corrective action included: (a) Reprimand!ng of the unployees involved; and, (b) Retraining of employees as to proper procedures. There is little indication, however, of a decrease in the number of such occurrences with time. Access to the interior of the containment building is necessary under some circumstances where containment integrity is dif ficult to main- ! tain. Provisions to allow access with reactor operator authorization should be available under controlled conditions such as those contained in Regulatory Guide 1,47 " Bypassed and inoperable Status indication for Nuclear Power Plant Safety Systems". All plant personnel should be instructed in their implementation. D-27
l i D-XlI FAILURES OF CONTAINMENT MONITORING SYSTEMS DUE TO ADVERSE ENVIRONMENTAL CONDITIONS l General Description
- Monitoring systems have temperature and moisture limitations assigned by their manuf acturers for correct and rellable operation, and designers of I
heating, ventilating, and air conditioning (HVAC) systems are constrained l to include these limitations in their design considerations. When HVAC systems fall to maintain spaces housing conta'nment monitoring systems within specified environmental conditions, the performance of these i monitoring systems becomes degraded and this may lead to total failure. l 1 Frequency of Occurrence l l A limited review of reports of monitors falling within containment l (including those designed to monitor post-accident conditions) because of i
- adverse environmental conditions showed that nine such events were reported I l during 1978. Seven of these involved post-accident radiation monitors - l 4 six of them at a single plant, which also had another monitor failure.
]
The f ailures were attributed to excessively high ambient temperatures in the spaces where these monitors were installed. During a nine-month period ending April 1978, one post-accident monitoring system was inoperable 55% , of the time. Failure of a post-accident hydrogen analyzer from similar )
- causes occurred at another plant in June 1978, and a containment atmosphere l
- particulate monitor f ailed due to high ambient humidi ty in a third plant in September 1978, implications Reajrding safety An instrument that fails because of high ambient temperature or other unfavorable environmental conditions cannot be depended upon for post-accident monitor ing (e.g., actuation of isolation systems may not occur);
and, without hydrogen analyzers an explosive mixture could develop with- 1 out the knowledge of plant personnel. I D-23 e'
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1 l l i
j Corrective Action ( The cause of these failures was attributed to a lack of adequate ventila-tion and the remedy chosen was to provide more air cooling and a reevalua-tion of systems performance. This study suggests a need for (a) a thorough j reexamination and reevaluation of the design parameters used for estimating 1 the potential environmental conditions to which such systems are likely to be subjected, and (b) a reevaluation of the procedures used to qualify j j monitoring systems to meet unfavorable, but realistic, environmental conditions. i l-1 i I 1 l l l 0-29 ( I
I l 3 D-XlIl INADEQUATE DESIGN CRITEdlA General Description An example of inadequate design criterla ls found in an event occurring i dt the Joseph M. Farley Nuclear Power Plant on Jar'ary 26, 197S. A nigh water level of the Chattahoachee River caused leaks through a hole in a pump base plate, a gland leak-off line, and cable penetrations, which in i turn caused flooding of the structure housing them. The reactor had to be shut down and emergency recirculation to the pond (the ultimate hear sink) ) had to be initiated. Tne designers had routed piping without fully I considering high river water level and the design regulroments associated I with it. In another case, galvanic corroslon of flange bolts on service water pumps by salt water caused high vibration of the pumps and required repair of alI eight pumps for this station (Salem, September 7, 1976). Another case in point is the f ailure of the condensate pipe at Browns Ferry which had the potential for serious flooding ef fects on critical equipment. LERs from San Onofre indicate a continuing problem wi th , failures of feedwater flow straighteners during normal operation. l F_requency of Occurrence The cases quoted above represent only a sampling of the many occurrences traceable to errors in design criteria due to the lack of consideration for important environmental factors or f ailure to identi f y a function that 1 the equipment wiii be called upon to perform. 4 Impiications Regardin LSafety in many cases it is apparent that the full implications of particular events are not appreciated. The implications regarding saf ety from this l class of events range from trivial to considerable. Inadequate design ' criteria can, for example, result in the unavailability of supposedly redundant systems or trains. Likewise, failure of Internal appurtenances in piping systems under normal operation raises concerns with regard to their structural adequacy under the dynamic forces associated with blowdown. Internal component detachment and subsequent relocation may result in the blockage of isolation valves or other safety systems which are called upon l fo mitigate an accident, j 1 l 0-30 0
r i I l l i ( ()CorrectiveAction
- a. The problems cited above have occurred basically due to inadequate design. The corrective action im many cases has been to design around the problem. This study indicates that this approach does not provide an effective solution.
- b. Experience gained f rom past errors should be factored into upgraded design criteria for systens and components. A review of criteria for the design of heat removal systems should be undertaken to ensure that proper consideration is given to all possible environmental conditions j and all functional requirements. Equipment weakness as revealed during the generally less severe conditions of normal operation can be important
. Indicators of how the equipment and systems might be expected to perform under potentially more severe accident conditions during which such failures might not be acceptable. 2 4
- c. This study also indicates that greater attention should be given to Improved design criteria to prevent non-saf ety system f ailures from ;
causing f ailures in safety systems. ; I i i i i D-31 1 j'
' D-XlV 4
LOSS OF HIGH-PRES $U'lE COOLANT INJECT;0N (HPCI) AND REACTOR CORE ISOLATION COOL ING (RCIC) SYSTEMS leneral Descrigt !,on_ There is a variety of problems associated with the operation of HPCI and
- RCIC systems. These include f ailures to isolate when called upon, as well as inadvertent Isolation. The former are due to a range of events Includ-ing valve leakage, improper valve lineup, and inadequacles in electrical supporting systems. The latter include set point delft and relay malfunc-tions. A review of LERs also shows that losses of the HPCI and RCIC systems can occur as a result of failures in air ventilating systems.
This type of failure is highlighted here. The basis of the problem is that the areas through which the steam lines to these systems pass are equipped with temperature sensors that are designed to isolate the systems in case there is a leak in the lines, if there is a malfunction in the ventilating systems for these areas, or a sudden change in the outdoor temperature ethich causes the sensor to indicate a steam leak, the two l systems are automatically isolated. Loss of ventilation due to failure of off-site power may also lead to similar events. The first problem is 4 unique to BWRs; the second is probably generic to all reactors. F_reguency_of Occurrence Nine events resulting from f ailures in air ventilating system were reported in 1976; eleven in 1977, four in 1978. Mplications Regarding Safety dith the systems isolated, emergency core high-pressure coolant injection is not available. Corrective Actton
- a. The cause_of the failure was clearly determined. The fix was to increase the flow in the ventilation systems for the affected areas.
This does not appear to address the basic problem, however, since temperature sensors do not respond solely to a leak in the steam lines. D-32
i i I
- b. A more permanent solution to the problem appears to be needed. This will probably require the development of leak sensors which are not dependent upon temperature to detect leaks, I
I I i 1 1 1 l i 1 l i i D-33
l l l l l D-XY i FAILURES IN AIR-MONITORING, AIR-CLEANING, AND VENTILATING SYSTEMS l General Description PWR and BWR installations are eaulpped with a variety of air-monitoring, air-cleaning and ventilating systems. LER data show that 14% of the LERs reported durIng calendar years 1976, 1977, and 1978 pertained to such j systems and, for BWRs, over hal f of the f ailures were in the equipment installed to monitor the performance of tne air-cleaning systems rathe, i than in the systems themselves; for PWRs, the fraction of failures per-
- talning to air-monitoring equipment compr ised over one third of the to tal .
Frequency of Occurrence l l
, Data show that in excess of 350 LERs pertaining to air-monitoring, air- l l cleaning, and ventilating systems are reported each year, 9ecause BWR I
installations have more such sys tems than are presen t in PWR installations, the average number of LERs reported per BWR per year is higher than for PWRs. Implications Regarding Safety
- The proper performance of air-monitoring, air-cleaning and ventilating systems Is important to the Integrity of primary containment and the con-trol of airborne particulate and gaseous releases from nuclear power plant 4
Installations. The proper performance of ventilating systems is also important to the operation of other key safety equipment such as the High Pressure Coolant injection and Reactor Core Isolation Cooling systems in BWRs and waste-gas processing systems in PWRs. The proper performance of air-monitoring equipment is essential to the avoidance of the bulldup of hazardous concentrations of gases, such as hydrogen, and to the assessment of the potential impact of environmental releases. Corrective Action Corrective actions for many of these f ailures appear to be inadequate i since the f ailures continue to occur. The f act that about 50% of them can be attributed to human error is of particular concern. The high frequency of air monitoring equipment failures clearly shows < a need for the development of more reliable monitoring systems. D-34 O l t i
P D-XVI SAMPLE LINE GLOCKAGE General Description For convenience, security, personner safety, and the need for a controlled environment, sample collection and analytical instruments are frequently located remote from appropriate sampling locations; sampling the gases emitted from tne top of a tall 3 tack i s a f ami l l ar examp l e . This neces-sitates the use of long sampling lines between the sampling point and the 4 collection / analytical point. Such lines are subject to failure from condensa-tion and freezing, corrosion, erosion, and blockage from solid materlais. Frequency of Occurrence Over 50 LERs submitted during the study period were concerned with failures of atmospheric sampling systems due to leaking or plugged sampling lines or an accumulation of moisture within them. This problem is more prevalent in BWRs; over 40% of the failures of sempilng systems in this type of f acility were due to this cause, implications Regarding Safety With sampling lines blocked or otherwise Inoperative, atmospheric monitoring systems can be disabled and unmonitored releases of radioactive material to the environment can occur. Corrective Action The remedial action taken in the case of most of the above occurrences was to unblock the affected lines by blowing them clear. However, the recurring nature of this type of failure suggests that the base cause is a design deficiency or fallure to follow established procedures. A more , permanent solution, based on a new or upgraded design for sampling lines, I is needed. ; 1 l t
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1 i l \ l l D-35
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1 f I
. -____ -_-___-__- -_- - __ -_ -_- - ________ _-- - _ ___ _-_ _ ___-__-____-___-._______ - - ____ - __________- - - - ________--_-___ _ -___l
D-XVII 1 AIRFLOW CONTROL DAMPER FAILURE Oqn_gr_a[20scr,[gt_{gq Dampers fall to open, fall to closo, or leaX after closure. The principal causes are improper Installation, nonfunctioning sensors and controllers, Incorrectl y pos itioned damper actuators and linkages relativo to control motors, and degradatlon of soaling surfaces by dirt and mechanical damage. 1 Fr_oguencL of Occurrence , Dampers are notoriously subject to failure. About 25% of LERs involving j f ailures in heating, ventilating, and air-conji tioning systems are dampor
- failures (e.g., 54 out of 222 events during 1976 through 1978). Ton of the 54 woro detected during reactor operation at 40% to 100% of full power (in spite of the fact that most testing of th!s class of equipment takes I place during scheduled shutdowns). '
Imglicarlons Regardinq Safot_y Safe ty implications are ser ious when nonf unctioning dampors are associated i with Engineered Safety Feature (ESF) systems. The safety Impi(cations aro i loss serious when the dampers f unction correctly but loak, the degree of seriousness being proportional to the leak rara. Half of the 54 LERs involving dampers were associated with ventilation of the containment and containment-associated spaces and a quarter were associated with control room and control building vontilation systems. When not in ESF systems, damper failures often caused inadvertent plant trips because of Instrument failures from excessive environmental condltions. Frequently, LERs result from excessive or deficient air flow caused by damper looks that may not j . have been sertous otherwlse. l Corrective Action l The corrective action has been to repair the defective parts. Provention. I however, calls for more frequent and more searching Inspection and main-i tenance procedures. i 0-36 j \ O
1 This is an important generic problem that calls for (a) the design and { production of more reliable dampers and damper controllers, (b) Im-
- proved methods for monitoring the precise position of damper blados tnat i are independent of the damper control mechanisms, and (c) improved handling i of this equipment during prelnstallation storage at the sito and during in-i stallation, i
i e I i i 4 2 s i D-37 k
L 1 l D-XVI11 l FAILURE TO RECOGNIZE AND CORRECT THE C4JsE OF AN EVENT l gener_al Description Frequently an LER will show tnat the licensee has f ailed to recognize the root cause of an event. The result is that the original probl em rema i ns . Sanetimes, as in tne case at the Mllistone event of July 5, 1976, the solution to tne original problem introduceJ a new problem. In this case the lwnediata corrective acticn was to raise the set point of the loss of power undervoltage relays with the result that on July 2I, 1976, attempts to load an emergency bus after diesel start caused Icac shedding witn i recovery possible only oy manually restoring of f-site power (by lower!nu l the undervoltage set point ajaln). Subsequent actions nave apparently cleareJ up most of the contributing factors, except that there does not appear to have been any attempt to determine if it is proper to have control circuit fuses of sucn size that failure of a contactor to close wilI blow the fuse and tnerefore toislly disable the contactor so that it cannot subsequently operate even if proper voltage is restored. In another case, an unintended discharge of radioactive gas was discovered by a routine gas volume calculation at Yankee Rowe. Tne cause was deter-mined to be a leaking diaphragm in a valve which the LER stated "... Indicated attack by a petroleum cased substance. The diapnragm was replaced in kind." Freauenct_of Occurrence Many LERs involve this type of problem. Abnormal cyr.b l en t conditions are frequently blamed for malfunctions when a closer examination would reveal that designs have not been suf ficien tly robusi to wi thstand them, in other cases, "operatcr error" was accepted as the cause when the design Invited such error. An outstanding example occurred at Vermont Yankee on February 17, 1976, when two RHR valves would "...not stroke electrically. Following manual operation freeing the seat, tne valve performed satisfac-torily repeatedly " Tne cause was stated as "... procedures were not followed correctly. Manual operation is required prior to auto operation as the seat normal ly sticks due to temperature of fects." A design that requires manual freeing of a valve before it can be operated electrically appears to be totally unacceptable. D-38 i
l 1 l 1 l l Implications Regardin_q_ Safety ,
}
The implications of such events can range from trivial to quite significant. I Corrective Action ! Replacement in kind is not an adequate corrective action when the failure was caused by an unexpected condition. The proper corrective action should include analysis of the origin of the condition and the development and implementation of measures to forestall its repetition. D-39
D-XlX FAILURE OF PROTECTIVE DEVICES ON ESSENTI AL EQUIPMENT General Descrigtlon A large number of LERs have reported failure or incapacitation of essential equipment as a result of failure of fuses or other devices Installed for the express purpose of protecting that essential equipment or Its services. The systems affected exist throughout the plant and include the plant control system, the plant protection system, and the engineered safety features. Particularly vulnerable are actuators that require power in order to drive motors, operate valves, and the like. The failures are not limited to overcurrent devices, but occur in equipment such as torque lim!ters, overspeed protectors, and other Interlocks and may be caused by improper applications or adjustments as well as component failures. Frequency of Occurrence No attempt was made to tabulate all such LERs, but a computer search by the Nuclear Safety Information Center listed $99 LERs In 1976, 1977, and 1978 that referred to protective devices and circuit breakers. There could be some overlap, but the number is substantial. Implications Regarding Safety Safety implications arise because the expected failure rate of essential equipment may be unjustifiably optimistic if reliability estimates do not account properly for failure of the protective devices. There have been several reports of loss of supposedly redundant devices through failure of fuses common to circuits Intended to be independent. If the failures are due to improper selection of fuse sizes or adjustment of other devices, there is an increased probability of common mode failure of redundant l vital services even though they are physically Independent. Corrective Action l The usual corrective action has been to replace the failed fuses or to I readjust the adjustable devices. Where the dlsabling of such equipment l l l l D-40 l O
may remove or substantiaily degrade vltal services, it appears that the basic criterla for protecting the equipment should be reexamined. Perhaps the rules for such equipment in nuclear power plants should be different than those currently used in standard electrical practice. O I l i I 0-41 , l lO '
I D-XX FAILURES OF OIESEL GENERATORS General Description The standby emergency power system in a nuclear power plant includes diesel generator units which are started on loss of off-site power, and are expected to provide the power required to shut down and maintain the plant in a safe condition in the continued absence of off-site power. Frequency of Occurreng
, A total of 470 LERs (1976-78), Identifled with diesel generator problems, were obtained from the data file of the Nuclear Safety Informatlon Center (NSIC). The incidence of diesel generator problems has historlcally been much higher than had been expected and has been the subject of at least two previous studlesl7,18 The large number of LERs, coupled with the components and types of fallure, makes the analysts sulted to computer techniques. For purposes of this revlew, the data from the 470 LERs were coded on the basis of plant, unit, manufacturer (when available), date of
, event, test or demand when f ailure occurred, mode of f a il ure, componen t involved, and the type of plant. These data were stored in a word processor system and subsequently analyzed. 4 The LERS were first grouped on the basis of plant, leading to an Identl-fication of those with an unusual number of LERs In this category. The data for a particular plant depend on the Technical Spectfications, the number of generating units, and the number of diesel generators. Big Rock Point, Dresden, Hatch, M11Istone, Calvert ClI f f s, and Farley reported relatively high numbers of LERs while tne number of LERs fran some 4 other plants were remarkably low. Additional follow-up 1s needed to determine the reasons for these differences. On a total basis, the year 1978 shows an increase of 34 LERs over the prevlous two-year average of 145. Additional study is required to deter-mine if this is the start of a trend. Contributing factors include additional plants on line and aging of older units. i D-42 0
g A total of 302 (64%) of the LERs was associated with Incorrect operatlon during tests of the diesel generators. A total of ill (24%) of the LERs was reported during maintenance Inspections. Fa l l ur e-on-demand ( act ua l operating conditions) accounted for 57 (12%) of the LERs. The fallure-on-demand percentage Is slmllar to the ll% reported in reference IS. The reported distribution of the failure modes was as shown below: Fallure to start - 105 (22%) Fallure to run - 162 (34%) FalIure to connect electrlealIy - 31 ( 7%) Not available (maintenance time over limit) - 144 (31%) Design problem - 28 ( 6%) Total 470 (100%) As may be noted, a large majority of the events are distributed over the start, run, and out-of-service categories. The extended maintenance time is of concern in that, under such conditions, the diesel generator would not be available if an emergency occurs. Recognition of these potential problems is one reason that redundancy of d!esel generator units is required, Within the NSIC computer bank, the LERs were further classifled into twelve groups according to the component involved. The distribution of failures was as shown below: Failures Percent Engine 52 Il Starting system 40 8 Fuel oli 47 10 Lube oil 27 6 Cooling system 39 8 Governor 33 7 Generator 21 4 Voltage Regulator 3 i Operator error 46 10 Swltching 46 10 Control system 50 ll Other 66 14 Total 4 76 IO6% As may be noted, with the exception of the excellent voltage regulator performances, the remalnder of the components were relatively uniform as contributers to failures, l D-43 l
In addition to providing an overall picture, the extracted LER data can be sorted on the basis of Individual plants and diesel generator units. Failure modes and individual component failures can be identifled and monitored for historical trends and devlations from the aggregate. Impilcations Regarding Safety Emergency electric power, supplied by one or moro diesel generators whlch are considered part of the engineered safety features of the plant, is required following loss of off-site power. Loss of both of f-site and on-site electric power will prevent engineered safety features from performing as expected. A significant number of malfunctions related to diesel generator systems has occurred. The consequences to date have been minimized because of redundant systems, but complete loss of AC power would be extremely serlous. Corrective Actton The performance of the diesel generators and associated systems during 1976 through 1978 was similar to that reported in 1974. Correctable problems (e.g., associated with maintenance) can be Idontlfled from the LER data both on generic and specific plant bases. There is evidence that test conditions do not always duplicate those encountered in a " col d" start. A review of test procedures should be Inlflated to determine 1f performance reports are biased. O l D-44 0
l D-XXl SET POINT DRIFT IN INSTRUMENTATION General Description An unplanned change in the set point of an instrument is referred to as set point drift. The effect of set point drift is to alter the actual ' value of the measured parameter at which a paritcular action is to occur. When the drift is of sufficient magnitude to cause the set point to be out of compilance with Technical Specifications, the event is required to be reported. Frequency of Occurrence There was a total of 775 such events reported during the three year period covered by this study (see accompanying table). These represent approximately 10% of ali reported LERs. Implications Regarding Safety Safety instrumentation channels are redundant and, in general, the LERs O stated that the set point drif ts of the associated redundant instruments were within Technical Specification limits. A review of each of the above ' ' 775 LERs would be required to determine whether there were any instances of simultaneous drift of redundant channels beyond Technical Specification limits. Such situations could have important safety Implications. Corrective Action For those cases where the set point drif t is caused by conponent f ailures which occur at random within an instrument, the corrective action is to make the necessary repair, recalibrate, and to restore the instrument to service. In other cases, the margin between the selected set point and the Technical l Specification limit is not sufficient to allow for the normal instrument inaccuracy. The corrective action for these cases is to increase the l margin between the selected set point and the Technical Specification limit to accommodate the inherent instrument inaccuracy. l l 0-45 f A
Licensee Even_t Report _s_ lnv_olvinq_ Set Point _Drlft For Operating Plant _s," 1976 1977 1978 Component involved TOTAL BWR PWR BWR PWR BWR PWR Nuclear in>+ruments 12 16 19 4 10 4 65 Radiation Monitors 3 0 3 1 5 2 14 Sensors, flow 8 7 23 13 13 16 80 Sensors, level 23 14 31 25 27 31 I5I sensors, pressuro 87 21 103 15 58 25 309 Sensors, temperature 7 7 13 5 13 3 43 Instrument, protective 2 5 2 0 2 1 12 instrument, miscellaneous 2 l 0 2 0 2 7 Relays 7 0 7 5 6 7 32 Valves 6 4 9 17 5 6 47 Miscellaneous I I 2 2 3 1 10 Subtotal 158 76 212 89 142 98 775 Total 234 301 240 775
- Based on ORNL/NSIC data, i
l D-46 O
l l l l D-XXil END-OF-LIFE AND MAINTENANCE CRITERIA General Description A review of events related to leakage of valves, pumps and seals shows I that the LERs often Indicate as the cause of failure: end-of-life, normal wear, and maintenance activi ty. The frequency of such events suggests that the criteria used for setting up maintenance periods and life expectancy require further review and evaluation. Frequency of Occurrence Of sixteen LERs related to leaks in valves, five can be classed in this category. Of ninety-two LERs concerning leaks through pumps and seals, thirty-one were related to causes such as end-of-llfe, normal wear, normal service, and inadequate maintenance. Implications Regarding Safety Unexpected equipment failure (due to lack of timely maintenance) could O lead to decreased safety margins. Corrective Action
- a. At present the response to most such events is repair or replacement of the particular components. At least one utility (Duke Power Company) carefully studies all LERs to identify maintenance needs prior to failure,
- b. Actual plant experience should be factored into the process of determining end-of-life statistics of various components as well as the statistics of maintenance periodicity. Exchange of such information and experience generated in this area should be encouraged and distributed to other plants with similar components. Some programs, such as those at Electric Power Research Institute and Southwest Research Institute, are directed towards accomplishing this goal.
