ML20148R014

Apps to Minutes of 243rd ACRS Meeting on 800710-12 in Washington,Dc Re Sequoyah Units 1 & 2 & FY82 Budget.Agenda Encl

ML20148R014
Person / Time
Site:	Sequoyah
Issue date:	11/25/1980
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-1769, NUDOCS 8101280487
Download: ML20148R014 (300)

Text

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A?CA 5- /& 9 APPENULXES TO MINUTES OF THE 243RD ACRS MEETING JULY 10-12, 1980

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1. hm.Te mm TABLE OF CONTENTS J'..f [P"i ]r.bN l...

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MINUTES OF THE .

[.'7 3 2 0 Ub I. Chai rma n's Report (0 pen to Publ i c) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 A. Reviewers .......................................................... 1 B. N o ti c e o f Aw a r d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 C. Agenda for Meeti ng wi th Comi ssioners . . . . . . . . . . . . . . . . . . ......... 1 II. Meeting on Sequoyah Nuclear Plant Units 1 and 2 (Full Power License) (0 pen to Public) ............................;... ............. 2 A. S ub comi tte e Re po rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 B. S ta tus o f NRC S taf f Revi ew . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

1. Schedul e of Seq uoyah Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

2. O p e n S a f e ty I tems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

3. Pres s uri zer Vent Repai rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

4. TVA Res pons e to NRC S taff Repo rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 C. TVA Presentations ................................................... 5

1. Status Report on Under-Clad Cracks in Reactor Pressure Vessel (RPV) Nozzles .................................. 5

2. Re l i ab i l i ty S tu di e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

3. Hydroge n Con trol S tudi es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

4. NRC S ta f f Re s po ns e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

0. Caucus ............................................................. 8 III. Meeting with Members of the NRC Staff Regarding the NRC's Reactor Safety Research Budget for Fiscal Year 1982 (0 pen to Public) .. .. .. .. .. . 9 A. S u bcomi *.te e R e po rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 B. NRC Staff Overview .................................................. 9 C. LOFT ............................................................... 9 IV. Meeting with Members of the NRC Staff on Recent Operating Expe ri ence (0pe n to Publ i c) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 i

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l TABLE OF CONTENTS (Cont.)

243RD ACRS MEETING 1

85 10 Three Mile Island 2: Venting of Containment to Release kr .....

A.

B. Brunswick 1/ Hatch 1: Potential Loss of Scram Discharge 10

,' Instrument Volume (SDIV) .........................................

.l I C. Browns Ferry 3: Failure of Control Rods to Scram Fully .......... 11 D. St. Lucie: Response of Primary Cooling System to Cool Down on Natural Ci rcul ati on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 12 V. Meeting with NRC Commissione rs (0 pen to Publi c) . . . . . . . . . . . . . . . . . . . . . .

A. Integration of Current Rul ema ki ngs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 B. Role of the Office of the Analysis and Evaluation of Operational Data in Accident Investigation . . . . . . . . . . . . . . . . . . . . . . . 13 C. Re s i dent NRC I ns pectors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 D. Policy Regarding the Standardization of Nuclear Plants ........... 14 14 VI. Executive Sessions (0 pen to Public) ..................................

14 A. F u tu re Sche d ul e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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

2. Schedule for ACRS Subcommittee Meetings and Tours .. .. .. .. .. .. 15 B. S ubconmi ttee Acti vi ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15

1. Wa s te Ma na g eme nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15

2. F i r e P ro te cti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 15

3. Probability and Risk Assessment ..............................

4. Rea cto r Operati ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 C. Establishment of Supplemental ACRS Fund for Support of Meeti ngs wi th Forei gn Di gni tari es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 D. S u bcommi ttee Re po rts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

1. Quanti tative Risk Accep tance Cri teria . . . . . . . . . . . . . . . . . . . . . . . . 16 E. Items for Next Meeting wi th the Commissioners . . . . . . . . . . . . . . . . . . . . 17 17 F. Deg r aded Core Acci dents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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. TABLE OF CONTENTS (Cont.)

243RD ACRS Meeting 17 G. ACRS Reports , Letters , and Memoranda . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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1. ACRS Report to the Commission on the Fiscal Year 1982 .. 17 j-E Reactor Safety Research Program Budget ......................

2. Sequoyah Nucl ear Plant Uni ts 1 and 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 l Sequoyah Nuclear Plant Units 1 and 2: Ice Condenser 3.

Capabili ty and Hydrogen Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18

4. Comments on the Reactor Coolant Pump Trip Requirement and High Pressure Injection Termination Cri teria . . . . . . . . . . . . .. .. - 18 C 5. Utili ty of I n-Co re The rmocouples . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 (

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6. Analysis of Barge Collision with Sequoyah Intake Structure ..... 18 v 18  !

1 VII. Executive Sessions (Closed to Public) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .

18 A. New Members ........................................................

B. Acti v i ti es of Membe rs . . . . . . . . . . /. . . . . . . . . . . . . . . . . . . . . . . .

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.s TABLE OF CONTENTS l APPENDIXES TO 1 243RD ACRS MEETING  !

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Appe ndi x I - A t te ndee s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Appe ndi x II - Future S chedul e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Appendix III - Schedule of ACRS Subconnittee Meetings and Tours .. .. .. .. .. .

Appendix IV - Award to 0. Okrent of the First ANS "Tonny" Thompson Award Background for Review ........................ A'35 Appendix V - Sequoyab 1 and 2:

Ope n Sa f e ty I tems . . . . . . . . . . . . . . . . . . . . . . . . . . . A- 50  :

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Appendix VI - Sequoyah 1 and 2:

Pressurizer Relief Piping Repair . .. .. .. .. .. A-54 ,

Appendix VII - Sequoyah 1 and 2: '

Under-Clad Cracking of RPV Nozzles .. .. .. .. A-59 Appendix VIII - Sequoyah 1 and 2:

Reliability Studies ......................... A-6,3  ;

Appendix IX - Sequoyah I and 2:

Hydrogen Control Studies . . . . . . . . . . . . . . . . . . . . . A- 72 Appendix X - Sequoyah 1 and 2:

NRC Staff Planned Review of Appendix XI - Sequoyah 1 & 2: Proposed Hydrogen Mi tigation Sys tems . . . . . . . .

Appendix XII - RES Program Endorsements - FY 1982 ........................... A-87 l

Appendix XIII - Proposed F C Reactor Safety Research Budget for FY 1982 . >

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RES Appendix XIV - Proposed NRC Reactor Research Budget for FY 1982 l Appendix XV - Proposed MRC Reactor Research Budget for R'-1982j j

' Appendix XVI - TMI-2:

Background Material for Venting of Containment .. .. .. .. A-164 85 Venting Systems for Release of Kr From the Appendix XVII - TMI-2:

Contai nment Buil di ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1 Degradation of BWR Scram Discharge Appendix XVIII - Brunswick 1/ Hatch 1:I ns trume nt Vol ume . . . . . . . . . . . . . . . . . l Bulletins Issued Regarding Appendix XIX - Brunswick 1/ Hatch 1: Potential Loss of Scram Dischange Volume Background Material on Failure of Appendix XX - Browns Ferry 3: Control Rods to Insert Fully . . . . . . . . . . . . . . . . . . . . .

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243RD ACRS MEETING Failure of all Rods to Insert Appendix XXI - Browns Ferry 3: Fully Foll owi ng a S cram . . . . . . . . . . . . . . . . . . . .

Response of Reactor Coolant System Appendix XXII - St. Lucie 2:

to Cooldown on Natural Circulation, Event of l J u ne 1 1, 19 80 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Appendix XXIII - St. Lucie: Pressure Instability During Cooldown on11, 1980 ............... A  ;

Natural Circulation, Event of June i Safety Implications of Upper Head Voiding ... .. .. A-240 Appendix XXIV - St. Lucie:

Appendix XXV - Summary of Connissioner Hendrie's Connents onTre atm:

Appendix XXV! - NUREG-0699, Comnents on the NRC Safety R Appendix XXV.'I - ACRS Report on Sequoyah Nuclear Plant Units 1 and 2 Ice Condenser Capability and Appendix XXVIII - Sequoyah 1 and 2: Hydrogen Control ...................................

Appendix XXIX - ACRS Comnents on the Reactor Coolant Pump T Appendix XXX - Use of In-Core Thermocouples for Detection of

' Appendix XXXI - Sequoyah Nuclear Plant Design of Intake Structure ..... ...

Appendix XXXII - Additional Documents Provided for ACRS' Use . .. .. .. ..

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. 15. 1980 / Notices Federal Register / Vol 45, No.124 / Wedn;sdry. Juna

.cgo.: will be permitted only during those Cilinsky regarding primary coolant portions of the meeting when a

• domestic producers to imports of pump tnp and termination of high. transcript is being kept and questfoes subsidized tomato concentratespressure and coolant injection during may be asked only by members of Me canned tomatoes from se EC. (Report at transients. and the use ofin-core Committee, its consultsnts, and Staff.

A-09). thermocouples in nuclear reactors. Persons desiring to make oral statements should notify the ACRS Condusions of Law Mday, July 11. N Executive Director as far in advance as A.The appropriate domestic Industry &JO o.a-12x p.ma Squoyah practicable so that appropnate egainst wiuch the impact of subsidized Nuc/cor Power Plant. Unit 1/Open)-- arrangement can be mac's to aUow the imports from the EC should be messured The Committee will hear and discuss necessary time during the meeting foe consists of those domestic facilities comments of its Subcommittee and such statements.Use of still monon devoted to the production of tomato consultants who may be present picture and television cameras dunes concentrates and canned tomatoes. regarding proposed operation of the this meeting may be limited to selected i B. The like product in question here is Sequoyah Nuclear Power Plant. Unit 1 at portiens of the meeting as determined tomato concentrates and canned full power.The Committee will alsa hear by the Chairman. Information regarding tomatoes as desenbed !n the reports from and hold discussions withthe tune to be set aside for this purpose determination without regard to the representanves of the NRC StaH and the may be obtained by a telephone eau to specific tomato type, C The domest2c industry is not applicant regarding this matter.thePortions ACRS Executive of this Diactor (R. will session Fraley) prior to the meeting. In view of F. be closed matenally injured or thnstened with as necessary to discuss Proprietary the possibdity that the schedule for materialinjury by reason of subsidized Information applicable to this matter, ACRS meetings may be adjusted by de i isiports of tomato concentrates and EMP.m-Omp.m Meetmg with NRC Chairman u necessary to facilitate We canned tomatou from the EC. Chairman and Other!GC conduct of the meeting, persons lan d Juneta m Comnussionen (Openkne Committee planning to attend should checi mth de By order of the Commientoa.

will meet mth the NRC Chairman John ACRS Executive Director if such F. Ahearne and other Commissionersrescheduling would result in maict Kenoeth R. Mason. who may have an interest to discuss laconvenience.

Secretary, safety related aspects of the NRC I have detennined in accordance mth in o=. aman w s.a.am == -4 regulatory program. - ProposedACRS Subsection 10(d) Pub. L 92-463 that tt is sumo corm 'am-ms - 2x p.m-&M p.n necessary to dose portions of this Reports (OperrNThe Committee will meeting as noted above to protect discuss proposed ACRS reports to ths Proprietary Informatico (5 U.S.C

' HUCLEAR REGUL.ATORY NRC regarding the proposed safety 552b(c)(4)) and to protect information COMMISSION research budget for FY 19a:: proposed NRC rule on fire protectfon criterf a for the nieue of which would consutute a Adytsory Committee on Reactor dearly unwarranted invasion of nuclear power plants: and cascade- personal privacy (5 U.S.C. 552b(c)(61).

Safeps; Muting failures in nuclear power plants. Further information regarding topics In accordance with de purposes of to be discuned. whethJr the meeting Saturdayddy 12.1980 See:fons 29 and 18:b. of de Atomic .

has been cancelled or nscheduled, the Energy Act (42 U.S.C 2039. ::32 b.). the &# o.m-Sx p.n Executive Session Chateman's ruling on requests for the Advisory Committee on Reactor (OpenN The Comm2ttee will complate opportunity to present oral statements Safeguards wd! hold a meeting on July discussion of and prepare its reports to and the time allotted therefor can be IM: 1980, in Room 1048. In? H Street, the NRC regarding matters considered . obtained by a prepaid telephone cau to I

NW. Washmgton. DC. during this meeting. In addition. the The agenda for the subject meeting the ACRS Executive Director. Mr.

Committee will hear and discuss reports Raymond F. Freley, or in his absence t willbe as follows: from its Subcommittees regarding ACRS Dupty Director. Mr. Marvin C )

activities related to deve!opment and Caske. (telephone 202/634-3:85).

,i Tbdan W % N use of quantitative nak cntaria. Class.9 tx a.n-9:13 a.at Opening Session acc: dents in the regulatory process, and between 4:15 A.M. and 5.00 PM. ' ECTT.

, /OpenNThe Committee wdl hear and other safety iseues. Dd; I""' 18 'IS'O' discuss the report of the ACRS The schedule for future Committee John C. Wa.  ;

Chairman regarding matters to be AcWsory Committee Management cvficer.

• activities will also be discussed.

considered dunna the meeting with de The Committew will discuss de in om sme N 64+ e =e -6 NRC Commissioners and ::nscellaneous qualifications of candidates proposedna_m o coce ne w n - )

mattters relating to ACRS activities. . (or appointment to the Comnuttee.

R13 a.1-!::X p.1 and 1:x p.n-4:x Pornons of this session wtU be dosed Wsay Committee on the Wedcal l

p.1; ACRS Report on NRC Safety as necessary to discuss Propnetary Uses of tsoWs; Pubuc Meeting l Research Budget (OpenFThe Information and also closed as i i Committee wtil discuss its proposed necessary to protect information the The Nuden .tegulatory Commission's  !

report to the NRC regarding the release of which would represent an l (NRC) Advisc / Committee on the proposed NRC safety research budgetunwarranted invasion of personal Medkal Uses r.f Isotopes wul hold a for Fiscal Year 1982. public meeting at 9 a.m. on Monday, Portions of this session wd! be privacy.Procsduns dosed for the conduct of andAugust 18.1980 in Versailles IV Room.

)l' as necessary to discuss Propnetary - participation in ACRS meetings were Holiday Mn. 81:0 Wisconsin Avenue.

Information. ProposedACRS published in the Federal Regisrae on Bethe Maryland.

4:x pta-e# p.n October L 1979 (44 FR 56408). In Th odowing agends is scheduled:

Reports to NRC Chairman / accordance with these procedures. oral t Trcining and Expenence Critena Camm;ssioners (open>The Committee or wntten statements may be presented for i crician Users. Including will discuss proposed reports to the by members of the public recordings NRC Chairman and Commissioner V.

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UNITED STATES

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, NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

. , WASHINGTON, o. C,20656 0, ; g 8 July 2, 1980 SCHEDULE AND OUTLINE FOR DISCUSSION 243RD ACRS MEETING JULY 10-12, 1980 WASHINGTON, DC Thursday, July 10, 1980, Room 1046, 1717 H Street, NW, Washington, OC OceningSession(0 pen)

1) 8:30 A.M. - 9:15 A.M. 1.1) First "Tvinmy" Thompson Award presented to Dr. David Okrent 1.2) Proposed Items for meeting with NRC Commissioners 1.2-1) Retponsibility of state /

local governments re.

Emergency Planning l

1.2-2) Standardization of Nu-clear Plants (NUREG-0642) 1.2-3) Role / responsibility of NRC Resident Inspectors l

ACRS Pecort to NRC on FY 1982 Safety

2) 9:15 A.M. - 12:30 P.M. Research Budget (0 pen) 2.1) Discuss proposed ACRS Report to t NRC (CPS /TGM/DZ)

, LUNCH 12:30 P.M. - 1:30 P.M.

ACRS Reports to NRC on FY 1982 Safety

3) 1:30 P.M. - 4:30 P.M. Researen Budoe_t,(0 t pen) 3.1) Discuss proposed ACRS report to NRC (CPS /TGM/DZ)

(Portions of sessions 2) and 3) above will be closed as necessary to protect Proprietary Inforniation.)

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l' I-July 1,1980

. Schedule Recent Ooerating Experience (0 pen)

4) 4:30 P.M. - 6:30 P.M. 4.1) 4:30-5:00: Hatch_/B runswick_

Nuclear Plants - potential loss of SDIV discharge volume 4.2) 5:00-5:30: Browns Ferry Nu-clear Plant Unit 3_ - 6/7/80 f ailure of control rods to )

fully scram l 4.3) 5:30-6:00: St. Lucie Unit 1_ - l response of primary coolant system to cooldown on natural .

circulation 6:00-6:30: TMI-2_ - Venting of 4.4) containment to release Kr-85 j

Proposed ACRS reolies to inquiries i

5) 6:30 P.M. - 7:30 P.M. from NRC Chairman and Commissioner Gilinsky re: (0 pen) 5.1) Delayed primary coolant pump i trip during SBLOCAs and cri-teria for HPI termination (memo / letter from Chrmn John F. Ahearne dated 2/22/80 and 4/1/80) (KSP/ALB)  ;

5.2) Use of in-core thermocouples j (letter from Comm. Victor l 3

Gilinsky dated 6/4/80) (00/PB) f l l

Friday, Julv 11, 1980, Room 1046, 1717 H Street, NW, Washincton, DC l Seouoyah Nuclear Power Plant Unit ~~1

6) 8:30 A.M. - 12:30 P.M. 10 pen) 6.1) 8:30-9:00: Recort of ACRS Suocormittee (JCM/RS) 6.2) 9:00-12:30: Discussion with NRC Staff and apolicant (Portions of this session will be closed as required to discuss Pro-prietary Information relating to this matter.)

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July 1,1980 Schedule l'

i LUNCH 12:30 P.M. - 1:30 P.M.-

Meeting with Chairman J. F. Ahearne

7) 1:30 P.M. - 2:30 P.M. and other NRC ConTnissioners_ (0 pen)

Procosed ACRS Reoorts to NRC (0 pen)

8) 2:30 P.M. - 7:00 P.M.

Discuss proposed 8.1) 2:30-4:30:

ACRS Report on FY 1982 Safety Research Budget (CPS /et al.)

8.2) 4:30-5:30: Discuss proposed ACRS comments re. proposed NRC nale on fire protection (MB/PST)

Discuss proposed 8.3) 5:30-6:30:

ACRS letter on " cascade' fail-ures in nuclear power plants (JCE/GRQ)

Discuss proposed 8.4) 6:30-7:00:

response to request from G. G. Sherwood, GE for clari-fication of ACRS report dated

' 4/16/80 on ATWS (WX/PAB)

Saturday, July 12, 1980, Room 1046, 1717 H Street, NW, .lashington, DC ConcludingSession(0 pen)

9) 8:30 A.M. - 4:00 P.M.

9.1) 8:30-10:30: Proposed ACRS Re-port on FY 1982 Safety Research Budget (CPS, et al.)

9.2) 10:30-11:30: Complete proposed ACRS reports to NRC on the Sequoyah Nuclear Power Plant Unit 1 9.3) 11:30-12:15: Discuss proposed candidates for ACRS membership (DWM/MCG) and activities of ACRSmembers(WK/RFF)

(Portions of this session will be closed as necessary to protect in-fonnation the release of which would represent an unwarranted invasion of personal privacy.) .

t July 1,1980 i

l Schedule j

12:15-1:15 P.M. - LUNCH I

9.4) 1:15-2:00: Complete proposed ACRS reports / comments on:

9.4-1) Use of in-core thermo-coupl es 9.4-2) Delayed pump trip and termination of HPI 9.4-3) NRC Rule on fire protec-tion 9.4-4) ' Cascade' failures in nuclear plants 9.5) 2:00-3:15_: Reports of ACRS Sub-committees on:

9.5-1) 2:00-2:45: Class 9 Ac-cidents (WK/GRQ)  ;

9.5-2) 2:45-3:15_: Quantita- /

tive Risk Criteria l (00/GRQ) 9.6)3:154:00: Future Schedule Items l 9.6-1) Anticipated full Committee Activity (MP/RFF) 9.6-2) Anticipated Subcommittee Activity (MWL/et al) 1 l

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Issue Date: l NOV 2 51980 l (FOIA EXEMPTION (b)5) l l

MINUTES OF THE ' 4 "

243RD ACRS TETING JULY 10-12, 1980

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j WASHINGTON, DC Ih The 243rd meeting of the Advisory Committee on Reactor Safeguards, held a  ;

1717 H St. N.W. , Washington, DC was convened at 8:30 a.m., Thursday, JI 1980.

Mr. H. W. Lewis was not For a list of attendees, see Appendix I. ,

[ Note: '

present on Saturday, July 12,1980.]

n The Chairman noted the existence of the published agenda for this m identified the items to be discussed. Advisory Connittee Act (FACA) and the ,

held in conformance with the Federal Government tively.

in the Sunshine Act (GISA), Public Laws 92-46 from the General Electric Company, and that time was being allotted for this No written statements from members of the public were received. l statement. '

He also noted that a transcript o'f some of the public portions of the meeting was being taken, and would be available in therepresentative The NRC's PublicofDocument General Room at

[ Note:

1717 H St. N.W., Washington, DC.

Electric did not appear to make the oral statement.

Copies of the transcript taken at this meeting are also available for I

[ Note:

purchase f rom Alderson Reporting Company, Inc. , 400 Virginia Ave. S.W. ,

Washington, OC 20024.

I. Chairman's Reoort (0 pen to Public) 1

[ Note:

Raymond F. Fraley was the Designated Federal Employee for this portion of the meeting.]

A. Reviewers The Chairman named J. C. Mark and M. W. Carbon as reviewers i

243rd ACRS Meeting.

8. Notice of Award The Chainnan noted that D. Okrent was the recipient of the first

" Tommy" Thompson Award from the American Nuclear Society (see Appendix IV).

C. Agenda for Meetino with Conrnissioners The Committee agreed that it wished to discuss the following them on matters with NRC Commissioners during the meeting with Friday afternoon, July 11, 1980:

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JisLY l'M2.1980 MINUTES OF THE 243RD ACRS MEETING intentions regarding the e Clarification of the Commi*sion's croposed Emergency Planning Rule, o

The NRC policy regarding the standardization of nuclear facilities, f inspectors, o The role and responsibilities of NRC resident e How the NRC plans to approach the current r developed, and e How the Commission separates and defines the four are rulemaking which appear to overlap:

accidents, reactor siting policy, and emergency planning.

(Full Power License)

II. seeting on Seoucyah Nuclear Plant Units 1 and _2 (0 pen to Puolic)

Richard P. Savio was the Designated Federal Employee for this

[ Note:

portion of the meeting.]

A. Subcommittee Reoort J. C. Mark, Subcommittee Chairman, noted that the Subcommittee m on July 8,1980, discussed with the The NRC Sub- Sta a report on the Subcommittee activity, see Appendix V). l

(

committee requested the Licensee and the NRC He said continuing significance or to raise difficult questions.

that presentations will not be made on the following items, but that members of the NRC Staff and representatives of the Applicant will be present to answer questions:

e flood protection e operation of Unit 1 before completion of Unit 2, a water sources, .

l the low-power test program (this program is to I e status of July 12, and the information that is currently begin about available has been discussed by the Committee previously).,

D. Okrent noted that the Applicant has responded to the ACRS recommendations that a reliability assessment be made of the pla and that consideration be given regarding a vented filtered contai ment.

The Applicant is prepared to discuss these matters at this meeting.

2

JULY 10-12, 1980 MINUTES OF THE 243RD ACRS MEETING of a request from Commissioner J. C. Mark noted the receipt Gilinsky that he be informed of the Committee's opinion regarding the capability of the ice condenser pressure suppression system used at Sequoyah, and also regarding the matter of hydrogen co in thethe small ice condenser Committee that thesecontainment.

two matters be addressed in corresponden to Commissioner Gilinsky, separate from any report made on the Sequoyah Nuclear Plant.

B. Status of NRC Staff Rev y

1. Schedule _ of Sequoyah Events s

C.

Stahle, NRC Staff, discussed the schedule of significant events to take place in the immediate future at Sequoyah Nuclear Plant, noting that zero power testing began July 5, and is sec .uled to be completed on July 11 or 12, the augmented low-power test will begin between July 12 a during the first week in August.

2. Ooen Safety Items _

C. Stahle discussed both the non-TMI safe 1 (see Appendix VI). He said that ofOf the 13 non-TMI issues the 40 TMI-2 accident i identified, 8 are currently resolved.

! related items identified,15 are resolved,1 is nonapplicable, be 13 are required by specific1 dates in the future,1 will is an Inspection and Enforce-resolved by NRC rulemaking, He noted, ment (I&E) function, and 9 are currently incomple tion of documentation, that agreement has been reached betweenthe and that the NRC Staff and the Applicant,these incompleted safety will be resolved prior to the issuance cf a full power license.

R. Tedesco, NRC Staff, said that the Staff plans to completeon supplement the Safety Evaluation Reporta request for Full Power License to the July, and to present He reiterated that incompl ete Commi ssion by early Augu st.

i safety items will be resolved prior to full-power operation.

j In answer to a quastion regarding the criteria used in determin-ing control room habitability during degraded conditions, H. E. P. Krug, NRC Staff, said that TVA The is e i

NUREG-0694, the latter providing the post-TMI-2 criteria.

review will be in agreement with the thrust of NUREG-0694, but 3

l 1

l

o l

JULY 10-12, 1980 MINUTES OF THE 243RD ACRS MEETING cannot address the complicated questions of the degraded core issue until the Commission completes its rulemaking on that subject.

R. H. Wessman, NRC Staff, noted that an I&E inspector will witness power ascension tests at Sequoyah.

3. Pressurizer Vent Repairs l

(For background information on this issue, see Appendix V.)

R. M. Gamble, NRC Staff, said that during the hot functional testing of Sequoyah Unit 1, a misplaced pipe hanger on the 6 inch vent line from the pressurizer hung up causing distortion of the pipe. The Applicant used the weld-bead method for straightening the pipe, which method consisted of grinding two grooves on the convex side of the distortion, filling the grooves with weld metal, regrinding the grooves, and rewelding, He discussed setting up stresses tnat straightened the pipe.

the necessary conditions that cause stress corrosion cracking,  !

l the NRC Staff's evaluation of the repair weld, and the conclu-  !

sions and licensing actions (see Appendix VII). The NRC Staff considers the matter resolved, but plans to require reinspec-tion of the weld during the first three refueling outages of Sequoyah.

J. Kalapatz, NRC Staff, disagreed with other members of the NRC Staff regarding the basis for the resolution of this matter.