0-47 0
D-XXllI l lNADVERTENT ACTUATION OF SAFETY INJECTION IN PWRs General Description Operator errors, Instrument malfunction, and reactor transients and trips have been reported as the cause of inadvertent actuation of the safety injection system. Frequency of Occurrence This review of LERs has identified at least forty cases of inadvertent actuation of safety injection. Approximately one-fourth of the events sampled were due to operator error. The problem is repetitive in nature; at several facilities the problem has a long history. The vast majority of events occurred in Westinghouse Nuclear Steam Supply Systems, whereas plants supplied by other vendors had few or no reported events. Implications Regarding Safety Safety injection systems are required to operate during loss of coolant accidents and other severe transients that require borated watar addition to the primary system. Inadvertent actuation of the system injects cold borated water into the reactor when It is not needed, subjecting injection nozzles to thermal stresses and requiring removal of boron fron the primary system before start-up. The present number of occurrences is probably not significant with respect to the ef f ects upon the primary system; however, operator response to { an inadvertent saf ety injection involves termination of the injection and re-setting of the injection signal. This generally occurs within one to eight minutes following the start of injection and follows a check of other plant status Instrumentation. Repeated operator exposure to inadvertent safety injection and its termination may produce en unacceptable response in cases where the injection is required to provide core cooling water. The NRC and utilities should study this problem to see if the number of inadvertent injections can be reduced without jeopardizing operation of the system when needed. The response of the operators to the Injection should be reviewed to assure that they have sufficient information avall-able in an easily understood manner to allow them to make correct decisions concerning early termination of these events. l D-48 O
( Corrective Action f The immediate action of plant operators is to correct, where possible, the l problem that caused the inadvertent Injection. in general, this approach ! does not address the root cause of the problem. in many cases the root f cause has not been identified. Abre work is needed in this area. ( l 1 i I l O l l \
\
)
l { l i l l 1 0-49 l l o ) 1
D-XXIV LIGHTNING RELATED MALPJNCTIONS General Description Lightning strokes generate current anc voltage surges which can directly or indirectly cause equipment mal f unction or failure. Because of the randon nature of lightning, its effects are difficult to control. The , effects of such events manifest thanselves in a variety of systems ranging from Instrument malfunctioning to grid Instabilities. Frequency of Occurrence This study revealed that lightning has led to the submission each year of 10 to 15 LERs for commercial power plants. The following are examples of electrical failures caused by lightning. Stroke Location LERs Malfunction Stack 8 Flow Indicators (6) Other instruments (2) Power Line 4 Off-site Power Loss (3) Relay Operation (1)
~~
Facility 2 Relay (I) Instrument (l) Unspecified 7 Breaker Trip (3) Instrument Loss (2) Off-site Power Loss (1) Fuse Fallure (I) l It is likely that not all of the lightning-related events were identifled in the LER files as explicitly caused by lightning. O-50 O
Impilcations Regardina Safety The potenti ally large curren ts and vol tages stress equipment to design limi ts, and beyond, and protection is di f f icult. Operating conditions often change suddenly, requiring the protective and safety features to operate correctly under emergency condi tions. Events of this nature may cause tripping of all units at a multi-unit plant and may have serious implicarlons regarding grid stability, availability of of f-site power and the adequacy of on-site emergency power. Other problems caused by lightning include on-site f ailures of trans formers, circui t breakers and s tack monitors; explosions in of f-gas systems; and f alse signals for actuation of HPCI and ROIC systems. Redundant diesel generator systems have prevented major problems to date. Corrective Action Stack flow indicator Instruments require better lightning protection if they are to function following a stroke to the stack, "Cause unknown" or " voltage surge" events should be critically examined if lightning is a suspected or possible cause. This can lead to improved protection, including installation of surge suppressors, for the equipment involved. Since loss of off-site power is a possibility during lightning activity, consideration should be given to starting at least one diesel generator and running It for the durat ion of a storm. This could avoid the starting failure of the diesel generator and associated equipment, and improve the rollability of the system. D-51
i O i AoPEN0lX E i
)
O STATISTICAL ANALYSIS OF LERs: A TRIAL STUDY O
introduction Approximately 8700 LERs were submi tted by the licensees of U.S. commercial nuclear power plants during the years 1976, 1977, and 1978. For several reasons, the number of LERs varies from unit to unit. These variations are important, because, rightly or wrongly, they are of ten viewed by government agencies and the public as indications of relative safety. While such variations may be Indicative of actual differences in safety among nuclear power units, they may have other explanations, it is therefore important to understand all possible explanations and their contributions to variations in the numbers of LFRci from Unit to unit. Certain dif ferences in the frequency of submission of LERs from unit to unit will occur as a result of the apparent randon nature of the evenrs being reported. Because of this " randomness", it is possible -in fact, probable--that, even among identical nuclear power plant facilities witn identical failure probabilities, there will be variations in the reporting rate for LERs. In reality, however, variations beyond those due to "randonness" wil l frequently be observed. The reasons for such non-random variations include the facts that: (l) Technical Specifications and Iicense provislons vary among nuclear power plant facilities, because of di f ferences in reactor suppliers, architect / engineers, and constructors, and changes in designs over the years. These variations cause differences in the reporting requirements among facilities. (2) There may be a tendency at some f acIIitles to report events more readily than at others in cases of marginal reportability. This consideration pertains to events other than obvious " reportable occurrences" (R0s), which all licensees must report l*. This tendency can also change with time. (3) The occurrence of an event may affect the probability of future events. Repair of a faciIity component or improvement of a defIciont procedure may significantly reduce the likelihood of an associated event. On the other hand, ineffective corrective action following an event may result in its repeated occurrence. (4) The mode of operation (e.g., on-line or shutdown) affects the frequency of various kinds of inspections and the susceptability of systems to random failures. The amount of reactor down-time, for example, may af f ect the frequency with which LERs are submitted.
*See reference list following Chapter 4.
E-l b________._._ ____ _ . . _
(5) Misinterpretations by licensee or NRC personnel involved in the preparation, submission, and processing of LERs can affect relative reporting frequencies among reactor systems. (5) At some multi-unit power stations, such as Oconee and Browns Ferry, events which involve plant systems or components common to all units, such as swing diesels and electrical switchyards, are filed in the NRC data bank under the docket number of the first unit. (7) The actual presence of more safety-related deficiencies in a sys tem at an Individual facility should result in more frequent submission of LERs. Differences in the number of LERs due to this cause would be a measure of relative safety. Although the above factors af f ect the frequency with which LERs are reported, their ef fects are of ten relatively small . Frequently the variations produced by these effects are too small to be distinguished from those occurring on a randon basis. For example, the Point Beach Nuclear Station in 1976 had 11 reportable occurrence LERs for Unit I and 16 for Unit 2. Does this necessarily indicate that one or a combination of the causes listed above produced this difference, or is it possible that a doviation of this magnitude couid have been expected If both units had the same average probability of occurrence of reportable events? Statistical analysis indicates that il and 16 in one year are both consistent with average occurrence rates in the range of one per 20 days to one per 37 days (10-18 per year). In fact, the pair of numbers, 11 and 16, is the most probable one-year outcome for two units witn an average rate of one per 27 days (13.5 per year). In 1978, the Zicn Nuclear Station had 85 reportable occurrence LERs for Unit I and 39 for Unit 2. In this case, the deviation in the number of LERs betwean the two units is too large to be attributed solely to random effects. I f randonness alone were involved, Unit I probably could not have had a reporting rate less than one per 5.2 days (70 per year), whlIe Unit 2 probabiy could not have had a rate greater than one per 7.2 days (51 per year). In fact, if both Zion units had identical probabilities of reportable events, there is no more tnan one chance in one miilion that a deviation this Iarge could occur by chance. Naturally, there are differences between the Point Beach units. Unit I is two years older than Unit 2. During 1976, Unit 2 produced lif more electrical energy than Unit I. The results in this example Indicate that one should not necessarily conclude that differences in the rates of LER submission between the two units are significant. At Zion, however, ! one should expect to find that the two units reported at significantly l di f ferent rates for reasons other than randomness. E-2 O
Methodology Methods from probability theory can be used to calculate the impact of randomness on the distribution of the number of LERs among identical nuclear power units. Of ten, probability tables from reference textbooks are sufficient to perform the analyses. Computer simulations are necessary for the more compilcated analyses, in interpreting the resulting data, it is important to note several basic ) facts: (1) Tne numerical size of expected random variations in event rates increases as the average event rate increases. Deviations of 10 or more are readily expected on a random basis for an average , yearly rate of 100, but are unlikely for an average yearly rate of I
- 20. The reIatIve size or percentage variation, however, decreases as the average rate increases.
(2) The chance of large random variations anong units increases as the number of units being compared increases. For two units with an j assumed average annual LER submission rate of 100, there is only a j small chance that one rate will deviate by more than 20 from the average because of randomness. For a comparison among 30 units, however, there is a good chance that at least one will deviate by l more than 20 from the annual average rate of 100 because of randemness. { I A selected set of LERs was used here to demonstrate the application of this methodology. The sources of the LERs were the 22 BWRs that achieved commer-cial operation prior to 1976. Records show that this group submitted a total of 27 LERs for 30-day reportable occurrences in auxiliary process systems during 1976, 1977, and 1978. Thus, for this group of units, the average was about one LER of this type for the three-year period, it is first assumed that all units j in the group were Identical with respect to their chances of generating LERs i of this type. Further, it is assumed that if a nuclear power plant experiences l a reportable occurrence in an auxillary process system, the chance of another i occurrence is unaffected. Throughout this study a Poisson distribution of ] events is assumed. Probability theory indicates that, while the average is one, it is very unlikely that each individual unit would experience in fact, tne probability that ali 22 units would each report l exactiy one, this number is less than one in ten billion. The most likely result is ; that about eight units will have no LERs, about eight will have one LER, I about four will have two LERs, and about two will have three LERs. Further, it is unlikely (8% chance) that any one unit will have six or more LERs. Comparison to actual LER data shows nine units with no LERs, seven with one LER, two with two LERs, one with three LERs, two with four E-3 i 0
- v \ f(
LERs, and one with five LERs. The distribution of Leas for tne 22 BWRs is consistent with the assumptions sta ted aoove. Thi s exa npl e, does not prove, however, tnat the 22 BnRs are identical to each other nith regard to the causes of auxiliary process systems failures, it simply indicates that one should not expect to find signi ficant di f ferences among these units, even though some submitted as few as zero and others as many as five LERs. The value of this analysis is that it provides a mornodology through which signi ficantly high deviations can be readily identi fied among a population of expected random deviations. Analyses For purposes of this study, the LERs from 67 nuclear power plants were reviewed. For purposes of analyses, these were divided into PW7s (total = 42) and BdRs (total = 25) and each of these groups was f urther separated into
" older" and " newer" power plants. In tnis case, " older" was arbitrarily defined as those power plants that went into operation prior to 1976 (see Table E-l). For this group, all LERs submitted during calendar years 1976 through 1978 represent events that occurred during commercial operation.
Data used in these analyses were based on the NRC computer bank and included reportabl e occurrences onl y. The R0s were separated into those required to be submitted on a prompt or two-week basis and those submitted on a thirty-day basis. Tnese were analyzed separately since there did not appear to be any correlation in the relative numbers of each type as reported by licensees at the 67 power plants. Lastly, the LERs were further separated according to the system to which they pertained. A listing of these systems is shown in Table E-2. The primary goal in the analyses was to identify significant deviations or varlaitons in the number of LERs reported f ran pl ant to plant and systen to system. A deviation was considered to be significant !f there was a 5% chance or less tnat it could have resulted fra, random variations. Conclusions On the basis of these analyses, the following conclusions and/or observations were made: (1) Tne frequencies of reportable occurrence LERS among the various nuclear power units wore significantly different. There were no identifiable groups of reactor units whose members generated the same average number of reportable occurrence LERs during each of the three years in the study.