His position was that while the weld repair may be adequate, the NRC Staff did not have enough information on which to make such a judgment. He recommended that the weld be mocked-up with identical or similar material, that the mock-up be welded, cut, and examined extensively metallographically. He noted that a mock-up which had been made by the Applicant using diff erent materi al s , showed i ndication of stress corrosion cracking capability, and would not meet current standards for this type of repair.

P. Van Doorn, NRC Staff, said that he witnessed all of the in-situ metallurgy as it was performed. He said that the mock-up made by the Applicant does not represent the field weld. Even so, very minimal sensitization was demonstrated.

He said that full penetration of the remaining rortion of the pipe wall was not achieved, and that no internal nxidation has been identified.

4

JULY 10-12, 1980 MINUTES.0F THE 243RD ACRS MEETING R. M. Gamble noted that this matter came to the attention of the NRC Staff when TVA requested an exenption from the require-ment of hydrostatically testing this weld. He said that such types of repair normally de not require Appif cant reporting to the Commission.

J. Muscara, NRC Staff, said that the tests that were performed on the subject welds were not relative to the question that is being debated. It has been shown through research results that little correlation is obtained between the level of sensitiza-tion that is measured on the outside of a pipe versus that on the inside of the pipe. During the laying of such a weld, the The pipe is less sensitized on the outside than the inside.

w only time there is good correlation between the sensitization on the outside and the inside is whenthanthere is needed a level of sensitization on both surf aces greater that to cause stress corrosion cracking. At levels that cause stress corrosion cracking, it is possible to have no sensitization on the outside of the pipe.

4. TVA Resconse to NRC Staff Report L. Mills, TVA, said that he has no comments on the NRC Staff's Status Report with respect to the open items. He said that he He is ready to answer any questions on the weld-repair item.

noted that with respect to the NRC Staff's review of the Sequoyah plant, the Applicant has responded more fully than any other applicant in history.

C. TVA Presentations '

1. Status Reoort on Under-C7;d Cracks in Reactor Pressure Vessel (RPV) Nozzles P. Timmons, Westinghouse, discussed the matter of under-clad cracking in the nozzles of PRVs manufactured by the Rotterdam Dry Dock Co. (see Appendix VIII). The cracks that have been identified so f ar, all less than critical size, are believed to have been caused by the method used in applying the cladding.

l Prairie Island has pressure vessels made by the Fromatome Co.

of France that use the same method of applying cladding to the nozzles as did Rotterdam. Prairie Island has committed to It examine its vessels later this year for under-clad cracks.

is believed that the cracks will not grow to critical size during the lifetime of the plant.

5

! . JULY 10-12, 1980 MINUTES OF THE 243RD ACRS MEETING

• D l

W. M. Mathis said that P. G. Shewmon has infonned him that he l has looked into the problem in detail, and sees no safety  !

problem. l

2. Reliability Studies i l

R. Christy. TVA, discussed the reliability studies made for the l Sequoyah Nuclear Plant as a result of ACRS requests.

He discussed the Systems Interaction Methodology Application l Program, the Reactor Safety Methodology Applications Program, the Auxiliary Feedwater System Reliability Evaluation, and the  ;

Plant Full-Scale Safety an< Availability Analysis (see Appendix f g

IX). He identified one problem uncovered by the studies: '

failure to remove drain plugs on the refueling floor following refueling could cause the core spray recirculation system to f ail . Administrative steps are being taken to prevent leaving plugs in place following refueling.

In answer to a question regarding the operation of the ice condenser system, J. Ballentine, TVA, noted that the ice con-dersers have been loaded for 1 1/2 years. There has been fre-quent inspection and maintenance work performed on this system.

In this time, neither frosting nor ice on the doors have been noted. An extensive inspection program is being carried out.

Ice is being lost at approximately the anticipated rate. He ,

noted, however, that since the ice has been loaded, the contain-ment area has been at operating temperature only once, and at other times was cooler. He noted his confidence that the ice maintenance program is satisfactory.

In answer to a question of whether a small LOCA could cause l partial melting of the ice and blocking of the cooling channels through it, R. Bruce, Westinghouse, said that this event has not been analyzed. He said however, that he could not envisage that a small break would melt channels through the ice.

W. Lao, TVA, said that Wettinghouse studied the matter of burn-through years ago, and concluded that there is no problem.

Battelle Northwest Laboratories has confinned this position via a computer study. Tests made at Westinghouse's Waltz Mill Laboratory show that there is adequate distribution of steam l

flow to prevent channeling.

J. Kudrick, NRC Staff, said that the Staff has considered this matter, and concluded that channeling is not a problem.

R. Christy said that the current study has limits, but that additional studies will be made to identify all the areas that need to be considered for plant reliability.

6

JULY 10-12, 1980 MINUTES OF THE 243RD ACRS MEETING G. Oilworth, TVA, said that the Applicant believes that it is  ;

doing more risk assessment and reliability studies than any He noted l other Applicant, and plans to expand this effort. l that R. Christy, in his presentation, was responding only to l the study contracted through the Kaman Corp. l R. Bernero, NRC Staff, said that TVA has committed to cooperate with the IREP studies of the NRC Staff, and to go further.

3. Hydrogen Control Studies G. Oilwccth discussed the hydrogen control studies made by TVA e over the past nine months (see Appendix X).

He noted that l I

after the TMI-2 accident, TVA reviewed the hydrogen issue for all of its plants, starting with Sequoyah, comparing them with NUREG-0598 requirements. The Applicant believes that Sequoyah is safe to operate at full power. He noted that reduction of overall ri sk can be achieved by providing for greater than design basis hydrogen generation, thus providing the He discussed even greater capability margin than is already available.

of the Sequoyah containment, the advantages and disadvantages of vented containment, both filtered and not filtered, as assessed by preliminary studies, the effect of coupled contain-ment, the effect of additional containment, the .effect of nitrogen inerting, consideration of a Halon combustion suppres-sion system, and a general discussion of a proposed distributed ignition system (see Appendix X).

In answer to a question, G. Dilworth said that all the effects of the use of Halon, or the costs, are not known; however, it has been demonstrated that the use of Halon can prevent hydro-gen combustion. He added that the controlled ignition system will not increase risk, since the TMI-2 accident showed Hethat an added, ignition system within containment already exists.

also, that inerting is not practical for an ice condenser containment, because of frequent entry for the large arcunt of maintenance needed for the ice system.

In response to requests made by D. Okrent, representatives from TVA agreed to provide the Comittee with additional documenta-tion of the following matters:

i e

TVA's filtered-vented contairunent studies, including the I dose calculations used to reach the estimated dose of 900 l rem as a result of venting from a postulated accident 1 with a degraded core.

e TVA's plant reliability study on Sequoyah 1.

l 7 l 1

l l

• W

P JULY 10-12, 1980 MINUTES OF THE 243R0 ACRS MEETING In answer to a question, G. Dilworth said that although the design of the system is preliminary, TVA expects to complete installation of the controlled hydrogen ignition system within 2 to 3 months, before continuous full-power operation of the l plant. This system is not safety grade.

W. Lao said that slow In the discussion of vented containment venting, in that if venting provides an advantage over fast venting occurs early in the accident, oxygen in the containment is used up, increasing the safe hydrogen level.

Members have suggested that TVA should consider the folicwing items in their containment studies:

e The use of thick charcoal filters, o

The use of a broader range of accident scenarios, o The use of a combustion turbine to strip oxygen from the containment atmosphere.

4. NRC Staff Resoonse W. Butler, NRC Staff, said that the Staff believes that Sequoyah can be authorized for full-power operation without additional requirements. He recognized that there is a poten-tial to improve safety margins, and that theHe Applicant is being briefly discussed encouraged to proceed along these lines.

the NRC plans for review of the Applicant's reliability and He noted that hydrogen mitigation studies (see Appendix XI).

user requests have been issued for research programs to be

~

carried out at National Laboratories.

f J. C. Ebersole recommended that the NRC Staff review a postula-ted accident in which a fully loaded Tennessee River barge tow collides at full speed with the new intake structure for Sequoyah Unit 2. This study would not be applicable to the l

licensing action for Unit 1, but should be completed prior to I operation of Unit 2.

D. Caucus l It was the consensus of the Committee that a favorable report on the TVA application for a Full Power Operating License for the Sequoyah Nuclear Plant Units 1 and 2 could be written at this meeting.

8

~

l

l JULY 10-12, 1980 MINUTES OF THE 243RD ACRS MEETING l

III.

Meeting with Members of _ the NRC Staff Regarding the NRC's Reactor _

Safety Reseach Budget for Fiscal Year 1982 (0 pen to Public) l Thomas G. McCreless was the Designated Federal Employee for

[ Note:

this portion of the meeting.]

A. Subcomittee Report C. P. Siess, Acting Chairman of the Reactor Safety Research Subcom-mittee for Preparation of the ACRS Report on the NRC reactor safety i research budget for fiscal year 1982, revicwed the status of budget ,

deliberations. He noted that each NRC User was asked to coment on that portion of the Reactor Safety Research (RES) budget applicable to his needs (see Appendix XII). He specifically noted a lack of F

interest by the NRC Staff and the Acting Executive Director for Operations regarding the research budget for fast and gas-cooled reactors, and further noted that these items are not likely to remain in the budget unless directed by Congress.

T. Murley, NRC Staff, informed the Comittee that In Congress has that budget, passed the FY 1980 Supplemental Research Budget.

the NRC had requested $26 million for research, of which $3 million was earmarked for waste management, and $23 million for TMI-related research. Congress appropriated $10 million of the $23 million requested, none of it for waste management. Upgrading of facili-ties to be funded by the increased appropriations may be delayed.

(For the fonn of the budget considered bx the Comittee, see Appen-dixXIII.)

l B. NRC Staff Overview \

l R. Budnitz discussed the differences between the RES Staff proposed l

budget, and the markup by the staff of the Acting Executive Direc- I tor for Operations, and noted the items and amounts for which He specifically reclamation was requested (see Appendix XIV).

defended that portion of the budget relatingHe to the saiddevelopment of that the press-codes for analysis of LOCA and transients.

ing safety concerns in these areas dictate the need for this continued code development. He noted also that RES is still having especially problems hiring personnel with the desired expertise, I with experience in high pressure thermal shock, man-machine inter-f aces, and human f actors.

C. LOFT G. O. McPherson, NRC Staff, provided a brief sumary of LOFT test ,

results and plans, and also provided proposed budgets for earlier

.> than originally planned retirement and decomissioning of the LOFT facility, including a phase out beginning at the end of fiscal year 1982 (see Appendix XV).

9 i

JULY 10-12, 1980 MINUTES OF THE 243RD ACRS MEETING The Comittee offered no disagreement with the position regarding LOFT as recomended by the Subcomittee: completion of all NRC-sponsored research and development by the end of FY-1982 and decomission it, or offer the facility to some other organization for continued research beyond FY-1982 at no cost to the NRC.

IV. Meeting with Members of the NRC Staff on Recent Ooerating Experience (0 pen to Puolic)

James M. Jacobs was the Designated Federal Employee for this

[ Note:

portion of the meeting.]

85 gp Three Mile Island 2: Venting of Containment to Release A.

(For background material, see Appendix XVI.)

J. Collins, NRC Staff, discussed the operation of venting the 85 Kr to the atmosphere, containment building of TMI-2 to release describing the systems and instrumentation used (see Appendix' He noted that at the time of his presentation, most of the gII). Kr had been purged. 18,500 cubic feet per minute were cur-rently being released, with only 6 Ci released during the last hour of the release. He expected that the purge would be completed Following completion of the purge, during the night of July 10. He the containment will be monitored each four hours for a week.

noted that no correction factors have been applied to . published doses. He said that the corrected value for the purged gas is now 43,800 Ci total . The measured offsite doses were well below the original estimate. The highest cumulative skin dose was calculated to be 0.3 mr. HeAll of the data from the monitoring program corre-noted that the weather was good during the period lated well.

of purging. He also noted that one error in the calculation was

~

that no correction had been made for the negative pressure of the in-stack monitor, since this matter had not been addressed in any of the manuals. He explained that the reason for the immediate shutdown after the original purge was that a high alarm was re-cei ved from an Eberline instrument in which two beta detectors alarg Xr because the program to be read omitted as a gas rather instructions than that allowed as a particulate. A for the multi-channel analyzer was installed that allowed for discrimina-tion during the later purges.

B. Brunswick 1/ Hatch 1: Potential loss of Scram Discharge Instrument Volume (SDIV)

(For background material, see Appendix XVIII.)

W. Mills, NRC Staff, discussed the events regarding the degradation

,' of the SDIV that occurred at Brunswick Unit 1 on Oct.19,1979 and i

l 10

1 l

l JULY 10'12, 1980 ,  ;

MINUTES OF THE 243RD ACRS MEETING l I 13, 1919. He also discussed the !&E at Hatch 80-14, Bulletin Unit 1 on JuneDegradation of Scram Discharge Volame Caoability, ,

issued June 12,1980 (see Appendix XIX).

1 E. Jordan, NRC Staff, said that the NRC Staff believes that a hydraulic event caused the trouble, but has not yet identifie event. He speculated that steam void, or may have been an overpres:ure The Committeeatreques-a temperature fo C

which the failed equipment was not designed.ted that t Staff has developed further information.

Browns Ferry 3:

Failure of Control Rods to Scram Fully

' C.

(For background material and preliminary reports on the event, see AppendixXX.) I l

W. Mills, NRC Staff, discussed the event All that '

99 inserted only partia31y after a scram (see Appendix XXI). ,

l rods that did not insert fully were on the east side of the core.

Insertion was adequate to bring the plant to near full shutdown.

Additional attempts to insert the rods further were successful and The mechanism of f ailure was all rods finally were inserted fully.

the presence of water in the scram discharge instrument volume, bu the cause of the presence of this water has not been determined l yet.

G. E. and the NRC Staff are working together with the Licen- '

sees who operate BWRs to detentine the nature of the event, and to further improve the reliability of this essential system. i j

M. S. Plesset requested that the NRC Staff return to the Committee ,

with further information when it is available.

. Resoonse of Primary Cooling System to Cool Down on D. St. Lucie:

Natural Circulation _

11, 1980 event, see Appen-(For background information to the June l dix XXII.) 11, 1980, E. Blackwood, NRC Staff, discussed the events ofHeJuneduri when the St. Lucie Plant, circulation conditions, developed pressure instabilities.

discussed the plant events and actions, discussed the design of the reactor coolant pumps, the primary coolant system, the pump seals, the ste indications, eratic pressure behavior, and a f XXIII).

11 f

JULY 10-12, 1980 l

MINUTES OF THE 243RD ACRS MEETING i attempting to Members questioned the advisability of the Licensh

, cooldown the p1 ant as rapidly as was done during the natural j circulation mode of operation.

E. Jordan, commented that if a Nuclear Data Link had been in opera-tion at the time of this event, the NRC Staff would have been monitoring the pressure aberrations, and could have recommended i that the plant return to higher pressure in order to eliminate the  ;

pressure instabilities.

R. Budnitz noted that neither the vendors' nor NRC Staff's analyti-cal codes currently are capable of predicting conditions inThese the stagnant water of the head region nor the upper head itself.This is one 3 items need to be coded separately.

said, that the RES Staff supports continued code development.

E. Jordan said that St. Lucie is the only plant to have cooled down on natural circulation, and that this is the second time that this plant has done so. He said that the NRC Staff has recovered and has noted the the pressure records from the previous event, same type of pressure instability.

B. Sherron, NRC Staff, discussed the implications of upper head voiding during transients and accidents, and concluded that such voiding is not considered a safety issue at this time (see Appendix He said also that NRR is considering requesting experimen-XXIV).

tal work in the Semiscale Facility to learn more about this type of transient.

D. Okrent suggested that a study could be made to detennine whether water can be " lost" from the primary coolant system without charg-ing water into this cooling system simultaneously.

V. Meeting with NRC Commissioners (0 pen to Public)

Raymond F. Fraley was the designated Federal Egloyee for

[ Note:

this portion of the meeting.]

Commissioners J. F. Ahearne, P. A. Bradford, V. Gilinsky, and

[ Note:

J. M. Hendrie were present for this meeting.]

A. Integration _-# Current Rulemak h

0. Okrent requested clarification regarding Commission plans to integrate degraded current rulemakings, including the considerati cores Class-9 accidents, proposed changes in the NRC nuclearIt was power noted plants siting criteria, and emergency planning.that the There is a need to define l Each of them interfaces on the others. l these interfaces. l 12 l

M

f . .

l l

JULY 10-12, 1980 MINUTES OF THE 243RD ACRS MEETING Comissioner Hendrie rephrased the original question to "How do all the rulemaking initiatives fit together?" He said that the NRC is The Comis-not making much headway in this type of coordination. He noted that sien does need to enunciate some general principles.

the Commission recognized that this may become a difficult problem, but that the Comission will have to proceed, and attempt to solve these problems as they are identified.

l W. Kerr raised the question of whether the NRC plans to resolve its l new problems in a mechanistic manner as in the past, or whether it '

will take a probabilistic approach.

V' Commissioner Hendrie offered his opinion that while detailed changes may be necessary, the current design basis concept for the j design of plants ought to h retained as a licensing standard. l Beyond the design basis events, the Commission should look at reasonable efforts to limit consequences, taking into account probability. To handle events beyond the design basis, he would A safety objec-recomend using a best engineering design basis.

tive is needed for the design basis, couched in terms of low probability for serious core damage. A further level is needed to deal with accidents beyond the design basis accident, and this can be handled by mitigating serious exposures to the public, such as-by evacuation.

(For a summary of Commissioner Hendrie's comments on the treatment of Class-9 accidents, see Appendix XXV.)

8. Role of The Office of the Analysis and Evaluation of Operational Data in Accident Investigation H. W. Lewis voiced concern that the Office of the Analysis and Data has been inhibited from taking a Evaluation of % rational meaningful part n. the investigation of recent incidents at reactors.

definition of role, Chairman Ahearne said that the organization, and manning of the AE00 has taken longer than was originally He anticipated, but noted recent strengthening of AE00 operations.

indicated that the charter of the AE00 will be published in the near future, and suggested that Members read this new charter, and inform the Comission if they still have reservations.

C. Resident NRC Insoectors functions of NRC resi-M. Bender voiced concern that the role at .

nuclear facilities do not appear to be well dent inspectors at defi ned. He suggested that it would be wise if the Comissioners inspectors in tenns that the public defined the role of resident can understand. l l

13 t

JULY 10-12, 1980 MINUTES OF THE 243RD ACRS MEETING Chairman Ahearne noted that the duties ofinspectors.

the residentAtinspectors this were derived from the role of non-resident time, there is no clear definition of the duties o tion.

D.

Policy Recarding the Standardization of Nuclear Plants _

colicy is M. Bender asked the Commissioners what He the not curreuhat there tsent time.

regarding standardization of nuclear plants.in this direction at t seems to be little thrust 6.

the Commission I t addressed Chairman Ahearne admitted that at TMI-2, becau- andardization this matter since the accident did not seem to be the most pressing issue. '

Connissioner Gilinsky said that he believes that if applications are day.

made for new plants, standard plants will be t ization, preapproved designs, and large scale duplication of to obtain economies of scale.with the nuclear industry, who mu detail before the start of construction.

W. Kerr offered the opinion that it may not be practicable to standardize a $1 billion item.

it may be appropriate to medf fy general D. Okrent suggested that He said that he design criteria to require standardized plants.has reac to improved safety.

Commissioner Gilinsky agreed that long term safety can be achiev by reducing variability in plants.

VI. Executive Sessions (0 pen to Public)

James M. Jacobs was the Designated Federal Exployee for this

[ Note:

portion of the meeting.]

A. Future Schedule

1. Future Agenda The Committee agreed on a tentative agenda for the 244th ACRS Meeting ( August) (see Appendix II).

14

JULY 10-12, 1980 MINUTES OF THE 243RD ACRS MEETING I

The Committee agreed to add to the schedule a meeting with the j Applicants for a power license for North Anna Unit 2 to discuss I the results of the augmented low-power test program, shift manning (e.g. technical advisor),

It alsoTMI-2 lessons agreed learned, to include in and the risk assessment studies. i schedule, if time permits, a review of the application of the l Fort Calhoun Station to operate at an increased power level ard to discuss with the Licensee of the Trojan Plant the impact on that plant of Mt. St. Helens volcanic eruption.

2. Schedule for ACRS Subcommittee Meetings and Tours j

A schedule for future ACRS subcommittee meetings and tours was l I distributed to ACRS Members (see Appendix III).  ;

l9 B. Subcomittee Activity i

1. Waste Management l of the S. Lawreski, Subcommittee Chairman, noted the intent ACRS Waste Management Subcomittee to evaluate theofproposed A set proposedNRC criteria on geologic disposal of radwaste.

criteria, incorporati ng public coments , is expected to be available by October 1980.

2. Fire Protection M. Bencer, Subcommittee Chairman, offered the opinion that if Branch Technical Position 9.5.1 is put into regul ation , it could cause confusion and more problems than it would solve.

More is needed to be known about the impact of the rules on the nuclear industry. The Edison Electric Institute is currently Understanding of fire compiling the industry positions.

protection principles and design needs to be understood well to protect public health and safety. He proposed that the Commit-tee defer action until after the Subcommittee holds its meeting in September, at which Branch Technical Position 9.5.1 will be considered.

3. Probability and Risk Assessment Review of systems interactions at Diablo Canyon, that might l result from seismic events, was assigned to the Probability This and Risk Assessment Subcommittee (O. Okrent, Chairman).

matter will be discussed at the Subcommittee's meeting in September.

15

a JULY 10-12, 1980 MINUTES OF THE 243R0 ACRS MEETING

4. Reactor Ooerations The Committee expressed its intent, during the sessio Evaluation of the NRC Inspection Program resident inspectors.

is the responsibility of the Reactor Operations Subcommittee (W. Mathis, Chairman).

C.

Estabishment of Supplemental ACRS Fund for Support of Meetings with Foreign Dignitaries The Committee agreed to establish a voluntary fund, contributed to by Members, for use to supplement NRC funds provided for the h entertainment of foreign dignitaries when visiting ACRS.

D. Subcomittee Reports

1. Quantitative Risk Acceptance Criteria D. Okrent, Subcommittee Chairman, reported on the July 1 Sub-Presentations were made comittee Meeting held in Los Angeles.

by E. Zebroski, EPRI, representatives from the Atomic Indus-  ;

trial Forum, Institute of Electrical and Electronics Engineers,A rev and General Atomics. ACRS fellows presented some new l tee report was considered. The Subcomittee l

. calculations for various types of conditions.

will try to have three connected draft memoranda available at the August meeting that might form a package for transmittal to the Commissioners as a beginning of communications in this area

l a sumary of past proposals, i i e '

l e the latest draft of the report, and e

a revision of the applicable studies on possible risk criteri a.

He suggested that since the audience of this report will be the .

Comissioners and the Congress, the main boctr of the report  !

should be in non-technical language, and the technical data should be appended to the report. He said that the Subcommit-  ;

tee would attempt to have this report in a semi polished form for the 245th Meeting (September), with a final draf t for L mittee approval at the 246th Meeting (October).

D. Okrent requested, and the Comittee agreed, that a scheduled time be provided at the 244th ACRS Meeting (August) for discus-He suggeste '

sion of the status of Risk Acceptance Criteria.

1 16 ,

______.-____-_______-__-________________-_______.__-__._.--__.._________-_-._______________a

1 JULY 10-12, 1980

• • MINUTES OF TliE 243R0 ACRS MEETING j

that experts in areas impinging on the subject of the a l tance including of risk Howard Raifa, Harold Greene, Lester Lave, aspects I

of and Arthu Discussions would include the legal l Kantrowitz. l Risk Acceptance Criteria, including the pos /

risk acceptance, etc.

E.

Items for Next Meeting with the Comissioners_  ;

The following items were suggested by Members for discuss

[

the Comissioners during the 244th ACRS Meeting (August).

The setting of priorities for requests made of Licensees by  !

E e Several Members were concerned that the large the NRC Staff.

number of requests made of licensees, al ,

I effect on safety.

l Several Members were concerned that e The Nuclear Data Link.from the nuclear industry has not Itbeen was fac-adequate input tored into plans for the proposed Nuclear Data Link.

suggested that the Comittee could discuss these matters members of the industry who will be present during the Comittee's discussion of Regulatory Guide 1.97 at. the 244t ,

ACRS Meeting (August), and the comments received from thes members of industry could be further transmitted to the Commi ssioners.

]

l F. Degraded Core Accidents '

the 244th ACRS Meeting The Committee agreed to provide time at(August) could be treated in the licensing process.

ACRS Recorts, Letters, and Memoranda _

G. j I.

ACRS Report to the Comission on the Fiscal Year 1982 Reactor l 3afety Research Program Budget The Comittee completed its annual report to the C  ;

on the NRC Fiscal Budget (see Appendix XXVI).

f'

2. Secuoyah Nuclear Plant Units 1 and 2_

The Committee completed its review and prepared a rep infonning the Comissioners that certain specified conditions, the Sequoya safety of the public (see Appendix XXVII).

17 W

JULY 10-12, 1980

" MINUTES OF THE 243R0 ACRS MEETING Ice Condenser Cacability_

3. Seouoyah Nuclear Plant Units 1 and 2:

l And Hydrogen Control I The Committee responded to a memorandum from Commissioner Gilinsky regarding the capability of the ice condenser contain-ment to cope with the steam from a large LOCA, and the matter l of hydrogen control in the Sequoyah Nuclear Plant Units 1 and (see Appendix XXVIII). i 4.

Comments on the Reactor Coolant Pumo Trip Requirement and H Pressure Injection Termination Criteria l

The Comittee responded to a memorandum from Chairm requesting additionalrequirements and high pressure injection temin pump criteriatrip (see Appendix XXIX).

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5. Utility of In-Core Themocouoles_

The Committee approved a letter to Commissioner Gilins regarding the utility of conditions and for use in accident detection of off-normal analysis (see Appendix XXX).

I

' 6.

Analysis of Barge Collision with Sequoyah Intake Structure.

The Comittee approved a memorandum from speed collision of a full regarding the analysis of a full barge tow with the new intake structure for Ce explosions and fires (see Appendix XXXI).

VII. Executive Sessions _ (Closed to Public)

James M. Jacobs was the Designated Federal Employee for this

[ Note:

portion of the meeting.]

New Members A.

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< . . .a JULY 10-12, 1980

.- v MINUTES OF THE 243RD ACRS .iETING L

1 l

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Saturday, July 12, 1980.