, E-4 Y
l till 1 r I i, 4 1
(2) Considering the three-year period as a whole, 5 units among the 29 older PWRs deviated signi ficantly from the others in terms of the total number of two-week R0s. The numbers of LERs from Calvert Cliffs-l, Palisades, Rancho Seco, and Three Mlle Island-l were high; Maine Yankee was low. The renaining 24 PWRs reported numbers of LERs consistent with an average of about 20 per unit for the period fron 1976 through 1978. (3) For the same 29 older PWRs, considered year by year, the data showed that the total number of two-week R0s steadily decreased in each successive year. The averages were ten per unit in 1976, six in 1977, and four in 1978. Signi ficant deviations from these occurred at Calvert Cliffs-1 in 1977, Palisades in 1977 and 1978, Point Beach-l in 1978, Rancho Seco in 1977, and Three Mlle Island-l in 1978. All had higher than normal reporting rates. Maine Yankee had a rate in 1976 signi-ficantly lower than normal . These results indicate that the high three-year totals for the four units listed in paragraph 2 above were basically due to high reporting rates in just one of the three years, while the rates for the other two years appear to be normal . (4) Further analysis of the data showed that the high totals of two-week R0s in four of the older PWRs were attributable to abnormally high numbers of LERs concerning speci fic systems. Calvert Cllffs-l had significantly high three-year totals for electric power systems and for reactor sys tems. Palisades reported high totals for the same two syst em s , in addition to engineered safety features. Rancho Seco O reported a high total for electric power systems. had high totals for radiation protection systems and for events Three Mlle Island-l classed as " systems code not appilcable." Many of the electric power system LERs were related to of f-site power systems and emergency diesel generators. Reactivity control systems were the source of most of the reactor system LERs from Pallsades. (5) Among the older PWRs with normal yearly forals for two-week R0s, some nevertheless reported significantly higher than normal totals of LERs for speci f ic systems. The number of LERs in reactor systems was higher than normal at Arkansas Nuclear One-1, Oconee-2 and -3, and H.B. Robinson-2. The number for Zlon-l was higher than normal for radiation protection systems. LERs for electric power systems were higher than normal at Fort Calhoun, Oconee-l and -3, Prairie Island-l, and Turkey Point-3. The systems mentioned here, however, did not contribute significantly to the total number of LERs, since LERs from engineered safety features and reactor coolant systems doninated the two-week R0s from older PWRs. As a result, deviations from normal in the less of ten reported systems did not have a significant impact on the total number of LERs fer these plants, i E-5
(6) The data show that newer PWRs, af ter they achieved ccmmercial operation, had significantly higher LER submission rates for two-week R0s than did older PWRs. The exception was indlan Point-3. As with the older plants, engineered safety features and reactor coolant systems were responsible for a large fraction of the LERs. (7) With regard to 30-day Ros, there were no identiflable units among the 29 older PWRs that deviated signi ficantly from the average totals for the three-year period. It is possible, however, to identi f y three separate subgroups among the units in th!s category. A first subgroup includes seven units with an average reporting rate
- of about twenty 30-day R0s for the three years. These were Oconee-2,
! Point Beach-l and -2, Rancho Seco, San Onofre-l, and Turkey Point-3 and -4. Another group had an average of about forty-f ive 30-day R0s for the three years. The 10 units In this group were H.8. Robinson-2, f Haddam Neck, Indian Point-2, Maine Yankee, Oconee-I and -3, Prairie Island-l and -2, R.E. Ginna, and Three Mlle Island-l. A third group of 5 units with a normal reporting rate of about 70 for the three-year period included Arkansas Nuclear One-1, Kewanee, Palisades, and l Surry-l and -2. Signi ficant deviations from these groups occurred in 1 7 units with high reporting rates. These were Calvert Cliffs-l, D.C. Cock-1, Fort Calhoun, Millstone-2, Yankee Rowe, and Zion-1 and -2. It is interesting to note that three of the five operating Combustion Engineering reactors are in this category. These are Calvert Cliffs-l, Fort Calhoun, and Millstone-2. In addition, this category includes all three of the older PWRs having a power level of 1000 M#e or more. These are D.C. Cook-l and Zion-l and -2. (8) The data show that the one-year totals for thirty-day R0s in older PWRs were similar to the three-year totals in that definite subgroups can be identifled, in general, a unit that was in a low or higher reporting subgroup in one year remained in the same subgroup in later years. The exceptions were Yankee Rowe, which was in a higher re-porting subgroup in 1977, but in lower reporting subgroups in the other two years, and Surry-l and -2, which were in a lower reporting subgroup during the first two years but in the higher subgroup in 1978. Several signi ficant correlations were found. Those units which tended to remain in the lowest reporting subgroups nevertheless in-creased their reporting rates for thirty-day R0s from year to year. The sum of their thirty-day and two-week Ros, however, remained essentially constant in time, since the two-week RO total steadily j decreased during the three-year period. Large units of 1000 MWe or more reported higher numbers of 30-day R0s, except when the plant factor for the year was low (less than one-third). Later Cmbustion l l l E-6 O l
~- .....__ _ _ _ _ _ _ _ Engineering units (not including Maine Yankee) also submitted higner numbers of LERs for thirty-day R0s, except when the plant availability factor was low (less than one-half). (9) Newer PWRs reported thirty-day R0s at rates consistent with the higner reporting subgroups among older PWRs. (10) The systems most responsible for the higher LER submission rates for thirty-day R0s in Combustion Engineering units were auxiliary process systems, electric power systems, instrumentation systems, and s team and power conversion systems. These units usually deviated from the normal reporting rate for these systems. In large units the l systems involving a higher than normal number of thirty-day Ros were l auxillary process systems, engineered safety features, instrumentation systems, and radiation protection systems. (ll) With regard to two-week R0s among the 22 older BWRs, eight units deviated from the normal reporting rate during the three year period. These were Dresden-2, Duane Arnold, E.I. Hatch-l, Fitzpatrick, and Peach Bottom-2 and -3, with higher rates than normal and Dresden-l and Lacrosse with lower rates than normal. The remaining units l reported an average rate of about twenty-four two-week R0s for the three-year period. The rate remained constant at about eight per year. O (12) E.1. Hatch-l reported two-week R0s at a comparatively high rate for each of the three years. The number of reports pertaining to nearly every system deviated from normal reporting rates for those systems. (13) Duane Arnold reported two-week R0s at a comparatively high rate in 1976 and 1977. The systems with higher than normal numbers of reports were related to electric power. For Fitzpatrick, the number of two-week R0s for 1976 was high. This unit also had a high number of R0s in instrumentation systems. For Peach Bottom-2 and -3, the number of two-week R0s for 1976 and 1977 was high. Unit 2 had an abnormally high number of R0s for reactor coolant systems and steam and power conversion systems. Unit 3 reported a high number in engineered safety features and for other auxiliary systems. Dresden-3 reported a higher-than-normal number of LERs in 1977. Further, this unit reported an abnormally high number of R0s in electric power sys+ ems. Nine Mlle Point-l reported higher-than-normal totals of LERs concerning instrumentation systems. Quad Cities-l reported a high incidence of two-week R0s in steam and power conversion systems. E-7 O t.. _ . . _ . _ _ _ _
(14) Among the three newer BWRs, only Browns Ferry-3 reported abnormally high numbers of two-weak ROs in reactor systems after the unit began commercial oper at i on . (15) Two BWR units, Fitzpatrick and Brunswick-1, reported abnormally high numbers of thirty-day R0s in nearly every system. As an extension to the above, LERs pertaining to set point drift were analyzed using as a data source the computer bank at the Nuclear Safety information Center (see Appendix D-Ill). These analyses showed that there was no significant deviation in the total annual LF.R submittal rate for setpoint del f t among older BWRs or among older PWRs. The average rate for BWRs, however, was approximately five times as large as that for PWRs. Six older PWRs reported rates higner than normal for the three-year period. These were Zion-l and -2, Fort Calhoun, Mi!Istone-2, Palisades, and Kewance. It is interesting to note that three of these are Combustion Engineering units. Among newer PWRs, four units reported at high rates in 1978. These were J.M. Farley-1, Indian Point-3, North Anna-l, and Salem. Three older BWRs reported set point drift events at abnormally high rates for the entire three-year period. These were Duane Arnold, Brunswick-2, and Nine Mile Point-l. Six older BWRs reported at abnormally low rates. These were Big Rock Point, Browns Ferry -l, -2, and -3, Lacrosse, and Monticello. Comment _ary This cortion of the study has clearly demonstrated the potential usefulness of statistical analyses in the evaluation of LERs submi tted by licensees. Such analyses make it possible to distinguish deviations in the numbers of LERs which would be expected on the basis of randomness from those that almost certainly would not. The latter can be used as a means for the identification of areas for possible further investigations, While the deviations noted in this study do not necessarily imply safety-related problems, they should nonetheless be pursued in order to determine the true impticatlons. It would probably be desirable to computerize these analyses for automatic processing of reports as they are logged into the LER data base. Utiliza-tion of the data base in this manner would make it possible to detect sign i f ican t deviat lons f rcm normal . Furtner, an automated system could be programmed to obtain detail beyond the system level, in order to identify reporting rate deviations for relevant subsystems and components. E-8 O
Table E-i Number of Reportable Occurrence LERs fron Commercial Nuclear Power Plants (1976-1978) GROUP 1: Older PWRs (commercial operation prior to 1976) Total = 29 Nuclear Reportable Occurrences Nuclear Reportable Occurrences Power Plant 30 dal 2-week Power Plant 30-day 2-week Arkansas Nuclear One-I 71 17 Point Seach-l 15 30 Calvert Cliffs-1 169 35 Point Beach-2 18 20 0.C. Cook-l 147 20 Prairie Island-l 51 17 Fort Calhoun 109 24 Prairie Island-2 36 18 H.B. Robinson-2 53 26 Rancho Seco 17 40 Haddam Neck 41 19 R.E. Ginna 44 24 Indian Point-2 57 26 San Onofre-I 19 11 Kewanee 75 19 Surry-l 79 19 l Maine Yankee 47 6 Surry-2 71 8 l l Millstone-2 l18 21 Three Mlle Island-l 44 41 I l Oconee-I 42 34 Turkey Point-3 24 11 Oconeo-2 21 26 Turkey Point-4 20 16 Oconee-3 41 21 Yankee Rowe 99 13 Palisades 64 55 Zion i 188 25 l l Zion 2 122 15 Average 65.6 E7 E-9
Table E-l Continued GROUP ll: Newer PWRs (commercial operation after January 1, 1976) Total = 13 Nuclear R_ ego _rtable Occurrences Nuclear Regortable Occurrences Power Plant 30-day 2-week Power Plant 30-day 2-week Arkansas Nucicar One-2 21 7 Indian Point-3 85 15 Beaver Valley-l 216 27 J.M. Farley-l 133 23 Calvert Cliffs-2 135 25 North Anna-l 93 29 Crystal River-3 154 32 S t . L uc l e- I 123 22 0.C. Cook-2 96 7 Salem-l 118 68 Davis-Sesse-I 220 32 Three Mile Island-2 42 17 Trojan 63 44 Average ll6.5 26.8 e i E-10 0
.[ Table E-l Continued GROUP lil: Older SW9s (commercial operation prior to 1976) Total = 22 NucIear Reportabie Occurrences NucIear ReportabIe Occurrences Power Plant 30-day 2-week Power Plant 30-dal _2-week Big Rock Point I05 31 Lacrosse 27 10 Browns Ferry-l 55 26 Millstone-I 80 27 Browns Ferry-2 33 18 Monticello 65 30 Brunswick-2 261 34 Nine Mlle Point-l 93 27 Cooper 122 18 Oyster Creek-l 56 35 l
Dresden-l 70 10 Peach Bottom-2 146 56 Dreaden-2 153 51 Peach Bottom-3 107 56 Dresden-3 109 29 Piigrim-1 103 25 Duane Arnold 120 88 Quad Cities-l 94 31 E.I. Hatch-l 94 162 Quad Cities-2 75 14 Fitzpatrick 181 41 Vermont Yankee 95 18 Average 102.0 38.0 1 E-Il l
/O U
Table E-l Continued GROUP lV: Newer BW9s (commercial operation af ter January 1, 1976) Total = 3 Nuclear Reger_ table _0ccurrences Pow er _P l_an t, 30-day 2-week Browns Ferry-3 58 12 Brunswick-l 2ll 9 E.1. Hatch-2 65 12 Average 111.3 11.0 O l E-12 O
/~
Table E-2 System Codes for LERs System S1 3, tem
- 1. Auxiliary Process Systems 8. Other Major Systems
- 2. Auxiliary Water Systems 9. Radiation Protection Systems
- 3. Electric Power Systems 10. Radioactive Waste Management Systems I
- 4. Engineered Safety Features ll. Reactor Systems
- 5. Fuel Storage and Handling Systems 12. Reactor Coolant Systems
- 6. Instrumentation and Control Systems 13. Steam and Power Conversion Systems
- 7. Other Auxillary Systems 14. System Code Not Applicable O
l E-13 l A V
l l l APPENDIX F G ACRS CHARTER AND MEMBERSHIP 1 l l l o J
COMMITTEE CHARTER The Advisory Committee on Reactor Safeguards was established as a statutory Committee in 1957 by revision of the Atomic Energy Act. The Committee was charged with the responsibility for review of safety studies and f acility license applications submitted to it, and to make reports thereon advising the Commission with regard to the hazards of proposed or existlng reactor facilltles and the adequacy of proposed reactor safety standards, and to perform such other duties as the Commission might request. Section 182b of the Atomic Energy Act requires ACRS review of the construction permit and operatinn license applications for power and testing reactors and spent fuel reprocessing facilities licensed under Section 103, 104b or 104c of the Atomic Energy Act; any application for a research, develoo-mental or medical facility licensed under Section 104a or e of the Act which is specifically referred to it by the Ccmmission; and any request for an amendment to a construction permit or operating license under Section 103 or 104a, b, or c which is specifically referred to it by the Commiss1on. The Energy Reorganization Act of 1974 transferred the AEC regulatory functions to the newly formed Nuclear Regulatory Commission and the ACRS operations were also transferred to NRC to assist in its regulatory functions. O in 1977, Public Law 95-209 added to its other duties a requirement for the ACR$ to undertake a study, making use of all available sources, of () reactor safety research and prepare and submit annually to the United States Congress a report containing the results of this study. I r i F-l a l
C0fIIIITTEE f tEMBERSHIP CHAIRMAN: Dr. Max W. Carbon, Professor and Chairman of Nuclear En-gineering Department, University of Wisconsin, Madison, Wisconsin VICE-CHAIRMAN: Dr. Milton S. Plesset, Professor of Engineering Science - Emeritus, California Institute of Technology, Pasadena, , California Mr. Myer Bender, Director of Engineering Division, Oak Ridge National Labor-atory, Oak Ridge, Tennessee Mr. Jesse Ebersole, Head Nuclear Engineer, Division of Engineering Design, Tennessee Valley Authority, Vnoxville, Tennessee (retired) Mr. Harold Ethe r ington , Consulting Engineer (Mechanical Reactor Engineer-Ing),.. Jupiter, Florida F rof. Willia- Ke rr, Professor of Nuclear Engineering and Director, Michi-gan Memorial-Phoenix Project, University of Michigan, Ann Arbo r , Michigan Dr. Stephen Lawroski, Senior Engineer, Chemical Engineering Division, Ar-gonne National Laboratory, Argonne, Illinois Dr . Harold W. Lewis, Professor of Physics, Department of Physics, Univer-sity of California, Santa Barbara, California Dr. J. Carson Mark, Division leader, Los Alamos Scientific Laboratory, Los Alanas, New Mexico (retired) Mr. William M. Mathis, Director, Planning, United Nuclear Intustries, Inc., Richland, Washington (retired) Dr. Dade W. Moeller, Chairman, Department of Environmental Health Sciences, School of Public Health, Harvard University, Boston, Massachusetts Dr. David Okrent, Professor, School of Engineering and Applied Science, ' University of California, Los Angeles, California Mr. Jeremiah J. Ray, Chief Electrical Engineer, Philadelphia Electric Com-pany, Philadelphia, Pennsylvania (retired) Dr. Paul G. Shewmon, Professor and Chairman of Metallurgical Eng inee ring Departant, Ohio State University, Columbus, Ohio Dr. Chester P. Siess, Professor Emeritus, Department of Civil Engineering, University of Illinois, Urbana, Illinois F-2
NRC FoRu 335 "
""3 U.S nucle AR REGULATORY COMMISSION NUREG-0572 BIBUOGRAPHIC DATA SHEET TLE AN O SU8 TIT LE (Add Volume No . rf accrocesste! 2 (L eave w a l EVIEW 0F LICENSEE EVENT REPORTS (1976 - 1978) 3 RE CIPIE NT S ACCE SSION NO
- 7. AUTHOR (S) 5 O ATE RE PORT COMPLE TE D ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
*kEPTEMBER 1979 9 PE R5 0RNONG ORG ANi2 ATION N AME AND V AILING ADORESS uncyve 2 0 Coael O ATE RENR f ISSA D ADVISORY COMMITTEE ON REACTOR SAFEGUARDS " '"
3EPTEMBER 1979 U.S. NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 6 ' ' " " " ' " ' S Iltave OlJ'81 12 SPONSORING ORG ANIZ A TION N AVE Mv D M AILING ADD RE SS It'*c vac l-o CodW . Same as above. y cc ,,uc1 su 13 T Y PE OF RE POR T *E a4 00 c o' t Ai t < ' 'v > a m ' 15 SUPPLEME N T A R Y N OTE S Ia 't -ar- r n 16 A8STR AC T C00
- o'os or reso This report was prepr ad in response to a request by the Chairman of the U.S. Nuclear egulatory Coninission for the Advisory Coninittee on Reactor Safeguards to review, with
( pecific objectives , the Licensee Event Reports (LERs) issued from 1976 through 1978. principal objective of the present study was to identify those events which have implications for improved reactor safety. Approximately 8,700 LERs were filed during the three-year period under review. These described events ranging from the trivial to those of major safety significance. This study has confirmed that LERs represent a source of valuable data and information. However, it must b e recognized that a detailed review of LERs will not identify all safety problems likely to be enc (untered in the operation of nuclear power plants. As a result of this review, a number of classes of events were identified. These classes were subsequently categorized into three groups: (a) those having potentially serious safety implications, (b) events that must be regarded as matters of concern, and (c) events of lower safety significance but with excessive frequency. 17 x,E Y WOR OS AND D OCWE NT AN A L YS55 1 74 DE SC H'P T O RS 17 t> IDE NTIFIE RS OPh N EN DE O TE RVS I h M AV AILABILITY STATE MENT 19 SE CU A l f v C LASSrry rporo 21 w 0; p A :g 3 UNCLASSIFIED UNLIMITED 20 SE cV AM y CL ASS UN oar / N PR M s N RC FORM 33s \111n
, UN11E D ST A TE S NUC LI A R REGULATORY CO M MIS $10 N W ASH 6NO 7 0N. O. C. 70$66 f ~l pot? AGE AND F E E S P AlO U.5 MUCLE A R N gGUL A TOR Y O F F ICI AL BUSINESS PEN ALTY FOR P AIV ATE USE, $300 C O M M ' SS6 0 N N A 1
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APPENDIX XXIII
Title:
Response to Questionnaire , , from NRC/TMI Special Inquiry Group
/ mcw%, UNITdD SEATES s ,
;- NUCLEAR REGULATORY COMMISSION
,!f' ' 1 WASHINGTON, D, C. 20655
_ '; > u . J
%% y August 7, 1979 In Reply Refer to:
, NTFTM 790807-04 ,
Mr. Raymond Fraley, Executive Director Advisory Comittee on Reactor Safeguards U.S. Nuclear Regulatory Comission Mailstoc H-1014 Washington, D.C. 20555 l
Dear Mr. Fraley:
One of the tasks of the NRC/TMI Special Inquiry Group is to assess the ef fectiveness of the NRC's licensing and review process in connection with the licensing of TMI-2. You can greatly assist us in this effort by your willingness to respond . to the attached questionnaire, which is being sent to key individuals ! in the NRC licensing process, including licensing and appeal board members, and selected staff. In order to minimize the inconvenience of this request, I we have tried to limit the questions to the absolute minimum necessary l to obtain what we believe will be useful information, and we are therefore hoping for a nearly 100 percent response. If you have any questions regarding the request, please contact Special Inquiry Group member David Evans at 492-8947. Completec questionnaires should be returned to him at mailstop AR-400 as soon as practicably possible. I appreciate your cooperrtion. Sincerely,
$ w Mitchell Rogovin, irector NRC/TM! Special nquiry Group
Enclosure:
Ouestionnaire ,
t NRC/TMI Special Inquiry Please answer the following questions to the best of your curront recollection: A. PERSONAL BACXGROUND
- 1. Your name, official title, office address and office phone number.
2.. How long have you been in your present position? What prior posi-tions have you held which involved nuclear licensing? B. INYCLVEMENT WITH LICENSING OF B&W PLANTS
- 1. In what proceedings have you been involved in which a Babcock & Wilcox design was at issue? (0conee 1, 2, and 3; Three Mile Island I and 2; Midland 1 and 2; Bellefor.te 1 and 2; North Anna 3 and 4; WPPSS 1 and l 4; Davis Besse 2 and 3; Greenwood 2 and 3; Pebole Springs 1 and 2- '
Erie 1 and 2.)
- 2. At what level (Staff, Licensing Board, Appeal Board, Commission) and l
stage (Staff review, CP, OL, post-OL) were you involved in B&W pro- ! ceedings?
- 3. What are the significant issues you recall having been raised in these proceedings?
() 4. Who raised these issues, what were the positions and responses of the other parties, and how were the issues resolved? S. Do you recall any proceeding at which B&W presented testimony or was present for questioning? Please describe the circumstances.
- 6. Based upon your experience with B&W issues, have any issues been raised more often than others? Which one(s)?
GENERAL OBSERVATIONS, REGARDLESS OF RESPONSES TO PART B. l C. 1
- 1. How would you rate the performance of the Staff in the licensing 1
p rocess? Specifically, now oc you assess: The role of the staff outside of the acjudicatory process, e.g. , 4 a. in review of designs and plans, and in contacts with the. utility and vendors; i
- b. The staff's role in adjudication;
- c. Staff documentation of its positions;
- d. Post-licensing staff action and its imcact on the process.
O 1 c _-.
i QUESTIONNAIRE O Page 2
- 2. How would you rate the licensing and review process in general?
Do you have any views on that process with respect to:
- a. Its consistency?
- o. Its effectiveness in implementing the statutory standards?
- c. Its control over vendors, architect-engineers, and constructors?
- d. Its effectiveness in ensuring safe design and operation of plants?
- 3. Do you believe the regulatory system needs reform? If so, what major step or steps need to be taken?
- 4. Do you believe the current system inhibits innovation on the part of applicants and vendors? If so, how could that be changed?
- 5. Please provide any comment you aay have on the following specific natters:
- a. Technical qualifications of licensees
(' b. Quality assurance
- c. Energency plans
- d. Backfitting and "ratcheting"
- e. The concept of the " design basis accident"
- f. " Safety-grade" vs. "nonsafety-grade" equipment
- g. Residual risk D. '33;IS FOR FURTHER SCRUTINY 4
- 1. Are there special areas of imoortance wnich you believe this Inquiry should examine with regard to TMI-2, B&W units or the regulatory system, in general?
- 2. Please explain any contact you have had witr " precursor events" in FW;s, i.e. , witn an issue or incident that was a cccconent of the accident at IMi-2 on March 28, 1979. An examole would be the Pilot Operated Relief Valve (PORV) which stuck cpen at TMI-2.
- 3. Are there particular people with information or knowledge, in the above areas that you believe the Special Inquiry Group shculd contact?
Please 1.ist. (Nare, organization, and address) O
1 LO ,,,.. . , . . .
.....,,,%R,
- a;e 3 E. COMiENTS
- 1. Do you have any aeditional coments you celieve would assist the Inquiry in assessing the accident and its iclications for the regulatory system?
1 e O i l l i . .I
l [p m,,,\ UNITED sT ATEs
!A *t NUCLEAR REGULATORY COMMISSION l 1 k 5 ADVlsORY COMMITTEE ON REACTOR sAFEGUAROs l.
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f" W ASWNGTON. D. C. 20555 September 13, 1979 ! Mr. Mitchell Rogovin, Director ! NRC/TMI Special Inquiry Group d U. S. Nuclear Regulatory Commission
- Washington D.C. 20555 i
Dear Mr. Rogovin:
The following information is provided in partial response to your question- )
- naire of August 7, 1979 which was sent to ACRS members and me. This infor-
- . mation is based on a review of ACRS records and provides information where l appropriate documentation exists. Members of the Committee will as available l to provide individual comments and opinions during a meeting scheduled for l October 6.1979 between the Special Inquiry Group and the members of the Com- 1 mittee as requested by members of your staff. I i A. PERSO.1AL SACK 3ROUND Ine personal background information requested in item A of your letter is in- l
, cluded in Attachment A. In addition to the current membership of the Commit- 1 tee. I have included those members who were active when the B&W facilities listed in Attachment B were reviewed. The terms of appointment of members , and Committee officers are as noted. In addition, all ACRS members serve as 4 subcommittee chairmen and members on a variety of standing and ad hoc ACRS j Subcommittees with responsibilities regarding specific project reviews and a l ] number of generic safety related matters. B. Involvement with Licensing of B&W Plants The ACRS must, by statute, review and advise the NRC on all applications for Construction Permits, and Operating Licenses for all commercial nuclear power plants as well as Preliminary Design Approvals, Manufacturing Licenses, etc., , ' for standardized plants. Consequently, the Committee has been involved in all such -proceedings in which a Babcock and Wilcox design was at issue. A list is included in Attachment B. The involvement of any individual Committee memoer , .in any of the projects listed would.have depended on the dates of his appoint-
- - ment and termination as an Advisory Committee Member.
The ACRS' review process normally involves discussions with representatives of the applicant, reactor vendor, architect-engineer, and selected sub-contractors and consultants as~well as the NRC Staff and its consultants. In addition, the 1 application as well as other related documents including the NRC Staff Safety l . Evaluation Report (SER) are.also considered. i n Following its review of each proposed project the ACRS provide a ra ort to the l 4 V Commission which includes ACRS recommendations regardig creed 60 changes and res-olution of outstanding safety related issuen
~
h j . i n . a
I O sr. xitcheii ao9ovin sentember 13. 1979 1 In most cases, ACRS Subcommittee and full Committee activity on project pro- l posals begins when the radiological safety review of the NRC Staff is largely complete and a Safety Evaluation Report (SER) has been written, but precedes the radiological safety portion of the AS&l.B hearing. Responses to Committee comments and recommendations in its report to the Commission are included in a Supplement to the SER which is issued by the NRC Staff and addresses the actions to be taken regarding ACRS recommendations. This supplemental SER is made part of the record of the proceeding so it is available to the Hearing j Board in its consideration of a case. Copies are made available to the ACRS ; for information. 1 Summaries of significant issues discussed during the ACRS review of proposals ' involving a B&W design are included in Attacnment B. The attachment also in- i cludes summaries of matters of particular concern to the ACRS as noted in its ! reports to the Commission, and their disposition by the Regulatory Staff as I described in Supplements to SERs, special memoranda from the Staf f, oral pres- ) entations at ACRS Full Committee and Subcommittee meetings, etc. ; C. General Observations The ACRS has recently begun its own in-depth examination of the NRC regulatory p policies and cractices as well as the basis and procedures for evaluation of Q proposed nuclear facilities. Since this evaluation is in its initial stages the Committee is unable at this time to respond in a comprehensive way to your questions in this category. Nevertheless the Committee is of the opinion that the NRC Staff's perfomance in the licensing process generally demonstrates a substantial degree of technical competence. At the same time, there have been specific issues about which the Committee nas not been fully satisfied. As was noted in the July 2S,1979 letter from Dr. Carbon to Mr. Rogovin, the Committee believes that responses to its recommendations concerning "instrumentaticn to follow the course of an accident" have not been adequate. There are several other examples about which the Committee would have preferred a more effective response on the part of the Comminion Staf f. Among the more significant of these are:
. Timeliness of Responses to %.h Recommendations - attached, as an example, is an excernt trom an OPE report, approved by the Com-mission, and t e E RS Reports on the Zion Station which were the basis for ;ne OPE report. (See Attachment C.1)
.Lonsideraton of Design Changes to Mitigate Sabotage - This matter, wnich goes beyond access control (the main thrust of the Commis-sion's anti-sabotage effort), was first. raised in a Committee report to the Chairman dated October 14, 1975. That and sub-sequent reports on the subject are included, as well as excerpts from the Committee's reports to the Congress on the RSR program of 1977 and 1978, indicating tnat a serious study of some aspects p) s of this matter has only lately been undertaken. (See Attachment C.2)
Mr. Mitchell Rogovin September 13, 1979
. Consideration of Accidents Beyond the Limits of the Regulatory Design Basis Accident - The Committee has, on a number of occa-sions extending over a long period, recommended consideration of the consequences of accidents beyond the design basis, and means of ameliorating them. Indeed, the question of instrumen-tation to follow the course of an accident, noted above, is just such an issue. Other matters in this category include means for retaining molten fuel, and the consideration of evacuation plans for people located outside the LPZ as defined in Part 100. Both issues are summarized in the material attached which contains excerpts from ACRS reports beginning in 1971 and culminating with the Committee's 1979 Comments on the E Safety Research Program Budget. (See Attachment C.3)
.More Widespread Use of Probabilistic Analysis - For some time i the Committee has oeen urging more widespread use of probabilistic !
I methodologies in the regulatory process. Examples are attached, the latest, an excerpt from the Committee's 1979 Comments on the ' l NRC Safety Research Program Budget, suggests that this goal has , not yet been fully realized. (See Attachment C.4) D. Issues for Further Scrutiny l The ACRS has provided the NRC with written and oral advice on some matters which it believes to be particularly significant in view of the accident at I Three Mile Island and its implications. Copies of the ACRS Reports on these matters are attached. (See Attachment D.1) With regard to events which might be considered " precursors" to the accident at Three Mile Island, the Committee had discussed the following such events j with the NRC Staff:
- 1. Oconee Unit 3 - June 13,1975 ICS failed to follow load demand below 1S*. power, PORV opened at 2255 psi and failed to close. Reactor tripped on low pres-sure; HPCI actuated and PORV isolatior, valve was closed to terminate depressurization. This was described by and dis-cussed with the NRC Staff during the 183rd ACRS meeting, July 10-12, 1975
- 2. Davis Besse Unit 1 - September 24, 1977 A half trip of the steam and feedwater rupture control system resulted in PORV opening, cycling, and sticking open. The operator manually tripped the reactor and ultimately isolated the PORV. Additional elements of this event included the formation of a steam bubble within the RCS, an upsurge of the
Mr. Mitchell Rogovin September 13, 1979 level in the pressurizer, and rupture of the rupture disc in the Pressurizer Quench Tank with consequent release of steam to the containment. This event was described by and discussed with the NRC Staff during the 210th and 211th ACRS meetings, October 6-8 and November 3-5. 1977. (See Attachment D.2) E. Comments As noted above, the Committee has recently begun an evaluation of the regulatory process and the key elements involved in this process. As Committee conclusions result from this study they will be made available to the Special Inquiry Group as well as the Nuclear Regulatory Commission. R. F. Fraley Executive Director
Enclosures:
A) Personal Background p B) Involvement with Licensing of 83W Plants d C) General Observations Issues for Further Scrutiny D) cc. ACRS Members O v
Attachment A fl A. Personal Background LJ
- 1. Name, Title, Office Address and Phone Number A - ACRS Chairman Dr. Max W. Carbon, Chairman Nuclear Engineering Department Engineering Research Building University of Wisconsin Madison, WI 53706 Phone: 608-252-5125 Appointed to ACRS April 1975 Vice-Chairman during 1978 Chairman during 1979 ACRS VICE-CHAIRMAN Dr. Milton S. Plesset Professor of Engineering Science - Emeritus California Institute of Technology 104-44 Pasadena, CA 91109 Phone: 213-795-6811, Ex.1176 Appointed to ACRS October 1975 P) q Vice-Chairman 1979 .
Mr. Myer Bender, Director of Engineering Oak Ridge National Laboratory P.O. Box X - Building 1000 Oak Ridge, TN 37830 Phone: 615-574-6455 Appointed to ACRS August 1972 Vice-Chairman 1976 Chairman 1977 Mr. Jesse C. Ebersole Head Nuclear Engineer, Div. of Engineerig Design, TVA (Retired) , 103 Newell Lane l Oak Ridge, TN 37830 1 Phone: 615-482-4184 Appointed to ACRS April 1976 ! Mr. Harold Etherington Consulting Engineer I (Mechanical Reactor Engineering) ! 84 Lighthouse Drive l Jupiter, FL 33458 l Phone: 305-746-3904 Member of ACRS March 1965 O oe""ers 19741 "er 1976-prese"t
l l l l Dr. William Kerr l Professor of Nuclear Engineering and Director, 1 Michigan Memorial-Phoenix Project 1 Department of Nuclear Engineering l Phoenix Laboratory - North Campus University of Michigan Ann Arbor, MI 48109 Phone: 313-764-6213 Appointed to ACRS August 1972 i Vice-Chairman 1974 I Chairman 1975 Dr. Stephen Lawroski Senior Engineer Chemical Engineering Division Argonne National Laboratory ! Argonne, IL 60439 j Phone: 312-972-4469 l l Appointed to ACRS August 1974 Vice-Chairman 1977 Chairman 1978 Q Dr. Harold W. Lewis Professor of Physics Department of Physics University of California Santa Barbara, CA 93106 Phone: 805-961-2670 i Appointed to ACRS May 1979 Dr. J. Carson Mark Division Leader los Alamos Scientific Laboratory l Theoretical Division i P.O. Box 1663 - MS 210 Los Alamos, NM 87545 , Phone: 505-667-7612 Appointed to ACRS February 1976 Mr. William M. Mathis Director, Planning, United Nuclear Industries, Inc. (Retired) 455 Mainmast Court Richland, WA'99352 Phone: 509-375-0359 Appointed to ACRS August 1978 l
Dr. Dade W. Moeller, Head Environmental Health Sciences Dept. School of Public Health Harvard University 665 Huntington Avenue Boston, MA 02115 Phone: 617-732-1169/1170 Appointed to ACRS December 1972 Vice-Chairman December 1975 Chairman December 1976 Dr. David Okrent Professor School of Engineering & Applied Science 5532 Boelter Hall School of Engineering & Applied Science University of California Los Angeles, CA 90024 Phone: 213-825-3259 Appointed to ACRS November 1963 Vice-Chairman November 1965 Chairman November 1966 Mr. Jeremiah J. Ray Chief Electrical Engineer Philadelphia Electric Company (Retired) 819 Lantern Lane Langhorne, PA 19047 Phone: 215-757-4183 Appointed to ACRS October 1978 Dr. Paul G. Shewmon, Professor and Chairman Department of Metallurgical Engineering Ohio State University 116 West 19th Avenue Columbus, OH 43210 Phone: 614-422-2491 Appointed to ACRS June 1977 Dr. Chester P. Siess, Professor Emeritus of Civil Engineering 3110 Civil Engineering Building University of Illinois
% ana, IL 61801 none: 217-333-3924 sppointed to ACRS April 1968
'! ice-Chairman April 1971 1] Chairman April 1972 l
O ACRS Office Mr. Raymond F. Fraley Executive Director Advisory Committee on Reactor Safeguards V. S. Nuclear Regulatory Commission Washington, DC 20555 Phone: 202-634-3265 Regualtory positions held: Assistant to Executive Secretary, AJRS 1959-1963 Executive Secretary, ACRS 1963-1975 Executive Director, ACRS 1975-present 1 l l () l l l I l O v 1
Attachment B t Proceedings Involving 8&W Designs
- 1) Arkansas Nuclear One - Unit 1 i
- 2) BSAR-205
- 3) Bellefonte Nucelar Plant, Units 1 and 2
- 4) Crystal iver Unit 'a
- 5) Davis Besse Units 1, 2 and 3
- 6) Erie Nuclear Plant, Units 1 and 2
- 7) Greenwood Energy Center, Units 1 and 3
- 3) Midland Plant Units 1 and 2 g) North Anna Units 3 and 4
- 10) Oconee Units 1, 2 1 3
- 11) Pebble Springs Nucle ir Generating Station, Units 1 and 2
[2 ) Rancho Seco Nuclear i_nerating Station, Unit 1
- 13) Three Mile Island Jn'.s 1 and 2 14 ) Washington Public Po r Supply System (WPPSS), Units 1 and 4
Attachment B - - AN O - 1 O v ( ARKANSAS NUCLEAR ONE - UNIT 1 Subcommittee Meetings Held Full Committee Meetings Held Augest 23, 1968 September 5-7, 1968 Septeder 4, 1963 Februa ry 8-9, 1973 July 26, 1973 August 9-11, 1973 November 3-5, 1977 May 7-8,1973 Topics Di; cussed: O Site Cha acteristics. 0 0uality Assurance Program. O Use of Stared Systems. Fuel Dens #ica ion. O l High Energy .ine Break Outside Containment. l O Feedwater Sy: tem Design l O Steamline Break Isolation U Common Mode 3ilures Related to Cable Routing. l 0 Emergency Cc e Cooling System Design / Loss of Coolant Accident Analysis. I O Pipewhip ' : lated Design Criteria. l 0 Hydrogen Control. 0 Appendi.x I Considerations. 0 Safety Related Valve Relability. O Pump Flywheel integrity. O Steam Ge erator Integrity, U Loose Parts / Vibration Monitoring Systems. 0 Ultimate Heat Sink. 0 Shipping Car.k Drop Accident. U Emergency Planning. Plant Security. U Training of Operating Personnel, o Containment Safety Margins - Structural Integrity / Containment Anchoring System. O G Design E:-sis Earthouake. O o Fuel Perforcance.
A GEMiSAS N'JCLEAR ONE - UNIT 1 ((,) 0 0perating Shift Size. UReactor Internals - Failure Due to Eccessive Vibration / Hydraulic Forces. ACis Comments were Provided in the Followino Letters (copies are attached) September 12, 1963 August 14, 1973 Summary of ACRS Comments and the NRC Staff Responses
'The Committee reiterated its concern on reacto" coolant pump overspeed.
Reactor coolant pump overspeed during a Loss of Coolant Accident is a unresolved ACRS Generic Item. (pWR vendors are arguing that these pumps will withstand Loss of Coolant Accident overspeed conditions.) i O( s ,! e Committee reiterated its belief that the instrumentation design
! ould be reviewed for common mode failures. This is being adressed in tne work on Anticipated Transients Without Scram. The Committee reiterated it concerns on tne Anticipated Transients Without Scram issue and recom-mended that the related studies be expedited and that design changes be l I
implemented as required in a timely f ashion. l l Ine resolution of the Anticipated Transients Withcut Scram issue is still pending, the NRC Staff is trying to resolve this on a f generic basis. l l l
' Toe Committee recommended that the Applicant give further attention to instrumentation to follow the course of an accident. l l
This matter has been addressed by Regulatory Guide 1.97. The ACRS has raised concerns as to the implementation of this Regulatory Guide. Tne NRC Staff has not yet made a decision as j n l 6") to how it will implement Regulatory Guide 1.97 l l l l
ARKANSAS NUCLEAR ONE - UNIT 1 ,
'The Committee recommended that the Applicant's Safety Review Committee include experienced personnel from outside of the Applicant's organization.
The Applicant has implemented this recommendation. The NRC Staf f has included requirements for Radiation and Health Physics and Nuclear Safety Consultants in the Technical Specifications.
'The Committee noted that the Applicant has adopted equipment and admin-strative measure to assure that acceptable fuel limits would not be exceeded and recommended that the Regulatory Staff establish criteria for establishing and implementing such measures.
The NRC Staff has developed and implemented Standardized G Technical Specifications, !V
'The Committee noted the existence of a proposal for additional operating limitations for that period of operation in which the temperature coefficient l of reactivity is positive.