The 243rd ACRS Meeting was adjourned at 2:30 p.m.,

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19 l- _

APPENDIX I

y ATTENDEES 243RD ACRS Meeting JULY 10-12, 1980 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Milton S. Plesset, Chainnan J. Carson Mark, Vice-Chairman Myer Bender Max W. Carbon Jesse C. Ebersole William Kerr Stephen Lawroski Harold W. Lewis William M. Mathis ,

Dade W. Moeller David 0krent Jeremiah J. Ray Chester P. Siess ACRS STAFF Raymond F. r, Executive Director Marvin C. L . Assistant Executive Director James M. Jaco , Technical Secretary l Herman Alderman Andrew L. Bates David E. sessette John Bickel Paul A. Boehnert Sam Duraiswamy Elpidio G. Igne David H. Johnson William Kastenberg Morton W. Libarkin i Richard K. Major )

Thomas G. McCreless John C. McKinley Robert E. McKinney Gary R. Quittschreiber Richard P. Savio l John Stampelos l Peter Tam Hugh E. Voress Gary Young Dorothy Zukor 4-/

.-. . ,- . . - . - - . - . ~ . - - - - - .

NRC STAFF ATTENDEES

, 243rd ACRS Mtg.

Thursday, July 10, 1980 G. Zech, NRR-

2. Speis, NRR A. Tardani, NRR W. Mills, I&E E. Jordan, I&E C. B1achard, I&E C. Nelson, NRR J. Hannon, NRR I&E R.

C . Wessman, S ta h l e , NRR R. Tedesco, NFR H. Krug, SD R. Gamble, NRR J. Halapatz, NRR J. Muscara, RES J. Kudrick, NRR W. Butler, NRR T. Murley, RSR R. Budnitz, RES G. McPherson, RSR J. Collins, DSE E. Blackwood, 1&E B on, NRR i

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PUBLIC ATTENDEES

~-

243rd ACRS Mtg.

Thursday, July 10, 1980 S. Babineau, A. Rept. Co. D. Thumbil, Phil . Bulletin R. Leyse, EPRI H. C. Pfeff. , General C',ectric R. Kalper, NSOC D. L. Peterson, General Electric Higgins, Atomic Industrial Forum R. Gridley, General Electric S. Harris, EEI J. Full, UPI Photas M. Crutsinger, AP P. Higgins, AIF E. Jennings, ABC H. Hamadt., TEPC0 T. Foster, ABC R. Whit, EPRI R. Mu nd, Westinghouse 0. Batum, S. Services P. A. iinson, A. Rpt. Co. T. Rogers, PG&E R. J. i.oss, Douc & Munt:ing J. Benton, ABC W. J. Armstrong, Boston Edison Co. I R. Borsum, B&W L. M. Mills, TVA J. M. Ballentine, TVA l F. Fancher, CNN  !

M. Miller, Oak Ridger S. Harris, EEI[

M. Casso, U. S. Senator Hein:' Office l

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1 NRC i,TAFF ATTENDEES 243rd ACRS Mtg.

Friday, July 11, 1980 G. G. Zech, NRR V. Clayton P. K. Van Doorn A. R. Hirel, I&E R. J. Serbin W. A. Taylor, RES J. C. Pulsipher, NRR R. M. Bernero, RES E. L. Jordan, I&E R. Bland, RES l

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l APPLICANT ATTENDEES

^

243rd ACRS Meeting Friday, July 11, 1980 Westinghouse Tennessee Valley Authority T. F. Timr.ons Wana Lau M. A. Siano g, Reeves Robert H. Faas Paul A. Evans R. A. Bruce Robert M. Jessee D. K. Goeset J. Ballentine W. H. Bamford L. M. Mills D. L. Lambert J. Lambert G. F. Oilworth Edwin A. Merrick Bob Christie R. Joe Hunt D. L. Williams M. J. Burzynski C. W. Delecoth

Public Attendees 243rd ACRS Mtg.

Friday, July 11, 1980 Leyse, EPRI M. J. Miller, Oak Ridger R. E. Lapp, Lapp Inc.

M. Silverstein, WRRPS P. Higgins, Atomic Industrial Forum K. Fortino, LNRA&T E. Solomon, Shaw Pittman D. Hollar, Debevoise & Liberman R. Kesp, NSOC

5. Harris , EEI M. J. Scaroa, D&M Lester Kornblith, Jr., National Nuclear M. Schwart:, Pickard, Lowe, Garrick M. J. Miller, Oak Ridger W. H. Rasin, Duke Power Co.

O. Batum, Southern Co.

M.Edwa rds, NLTECH P. R. Davis, ITI F. Hamada, TEPCC T. C. Houghton, KMC

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, APPENDIX 1 I FUTURE SCHEDULE l

AeSuST (244th Meeting)

Hours

• Tnree riilo Island Unit 1 -- Restart and Full Power Operation Nortn Anna Unit 2 -- Aug.unted Low Dower Test Results, Shif t i manning (e.g. Tecnnic al Asvi ser), Tril-2 l Lessc s Learne_, Risa Assessment Stucies  !

4 Quantitative Risk Criteria Factorin; Of Degraded Cora Considerations into tne Licensing Process 1 1/2 l

• Regulatory Activities (Rag. Guide 1.97, et.al.) 3 1

Macting with Co.amissioners Overloading cf industry witn Requests Respcasicility Of State and Local Governments witn regarc to Emergency Planniny  ;

1

• ascade Failures in Nuclear Plants dew Members 1/2 1

(:iote: Additional items to be seneduled if time permits include: '

• Fort Calhoun -- stretch povar Trojan Plant -- 1mpact of tit. St. rielens eruption) l SEPTEM.5ER (245th Meetin;)

"Fi re Protection r,ulemaki ng Hatch /3runswich/ Browns Ferry -- performance of shutdown systems GCTOBER (246th Meeting)

Crystal River -- implications of incident of Feb. 26, 1980

• ACRS reports / action has been reques;ed by :iRC

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deleted as l

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[a =% 'o UNITED STATES

!" 5. g n NUCLEAR REGULATORY COMMISSION 5

• "/ $ ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 1

%, ' w [ff o, W ASHINGToN, D. C. 20555

., APPENDIX III Schedule of ACRS Subcommittee Meetings and Tours ACRS Members SCHEDULE OF ACRS SUBCOMMITI'EE MEETIES AND TOURS The following is a list of tours and Subccanittee meetings currently scheduled, subject to the approval of the Advisory Committee Management Officer. If you are listed and cannot attend a meeting, or if you are not listed but would like to attend, plea =? advise the ACRS Office as soon as possible.

Most hotels currently being used by ACRS Members in the downtown Washington and Bethesda areas require a guaranteed reservation if arrival is scheduled after 6:00 p.m. Failure to use a room under these conditions involves forefeiture of the cost. Please advise the ACRS Office as soon as possible if you cannot attend a meeting for which you are scheduled so that reservations can be cancelled in time to avoid this.

M. W. Libarkin, Assistant Executive Director for Project Review cc: ACRS Technical Staff M. E. Vanderholt B. Dundr R. F. Fraley M. C. Gaske J. Jacobs i

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SCHEDULE OF ACRS SUBCOMMITTEE MEE'i1N35 l

AUGUST 4 Reactor Operations (Major) Mathis, Ebersole, Etherington, Moelier, Kay.

Purpose:

To review the application for a STRETCH Power increase for 0maha Public Power Dist. ft.

Calhoun and includes a discussion with I&E regarding the duties and responsibilities of resident inspectors.

5-6 TMI-l (Major) Moeller, Etherington, Kerr(5th only),

Lawroski.

Purpose:

To review the modifications made to TMI-l in preparation for a restart following the TMi-2 accident.

o Regulatory Activities (Curaiswamy) Siess, Kerr, Mathis, 5encer, say. Furpose: To review RG 1.97 Rev. 2, et al.

6 Safety Philosoony and Criteria (Savio) Okrent, Bender, Emerscie, : nerington, Kerr, Mathis, Ray.

Purpose:

To review the progress made in setting requirements for NTCP plants.

7-9 244th ACRS meeting 19-20 Fluid Dynamics (Sar, lose, CA) (Bates) Plesset, Ebersole, Ma;nis, 5encer, Siess.

Purpose:

To review the Browns Ferry CRD insertion f ailure and the present status of the Mark I, Mark 11 containment programs.

21 Reactor Fuel (Idano Falls,10) (Bochnert) Shewmon, Mathis,

• Okren;. Furpose: To continue review of the NRC Fuel l Behavior Research Branch (FSRB) programs.

SEPTEMBER 3 Regulatory Activities (Duraiswamy) Siess, Kerr, Bender, Ray, Furpose; io review .ask No. RS705-4, Task No. SC 704-5 and Task No. RS 809-5 l 3 Reactor Fuel (Boehnert) Shewmon, Mathis, Lawroski, Mark.

Purpose:

io conclude review of the " cladding Swelling and Rupture Models for LOCA Analysis," NUREG-0630.

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SEPTEMBER (Continued) i 4-6 245th ACRS Meeting f

18-19 RSK (FRG) Pasadena, CA (Fraley) Plesset, Moeller, Okrent, IGy

Purpose:

To discuss NRC Siting Policy Design of f

)

B&W Reactors.

18-19 Combination of Dynamic Loads (Igne) Bender, Siess, .Etherington Purpose; io continue the review regarding the use of dynamic load combinations as a design basis for nuclear power plants 26 Crystal River 3 (Igne) Lawreski, Mathis, Kerr, Moeller, (tent.) UKrent, Plesset.

Purpose:

To review the Crystal River 3 transient that caused flooding of its containment and to ,

review NRC Staff recommendation on the Crystal River 3 fix.

i OCTOSER 8 Safe +y Philosophy & Criteria (Savio) Okrent, Bender, Ebersole, ,

Eiierb y cn, Kerr, Matnis , Ray

Purpose:

To review seismic-related ystems interactions for Diablo Canyon.

9-11 245th ACRS meeting

? ECCS (Idano Falls) (Bates) Plesset, Mathis, Okrent, Ebersole.

eurpose: To review new data on RCP Trip & Semiscale Program i

? Imoroved Safety Systems (Duraiswamy) Siess, Okrent, Plesset, l Moel ier, Lawrosa .

Purpose:

To discuss NRC and DOE's l

)

research program plans to improve reactor sa#ety, etc.

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Note: Subcommittee meetings on the Waste Disposal Feasibility Rule and on the proposed geological disposal criteria are being clanned.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING SUBCOMMITTEE STAFF ENGR. & MCMBERS DATE Reactor Operations (MAJOR) Mathis, Ebersole, August 4,1980 ' Etherington, Moeller, Ray

' t' '  !

LOCATION: Washington, D.C.

I BACKGROUND:

. Who proposed action: NRC Staff

Purpose:

The primary purpose of this meeting will be to review the applica-tion for a STRETCH power increase for Omaha Public Power District's Ft. Calhoun Nuclear Station, Unit No. 1. If approved, Ft. Calhoun's As licensed power level would increase from 1420 MWt to ISOC MW:.

time permits, a second goal for the meeting will be to review the various ACP.S genaric items assigned to the Reactor Operation's Subcommittee to monitor. These include:

1. NPSH for ECCS Pumps

11. Quality Assurance During Design, Construction and Operation

18. Criteria for Pre-Operational Testing

19. Isolation of Low Pressure from Hign Pressure Systems

51. Maintenance and Inspection of Plants

65. ACRS/SRC Periodic Ten-Year Review of All Power Reactors

69. Locking Out of ECCS Power-Operated Valves PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

An SER on Ft. Calhoun's STRETCH Power increase is expetted in early to mid-July.

A description and status report on those generic items assigned to the Reactor Operations Subcommittet will be included in the project status report if it appears the Subcommittee will have an opportunity to review them.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING i

STAFF ENGR. & MEMBERS SUBCOMMIUEE DATE Three Mile Island, Unit 1 (MAJOR) Moeller, Etherington

• - August 5 & 6 Kerr (5th only), Lawroski,  !

Cons: Catton, Lipinski l J

Zudans, Grendon, LOCATION: Washington, D.C.

l BACKGROUND: \

Who proposed action: Moeller/ Staff  !

j

Purpose:

The purpose of this meeting will be to review the modification made <

to TMI-1 in preparation for a Restart following the TMI-2 accident.

A previous Subcommittee meeting on the TMI-1 Restart During was held in this previous Middletown on January 31 - February 1,1980.uation of Licensee's Compli-meeting the " Status Report on the Er ance with the NRC Order Dated Auguts; ,,1979, Metropolitan Edison 11, 1," January Company, et al, Three Mile Island Nuclear Stacion Unit 19S3 served as the basis for discussions. Briefly, the eight Order items were:

la. EFW Timeliness and Reliability; lb. Independence of EFW from ICS; ic. Reactor Trip on Loss of Main Feedwater; ld. Analysis of Small Breaks; le. Retraining of Operators;

2. IE Bulletins;

3. Emergency Preparedness;

4. Separation of TMI-2 Operations from TMI-1;

• 5. Waste Management Capability;

6. Managerial Capability and Resources;

7. Financial Qualifications; and

8. Lessons Learned Category A Items.

There are also four Long-Term Actions considered:

1. ICS Failure Mode and Effects Analysis;

2. Small Break Analysis and Procedures;

3. Lessons Learned Category B Items; and

4. Emergency Preparedness.

By June 20, 1980, the Staff expects to have a rewrite of the

" Status Report on the Evaluation" available. This document will basically cover the same ground as the Janaury 11, 1980 version, This although conclusions reached or form of review may differ.

document will not cover any items outside of those specified in the order. (A major comment from the Subcommittee at the pre-vious meeting was along the lines of, what else, besides the order f

-/

2-requirements, does the Staff believe needs to be addressed.) The LPM believes that an additional list of items which must be ad-

, l dressed will be produced about a month after the re-write of the ,

. Status Report is available. (Unofficially, the list of additional l

i items for the TMI-1 Restart may be similar to the NTOL list.)