The NRC Staff has reviewed these operating procedures and has l incorporated limits in the Technical Specification. l l i
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L., ADVISORY COMMITTEE ON REACTOR SAi EGUARCS
/"< UNITCD STATCS ATOMIC ENERGY COMMISSION -
WASHINGTON. D.C. 20543 1 August 14, 1973 i a lionerable Dixy Lee Ray Chairman U. S. Atenic Energy Cor:nission Washington, D. C. 20545
Subject:
REPORT ON AR'<ANSAS NUCLEAR ONE-UNIT 1
Dear Dr. Ray:
During its 160th meeting, August 9-11, 1973, the Advisory Coenittee on Reactor Safeguards completed its review of the application of the Arkansas Pcuer and Light Company for a license to operate Arkansas Nuclear One-Unit 1 (fo=2rly Russellville Nuclear Unit) at pcwar levels up to 2565 1%(t). The site was visited by a subcomittee en May 4,
,, 1973, and the project considered durinE a Subec=mittec meeting held in '( Washington, D. C., on July 26, 1973. In the course of the review, the Cc=ittco had the benefit of discussions with representatives and cen-sultants of tha Arkansas Power and Licht Conpany, the Babece': and Wilcox
. Company, th'c Ecchtel Corporatten, and the AP.C Regulator / Staff, and of the doce onta listed. The Cc=ittee last reported to the Co etission on the construction of this unit in its letter of September 12, 196S, and on Unit 2 in its letter of February 10, 1972. Arkansas Nucicar One is locatad about six miles from Russellville, Arkansas, on a peninsula fomed by the Dardanclle Reservoir on the Arkansas River. The application for a cons truction perait preposed initial operatf.en at power levels up to 2452 L'(t), the sa:a as the cons truction pc=it peuer Icvc1 of Oconec Nuc1 car Station Unit I which employs a similar reactor. Safety studies and perfomance analyses have been made for a power icvel of 2568 FM(t) for Arkansas Nuclear One-Unit 1. Tbc Committee believes that review of the operatien of Oconce Nuclear Station Unit 1 by the Regulatory Staf f should he completed and satisf actory performance of Oconce Nuclear Station Unit 1 should be demonstrated before Arkansas Nuclear One-Unit 1 is operated at full licensed power. 'i l
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Honorable Dixy Lee Ray August 14, Ic73 () jx- \ The hot functienal testing of Oconce Nuclear Station Unit I which was conducted in 1972 caused damage of soro components, including reacter vessel internals . The design changes which ture made for Oconec Nuclear Station Unit I have been applied to Arkansas Nuclear Cne-Unit 1. The ; Co==ittee believes that these changes are acceptable. l The applicant has been responsive to the Cc==ittee's reconnandation j that suitable instrumentation be sought to conitor for loose parts and i j for vibration; such instrumentation has been designed and 5111 be utilized. 1 I The applicant stated that he will prepose appropriate addi:icnal operating limitaticas if, at any tima during operatien, the ecdcrator temperaturc coefficient of reactivity is pcsitive . This matter should be resolved in a canner satisf actory to the Regulatory Staff. ; 1 The Regulatory Staf f has been investigating on a generic basis the problems l associated with a potential reacter coolant punp overspeed in the unlikely l I cvent of a particular type of rupture at certain locations in a main coolant pipe. Seme additional protective measures may be warranted rad this matter should be resolved to the satisf action of the Regulatory Steff. The Coa-cittee wishes to be kept informad. ! (' / The Committee reiterates its previous com cnts on the need for further j study of means for preventing commen mode failures from negating reacter e scran action, and of design fectures to take tolerabic the consec.uences of f ailure to scran during anticipated transients. The Cc=mittee believes it desirable to enpedite these studies and to i=plenant in timely f ashien such ) design modificatiens as are found to improve significantly the safety of th plant in this regard. The Committee wishes to be kept informed of the reso-lution of this catter. The applicant should assure himself that instrumentation for determining the courac of pctentially scricus accidents, on a time scale that will permit appropriate emergency action, is previded at the station and that appropriate calibration =atncia and calculated bases for interpreting instrument responses are availabic. In view of the important role of the applicant's Safety Review Cenmittcc in providing continuing reviews , and in updating and implementing safety measurcs, the ACp,S recommends that the Safety Review Cc=mittee include additional experienced personnel frc: outside the corporate structure as voting ec=bers. m
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Honorable Dixy Lee Ray - 3- August 14, 1973 l The applicant has proposed measures, including alarms 2nd administrative , procedurcs, to prevent operating under conditions which might resule in l exceeding acceptable fuel limits established from accident studies and ! other considerations. -The current review has been confined to the first " i fuel cycle and the analyscs have been based on the as-built fuel. The ACRS rcccmmends that the Regulatory Staff establish suitable criteria fer l t.bese measures, and provide suitable bases for evaluating f uture loadings . j The Committee wishes to be kept informed. i The Committee recognizes that re-evaluation of operating limits may be j necessary as a result of possible changes in the acceptance criteria for emergency core cooling systema. The Con:ittee wishes to bc kept infor 2d. Other problems relating to large water reactors which have been identified i by the Regulatory Staff and the ACR3 and cited in previous reports should be dealt with appropriately by the Regulatory Staff and the applicant as suitable approaches are developed. ll The Adviscry Cennittee on Reactor Safeguards believes that, if due regard in given to the items mentioned above, and subject to satisf actory ccm-pletion of construction and preoperational testing, there is reasonabic assurance that Arkantas Nuclear Ona-Unit 1 can be operated at pc.2r icvels , up to 2568 Mi(c) without undue risk to the health and safety of the public. I ( Sincerely yours,
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H.G.Mangelsd[jf d t Chairman J References attached. 4e a e. . e s
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- 4- August 14, 1973 Honorabic Dixy Lcc Ray 1
Referencen - Arkanons Mucicar one-Unit
- 1. Final Safety Evaluation Report, Volumes I through IV
- 2. Amendments 21 through 39 to the Application
- 3. Arkansas Feuer and Light Company (A?6L) letters dated October 2 and 25, 1972, transmitting lists of Bus Topical Reports for ASO-1 i
4
- 4. AP6L letter dated February 2S,1973, notifying AEC of its intent to incorporate the Winter 1972 Addenda of AS:2 Sectica III into the requirenents of a valve purchase order for A: 0-1
- 5. AP&L letter dated March 13, 1973, regarding requirements A:0-1 in
' clectrical instrumentation ann control systens at
- 6. AP6L letter dated April 11, 1973, furnishing information regarding engineered safeguards control circuits
- 7. APLL report dated April 1973, " Interim Report on Fuel Densification for AS0-1"
- 8. AP&L letter dated April 23, 1973, furnishing informatica on stress profiles for the main steam and main feedwatcr lines
- 9. AP&L ictter dated May 11, 1973, furnishing respenses to AEC require-ments for electrical instrumancation and centrol systems
- 10. APil letter dated May 11, 1973, furnishing responses to AEC require-ments to modify design of emergency cooling reservoir at ASO-1
- 11. DL Safety Evaluation f or A50-1, dated June 6,1973
- 12. DL Technical Report en Densification of BLU Reactor Fuels, dated July 6, 1973
- 13. Lotter from Mrs. Robert h. Douglass , Russellville, Arkansas , dated July 17, 1973, regarding ANO-1 and Subcccmittec Maecing July 26, 1973
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Attachment B --- WPPSS-1&4 ba ) \ WASHINGTON PUBLIC POWER SUPPLY SYSTEM (WPPSS), UNITS 1 AND 4_ Full Comittee Meetings Held Subcommittee Meetings Held June 5-7, 1975 May 16, 1975 Topics Discussed: Mangement Organization and Qualifications / Quality Assurance Organiza 0 National Circulation Capaility and the Proposed Natural Circulation Demonstration at Oconee. U Probability of Loss of Off-Site Power. O !ndustrial Security. 0 Environmental and Seismic Qualification of Instrumentation Equipment. OSite Characteristics. 0 0ff-Site /On-Site Power Supplies. 0 0conee Natural Circulaton Tests. oLoss of All AC-Power. 0 Emergency Planning. O
!nstrumentation to Follow the Course of an Accident.
UPlant Security. ACRS Co m ents were Provided in the_ Following Letter (copy is attachedl June 11, 1975 Summay of ACRS Comments and the NRC Staff Resoonses C ine Committee noted the previous recommedation for improving the Emergency Core Cooling System and recommended that the applicant c: studies directed towards improvements in the reliability of and capability I of Emergency Core Cooling System. This is being followed by the ACRS Subcommittee on Emergency \ Core Cooling System. l r O ,
-) WPPSS - 184 t(
U it would reserve judgment on the adequacy of The Committee stated that ification the Mark C fuel design until the Committee reviews the fuel qual program. The NRC Staff has reviewed the Mark C fuel as a generic item and has concluded that it is acceptable. O Ine Committee commented on the use of the hybrid reactor protection system (RPS-II) indicated that the remaining issues associated with the use of this system should be resolved in a manner satisf actory to the Staff. Tne NRC currently has this system under review. A V oTne Committee commented on the environmental and seismic qualifica I Class 1 instrumentation and electrical engineering and noted that an important aspect of this testing would be One development of an ac aging qualification program. Tnis' issue is being addressed by the NRC Staff in a generic fashion under Task Action Plan A-24. it C Tne Committee commented on the asymmetric load problem and indicate should be resolved. The Tnis asymmetric load issue is an unresolved ACRS generic item. NRC Staf f is trying to resol se this on a generic basis. I
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMIS$10N W ASHINGToN, D. C. 20555 (' June 11, 1975 Honorable William A. Anders Chairman U. S. Nuclear Regulatory Co~nission Washington, D. C. 20555
Subject:
REPOR" CN WASHDW1 PUBLIC PGER SUPPLY SYSTEM NUCLEAR PGER STA"'IC*lS hh"P 1 and 4 Eear Mr. Anders: At'its 182nd meeting, June 5-7, 1975, the Advisory Comittee on Reactor Safeguards completed%its esereview plants were of the the Washington Nuclear Power (WP) Stations 1 and 4. previously considered at a Subcomittee meeting on May 16, at Ric 15, 1975. Washington, and the site was visited on May n the Comittee had the benefit of discussions with reeresentative of Washington Public Power Supply System and consultants, thet e Comittee Babcock and Wilcox Comeany (BMT), and the NRC Staf f.
- also had the benefit of the documents listed.
We WNP Station site is located on the Energy Research and reveloprent Administration's Hanford Reservation in Benton County, Washington, eight miles north of Richland,The Washington, the nearest exclusion radius population center (1970 population 26,290) .We low population zone is 6400 feet. In 1970 there were 38 residents within the low populationReactorzone. We Fast Flux Test Facility and WPPSS Hanford-2 OCTP-2) are the only installations within the low population zone. We safe shutdown earthquake is 0.25g horizontal acceleration at
%e operating basis earthquake is 0.125g.
the foundations. For shutdown heat removal the plant has two sources of water, the operating water supply from a river intake on the Colu-bia River, which is not Seismic Catecory I, and Seismic Catecorv I spray pends designed to provide a 30 day energency water supply for each unit. W 4 , v
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( ( Honorable William A. Anders June 11, 1975 I ne nuclear steam supply system supplied by B&W is identical in design to that of Bellefonte Nuclear Plant, Units 1 and 2, previously reported on in the ACPS letter of July 16,1974. We design operating power is 3600 MN(t) . We reactor coce will use 205 B&W Mark C (17x17) fuel assemblies. %e Committee recomended in ii:s report of January 7,1972, on Interim Acceptance Criteria for ECCS, that significantly improved ECCS capability should be provided for reactors for which construction permit applications were filed after January 7,1972. %is , position was repeated in its report of September 10,1973, on 1 Acceptance Criteria for ECCS. % e Mark C fuel asserblies are ; responsive to this recomendation. %e new fuel assemblies will i be operated at lower linear heat generation rates and are expected to yield greater thermal margins for fuel design limits , I and improved safety margins in the analyses crf the loss of coolant accidents. An extensive program has been initiated for j determining the mechanical and thermal / hydraulic characteristics l of the new fuel assemblies. A program of control red tests also j is proposed, includim testing of trip times and control rcd , wear. Should modifications becone necessary as a result of the ! control rod tests, retesting of the entire control rod drive would ' be undertaken. Nhile many of the details of the proposed design are available, complete analyses of the perfomance of the Park C i fuel are not yet available, and the NRC Staff has not completed l O, its review. We Committee reserves judgment concernirs the final I design until the required performance information is presented i l and has been revie W . % e Cormittee recomrends that the applicant continue studies directed at further improvement in the capability l and reliability of the ECCS. We Comittee wishes to be kept informed. l We tac Staff has determined that the ECCS performance evalua-tion for WND Stations 1 and 4 reets the Interim Acceptance Criteria f of June 1971. In addition the Applicant's ECCS perfomance evaluation, using an approved B&W model, to show compliance with the Final Acceptance Criteria of 10CFR50.46 and Appendix K, must be reviewed and approved by the NRC Staff. We Cocinittee wishes to be kept informed. We applicant proposes to utilize a new reactor protection system designated as RPS-II. %e system, a hybrid usire both analog and digital techniques, represents an evolution from the analog system, RPS-I, currently in use in the Cconee reactors. We applicant has proposed a series of environrental, reliability, and in situ tests for qualification of this system prior to l l O l i l l
. i June 11, 1975
! Honorable William A. Anders
%is its use in Bellefonte Units 1 and 2, the lead plant.
ratter should be resolved in a manner satisfactory to the NRC Staff. mental and seismic qualifications of Class I instrumentation An important aspect is that of and electrical equipment. defining what represents an acceptable aging procedure for
% is issue should be resolved by multi-co ponent syste~s.
the applicant and the imC Staff. - %e Comittee wishes to be kept informed. , A question has arisen concerning loads on the vessel support structure for certain postulated loss-of-<:oolant accidents @ is matter should1 in pressurized water reactors. l for the hWP 1&4 Plants, in a manner satisfactory to the NRC Staf f, l Generic problens relating to large water reactors have I been identified by the imC Staff and the ACPS and discussed . j in the Comittee's report dated March 12, 1975. %ese problems should be dealt with appropriately by the imC Staff and the applicant. i te Mvisory Cottnittee on Peactor Safeguards believes that the items mentioned above can be resolved during construction and that, if due consideration is given to the foregoing, the Washington public B]wer Supply System Plants WNP 1 and 4, can be constructed with reasonable assurance that they can be operated without undue risk to the health and safety of the public. Sincerely yours, M William Kerr , Chairman "Merences Attached. O
l g . * ( june 11, 1975 Honorable William A. Anders ( Peferences 1. Preliminary Safety Analysis Peport, Washington NJCleat Projects 1 and 4. (Including Amen &mnts 1 thru 17) . l 2.
" Safety Evaluation of the Washington Nuclear 50-460, 50-513, Projects 1 and May, 1975, l
4", PUREG - 75/036, Docket tbs. QPR, U. S. Nuclear Pagulatory Comission, Washington, D. C. i 1 3. WPPSS Intter dated May 14, 1975, J. WPPSS J. Stein to Angelo tKx: lear l Gia-busso, DRL, C PR, US!!RC,
Subject:
l Projects 1;os.1 and 4, On-site Meteorological data.
- 4. Supplement 1 to the Safety Evaluation Peport, Istter from Voss A. tbore, Asst. Dir, for Light Water Peactors, Group 2 Division of Peactor Licensing, US!!RC to Dr. Willian Kerr, ,
Chairran ACRS dated June 2, 1975. , i O - l l l l l l l l
Attachment C, I v 4 UN11 E D ST AT C S Ci .[ ,.peute.9 ,\ NUCLE AR REGULATORY CO'.*Ml",SION j W ASHING TON. D. C. 2005s
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,1[*x T $'l f' ! April 20, 1970 fl0TE TO: Stephen Lawro ni, Chairman, ACRS (20)
Lee V. Gossici., Executive Director for Operations (2) Stephen l!anauer, Technical Advisor (1) Edson Case, Acting Director,llRR (3) Roger Coyd, Director, DP:t (?) liarold Denton, Director, DSE (2) Roger "attson, Director DSS (2) Victor Stello, Director, D3P, (2) James Yore, Chair:ran, ASLCP (2) Alan Rosenthal, Chairran, ASLAP (2) Cli f f ord Smi th , Direc tor , ("'.SS (3) Saul Levine, Director, RES (3) Robart "inogt.e, Director, SD (3) Ernst volgenau; Director, 1&E (3' Ho'crd Shapir,lELD (2)
/Q (2) t'orcicn lialler,
\ l, lDi rector , .PA FR0": Ken Pedersen: ~
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St!CJECT : REPC2', "FOLLO.l-UP C'l ACRS l.ETTERS" In view of the Ccuaissien's recent approvcl of this report,* I enclose a copy fcr ycur inforction and reference. The report was prepared at Cc=iissien request, by an OPE-led task force. Tne participants, who, beside OPE, represented the ACRS, NRR, and other I;RC g oups , are identified in Appencix li.
' ttemarc.ndun, "impicrenta tion of Recon =cndations f or 'Folle,t-Up on ACRS Letters,'" Samuel Chilk, Secretary, April 20, 1970.
Enclosure:
As stated CONTACT: ; A George Sege (OPE)
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The ACRS, at the Ccemission's request, has recently begun to provide j a periodic report to the Commission of items deserving Com attention. meetings between the ACRS and the Ccmmissicn to discuss matters of (It has oeen used
- i particular interest and/or concern to the ACRS. This has at the May and September 1977 Commissien-ACRS meetings.) ,
already helped to focus staff attention on items that are of cencern to the ACRS regarding staff policies, practices and priorities.
- 5. CRITICUE OF PRESENT PRCCECURES The present working arrangements between the ACRS and the NEC staff l with respect to folicw-througn on ACRS. advice are not entirely satis-factory to either group. This statement should not be construed to indicate tnat relations are strained or that there are serious cifficulties encountered; rather, in the view of each group (though these views vary), certain procedural changes could be made which would improve the process.
No major differences are distinguishable between the procedures In past l used for handling OL reviews as compared to CP reviews. e prcctice, both the NRC staff and the In generalACRSterms, have each conducted grouc a complete review at each stage of the process. has reccgnized that by the time of the OL review, many matters l related to plant layout and hardware have been committed, so that there is 1ittle flexibility to make substantial changes in piant design. Nonetheless, each group has continually pressed to have each facility updated as much as practical during the OL review. The ACRS is less directly involved with individual operating facilities. The Committee maintains a subccamittee with responsibility for each plant, and on occasien, the As ACRS a matter hasofasked course,for themeetings NRC staffregardinc one or more operating facilities. keeps the Ccemittee informed of significant occurrences at the operating facilities. Regarding the latter function, sucjectsd the
, for discussion with the ACRS are propcsed both by the AC NRC staff.
Zicn, the active role of the ACRS for each plant is much reduced when stage. the Committee issues its report to the Ccmmissi to an operating plant, seme ACRS members stress the need for systematic early dccumentation of staff plans in respense to the (The staff Special Report on Zion was Committee's comments. submitted about one year after the ACES report to which it respended . t a I i [ . . _ , , . , _ ,,. , , , , , _ _ , , _
s 6 t O The' result has been that in the eyes or a numcer or ACRS members, x the staff has not been as diligent as it shculd have been in
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pursuing resolution of the generic issues icentified by the
- Committee. Current staff efforts to identify, categori:e, and establish schecules for resolution of all generic ma :ers, including those iden:ified by.the ACRS, shoulc do much to resolve i this ma :er. When the final staff listing has been ;ublished, both the ACRS and the staff will, for the first time, have an
- identified ordering of the various technical generic matters and ;
t some idea of when the issues will be res:ived. Fue:her, :ne staff will have the means for applying its available rescurces to rescive the nost significant mat;ers en a pricrity basis. A particular scurce of disagreement between the staff anc the ACRS , has been the matter of adequecy of the emergency core ccoling l systems. The ACRS has con:inually pressed for improvemen:s in l the ECCS while the staff feeling is that i f an applicant meets the requirements of 10 CFR 50.46 and Appendix K, the ECCS is i adequate. Staff statements that they cannot require more than the l rules call for are countered by ACRS members by pointing out that the Ccmaissica can change rules when necessary or advisable. I On scme matters, the staff disagrees witn the ACRS on the significance , of generic matters identified by the Commi::ee. One example rela:es ) O to instrumentation to detect gross fuel failure; another, to the matter of a seisdic scram. The s:aff considers bcth of these items I to be of only marginal safety significance, but there nas been r.o established mechanism in the past :0 rescive such a difference in views explicitly wi-h the ACRS. Rather, pas: practice has been for the staff to give merely lip service to the i: ems of perceived lesser i safety sicnificance while it devoted its atten:icn to matters of greater concern to it. he present errorts to iden:ity anc cate;crire i as to safe:y significance all technical matters of concern should i 4 .largely resolve this problem. 1
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For its part, the ACRS has not in the past attempted to rank the varicus generic ma;ters as to safety significance. Curing the 4 August mceting of the Ccmmittee, hcwever, there were indications
; of member sentiment to establish some such ranking, inis shculd help the staff in the ordering of its priorities.
Effective . staff follew-through rests in part en a good mu;ual under-standing of views between ACRS and staff. Logistic problems stemming from geographical separa:icn between the staff offices in Se:hesca and the ACRS's dcwntown meeting rocm, together witn the difficulty of I a . e l l l l i i i
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t ( l adhering clo d a a pre-set discussien schedule at meetings, Some
,~ sometimes stano in tne way of achieving #w 4 c:71rsuch
- underst .
offices at ACRS meetir.p in order to be present for au t % M n tnat it minutes when their subject comes en-/ mb matter is the frustration micht come uo. Tha
- of tne Comitta ,.:@n .us que:ticns lead into a subject area where qualified cemser present to respond to the the staff has A rio closed circuit t A /ition link has been proposed d questi~cn.
which would help this matter, te. co long as the staff is loca:e in Bethesda and envirens while the Cc=ittee meets at H Street, this problem probably is not totali, n;tclvable. cceptability, or deficiencies, The Committee has rarely ccamented Copie: on a " SER supplements, in of staff respenses to its reports. _atineiy, provided to which the staff respenses are presented, arefor sy;mic Comittee reaction h'cwever, a mechanism the Committee, has never been established. A number of distinct problems for the staff arise to m Scme significant specific types of probin n1 the Cc=i s sicn . manner in which they have been handled are as folicws: 1. The prcblem wnere the staff is uncertain of the intent c' ne
. ACRS regarding one or mere of its cc=ents in its report toIn the Ccmmissicn.
4 the staff will request clarification from the ACRS In certain staff or instances, frem the ACRS Subcccmittee Chairman. where the uncertainty may involve points of major concern, the staff, by letter, formally requests ACRS clarification. The ACRS respcnds in a supplementary report tc the C or to the staff. l 1 of Appendix E. l
- 2. The problem where the staff believes a positicn reccmmended i 4
by the ACRS in its report to the Ccmissicn is unnecessary i or ' unwarranted or not in the best interests of the if program. i view and codifies its procedures anc/cr requirements acccedingly, In rare inctances, when the staff believes the ACRS made its
, recommenda'.fon on the basis of a misunderstanding, the staff ;
will discuss the matter with the Corrmittee at a subsequent l i meeting sta ff requirement. and obtain its agreement to proceed with tn l 4 Supplement to the staff's Safety Evaluation Report. (d ' example is discussed in Item 2 of Appencix E. Sheuld the staff continue to disagree with an ACRS recc= ended l 1 position that is more conservative and directly ocposed to a position taken by the staff, the staff would inferm ne
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.v.cm1ssicn of its deliberations en the matter and facter any An illustrative 'ss comments received int: its final decision.
example of a 3;ecific ins ance of this precedure is discussed ' in item 3 ci- A endix E. Nc withstancing cccasiencl cisagree- } ments, the staff recalls no instance wr.en the s af" pecceedad in ep;csiticn to a more conservative p sitica rec:mmandec by the ACRS,
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i.. r2 e e. c , ,. e. . ., s .. .<as. the Safe:y Evaluaticn Report usually respcnds to the ACRS - p..r2,a. r e., . . . ,. . . ,n.. . sn.. a.. .z..s1 w.n a c ,m- .1 ...=.n.... c n.... .. .. a n. s .an .s. The A RS resciution within its eng:ing long-term program. and staff reports lack specifici .y on these =2*:ers and :nare are c:ntinuing ccm ents from intervencrs, At mic Safety and Licensing Boards, and fr:a the public requesting mcre explicitness in these re;crts. Illustrative examples are discussed in Item a
,/~' cf Appendix E.
ro4.*,ivn ". n. .r . s.r~d $ d. k /' 4 T".. o. -g . . h. i s . . w h.e- r o. ' . * . ^ .. c '. .= f #.*e'ia.ves
. .a the ACRS in its re;ce cannot be accepted fully by the staff The because cf restraints ime csed by Commissicn reguia:icns.
ECCS mat:er discussec previcusly is an example of this ty:e cf pecblem. I' is cnly resolved by an eventual change in the regulations cr in the Commit:ae's cu:lcck en :he matter.
.s re:cr at a re,iatively late ine preciem o receip: o,, an cC<,
- . l pcin t in the schedule. While tne ACR$ is ccnsidered to c r. duct an ir.de;cr. dant review of an 2 plication, i: has by traditicn established a cra:: ice of net taking a ;csiti:n en a matter j
un:il the staff has firs; taken its posi:icr and pr:vided a j report to the Cecri :ee en that positicn as well as :he bases i The review by the Ccamittee at the j upon which it is founded.
- conclusien of the staff review is a necessary censecuence of !
- the value that.the ACR$ places on the staff revicu as ingut
, ._ M _the.CcEmittee's a n review. I
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O$ llCu# 4 fi I UNIT E D u.u i ., ( t ? j , f' t
; NUCLEAR REGULATOf: Y C ').'.P.i l S S I O N
(.,i Q:j'y [ [ q > 1., j ADVISORY COMMITTEE ON HL ACTOn SAF EGUAHOS w^smscion. o e m,3 JUlit 17, 1977 f llonorable l'accus A. Ro.rlen
. Chairsaan U.S, 1:uclear Pagulatory Cc r ission Washington, C.C. 20555 SUOJECT: P.EP3RT 0:: TF 21C.'; STATIC:1, U::I'i.; 1 A'O 2 Dent t . Poden: .
During its 206th m eting, June 9-10, 1977, the Advi:Ory Cor 'ttee on g
, R: actor Safecuards continued its revic' os the c'uration of the Zion ,
Station, Caits 1 & 2. The Co. 7ittec h..i previously discus ed operation j of tne Zion Station in its reports of /..:;u::t D , 1972, fi y 17, 1973, D:ceta r 9, 157':., and June 9, 1976. A Si.cymitte: reeting was h210 in K rosh:,;liccantin, on Mai 17, 1977, fulla.,i ,7 a rcur of the Cta.tlon by Comittee m:.-barc. DJring its revi: c, tne Cow.ittes h?.0 the benefit of di::;: icn: with represen;ativos of Un Ccnr...anith Edison C many
- 3) (Licen:.ae) , i.~estinghou. 2 Electric Corpration, anJ tho :: c} car i.o;;la-tory Cc'nisci:n Staff, t'e Cc rittee al:o haj to2 hen 0 fit of tne d cu-men s listed.
In 'ts June 9,1975 te:cr t, the Co .ittoc id.ntifiad ten carcrolved mat-ters Ust should b: dealt with in a tivly farJaion if the Zion re ::nts are to continaa to 'c.2 cpacted at full re .ar over the lifctime of the plant. The tcn matters idatified arc aa follo./u: (1) A revie. of the entire Sation for ryro m.s ink: actica that might lead to cignificant degrantion of safcty. ( ?. ) A review of the Statien with ri'.prd to dif ferenc?c fren
- current criteria, and judgm:nts cancerning roccible back-I fitting reqJire 2nts.
(3) '2he impicmantation of inntnrntatica to provide early inforuation concerning the cout ce of a full ranqa of ! postulated carious accidentn, anJ precedurco for inter-proting and relating this infot nation t.o emergency plcnc.
. (4) Installation of a loone-parts monitoring system aa soon as practicable.
f~ Evaluation and prc:pt implement at ion of improvements in g)
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{ ,l, (5) fire protection ccocbilit t y, an n 'cenCOy. 4-p 4 . , , ,,
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(J) June 17, 1977 ( Honorable Marcus A. Rowden l' (6) Tirely iralemnt at ion of tredi t icat ions required in connec-tion wi th the recolut ion of /6'..:1. i (7) Cont inued studico di rected t o <.nh.,nec.x;nt of the et.pabili y and .eliacility of the C.frgenc/ Core Cooling System. . l (8) Daronstrat ion of the relichilit y of the dienal g 'ncrctort to opcrate with luad for extcr iad pariods c tire.
. (9) Assoc::: cat of the cafety signi:icarn of the large nu:Ler of repor t able e vent s e>:perien. -
i at 'hc station, and l proret iO1c.unt at icn of cini cicant ircrovements in '
! operctional quality assurcnce.
?
{ (10). Prompt irglerrentct ion of i. npro c r ent c in indust rial security as appropriate. l The current st atus of these iters was the principal ratter of this l ) latest revie./ of the Zion Station.
/T The Co.:Tittee b lieves that little prcere..., hcs b.scr rade to.tard res-V Discu# ion with th: Licens:2 and tne ;n Staff olJtion of itc 1.
cu n ut.s that. thi. has ixen the result, c.t leaat in part, of a lack of undarstandi .g by the Liccr.sce and the W. Staf f cf just what is acant b.' "Evs t er; i r.i cract ion . " In thi~ rec.r:ct, th0 Cer.iittee c;115 attenticr to i t s let t cr of 1;0ve;x.2 r 6,197.;, t o L. " anni.g ; .nt zing , Di rcct or of Regulatica. 2.e Cori t t ee reco .renda that , wi t hir nine trenths , the Licenare scr it to t he .Tf. Staff, as a rini: : n, the results of a stuay of ryst er.= int eract ica relat ing to t h.' co .nibili ty t r.at failurt of catet) 2. i nan-cafety cyster.3 ti l 2 int'cre 'ith the plant crcratcrs' ahility to aceeg)ish shatdo.m heat renca'. , t.cp her wi th a plan cr.d schc Me for rt adies of ot Scr sycten int. ract icn; of potent ial safety significance ts t he '.',i on St at i on . Wit h respcel tc item 2, for which lit t ]c han tren danc, the Comittee recca:rendo that , during t he next twelve mnt hs, t he Licensee review t he Zion St at ion for poscibly signi ficant di!!crencec from current criteri1, and t h at the NR St af f evaluat e t hin revir u anJ repor t to the AOG its conclusion conecrning runiolo bacsfit t ing re.:;ireants. I { The Cor.M t t ee wi ches t o revi ew t he cl at un of i t en;c 1 and 2 wi t hi n t he t next eight een mant hs. Items 3, 5, 6, 7, and 10 are concidered by tot h t he ACIG and t he NI C (j St af f Io b2 unrenalved mat ters generic to all operat ing light wat er v e i - - - - ====e.-~==-~~~==e-' * * * *
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1 i i 1-1 i (' 'i 1 ( 'l ilonorable Marcuc A. novJen -3 June 17,1977 l l (~~^ '
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! reactors. During the pact twlve mnan, com prcgrecs has been made 1 I toward the rccolution of thoc? 9:neric inr: and in the planning fnr the application of appropriate colu' i. to thn '2cn Units. The Con-I mittee reco= ends that t.he NR: St:M i. ' t.icanric urgently seek treano I
to c;qrediw colutior.s to outstanding . r ic it.r.rq and the inplementa-
.' tion of solutionc, when fec3ibl2, to : '
.nn Sta' ion. i The Licencee hac mado a c --"itTent t , r all lon c.-parts monitoring cyctems en th: tw Zier St: tion rea tm c,: ring U.e 1978 re fueling out-I ages. %c NRC Staf f considers item a r. :alvcd. %e Committee concurs. )l t
; In resp:nse to itm 3, one of the di;. .l yncrators was run continuously l
- for seien days at a con .rcl103 m..;r o... ..;t e.uit. Aent to th? ECCS loa]. l k'hereas the unit ran sotisf actcril . d. n.] thi p riod, the -ignificance of the test recultc in confirming the c : bility of the emerge"cy o0wer l e . ., . ... .c ,,
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-; . .., aC . ~.: L~o uw b. w ^. l sever.th day, an op3 rater er ror led to a larce cat ge in the lon and l l the doctruction of the generator by fire. C: gentr: tor failure van l the result of an unanticic ted int;.2;t m bet.ceec t're main electrical .
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work in p cgrenn, and tne ct:rr grou,' l' it ntif cr.d resalve proL'1;ms l prorptly. The 'm: Staf f hac also ir : si c an en s to the ~. ion licon o which revisa the entire aJ-inistratiw c >u rol c..ction of the Teeb '.ical l Sycifications. We Corc.aittec conclu b that t!c: . actionc of the Nn: Staff and Liconq,2 are res pasive to i*r o and encaJrages the Licomm ? to coat.inue te seek forther it.proverc 'at in thena areac. Tne Mvisory Co: ::ittee on R. actor Safegualdo N1ievec that, if due regard l is given to the itetn untioned a'cov , t h 'rc' ir reanona' ale assur ance that the ;'. ion Station Units 1 and 2 can coat ina) to e n ate at full power , l 3250 N , without undue tick to the h.c.ilth and cafety of the pu'alic. . 1 i 1 , l Sincerelv, - ! t t i / \ l
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-4 June 17, 1977 lionorabic Marcus A. Powden (v')
i References
- 1. Report to the ACG by t'ic Of fice of : -lear Reactor Regulation, U. S. Nuclear Regulatory Co;,;aiocion in de :1;.Lter of Cccumtea.' th Edison Cor. pan / Zion Station, Units ] a. . 2, dated I.ptil 22, 1977.
- 2. Sapplemnt to the Report by the CCCico of Nucleat near. tor nogulation, NRC, to the ItCPS concecning Zion St.i'. inn, Units 1 and 2, dated Jur.e 8, 1977.
- 3. Letter fre Co..mzealth Mi90n Co. to the Of fice of Nuclear cactor 20nitoring Regulation, 3C concerning ir,stallt.t i;n of a loose p;rt:
, system, dated January 1,1977.
4
- 4. Letter frc cot.ronvisalth Edison Co. to Um Of fice of ':uclear Ge30 tor i Regulation, GC, concerning ir.for.ation T on gaccous releases f rot Zion Station, dr.ted April 1,1977.
I i t 5. Letter fro.: Ccaun.lealth Edison Co. to the Of fico of ':J: lear Ranctor
. ~< Regul: Lion, NRC, conccrning a su- :..ry oC the diencl ganoratcr tast, l
dated April 6, 1977. . C Letter f rom Comnsealth Mica". Co. to ui. C:Cice of ::u len Kn: tor Y 6. Regulation, ':r, enncerning a recxt en fire protcetica, datt.d April 23, 1977. i I i
'b o
V i I f . . _ _ . . _ . . . _ . _ , _ _ _ _ _
ADVISORY COMMITTEE ON REACTOR SAFEGUAROS ' NUCLE AR REOUL ATORY CO .* MISSION [ W ASWNGTON. O. C. 20Z5 7- v q , (v JU.iE 9, 1976 t. 6 I Honorable Marcus A. ParJen Chairman U. S. I'uclear Regulatory Co missica h'achington, CC 20005 RCPC:!f ON ZIO:' S~'ATIO:' U7ITS 1 IJO 2 SLP_J Es.'T : i i
Dear Mr. Rowden:
I
! During its 194th meeting, June 3-5, 1976, the Advisory Co r.ittee on Reactor Safeguards ccTpleted its review of the proposal to increrse the maxi.T.um rc:ctor power of the' Zion Station Unitc 1 and 2 fror. 27G0 l :t (85% of full power) to the rated po.er of 3250 L.it. The Co mittec had
~
previously diccussed operation of the Zion Station, in its reporta of August 17, 1972, May 17, 1973 and December 9, 1974. A Subco- ittee M1976, et-
) '
ing on the current promcal was held in Mcnocha, Uisconsin on l'ay 27, (T subcequent to a visit to the site en May 20, 1976. During its revia.;,.an- the
'y' Co- tittee had the benafit of discuccions uith represanratives of Cc I
wealth Edicen Corpany, Uestin9.ou30 Electric Corporation, the :;uclear Regulatory Co =ission Staff and of the docuants listed, as well as com mants from memb2rs of the public. In its previcus letters relating to the Zion Station, the Comitta : listed I I i l
' several concerns which, in its opinion, mitigated againnt the full pc. ar l operation of these larce reactors at a site having a significantly Ingcc '
than average populatica density. Three of these concerna, relatira to fuel behavior, centrol of core po ar distribution, and relitility of diesel gen rator etcrtup, have been resolved to the satinf actica of the NE Staff, as reported in its Safety Evaluation Report date] (:ay 20, 1976. ; 1 I The ACRS believes that with the resolution of these three matters it is I acceptable for the Zicn Station reactors to be operated at full pmiar. '
!!cwever, the Cor,ittec Lwlieves that other matters should be dealt with in a timely fashion if the Zion rcactors are to continue to be op: rated l at full power over the lifetime of the plant. The ACCS reco.r.nen.b that !
these matters, as set forth belo.i, be addressed by tha Applicant and tha l tmC Staff during the next year: j
}
\ n .] U . i l l _ , .
Eonorable Marcus A. RouJan June 9, 1975 O v\ s (1) A review of the entire Station for systems interaction that night lead to significant degradation of safety. (2) A review of the Station with regard to differences from current criteria, and judgments concerning possible back-fitting requirements. (3) The brplcmentation of instrumentation to provid2 early information concerning the cource of a full range of postulated serious accidents, and procedures for inter-preting and relating this information to emergency planc. (4) Installation of a loose-parts monitoring system ac coon as practicable. (5) Evaluation and prompt implementation of improverents in fire protection capability, as nec2ssary. (6) Timely implcrentation of modifications required in i connection with the resolution of AT.:S. (7) Continued studies directed to enhance ent of the cacability and reliability of the B:ergency Core Cooling Systers. ,
- f
(_s/ (C) Demonstration of the reliability of dia diesel generators : to operate with load for extended p:riods of time. l (9) Assessment of the safety significance of the large number l of reportable events experienced at the station, and ! j prorpt implementation of significant improvements in l
; operational quality ccsurance. l
, 1
, (10) Promet L olementation of improvements in industrial l security as appropriate.
! l l
Cth r ganaric prob 1cm3 relating to large water reacters are discucced in e l i j the Committee's report dated April 16, 1976. As solutions to those prob- i Ic .3 are found, both the Applicant r.nd the : fC Staff should give a high priority to their prompt h:plementation at the "ica Station. The ACPS wichas to review the status of those matters by June of 1977. The Advisory Cc mittee on Reactor Safeguards b?lieves that, if due regard i l ic given to the items contioncd above, there is reasonable ascurance that { the Zion Statica Unit 1 and 2 can bo op3cated at full powcr, 3250 l' cit, l without undue risk to ute health and safety of the public. As noted above l (~) f')1 x.s 1 . i \
- - - - - - - ~ ~
June 9, 1976
~- '
.