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY: l

" Update.on the Status Report on the Evaluation of Licensee's Compliance with the NRC Order Dated August 9,1979" issued on June 1e,1980.

A report on items in addition to those considered above will be available approximately one month after the Status Report.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE STAFF ENGINEER & MEMBERS Regulatory Activities (DURAISWAMY) Siess, Kerr, Mathis, 8/o/80 Bender, Ray CONS: Z. Zudans, W. Lipinski, I. Catton i

PURPOSE: To review the following items:

Post-Comment Item

1. Regulatory Guide 1.97, Revision 2 " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" Pre-Comment Items

2. Proposed Regulatory Guide 1.8, Revision 2, " Personnel Selection and Training"

3. Proposed Amendment to 10 CFR 50, Appendix A to Reference

' 10 CFR 50, Appendix B s

4. Proposed Regulatory Guide (Task No. RS 801-4), " Periodic Testing of Torque-protected Motor Operated Valves Important to Safety"

5. Proposed Revision to 10 CFR Part 50.54, " Control Room Staffing and Working Hour Limitation"

6. Proposed Regulatory Guide 1.33, Revision 3 " Quality Assurance Progt im Requirements (Operation)" l Copies of these items will be transmitted to you as soon as they are available.

NOTE: The following Regulatory Guides, previously scheduled for discussion at the August 6, 1980 meeting have been deferred. These items will be sub-mitted for the Subcommittee's review in one of the future meetings:

1. Regulatory Guide 1.136, Revision 2, " Material for Concrete Containment".

2. Regulatory Guide (Task No. SC 704-5), " Functional Specification for Safety-Related Valve Assemblies in Nuclear Power Plants".

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Cont'd Reg Act Mtg 8/6/80 l l

lL STATUS: The status of these items are as follows:

Item 1 - Revision 1 to this Guide was issued as an effective

' Guide in August 1977. As a result of the TMI-2 accident, several questions arose with regard to the reliability and adequacy of certain essential instruments during and following an accident. The proposed Revision 2 to this Guide reflects consideration of the lessons learned and questions raised subsequent to the TMI-2 accident.

A previous version of Revision 2 to Regulatory Guide 1.97 was reviewed by the Regulatory Activities Subcommittee at the November 7,1979 meeting. The ACRS full Committee was also briefed by the NRC Staff on the technical contents of this Guide during the November 8-10, 1979 meeting. The ACRS full Committee agreed with the NRC Staff's plan to issue this Guide for public comment. It ,

was issued for public comment in December 1979. Numerous i comments were received by the NRC Staff from the industry l and other technical communities during the public j comment period of this Guide. The present version of l this Guice reflects consideration of public comments. l The NRC Staff requests ACRS concurrence in the Regulatory Positions of this Guide.

item 2 - This Guide endorses ANS!/ANS 3.1-1978, " Selection l and Training of Nuclear Power Plant Personnel".

l It describes a method for complying with the Commission's Regulations with regard to personnel selection and training for nuclear power plants.

A previous version c' this Guide was reviewed by the Regulatory Activities Subcommittee at the December 6, 1978 meeting and was issued for public comment in February 1979. Af ter the TMI-2 accident, this Guide was reissued for public comment in May 1979 with the intention of getting public reaction on the personnel qualification requirements delineated in this Guide.

Subsequent to the TMI-2 accident, ANSI /ANS 3.1 Standard which is endorsed by this Guide, has been revised ex-tensively to incorporate the lessons learned and to provide improvement in the personnel qualification area.

Consequently, this Guide has also been revised to in accordance with the Draft ANSI /ANS 3.1 Standard, dated December 6, 1979.

Af ter the Subcommittee's review, the NRC Staff may reissue this Guide for public coment.

Cont'd Reg Act Mtg 8/6/80 -

Item 3 - This is to initiate rulemaking to revise Appendix A to 10 CFR Part 50 to state that the criteria for the quality assurance program required by Appendix A are those criteria listed in Appendix B to 10 CFR Part 50.

This proposed revision has been initiated by realizing the fact that there has been inconsistent treatment

~

with_ regard to which of the various nuclear power plant structures, systems, and components should come under the quality assurance program requirements of Appendix B. Previously, several attempts were made in vain to resolve this inconsistency and confusion.

The NRC Staff believes that the proposed Amendment to Appendix A to 10 CFR 50 will clarify the relation- ,

ship between Appendices A and B. and will ensure appli- l cation of the appropriate quality assurance program requirements to each structure, system, and component important to safety for both operating and planned nuclear power plants.  :'

Subsequent to the Subcommitteas' review, the NRC Staff may issue this item for public comment.  !

Item 4 - This Guide provides guidance for in-situ testing of torque-protected motor operated valves important to safety to make l sure that these valves perform their intended functions.

7 1 I

Subsequent to the Subcommittees' review, the NRC Staff may '

issue this Guide for public comment.

Item 5 - This proposed revision to 10 CFR Part 50.54 is to incorporate the following recommendations that are delineated in SECY 80-230, dated May 2, 1980:

1. A technical advisor to the shift supervisor shall be present on all shifts and available to the control room within 10 minutes. (This requirement shall be met be-fore fuel loading.)

2. The minimum shif t crew for a unit shall include 3 operators, plus an additional 3 operators when the Unit is operating. Shift staffing may be adjusted at multi-unit stations to allow credit for operators holding licenses on more than one unit.

Administrative procedures shall be established to limit maximum work hours of all personnel performing a safety-related function to no more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of continuous -

duty with at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> between work periods, no more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in any 7 day period, and no more than 14 consecutive days of work without at least 2 consecutive days ori.

/) - / V

Cont'd Reg Act Mtg 8/6/80 . . .

(These requirements shall be met before fuel loading.)

. Subsequent to the Subcommittee's review, the NRC Staff may issue this for public comment.

. Item.6 - This Guide describes a metned for complying with~ the Commission's regulations with regard to overall quality assurance program re-

- quirements for the operational phase of nuclear power plants.

The previous version of this Guide endorsed, with certain ex-ceptions, ANSI N18.7-1976/ANS 3.2, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants".

A previous version of this Guide was reviewed by the Regulatory Activities Subcommittee in March 6,1979 and was issued for  ;

public comment in August 6, 1979.

Subsequent to the TMI-2 e.ccident, ANS 3.2 Standard has been revised to reflect the lessons learned from the TMI-2 accident.

Consequently, this Guide has been revised in accordance with the Draft ANS 3.2 Standard, dated February 1983.

Subseouent to the Subcommittee's review, the NRC Staff may issue this Guide for public comment.

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Sa#ety o n '::s::'y and Cr'teri a (So.nc} Ok rent , Be ncer, i Aug. 6 Ebersole, Etneringto ,

Ke rr, Mat tis , P.ay

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Par::se: The ru- :se Of th's mee*.ing w'll be to re /i ew the Dr09-ess ma de i i n set *. ' r.: reca'-ements for NT'? Diants. The Sta## e x : e:*. s I

i have issued a 00--ission paper en NT'.05 and 9.Ls unien will be a re:;'#e ent's a:Lage that int:r; orates comments from tne ore-vi cus *::.5 Su :cm-' tee meeting and NT' a0:licants. The Sta##

wil' also have su: le9 ental des:-i *.i ons c# Action Plan se:*.icns that 0:ntain recuirements for NTCDs. The NTCP a0:14: ants w001:  ;

like to ciscuss studies performed D.v their consultants weien u Orct;;e interi a:;*ea:hes to resolution on several :' icy l Ques'.i cns w* i:n rel at e t: how and whe*.ner Constru: tion De-mi*.

a -li ations w'11 be pro:essed in :ne near-term. Ine tw policy issues to be addressed will be Siting and De; 3:e: 00-a Conditions. Tne Staf f will mment 0- the aorlicants' studies.

3 ,...n....:

p.3....,c. ... -. . ,.-,.,.,:

.t .o i. ,J r ,. .. ...h...,:.  :.L. .

Com-ission Paper - S;::l e e-tal A:ti:n Plans - ea-ly June.

N'. '.D. n-'.4...*.*.'

. S ~. ". ". *. =. c. ~~4

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING l l

OATE SUBCOMMITTEE STAFF ENGR. & MEMBERS Fluid Dynamics (Bates) Plesset, Bender, l 8/19-20/B0 Ebersole, Mathis, Siess i San Jose, CA i

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I BACKGROUND:

Who proposed action: M. Plesset

Purpose:

To review the Browns Ferry CRD insertion f ailure and the Brunswick and Hatch scram discharge volume instrumentation failures. In ad- l l

dition, the present status of the Mark I and Mark !! containment activities will De reviewed.

J Pe rti ne nt l Publications: Appropriate information and data will be supplied prior J

to the Subcommittee meeting.

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SCHEOU'.E OF ACRS SUSOC""ITTEE MEETIN3 STA:F ENG:.. & MEM3EES SUE ~0M*:TTEE DE

.__ (BOEuNE;T) Shee.on, Mathis, Rea: tor Fuel

'B/21/80 Ok.re n t Icah: Falls, ID BA:v.3E0'JQ:

Sh: propcsed a: tion. F. Snee.:n/NR -Resea-:h Purp0se: Te : .tinue reviee of the N?.C Fuel Behavier Ftesearch Branch (FSE5' progra s.

At the Aoril 29,195: Subcorr-ittee meeting, the Fuel Subcomittee bega revie cf the F5:.E's programs for the Comittee's annualThe subje:t mee repcet t: the Com-ission and Congress.

continue review of the PSF and 0:ne- In-File research progra: 5.

PE:~? NEN* PUE'.::fT:0NS At;: THEIR AV A!'_ AS:.!TY :

Wi'.1 be Or: viced i- t6e near future when available.

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Date Subcommittee Staff Encineer & Members Regulatory Activities (DURAISWAMY) Siess, Kerr, 9/3/B0 Bender, Ray CONS: 1. Burge- Johnson PURPOSE: To review the. following items:

1. Regulatory Guide (Task No RS 705 4), " Lightning Protection for Nuclear Power Plants" (post comment).

2. Regulatory Guide (Task No. SC 704-5), " Functional' Specification for Safety-Related Valve Assemblies in' Nuclear Power Plants" (post comment).

3. Regulatory Guide (Task No, RS 809-5), " Qualification Tests for Cable Penetration Fire Stops for Use in Nuclear Power Plants" (post comment).

Copies of these items will be transmitted to you as soon as they are available.

STATUS:

The status of these items are as follows:

1 Item 1 - This Guioe describes acceptable criteria for the design, application and testing of lightning protection systems to assure that electrical transients resulting from lightning phenomena do not render inocerable or cause spurious operation of systems important to safety.

A draft copy of this Guide was reviewed by the Regulatory Activities Subcommittee at the May 3,1978 meeting. During this meeting, it was brought to the attention of the Subcommittee that there were I some di fference of opinions among the NRC Staff with regard to the I technical adequacy of this Guide, The Subcommittee suggested that l the NRC Staff try to resolve these differences and resubmit this Guide in one of the future meetings for further review.

A revised copy of this Guide was reconsidered by the Subcommittee at the January 3,1979 Regulatory Activities Subcommittee meeting.

At this meeting, the Subcommittee was informed that all of the written technical comments based on dissenting views have been resolved with the exception of those from Mr. Rosa. The Office of .

Standards Development stated that they have reached a compromise with Mr. Rosa that his views will be included in this Guide as an

, alternate Regulatory Position and public comments will be solicited c, on the alternate approach.

This Guide was issued for public comment in August 1975. The

- present version of this Guide reflects consideration of public comments. The NRC Staff requests ACRS concurrence in the Regula-tory Position of this Guide.

EEk b

l Reg Act Mtg of 9/3/80 -,

l Item 2 - This Guide delineates a procedure for implementing the regulatory i requirements with respect to detailed specification of information pertinent to. definition of operating requirements for valve assemblies  ;

whose safety related function is to open, close, or regulate fluid l flow in light-water-cooled nuclear power plants . It endorses, with certain exceptions, ANS! N276.1-1975, "Self-operated and Power Operated.

Safety-related Valves Functional Specification Standard".

A previous version of this Guide was reviewed by the Regulatory Activities Subcommittee at the January 4,1978 meeting. Subsequent to that meeting, it was issued for public comment. The present version of this Guide reflects consideration of public comment.

The NRC Staff recuests ACRS concurrence in the Regulatory Positions of this Guide.

Item 3 - This Guide endorses , with certain exceptions , 'IEEE Standard 634-1978, "IEEE Standard Cable Penetrations Fire Stop Qualification Test".

This Guide provides guidance for establishing a standard type test for qualifying cable penetration fire stops used in nuclear power plants.

A previous copy of this Guide was reviewed by the Regulatory Activities Subcommittee at the April 4,1979 meeting and was issued

, for public comment in July 1979.

The present version of this Guide reflects consideration of public comments received during the public comment period of this Guide.

The NRC Staff requests ACRS concurrence in the Regulatory Position of this Guide.