- Monorable :4 arcus A. Fr. den l g
V '.-) ( - the Comnittee will again review the oTration of the Zion Station in
! approximately one year.
I Sincerely yours, p.AA $'// b Dade W. Moeller Chairman REFETZ:CES
- 1. Safety Evaluation Report on Zion Station, Units 1 and 2, dated May 20, 1975.
- 2. Supplement Mo. 1 to the Startup Test Report Iczued !!ovizber 1974 l on Zion Nuclear Power Station, Unit 1, dated 7.pril 16,1976. .
- 3. Letter from E. Jenkins to ACTS concerning full pcwar operation of the Zion Station, dated June 1976.
Letter from D. Comey to S.G. Case, concerning reliability of 4. emergency diesel generators at Zion Stations, dated Auguat 28, 1974. r~ ()s (-)
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Attachment C. 2 f i O NUREG 0496 l i 9 i i f I i ! 1978 REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION
- SAFETY RESEARCH PROGRAM s
l A Repod to the Congress of the United States of America lO i i i l 1 Manuscript Completed: December 1978 I Date Published: December 1978 , i 1 O Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission Washington, D. C. 20555
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4' large progra s in these areas, resulting in greatly i proved equir-ment and te:".iqJes. ~hese develop ents are available to the NC.C and are being taken i .to at ount het additional work is needed. The h"RC p r og r a- is addressing reans to ensure that the calibration ani handling cf instra ents and prore"ures used by licensees for the ,easure e.: o f TP' w!!1 provide results which are reliat'. related to abse'ute standards. . 10.4 Additional :esear:P Needed Particular questions related to possitie alternative fuel cycles are not bei q dire::1y addressed in the present prccra . An exa ple e r sa:" a proble- is that of the measurrent standards and pr ocedur es V.!:h nay be neede" for .aterial control and inspection. Althou ' work undertaken new may never be used if a pa rticul e r fuel cycle
-ion is net adepte d , some of the require-ents in this field involve q :i te finick, an? necessa rily slow-neving wo r< which will take ,
several years :: co..plete af ter a program has been initiated. Sough d it na', be the proper Eclicy at the no ent te defer such sta ies u .til decisions are rea:hed as to @ich new fuel cycles are to be con-d sidered for actual use, there will be a strong naed to ad sur' studies to the present progra . as soon as fJe1 Cycle decisions are mad e.
.O M 'as . year'r. repc rt , the A^RS re:c- enied that aiditienal ef fert be dire:*a" cecera' items, including:
. (1) alternative fuel cy:les: (2) alternative (below grade) locatic .s for spent fuel pc:1s; (3) dedi cat ed heat renoval systems: (4) increase" sepa r a tion of ra."-'..~".=...*. s='a..",. .-a.'
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te exa-inad #.e- feasi'le; (3) is included in 5 C's new prx ra of research on i-: roved safety systems; (4) is receivin: sore attention
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10.5 Evaluatic. an? Co-- ent
-'he particular proble-s being addressed, and those planned for FY 79, are all relevant to NRC needs. Though it has not been pessible to acco- cdate all the items deserving attention within the present at e ara , the sele
- ion of topics for current funding would see . te be cuite reasenable. The centractors chosen appear to be fully capable.
Good use is being made of info:- ation available from other sources. With one potentially werrisene exception, discussed below, the research progra en safegua rds and security arrears to be well for-.u-lated and wil nana Jed. O v 10-3 l
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9 (J / t t 8.3.3 Pro:ress and Resalts Tne ACPS concludes that satisfactory progress and results have been o:::ained thus far by the NRC on its safegJards research. Hewever , the ACPS believes that it vauld be desira:1e to develop scenarios and protective measures that ta<e into account the exporience, eriteria, and procedares of law enforcement and military organications and app!y this information to NRC's regulatory needs in lieu of the independent model develo; rent and validation by NRC. For soce tire the NRC has strongly supported wer < to i prove the accuracy and reduce the tire regaired for SNM reasurecents. Tnis has resalted in better raterials accountability which is an Lportant co ponent in one of NRC's regulatory roles, na.ely, Working coopera-refutation or confirration of SNM diversion. tively with DDE, the hr is supporting studies on newer techni-qJes for materials accountability wnich have the objective of real-t1.re essentially instantaneous accounting of SNM so that inventory of SNM can be obtained in licensed facilities. 8.3.4 Naeds for Additional Research The ACPS believes that it would be desirable for FIS to have additional research performed for purposes of assessing the effectiveness of techniques being developed by otner organi- l zations for procpt detection of atte pted diversion of SNM. Confirmatory research on the applicanility of new techniques to accom:date alternative fuel cycles should alsc be under-taken. In add ttion, the ACPS celieves that conceptual design studies l and'or probabilistic risk assess ent studies may be warranted ) in connection with the following exa ples of design changes that could lead to improved safety: In its review of the physical security of nuclear ;mer plants, the ACF3 has noted and expressed concern on many occasions that storage pocis for irradiated fuel are frequently located at ele-vations, relative to grade level, such that breaching of a pool wall would result in partial or total loss of shielding and cool-ing water. This could lead to reiting of spent fuel and, perhaps, even to some volatilization and dispersal of fission products.
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With regard to providing better assarance of shutdown heat removal in the event of plant accidents or sabotage, a bunkered, dedicated, shutdown heat removal system with independence from components of the ragular heat removal system might signifIcantly enhance reltability. Increased separation of redandant safety-related cogonents or fa-cilities co;1d contribute to plant safety b,. reducing tne .aalner-ability of sucn facilities or cogonents to possible un-anvisioned plant accidents or to acts of sabotage. 8.4 Conclusions and Pecomendations 6.4.1 Conclusicns Tne ACF3 believes that the NRC has a satisfactory safeguards research program and that it generally has been well formulatad and co petently managed. Thus far, the absolute and relat tve l levels of funding for safeguards that have been provided in the I total FIS budget appears reasonable. Modifications i guards researcn pr:cra- including some increase, may beco .n thaesafe-neces-sary snould the Administrat ton's non-proliferation policy te i. pie-mented by adoption of alternative nuclear power reactors and fuel cyc'es tnat differ significantly from present LWRs and f;el cycles.
\
3.4.2 PPCor andations The research efforts in regard to the protection and accountariltty of 3 N snould ca realigned i. order to allow more support of work to enhance tne evaluation and application of tecnnig;es which are capatie of pro ptly detecting diversion of SNv.. Studies of tnese technig;es sno;1d te encouraged to assess their reliacility for use in locations sucn as ingress and egress points of nuclear f acilities whicn handle large quantities of SNM in various physical and enemical forms. The SRC should also support studies to determine the feast-bility of locating diverted 3:F. at locations such as open terrain, cities, rivers, lakes, etc. The ACPS believes that tne NRC should support research on design concepts that make sacotage more difficalt to perfor.T and less likely to be harmful to public health and safety. So. e spacific exa. ples of design features which the AC?S te-lieves have the potential for improvements in th? designs and for which the NRC snould fund conceptual design studies of a ge-neric nature are: (1) alternative locations of the irradiated t l /~' - d'4 - l C 1 E t Gy n!: r ; . *. . .;;,.: i.. i_ an-- __ ,b ,- --
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I t h' * %g 3 UNITED STATES [ 3, e ([' I, NUCLEAR REGULATORY COMMISSION
, ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
*[% , ' ' [#-[ k @O YSepte.ber
?-
f j WASHNGToN, O C. 20555 16, 1976 Honcrable P. arcus A. Ro den Chai=an United States Nuclear Regulatory Co=ission Washington, EC 20555 CIAPlFICATICU OF AUGUST 17, 1976 ACRS REPor Gi DESIG1 PN/ISIG:5 rTR PROI'EC'"ICN AGAIK5T SA3mGE
Dear Mr. Rowden:
In order to provide clarification of its report of August 17, 1976, on Design Provisions for Protection Against Sabotage, the ACPS offers the followi ; additional infer: ation. Project reports in which the ACRS has recently co=ented on the need for stMies of design provisions to protect against sabotage are listed in t Attachme. A. In these reports the Co=ittee has reco= ended that the NRC Staff and the Applicants review the proposed pla-ts for desip featurcs that could reduce the possibility and conseqJences of sabotage. In its report of Cctober 14, 1975, on indastrial sabotage the Cc=ittee noted:
"Tne NRC Staff has taken steps to reduce the possibility of industrial sabotage at nuclear power plants, particule.rly with regard to control of unauthorized access, although detailed criteria have not yet been developed and a considerable vari- l ation exists among plants currently proposed for construction permits. We Office of Regulatory Research has also fucded useful studies concerning possible medes of sabotage by individuals or groups external to the operating organization. Some reco= endatiens regarding possible design changes resulted from these studies.
1 Be ACRS has recorrended in several project re;crts that deliberate ) attention be given to aspects of desi, that could irgrete plant l security. We Committee believes that, at this time wt e; phasis ; is being placed on standardized plant designs, increase attention I should be given to design reeasures which would further protect against indust. rial sabotage or mitigate the consegaences ther ,f . We ACPS would be willing to cooperate actively with any special NRC-s:cnsored workinq group in order to veelerate achieve ent of O
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t8ese ohsectives.- 1 l l l
O Bonorable Marcus A. Ro den September 16, 1976 Tne studies made by and for the NRC Staff to which the ACPS was referring are those recently made by the Sandia Laboratories on the vulnerability of nuclear power plants to sabotage. Tne phrase, " bunkered,. dedicated syste s," means systers whose sole fu*.ction muld be to assure decay heat re.cval under emergency conditions. Tnese systems would be physically secured against unauthorized access and wuld include all necessa.ry instre.entation, control, power, and cc plete cooling capability. 4 By " spatial redundancy," the Co=ittee intended to convey a degree of physical separation which would make difficult and unlikely the cc plete loss cf a capability currently provided by physically redundant, although , not necessarily adequately separated, syste .s. S' erel" yc" s, J 4,/. ,
/ } '.
Dade W tbeller Chairman Attachent: A V
1 Attadr.ent A January 13, 1976 Interir. Paport on Koshkeneng Nuclear Plant, Units 1 and 2 l February 11, 1976 Interi.. Peport on Pebble Springs l i Nuclear Plant, Units 1 and 2 I February 11, 1976 Report on S*,.TSSAR-PI,
. Stone and 4 Webster Engineering Corporation Balance-of-Plant Eesign April 16, 1976 Reprt on Washington Public Power !
Supply Syster. Nuclear Projects ' No. 3 and No. 5 l May 12, 1976 Report on Kochkenong Nuclear Plant, Units 1 and 2 May 13, 1976 Report on Eartsville NJ: lear Plants, Units A-1, A-2, B-1 and S-2
- June 11, 1976 Report on S;.TSS?+P1, Stone and j
Webster Engineering Corporation Balance-of-Plant Eesign as Applied to Cce.bustion Engineering, Inc. CESSAR-30 July 14,1976 Report on Westinghouse Electric Corporation Reference Safety Analysis Report, PISAR-3S A; gust 18, 1976 Report on ST:SSAR-P1, Stone and Webster Engineering Corporation Balance-of-Plant Design as Applied to the Westinghouse Electric - Corporation RESAP-35 NSSS Design Septecter 15, 1976 Interir. Report on Montague Power Station, Units 1 and 2 O U l
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
'I NUCLEAR REGULATORY COMMISSION W A$NINGTON, D. C. 20555 Aui;us t 17, 1976 Honorable Marcus A. Pa.rJen Chairman U. S. Nuclear Regulatory Co=ission Washington, EC 20555
Subject:
DESIG; PRCT/ISIO!S POR PRCfrECTIO; AGAINST SA57? AGE
Dear Mr. Rcwden:
The ACRS has reco: mended in reports on several projects, as well as in a re;crt to the Co = ission dated October 14, 1975, that deliberate attention O be given to aspects of design that could i prove plant protection against ( sabotage. In its continuing review of this matter, the Co=ittee has had the benefit of studies made by or for the NP/; Staff, as well as other information. The ACRS believes that it can now make some specific
. reco=endations regardc g studies that shoald be initiated, as follem:
- 1) studies of the location of the irradiated fuel storage pool with regard i to protection against overt action which could lead to a loss of water for cooling and shielding;
- 2) stu3ies of bunkered, dedicated syste s cap.1ble of accomplishing shutd:.c J heat removal for many hours;
-3) . studies of spatial redundancy and the possible merits thereof. j Sincerely yours, L'r
> -Dade W.140eller i Chairtan l 4 -.
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i i NUCt. EAR REGULATORY COYY!1SION ! W ASHING TON. D. C. 2c155 O . o toeer 14, 197s - l 4 l l Bonorable William A. AMers 1 Chairran ! U.S. Nuc] ear Pegulatory Ccenission Washington, D.C. 20555 _. SC1TDCT: M.g_CU DCXLTAIAL SABTAT_,. __ , Dea- Mr. Anders:
%e ~.:FC S"aff has taken stepc to reduce the possibility of iMustrial sabotage at nuclear pc w r plc.nts, particularly with regard to centrol of t.r.authori:ed access, althous, detailed criteria ha/e not yet been develo, red and a cenciderable variation exists amorn c]nts currently i
properd fec construction per .its. Be Of fice of Prgu>atory Researc.h has also f' nded useful studies concernirn possible rnodes of cabota:c by individuals or groups external to the o,: crating organi:stion. Sem recet ervlaticas regarding possible design cha:7;es h- resulted frcr. these studies.
; he ACRS has rccer mnded in severa1 project reports that delib' rate attention be given to aspects of design that cou.16 i::. prove plant security. ?..e C.. =ittee believes that, at this ti.m2 when e phasis a
is beira placcd on standarized plant designs, increased attention should to given to de:igr masures stich kculd fur t.her protect
; against industrial sabot ge or mitigato the consec3. >:nces thereof.
j 2e ACF:S would te willing to ccc:erate actively with any special 1mC-spennored workirrg group in ordar to accalerate achievement of those objectives. t Sircerely yours, i W. Kerr Chair:an
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4 NUCLEAR REGULATORY COMM!SSION IT ASHING TON D. C. 20556 _) . Cttober 14, 1975 ' Ibnorable William A. A.-ders Chairran U.S. Nuclear pagulatory Cctr.ission Washington, D.C. 20555 SC2 JECT:
.R_%._ _E. ._.Q !. _I.?C.CS__TR_IA_L S.AB. CG__T d _ . . .__ _ .
Dea- Pz. /Mers: Ee :FC Staff has taken steps to reduce the possM1ity of iMustrial sabotage at nuclear pc-sr picnts, particularly with regard to centrol of tnauthori:ed access, althoup detailed criteria hate not yet bcen developed and a censiderable variation exists amore ola tts current.ly i
' proper'd fcc cortstructica per~its. 2e Office of F.uguia tory Fesearch has also f' nd0d useful studies concernire possible mcdes e
of cabotage by individuals or groups external to the operati.m organi:stion. Sc e recor erdations regardirr; possible design changes O . resulted f:cr. these sttdies.
; 2e ACRS has tw . ended in several project reports that delib2 rate attentien 62 given to acx<ts of design that could i . prove plant recurity. T..o Cc r.ittee believes that, at this ti. c sten e phasis I
is beim placed on standari:ed plant designs, increased attention shou.!d to given to de:ig rea:ures which would fw ther protect l against industrial satet.:ge or mitigate the. consc<.mnces thereof. . j te ACES wuld te willing to ecc,crato actively with any spccial
. NFC-s;crected workirr; group in ordar to accelerate achiever. ant of those objectives.
; Sincerely yours, i
W. Kerr
% air:an t
I e I l t
l l NUCLEAR REGULATORY COYY!SSION n AsHINGTON. D. c. rosH October 14, 1975 ' i Bwarable William A. Arders Chairr.in U.S. Nuclear Pegulatory Cctr.ission Washirgton, D.C. 20555 , SCSTCCT:
.Pl.~:G_FT.- CI:..DC.G TRIAL SM.C:X d ..
Daa - Mr . Inders : 1 m? E Guff has tahn steps to reduce the posslility of inSustrial sabotage at nuclear pcwr plcnts, particularly with regard to centrol ' of unauthori:ed cccess, alt.%v. , detailed criteria ha/e not yet tx-e.n developed and a censiderable variatien exists amorn cints currently i
' propard fec construction per-its. We Of fice of R=cuia tory Fesearch has also f' nded useful studies concernity possible rxv3es ,
of sabotage by individua.ls or groups external to the operatirg ' organi:3tien. Sc e recet erdations regardire possibic design changes resulted fra- these studies. h
; 2e ACFS has roccmended in several projcat rep:rts that deliberate attentien be given to a..aects of design t. hat cou.1d i:. prove plant security. T..e Cc .ittee believes that, at this ti. e when e phasis i ; is being placed on standarired plant desiens, increased attention should to given to decigr reasures stich v;uld fu ther protect l l' against industrial sabotzge or mitigate the cor.x<uences thereof.
j te ACT:S teuld te willing to eco;erate actively with an sgecial NFC-spcrcored workiry group in ordar to accalerate achiever.2nt of these objectives. I Sircere.ly yours, t Y W. Kerr 1 Chair an l l l l l
, 1
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Attachment C. 3 NUREG 0603 O COMMENTS ON THE NRC SAFETY RESEARCH PROGRAM BUDGET O Manuscript Completed: July 1979 Date Published: July 1979 Advisory Committe on Rea: tor Safeguards U. S. Nuclear Regulatory Commission Washington, D.C. 20555 $
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w . . . . . e., .s. Ec s... . , '.' a '. ed a . . .i .de . .*. s .i n.% ^ .' 'v' . i .nc, * *., r e 3 fr a cr0ad range of land-based reactor sites '.atOse Characteristics are reason-s;tes in use, pre e:tec for use, er c., c, at., representative . Suct an ef fort has already pcte.tiai interest in lenc-ter . cianning . The de::th of the teen initia ed as c. a r t of tne NE research revide the eroc.ra ...d inferration needed cackgrou Cr:cra.- sh:Jid te s;f ficient to e
.for the pess:bie deveicprent of hydremgic siting criteria -tich allow -
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A stad should be made of the relative and absciute a:0ident risks, with uncertainties, fer a wide rarse of potentially suitable sites. The study chould exa-ine the costs and benefits associated with dif ferent types of i;tes and should in:1cde the Ecssible interactionThe of aintent sericus of accident the study in one rea:ter en other reactors at the site. stu.d be to erevide insight into the relative advantages and disadvan-tages of more remote siting and pcwer prks. f 1 l
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NUREG 0496 1978 REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION SAFETY RESEARCH PROGRAM 1 I A Repod to the : Congress of the United States of America l
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Gl Manuscript Completed: December 1978 Date Published: December 1978 A f l l l l Advisory Committee on Reactor Safeguards h U. S. Nuclear Regulatory Commission Washington, D. C. 20555
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n (_ e me* C 3.6 Release and Transport of Molten Fuel and Fission Preducts Although accidents that could cause a significant fraction of the reactor core to melt are extremely improbable, the potential danger As a to the public is substantial if such an accident should occur. l result, the N3C believes it prudent to obtain better information j to scope the consequences of accidents of this type. In addition to l the NRC prog ra.9, the NRC Staff follows the work of others, in par- . ticular the substantial FRG research program in core-melt / concrete l interactions, fission product release from rnoiten fuel, and the
, interaction work in many countries. The NRC is cur- -
fuel-coolant The ACRS rently planning a reduced level of effort in this program. the believes this area of work is important and is concerned that NRC effort in this area not be phased out. The ACRS believes that '
; careful consideration should be given to the needs of the licens- risk ing staff, and to the usefulness of such studies for future assessments and studies on improved light-water reactor sa f ety.
3-3 1
h ADVISORY COMMITTEE ON REAC' TOR SA EGUARCS NUCLEAR REGULATORY COM.YlSSicN
' W ASHINGTcN, Q. C. 20$$5 ..
De c er.b e r 10, 1975 -
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D:ecutive Director of Cperations U.S. ' uclear Fagu' atory Cc=.issica h*ashi.vg.cn, CC 20555
Subject:
FIKP:' CM Rrl~r. CF Sr!2K; KLICM KR LICD:52;G !TC"d?. _ FACILI"~.J:S :- Cear P.r. G:ssick: . s .1...
~.n respense to a request f:c= the CC Staff, de Siti:rg D aluation Subc.....iittee of the ACPS met with at-bars of the Civision of Sit _i.rg, ;
Realth a-4 Safeguards Standards en rece-ter 2,1975, to discuss siting -:clicies. Ois ratter was the subject of f=$er delibera- M tien.s bv ce-hers of the Siti:n. Ivaluatien Sdec=.itte- en rece-ber 3,
. i W
1975, and was discussed during de 187th and 1882 Meetings of the - full C- =ittee,1::verter 6-8, 1975 a.rd recche: 4-6, 1975. & i Se Cc=.ittee u .derstands that the pri ary cbjective of the current ,
- z. l review by the ::IC Staff is to dete=.ine if c'.arge: in c= rent sitirs l
4
.=clicies are ces:rable. ?
l In reviewing e.xisting criteria fc: possible i, p:cve ents, the Cc=.it- . tee recc= ends that the fellcwing iters be cencidered: _ t
- 1. D:ter.sion of current siting criteria to include reacter ty:es i other than '.S?s (for expla *m and EFE.?s) , as mil as -
t i floating nuclear pcwer plants.
- 2. Se 1: pacts of major accidental radienuclide releases on t,ater ,
resources, such as underg cund agaifers ard nearby lakes.
- 3. Peevaluatien of the bases used for setting dose limits for the q-resign Sasis Accidents. .
,. revele=ent of suit.dle criteria fc: acceptable risks to pecple . ;.
h (both ' individually and ecliectively) living in the vicinity of a site. tis effort will also require censideratien of asso-i 4 cicted benefits. I 1
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e P.r. Iae V. Gossick Dece:ber 10, 1975
- 5. Cesirability of specifying a minim = sice for the exclusien area, a - ~I minimum radius for the I?:, and a maxi == population distribution
~
in the area surrounding a site. A pctentially valuable input micht be the siting policies and experiences developed in densely populated countries where a sucstantial level of nuclear pcwer is in place or planned (Ger any, United Kingdem, Japan, and France) .
- 6. A " figure-of-cerit" for population dist.ribution, and perhaps for mete- ,
orological characteristics, for use in site selection. A plication of the consegence rcethcdology of iGEd-1400 to a range of siter . ay prove ! to be useful in cbtaining an i. proved basis for choice of a figu:e-of-merit.
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- 7. Inclusion of seismic considerations. .
- 8. Assessment of interactions between approved sites and changes in the surrounding environment. Fcpulatien increases and chances in the .
characteristics of industrial, cv.adreial, or recreational activities ! in the neighbothecd of an cperating nuclear facility can have signif-lh icant i. pacts on the centinuing acceptability cf the site. Fe aa.s [' I for predicting and dealing with such changes should be investigated. ; te development of me=cranda of understanding between the hAC and cther ; govern ental agencies, both federal and local, might be an effective i acoroach. r.
- 9. Extent to which require ents for end-cf-life deccer.issicninc and -
potential plant replace ents affect siting policies.
- 10. Short and long term censegences of a .ajor accident in a nuclear ;
installatien en other operations at a mulci-unit site such as a nuclear pcwer park. ; l
- 11. Se nu-ber of sites recuired within a given re;ica over a specified !
pericd of time as a factor in deciding on the type of sites that might have to be accounted for in the siting criteria. i
. f
- 12. Considerations of national defense and industrial security. ;
13., Evaluation of the potential i. pact of any newly developed siting -
.M policies en existing nuclear installations. A particular subject r*
to be considered is the effect of any changes on sites already ..._~z approved. .
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[ ACF3 Concents on a Core Retention System to Mitigate the ConseqJences I of a Core Meltdownr 1/11/71 J The Advisory Comittee on Reactor Safegaards appreciates the meeting
;- with you, members of your staff, and representatives of Battelle ; Me crial Institute and E. I. du Pont de Nercurs and Co pany, on &
5 January 8,1971, to discuss the matter of postulated core meltdown W E' s accidents. As you know the Comittee has had a continuing interest
? % in this ratter.
p Following its meeting with you, the ACPS further reviewed the use-
'y falness and feasibility of a core retentien system to mitigate the 5.u consequences of a core meltdown. 'Ihe Comittee agrees with you that
- f. quality assurance, including assurance of proper functional perfor-
'ri ma:.ce, of present systems having safety functions is of prirrary im- % pottance. The Cox.ittee believes that the probability of meltdown 9, with present systems is very low, and that rmre stringent application
- f. of principles of quality assurance will make the probability still F lower. However, imorovements to systems and system quality cannot f lead to continued s'ignificant increase of safety with6uT limits;
% external phenocena, unforeseen events of very low probability,
- k, comern mode failures, and human error will set a practical limit g,. to system reliability.
b V The Comittee believes that even though a core retention system may not be effective for all causes and modes of core treltdown, it could,
.UT as an independent backup, decrease the probability of an untenable
'7 fission-product release to the environment t*/ at least an order of !sh, magnitude, a result that becomes increasingly difficult to achieve by lp- refinement of systems designed to preserve core integt ty within the l ij ? reactor vessel. lv*
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6. I 1 g ( CORE MELT ACCIDENT (CORE RETENTICN SYSTEM) { B C I' ACFS Coments on a Core Retention System to Mitigate the Consequences
- of a Core Meltdown: 1/11/71 (Cont'd)
L y Tne Ccarittae has found the work by Battelle and du Pont to be very helpful in its considerations. Both groups have separately concluded k that it appears technically feasible to ritigate the consequences of p a core meltdown accident. Both groups have recomended that, if work r ' in this area is to be continued, the logical next step is to choose
, one or possioly two design approaches which appear to have the best
- j. potential of success, to evaluate this design in greater depth, and e to pursue an associated research and development program organized 9 so as to obtain information vital to the success or failure of the h particular design approach. The ACES believes that the results of E. the studies thus far are encouraging.
[. The Comittee recognzies that physical and physicochemical properties 't' of the ac1 ten fuel and structural raterials would be required before P a good con:eptual design could be made, but it is believed that present si kncwledge of these properties may be sufficient to establish basic I[ feasibility. However, there appear to be other major uncertainties g that do affect basic feasibility. For exa ple, sudden admission of a stren of molten fuel into water (even hot water), especially in fg i a manner that can trap water under the fuel, as in steel retaining g
- channels, could lead to a steam explosion of such violence as to make the p w s.a, such problers should be explored qualitatively; for exa.ple, with raterial that can be readily aelted in conventional j furnaces, but using quantities that are large enough to given con-r fidence in the results, h The Comittee believes that a pecgram of conceptual design studies p and analyses for a core retention system, coupled with the kind of n exploratory experiments cited above, should be undertaken. One to f, two million dollars over a period of two to three years might be a p reasonable estimate of the ef fort and time scale to acconplish this p step. The Comittee recomends that a pecgram of this type be 9 under taken with a corpletion goal of 1973-1974 for this phase. The -
%. Cccmittee believes it important that the group undertaking the task have considerable background and resources in pratical engineering and retallurgy, as well as a strong research and development cap-ability.
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XD [LHEE e X> 52T r?mg g - , EtR.;BNCY PLANS Newbold Island 1&2; 7/17/73 The Applicant has prepared a preliminary emergency plan which considers, arrong other things, the feasibility of evacuating the population within the Lcw Population Zone (LPZ) in the unlikely event of a major accidental release of radioactivity frco the plant. The Applicant has also descrited studies of the feasibility of evacuating an area extending as much as three i miles f rce the plant, assuming the projected population that would result
; the full development envisioned by the M.IC, and has concluded that such e evacuation is feasible. Detailed e:rergency plans, to be developed by the State of New Jersey and the Ccuronwealth of Pennsylvania, have not yet been completed.
The Ccanittee concludes that a suitable emergency plan can be developed for the Newbold Island site. The Com.ittee believes also that plans for e appropriate protective measures should extend several miles beyond the
- proposed LPZ radius of one mile. It is essential also that plant per-sonnel be provided with those instruments, indicators, and treasurerrents ,
that will define clearly the nature and course of an accident so that off-site e.tergency plans can be initiated at a level and on a time scale consistent with the severity or potential severity of the accident. O y N h E-1.2 4G.w.m wTmrx.manmaus1%>.LnQLM: w.: 3' IGE"i :iQ?2@gC
Seabrook 162; 12/10n4 s h ne Seabrcok Station Units 1 and 2 will be the first comercial nuclear ' power plant in the State of New Hampshire. Fbr this reason, the Comittee reccr. mends that the Applicant and Regulatory Staff give particular attention to assuring proper coordination with appropriate state and regional agencies in the develo;nent of effective emergency plans for this facility. Because of the proximity of the Seabrook Station to the beaches on the coast and because of the nature of the road network serving the beaches, the Applicant has given early attention to the problem of evacuation. %e Comittee be-lieves, however, that further attention needs to te given to evacuation of residents ard transients in the vicinity even though they reay be outside the LPZ. 5-1.~ 0 m O
Attachment C. 4 NUREG 0603 COMMENTS ON THE NRC SAFETY RESEARCH PROGRAM BUDGET O Manuscript Completed: July 1979 Date Published: July 1979 Advisory Committe on Reactor Safeguards U. S. Nuclear Regulatory Con. mission Washington, D.C. 20555 g
t O l' 1 1.2.11' Application of Probabilistic Methodology The A35 re :ccends emphasis on the application of probabilistic and other methodology to an evaluation of the adequacy of the single failure criterion and to studies of alternate design approaches to systems and groups of systems important to safety in order to provide a better basis for decision makire concerning the optimization of plant design for safety. l. 4 i 4 t t i O i 0 i I 1 i j l s O , l
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NUREG 0496 h 1 1978 REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION SAFETY RESEARCH PROGRAM A Report to the Congress of the United States of America O l l l l Manuscript Completed: December 1978 Date Published: December 1978 . l t Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission Washington, D. C. 20555
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// O investigation. Computer code develerent procra-s, altho gh even- l tually respending t: %C needs, is prinarily in this category. The I
PAS work to collect, correlate, and evaluate perfo.mance data is also being done primarily as a result of PA5 initia:!ve. Pesear:" on flood risk analysis, fire risk assessnent, and the analyses of Class 3-9 acciden'.s for use in environrental reviews is i in direct response to recuests fro . various other groups within mC. I
?e recent increase in professional staf f represents an in:rease in level of activity connensurate with increasing applications of ,
I risk assessnent in the licensing and regulatory activities of the l l W C. These applications are likely to increase. It is important that i j i the PAS continue to recogni::e that risk assessment is not an end in i itself and that, although the PAS will continue to be resconsible for l l initiating and assisting in the develo;r.ent of new projects, methods ., nast be taken over and used by other divisions as soon as feasible. I 11.4 Progress and Pesults of special note are the activities of the PAS in inproving the ; methods first developed in W.SH-1400 for predicting consequences of the release of radioactive materials in reactor accidents. Various i aspe:ts of this part of the Reactor Safety Study have received serious criticisms, and a major ef fert is being made by the PAS to improve the method. The basic vehicle now being developed for l conseqJenre prediction is Called the CRAC Code. It is designed to sample statistically a large population of atnespheric situations and to nodel a large nu .ber of atmospheric phenomena and site character-istics. Results are expected to predict consecuences in some repre-sentative situations. Although progress is being naie in improving ' l the nodel, there are indications that it still has deficiencies that recuire further effort. This is an activity which should be pursued with diligence. The PAS is nearing completion of a study that extends the effect of liquid-borne activity on reactor acciient consecuences beyond that carried out in WASH-1400. Another study extends the WASH-1400 study to light-water reactors of different designs. This new study includes a reanalysis of the ' dominant accident sequences using improved models. Special attention is given to analysis of systems designed to mitigate accident conse-quences and to accident analyses which provide a nore advanced treat. ment to release magnitudes. In addition, attention should be called to the beginning of a progran to assess the risks associated with deep sea bed disposal of wastes. O 11-3
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REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION SAFETY RESEARCH PROGRAM ! A Repon to the Congress of the United States of America O l l i 1 Manuscript Completed: December 1977 Date Publishw!: December 1977 l r l l l i Advisory Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission Washington, D. C. 20555 1
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\g I Further extensions of the applications of the methodolegf to areas not having data bases and facility operation experience comparable to LWRs, should be preceded by the developnent of well-coordinated sets of general criteria for establishing and documenting the suit-ability of: (1) research teams organized for specific research projects; ( 2) basic input information such as, expanded data bases and facility designs and operating experience; (3) quality control and assurance procedures; and (4) program results.
RE5 should provide for greater participation of other NRC offices in establishing the needs, programming, and suitability of results of future PAS research. Also, probabilistic analysis programs in support of regulatory activities should begin to be performed by the user group as work loads and analytical capabilit tes develop. Consideration should be given to expanding representation on PAS Program Review Groups for the larger independent risk assessment programs, such as a second PSS, to a broader segment of society. O
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July 14, 1976 Mr. L. V. Gossick Executive Director tor Operations U. 6. Nuclear Regulatory Co= mission Washington, DC 20555
Dear Mr. Gossick:
In its report of April 8, 1975, to Chairman Anders on the Reactor Safety Stray (RSS), WASH-1400, the ACP.S expressed its belief that many of the f.echniques used in the RSS can and should be used by reactor designers to improve safety and by the NRC Staff as a supplement to safety assessment. During recent reviews, the Co=ittee has learned of ongoing efforts in the NRC Regulatory Research programs to test fault-tree methodology
'for possible application to an evaluation of the single-failure criterion, redundancy requirements, and testing and caintenance g
requirements. The Co=ittee reco= ends that this methodology be applied in inves-tigating the areas listed below: (1) In recent acnths varying approaches to syster design for residual heat removal (RHR) and ECC syste=s have been proposed by the same l PWR reactor vender for different models (e.g., two 1007. capacity systems versus three 507. (or possibly larger) capacity systems). It appears that RSS techniques afford an opportunity of providing I an assessment of the relative and absolute reliability of such i alternative designs, albeit with uncertainties. (2) The ACRS has received a report that current regulatory requirements for ECC systems in Get nan PWR's require an additional redundant i system (which may be representative of a system under maintenance). The ACRS recornnends that an assessment be made of the merits and l costs as well as potential disadvantages of such a requirement l compared to current U.S. requirements. l l Ol
Mr. L. V. Cossick 2 July 14, 1976 (3) There are cany systems which are norr. ally in operation, and which are required in varying degree and for varying ti=e periods to maintain safe shutdown (for example, the component cooling water syste ). Many of these systems currently empicy two trains in order to meet the single failure criterion. The ACRS recoceends that the methodology of RSS be applied to such systees, and coupled with other considerations, an assessment be made of whether the single-failure criterion as currently applied is adequate. This assessment should allow for the frequency and time periods of possible downtice for one of two vital systems for repair, and should include an assess =ent,of the adequacy of measures for repair. (4) The ACRS recer. ends that the reliability of isolation of the high pressure pri=ary syste: fro: connecting low pressure systens be assessed for existing plants and for proposed plants, in the light of the potential consequences of a loss of such isolation. The ACRS believes that by examination of these (and si=ilar) systems by the methodology of RSS, further insight into potential problets or into
* - improvements in safety can be obtained, and that an improved basis for the setting of priorities and the allocation of resources cay be provided.
The Co=.ittee suggests that such efforts be pursued both by NRC and by the appropriate groups in the nuclear industry. Sincerely yours, V Dade W. Moeller Chair =an B l l l
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*+..* September 10, 1979
, APPENDIX XXIV
Title:
Letters to Members of Congress Proposing Rescheduling of The ACRS Annual Report on the NRC Safety Research Program The Honorable Morris K. Udall, Chairman Subcommittee on Energy and the Environment Committee on Interior and Insular Affairs United States House of Representatives Washington, DC 20515
Dear Chairman Udall:
Based on discussion with the Staff of the Senate Subcommittee on Energy and the Environment, it is our understanding that the ACRS annual report for 1979 on the NRC Safety Research Program (Public Law 95-209) will be more useful if it takes into consideration the NRC research proposal
, actually submitted to Congress and can be provided to the Congress no later than mid-February,1980.
It is the plan of the ACRS to modify the schedule for preparation of its 1979 report consistent with the above request from the Staf f of the Sen-ate Subcommittee. If this plan of action is not acceptable, we would appreciate notification as soon as possible. Sincerely, a Max W. Carbon l Chairman cc: J. Hendrie, Chairman, NRC q
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# 'o, NUCLEAR REGULATORY COMMISSION f* , ' ,'# f' ,gc
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (*.
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i i ( The Honorable Gary Hart, Chairman Subcommittee on Nuclear Regulation Conrnittee on Environment and Public Works l- United States Senate - Room 61B 128 C Street, NW Washington, DC 20510
Dear Chairman Hart:
Based on discussion with the Staff of the Senate
<"e ^cas emmee1Succommittee regere o O
' eme the emv1<emmeet, for 1979 on the NRC Safety Research Program (Public Law 95-209) will be more useful if it takes into consideration the NRC research proposal actually sutxnitted to Congress and can be provided to the Congress no later than mid-February,1980.
It is the plan of the ACRS to modify the schedule for preparation of its 1979 report consistent with the above request f rom the Staff of the Sen-If this plan of action is not acceptable, we would ate Subcommittee. appreciate notification as soon as possible. Sincerely, Max W. Carbon Chairman cc: J. Hendrie, Chairman, NRC '1 0 - 3 i
- )
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APPENDIX XXV
Title:
Additional Documents Provided for ACRS Use l t'N) (_/ APPENDIX XXV Additional Documents Provided for l ACRS Use l l
- 1. Letter, Representative M. K. Udall, Chairman, House of Representatives Committee on Interior and Insular Affairs to J. M. Hendrie, Chairman, l NRC, dated September 4, 1979. I
- 2. SECY 493, NRC Action Paper, Report of the Siting Policy Task Force, dated August 16, 1979.
- 3. Memorandum, C. P. Siess, ACRS to Subcommittee Chairmen / Chapter Authors for 1979 Research Report, Research Priorities, dated August 29, 1979.
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