+

i-SCHEDULE OF ACRS SUB00MMITTEE MEETING STAFF ENGE. & ME".BERS

- DATE m .=

.c= ' ef...-

. - --o r Reactor Fuel (BOERNERT) Shewmon, Mathis.

Lawreski, Mark j(3[jd BA C KGRDUND_ :

Whe proposed action: P. Shewmon, NRC Staff

Purpose:

This 1/2 day meeting will be held to conclude review of the NRR NUREG report " Cladding Sweiling and Rupture Models for LOCA Analysis" - NUREG-0630. This item had been scheduled for Com=ittee discussion at the March meeting but was postponed due to schedule constraints. .

Prior to the April ACRS meeting it came to our attention that NRR and RES had substantial differences of opinion over espects of the NRR models. Accordingly the subject meeting will be held to discuss these differences.

! OUTCDuE:

It is anticipated that the NUREG will be reviewed at the September ACRS meeting.

! pEETINEN~ PUELICATIONS AN: THEIR AVAILAEILITY:

j ill be provided when available.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETINQ STAFF ENGR. & MEMBERS.

SUBCOM"ITTEE ~

DATE (IGNE) Bender _, Siess

=s'*" Combination of Dynamic Loads Okrent, shewmon, Plesset, Etherington qJi;I/jo f

BAC'K3ROUND: - I' Who proposed action: M. Sender

Purpose:

The purpose of the meeting is to continue the review regarding the use of dynamic load combinations as a design basis for nuclear power plants.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:_

Documentation on this subject matter Will be received from the NRC Staff at the end.of July or early August 1983.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING L

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j DATEL SUBCOMMITTEE STAFF ENGR. & MEMBERS 1

-U-da'*' ,;d Crystal River 3 (IGNE) Lawreski, Mathis, Kerr, Moeller, Okrent,

. q q c .gg) Plesset of the TM' Subc..

will be invited.

BACKGROUND:

Who proposed action: S. Lawroski

Purpose:

The purpose of the meeting will be to review the Crystal River 3 transient that caused flooding of its containment and to review the NRC Staff recommendation on the Crystal River 3 fix.

The plant is , scheduled to start up about late June or early July.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

The Staff Safety Evaluation Report is due to be released on June 10, 1980.

The ACRS-Task Force on Crystal River will review the SER and recommend to the Subcommittee its findings.

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SCHEDULE OF.ACRS SUBCOMMITTEE MEETING-l l

DATE SUBCOMMITTEE STAFF ENGR. & MEMSERS l 10/8/80 Safety Philosophy & Cri- (Savio). Okrent, Bender, Eber-teria sole, Etherington, Kerr, Mathis, Ray BACKGROUND:

Purpose:

To review seismic-related Systems Interactions at Diablo Canyon.

The ACRS completed its review of Diablo Canyon for a full-term operating license in July 1973 and issued its. report on July 14, 1978. Subsequent to tne ACRS review, PG&E undertook a systems interaction program for

, seismically induced events'to assure that no systems required for safety would be disabled by the seismically induced f ailure of nonsafety grade systems or equipment. A description of this program was published on June 10, 1980. The review of this effort by NRC's Systems Interaction Branch (J. F. Stol:). is due to be completed in early August with an SER supplement due to be issued in mid-August. A Subcommittee meeting to - )

review this effort would appear to be appropriate in about mid-September.

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SCHEDULE OF ACRS SUBCOMMITTEE MEET t!G

~

SUBCOMMITTEE STAFF EN3R. & MEMBERS DATE October ? ECCS (Bates) Plesset, Okrent, Mathis, Ebersole BACKGROUNO:

Who proposed action: Dr. Plesset

Purpose:

To review the new infomation available on RCP Trip and one progress of the Semiscale program.

Pertinent Publications Appropriate documents will be provided prior to the meeting.

h 9

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Subcommi ttee Staff Enaineer & Members Date Improved Safety (DURAISWAMY) Siess, Okrent, (To Be Scheduled Plesset, Moeller, Lawresai in October) Systems PURPOSE: To discuss the following:

1. NRC's Improved Safety System Programs (current status ,

achievements , results of the completed programs , if any) .

2. DOE's Light-Water Safety Technology Program Plan.

3. NRC-DOE Coordination.

4. OMB attitudes / policies toward NRC & DOE research programs pertinent to Improved Reactor Safety Systems.

5. Response from Congress (if any) on the ACRS recommendation contained in the FY 79 ACRS report to Congress (NUREG-0657) witn regard to Confirmatory Research vs. Exploratory Research, etc.

6. Passive Containment System (PCS) Concept by NucleDyne Engineering Corporation - DOE's evaluation of PCS concept.

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APPENDIX IV

- AWARD TO D. OKRENT OF THE FIRST ANS I - ' TOMMY" THOMPSON AWARD r# su fljfb-u?Eff w nc.

3.

INTERN ATION AL CO N SU LTANTS

-* " 13 m 3 4 suae a 2. Tasao'. %.<,

17 June 1980 n,y as: c a a Q ,,, Cf

, ,.fg.y$$$ *~~ ** too "

~-

-S Mr. Raymond Fraley sos.es2. ass: ,

Advisory Comittee on Reactor Safeguards U. S. Nuclear Regulatory Comission

• Washington, D. C. 20555 Dea- Ray, The first "Tomy" Thompson award was presented to Dave Okrent by the Nuclear Reactor Safety Division at a Division luncheon in Las Vegas on June 9,1980.

Dave was present, with his wife, at the luncheon and was given a framed citation and a check for $300. The citation read:

"This first "Tomy" Thompson award is presented to you for your outstanding ,

contributions in the field of nuclear safety. Your work at the Argonne National (

Laboratory and at the University of California, and your long and continuous service on the Advisory Comittee for Reactor Safeguards have been characterized by exceptional dedication, insight and creativity in understanding and evaluation i of nuclear safety, nuclear risk and nuclear plant designs. You also were instru-mental in organi:ing and directing the initial operations of the Society's Nuclear Reactor Safety Division."

In making the presentation, I made the following remarks:

"It is a real honor for me te represent the Nuclear Reactor Safety Division in presenting the first "Tomy" Thompson Award. The Executive Committee of the ,

Division decided last fall that the Division should present such an award on an j annual basis for outstanding contributions to the field of nuclear reactor safety.  !

I must say that I am also very pleased with the selection of the name of the award, since I knew and admired " Tommy" Thompson from the days when I was a reactor in-spector for the AEC and recomended in 1958 that the MIT research reactor be issued an operating license. I also was among those who strongly urged Tommy to serve on the Advisory Committee on Reactor Safeguards where, among other things, he played a leading role in the safety review of the N.S. Savannah; and also as chairman.

Tomy later became a Comissioner of the AEC and was one of the very few who under-stood from direct experience wiat reactor design, operation, and safety were all about. In a humorous vein, though, I can remember Admiral Rickover telling Tommy at an ACRS meeting that he knew that Tomy was asking questions about the Navy program just to learn scmething he could then teach at MIT.

I'm also very pleased with the selection of the first recipient of the Tomy Thompson Award. My first exposure to Dave Okrent was when he was a consultant to the ACRS on fast reactors, also around 1958, and I have respected his intellect and dedication to reactor safety ever since. He, of course, then became a member of the ACRS and has served as Chairman. In fact, Dave, who is still on the ACRS, 3/

lh[ NNbb. Ut.I.F.l ine.

Mr. Raymond Fraley 17 June 1980 Page No. 2 a

has served more years on the ACRS than any other member. It probably isn't appropriate to single out any particular accomplishment of Dave's for special recognition, but the one that stands out in my memory is his determination to get adecuate review of loss of coolant accidents, clearly a very foresighted objective. Finally, I think it is appropriate that Dave be recognized for his leadership in organi:ing and directing the early operations of tne Nuclear Reactor Safety Division, now the third largest division in ' ANS, with approxi-ma tely 2,500 members."

Best regards, 2

Peter A. Morris Executive Vice President PAM:jw l

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SCHEDULE ruR JULY 11, 1980 DISCUSSIONS ON THE SEQUOYAH NUCLEAR PLANT 8:30 - 9:00 SUBCOMMITTEE REPORT (J. C. MARK) 9:00 - 12:30 DISCUSSIONS WITH NRC STAFF AND APPLICANT 45 Min ,

. 9:00 - 9:45 1. NRC STAFF REPORT 1 I

a) Review Schedule b) Status of non-TMI Open Items l c) Status of TMI Open Items l

d) Discussion of Pressurizer Vent l Pipe Repairs

e. ) TVA Response

2. REPORT ON SEQUOYAH N0ZZLE CRACKS (TVA) 15 min 9:45 - 10:00 SREAX 15 min

3. STATUS REPORT ON ICE CONDENSERS 45 min 10:15 - 11:00 a) Discussion of TVA Risk Assessment Work (TVA) b) Discussion of PAS Oraft Study on Ice Condensers (NRC)

4. 60 min 11:00 - 12:00 STATUS REPORT ON HYDROGEN CONTROL STUDIES a) Discussion of TVA Studies and Proposed Interim Measures for Enhancing Hydrogen Control Capabilities (TVA) b) NRC Response / Status Report on NRC Work EXECUTIVE SESSION 30 min 12:00 - 12:30 ll- 37

PRCPOSED MEETING SUM $ARY.0F THE JULY 9, 1980 ACRS SUBCOMMITTEE ON THE SEQUOYAH NUCLEAR PLANT, WASHINGTON, D.C.

s PURPOSE:_ The purpose of the Subcomittee meeting was to discuss TVA's applica-tions for a full power operating license for the Sequoyah Nuclear Plant.

PRINCIPAL TOPICS-OF DISCUSSION: The principal topics discussed were the imple-l mentation of the TM1 NIOL requirements, interim measures for hydrogen control, )

risk assessment for ice condenser plants, and the status of the NRC's filtered l 4

vented containment studies.

ATTENDEES: '

NAME AFFILIATION J. C. Mark ACRS Member i

W. Mathis ACRS Member

1. Catton ACRS Consultant W. Lipinski ACRS Consultant

2. Zudans ACRS Consultant R. Savio ACRS Staff l

C. Stable NRR '

G. Zech NRR R. Tedesco NRR A. Schwencer NRR

0. Parr NRR R. Wessman ILE J. Buzy NRR R. Leyse EPRI M. Burzynski TVA J. Ballentine TVA '

L. Mills TVA V. Esposito W R. Gamble RRC G. Di11 worth TVA D. Lambert TVA

0. Williams TVA G. Reed TVA W. Lau TVA W. Johnson W_

MEETING HIGHLIGHTS AND AGREEMENTS:

1. TVA brought Sequoyah Unit 1 critical on July 5, 1980 and expects to begin the special low power testing on July 12, 1980. The schedule calls for completing of the special low power tests in three weeks. The NRC Staff expects to have a draf t SER available to the Comittee by July 28, 1980. The Staff's schedule projects licensing of the Sequoyah plant near the end of August. The NRC Staff requested that the ACRS, if possible, write a letter on the Sequoyah plant during the July 10-12, 1980 ACRS Committee meeting and indicated thst if it would not s be possible for the ACRS to write a letter or, the full power operating license they would request an ACRS letter on hydrogen control for the Sequoyan plant.

W'

/]-3C  !

l

Sequoyah Mtg. Sumary.

2. The NRC gave a status report on the NRC review of the full power issues.

These era either resolved or are expected to be resolved within the next two to three weeks. A list of the outstanding issues is included as Attachment A.

3. TVA discussed the measures taken as protection against the design base flood.

The PMF is based on a 16.8" three day rainfall distributed over the water shed (21400 sq. mi.) preceded three days earlier by a 6.7" three day rainfall. The seismically induced flood is caused by the failure of the four major upstream

. dams coincident v:ith t. river crest corresponding to one-half of the PMF. The maximum' flood with wind driven waves is approximately 20' above plant grade.

The auxiliary and turbine buildings would be flooded under these conditions.

Safety grade equipment is protected with water tight doors and sump pumps powered by the diesel generators.

4. TVA reported on the Sequoyah nozzle cracks. The nozzles in the Sequoyah Unit I have been inspected and underclad cracking has been found. The cracking ,

is extensive in some parts of the nozzle but all crackr :re below the critical i flaw size. The cracking phenomena is believed to be a.c niated with the method of installing the weld cladding. The cladding in the Prairie Island vessel was installed using similar techniques. Inspection of this vessel is i expected in about six months. The object will be to obtain infomation on crack growth. Prairie Island has been in operation for about nine years.

5. Various ice condenser orientated risk assessment studies were discussed.

A systems interaction methodology applications study has been performed by i

- Sandia using the Watts Bar plant as a model plant. The Reactor Safety Study Metnodology Application Program (updated WASH-1400) sponsored by PAS has used the Sequoyah plant as the model me condenser plant. TVA has sponsored a reliability evaluation for the auxiliary feedwater systems in the Sequoyah l plant and will sponsor a plant safety and availabilit,y analysis. The latter l study is orientated toward both safety and plant availability and will l consider all plant systems and not be confined to safety systems. The PAS l sponsored study is not yet completed. The tentative conclusion, however, is that the risk associated with the ice condenser plant is, within the uncertainties of the study, comparable to that associated with the WASH-1400 ,

plants.

6. The status of the hydrogen control studies for ice condenser plants was discussed. TVA has concluded that Sequoyah can withstand a hydrogen burn equivalent to approximately a 25% metal-water reaction (using the ultimate strength of the material as a failure criteria). TVA indicated that their studies had addressed filtered vented containments, ignition sources, and Halen suppression. Ignition sources were selected on the basis of the study

s a nost promising means of interim hydrogen control. On the basis of these studies TVA has determined that filtered vented containments and inerting are not acceptable means for hydrogen control. TVA is proposing comercially available ignition sources as an interim means of hydrogen A-37

Sequoyah' Mtg. - Sumary  ;

control and the TVA studies will continue to address improved ignition source !,ystems and Halon suppression. The NRC Staff indicated that they would evaluate the proposed interim ignition source system. TVA would be permitted to install the system but would not be permitted to operate it until the NRC review was complete. The Staff feels that the operating li-cense need not be withheld until the completion of this review. It is ex-pected that the system could be installed by TVA and evaluated by the NRC within'three to four months.

7. The status of the NRC's programs on filtered vented containments was di scussed. The effectiveness of systems for Zion and Indian Point are cur-rently under study. The proposed systems would contain a suppression pool, with the possibility of using sand and gravel filters and HEPA/ charcoal filters for further dose reduction. Containment openings of about three feet in diameter appeared to be of sufficient size to cope with most ac-cidents.

8. Mr. Joseph Halapatz of the NRC Staff appeared before the Subcomittee and expressed his concerns regarding the repairs made on the Sequoyah Unit 1  !

pressurizer relief pipe. Weld drawbeads were used to strengthen the re- j lief pipe. The affected section of pipe is six inches in diameter and cannot be isolated from the primary systems. Mr. Halapatz's principal concern is that the pipe material has been sensitized in the region of the repair welds. l

-- FUTURE MEETINGS: The TVA application to operate Seouoyan Nuclear Plant, Units 1 anc 2 will be discussed on July 11, 1980 during the July 10-12, 1980 ACRS full Committee meeting. Subcommittee and full Comittee dis-cussions will be held as required after this.

Attachments:

1 Charts i

l W

i

• • b 4e

e ~

s COMPLETE (FULL POWER) Till ISSUES ON SEQUOYAll UNIT N0.1 f

1 1. REACTOR INSPECTOR AT ODERATING REACTORS l 2. Sil0RT TERM ACC. ANALYSIS AND PROC REVISION

3. NSSS VENDOR REVIEW OF PROC <-('--' f' 11 . PILOT MONITORIflG 0F SELECTED EMERG. PROC FOR NT0L APP. - W g 5. LOW POWER TESTING TRAINit!G 1 5. Pl)NT SilIELDING g g Pe- c#

d 7. EMERG. POWER FOR PRESSURIZER llEATEP.S - # E#

8. PRIt1ARY COOLANT SOURCES OllTSIDE CONTAINMENT 2'E

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C/ INCOMPLETE (FULL-POWER) TMI ISSUES ON SEQUOYA!! UNIT NO.1  !

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1. SH I FT IECH ADv i SOR - 1/81_ . . _. . -167--E0r TAiriciEr4T DEntc*TelrPrmETantics - 1/&

2. Ilit4ED. UPGRADE OF SRO & R0 QUAL.-8/80 17. CONTAINMENT ISOLATION DITENDABILITY ,

13. ADD. ACC. MONITORING Ifl3IRUMENTAT1ON -

3. ADrilN. OF IRAINING PROGRAM FOR LICENSINC- EXAMS - 8/30 1/81 wM II. REV. SCOPE & CRITIERA FOR UORMAL 19. ItlADEQUATE

=e + < - ~ 'COF~E-~_ COOLING,' < ^ IUSTRUMENTS-

" 1/81  !

LICENSING EXAMS - 8/80 20. FINAL REC 0ti. OF B&O TASt FORCE

• j 5 REv. SCOPE & CRITERIA FOR SitiUL.

l. 21. UPGRADE EMERGENCY PREPARDilESS E  ?

22. UPGRADE EMERGENCY SUPPORT-FACILITIES .1/81 g 6. - P FOR VEIRIFICATION OF 2fl . COMMUNICATIONS / e is- -

7 < -- g -  %

i CORRECT PERF. OF OP. ACTIVITIES CONTROL R00ti DESIGN REVIEW cS%,'u) 25. IMPL. Or t!RC AND FEfti RcSPort.

7.

- 8o REACTOR COOLANT SYSTEMS VENTS- vi 1/81 26. OFFSITE DOSE MEASUREMENTS

.- n.~.-~- -n w i p('? 27. IN-PLANT RADI ATION MONITORIrlG- 1/81

9. POST-ACCIDENTSAMPLING(. 1/8

28. CONTROL RO0tiIIABITABILITY p. + . 4 [

10. TRAINING FOR MITIGATitlG CORE DAMAGE

• "0" O 29. POWER - ASCENSION IEST' Cll'? AtlALYSIS OF !!YDROGEN CONTROL

12. DEGRADED CORE - RULEMAKING ~

13.. RELIEF AND SAFETY VALVE TEST REO.- 6/81 ( 6 b '

d & 'n h l't . AFW RELIABILITY EVALUATI0tl u <4. O - E J c ep

15. AFW INITIATION AND INDICATION -

1/81 .

v .

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' INCOTLETE MON-TMI ISSUES ON SEQUOY.tl UNIT NO. I 1

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(_1_. SEISMIC AUDIT PER ACRS LETTER 16'et 8 ATWS - REVIEW AND APPROVE PERATING PROCEDURES yrg b 2-3 d j

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- 2. POSITION REculRED REGARDING 0MPLIANCE OF IE BULLETIN 9.

F0uDATION flor:ITORING ON SETTLEMENT 7 -27, LOSS OF UON-CLASS lE

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3. POSITION REQUIRED ON CONTAINMENT p /q

' NSTRUMENTATION 1 CONTROL MA SUMP DEBRIS bd yi:

SYSTEM DURING OPERATION )

ECCS EVALUATION N DEL CONCERNING o "g l .

u. s te t 1. IESEL 6ENERATOR REllABILITY -

PUEL CLAD SNELLING c <~y M' COMPLIANCE NITH R.G. 1.193

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'% 5. POSITION REQUIRED REGARDING PROCESS AND NUREG/CR-0560 m i a ~ l' ~ m 0"%

X CONTROL PROGRAM o-~44# 11 TOPICALREPORTSWCAP-$226, N EQUIP. QUALIriCAT'10NS COMPLY WI H f7g/ .. 9230 AND 9236 3 ELATED TO THE SUIDtILINES OF *1UREG-0538 -f g ,9,2 c,,j ej, MAIN STEAM & FEEDLINE#BREAK ACCIDENTSy 07~ ) W# "h

7. PAD 3-3 PERFORMANCE CODE - COMPLETE EVALUATION REGARDING RESTRICTION IN 12. Q-l!$T COMPLETE REVIEW OF THE USE OF THIS CODE - e A -

• Q-LIST" REQUIREMENTS - e b -

- 13 COMPLIANCE OF OIE BULLETIN l b 80-05 RELATED TO BY-PASS,

> f, 6 / OVERRIDE, RESET EIRCulTS V (f g/ @l c9l-cf n ~n n s.~

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• • • • • June 16, 1980 l

Docket pos. 10-327/328 MEMORANDU". FOR: S. S. Pavlicki Chief l Materials Engineering Branch Division of Engineering FROM: J. Hal apat z Material s Engincering Branch

$UBJECT: EXPRESSION OF D1FFERINO PROFESSIONAL OPINION IN THE MATTER OF THE ADEQUACY OF SENCYAH UNIT ONE 'aTLD DRAWPEAD REPAIR OF PRESSURIEER RELIEF P1 PE The author of this me+erandum, hereinaf ter referred to as the minority, herewith expresses his minority opinion in the matter of the adequacy of the weld drawbead repair of the Sequoyah Unit One pressurizer relief pipe. The minority expresses its dif fering professional opinion in accordanea with Section II. A. 3.J of the memorandum, Samuel J. Chilk to William J. Circks, da ted May 1, 1980, subject, "FY 1982-86 Policy Planning and Program Guidance

( P P FG ) . "

Non Conformance 7ieport NOR Sa'P-79-5-8 disclosed, that during the hot functional testing of Sequoyah Unit One, 1-ROH-93 pipe support f or t h'::

pressurizer relief piping failed to slide in the vertical direction as the pressurizer expanded during heatup of the reactor coolant system. As a result l the 6-inch, schedule 160 (nom. .718 wall), Type 316 stainless steel ..

pressurizer relief pipe was bent. The related safety implicatilon was that f ailure of this piping could lead to an uncontrolled blowdom of the reactor coolant system.

! As corrective actions, T7A had two options. The fir st option wts to cut out the damaged pipe anc replace it. This option, ho.ever, would require a system pressure test in accordance with ('77) ASME Code Section XI IWA-4400(a),

which requires that af ter repairs by welding on the pressure retaining boundary that a system pressure test be performed. The s?cond option wa s to straighten the pipe by a repair procedure which would be eme:pted f rcrn system hydrostatic testing. TVA, to avoid cutting out the damaged pipe, rought this exemption through IWA-4400(b)(3), which exempts f rom hydrostatic testing repairs by welding on the pressure retaining boundary provided that the repairs did not penetrate through the pressure boundary.

The corrective action used by TVA to straighten the pipe vas the s~ald drawbead -

technique. Two 270* grooves were ground in the pipe opposite to and straddling the kink. The grooves were filled with weld metal, reground to -

remove that weld metal, then filled a second time with weld metal. Weld tretal shrinkage provided the stressing to plastically straighten the pipe.

N 1

I S. 5. Pawlichi l l

The repair was acceptad by the Mtterial; Ingineering 3 ranch via the memorandam, Pawlicki to Rabenstein, dat"d Dece-ber 4, 1979, subje:t,

" Tennessee Valley Authority, Sequoyah M iclear Unit No. 1." TVA justifini the exemption from hydrostatic testing of the system af ter the repair on the bisis  ;

of T4A-4400(b)(3), claiming that the process of welding to re ali gn the pi pa did not result in penetration of the reactor coolant boundary. The ninority challenged acceptance of the repair on the basis that more in foiva tilon wa s needed.

The memorandum, Gusta f son to Pawlicki, dated January 25, 1980, subj ec t , " Trip Beport of Visit to Tennessee Valley Authority Sequoyah Nuclear Plant, Unit-1,"

which reported on a visit to the sequoyah site, found the repiir a:ceptabic.

The minority, af ter review of this memorandur and do:umentation related thereto, recommended in the mcmorandum, Hilapatz to Pawlicki, dated February 27, 1980, subject, " Sequoyah Unit One Weld Drawbead Realignment of 6" Pressurizer Relief Pipe," that the Materials Engineering Branch defer acceptance of the repair pending the development and review of additional information. The minority was then advis*d by his assistant director tha t he was to personally examine the wcld mo:kup used to qualify the repair which had )

been mode. The memorandum, Pawlicki to Rabenstein, dated February 28, 1080, subject, " Tennessee Valey Authority, Se quoyah Muelear Pl ant, Unit No. 1, i Realignment of Pressurizer Relief Pipe," then reiterated acceptance of the repair and recommended that the minority meet with TVA personnel and examine metallographic samples. On turch 5 and 6, 1990, the minority visited TVA at j Knoxville and performed a metallurgical examination of the mockup used for the qualification of the weld drawbead realignment of the Sequoyah Unit One pressurizer relief pipe. Metallographie evidence was do:umented whi:h shosed I

that the mockup weld was fully penetrated. Pull penetration of the mockup weld, which was supposed to represent the weld repair of the da aged pressurizer relief pipe, obvously did not de :nstrate co pliance w;.th Section XI IWA-44 0 0 ( b ) ( 3 ) . This finding, in itself, provided cause f or denial of exerption from hydrestatic testing of TVA's wcld drawbead repair of the ,

pressurizer relief pipe which had been made. Other inconsistencies were noted between the mockup and the actual relief pipe. Por example, a different material was used in the mo:kup. Further, while the mo:kup had only one weld groove, the actual relief pipe repair used two weld grooves. In a.M ition, metallographic evidence was documented which sh wed through-wall sensiti ation to a significant degree, indicating that a potential through-sall crack propagation path existed. Since the propagation of cracks through the pipe I

wall is the essential concern with respect to the integrity of the reactor coolant boundary, it is the minority opinion that intergranul ar corrosion tests which would expose to the test environnent specimens which represent the through-wall microstructure should be performed. However, only terts of ID specimen surfaces were performed.

Given that the mockup weld was fully penetrated, the minority concluded that TVA had not qualified its exemption to system hydrostatic testing.

Disclosure of the above in formation' let to a meeting of TVA and NRO on ?brch 13, 1980. It was agreed that TVA would perforn in situ metallography to evaluate sensitization in the actual relief pipe repair and re-radiograph the repair to determine whether or not the pressure boundary had been fully

^ ( i?b

. .._.11- .. - ..-....... .-

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• S. S. Pawlicki l penetrated. The examination, reported in the re . aran.tum Mills to 0 9eilly, l l

l

• dated April 11, 1980, subject, "Scqaoyah Nuclear Plant Unit 1 Tn.ssurizer Relie f Piping Support - N R SWP 79-5 Suppl eacntal In f orm 4 tion" foani the l weld heat af fected zone to be unsensitized and therefore, that sensitized Lose metal underlying the weld did not encroach on the pipe ID. In addition, on the basis of radiographic examination of the repair, it was concluded that the weld did not encroach on the pipe ID, i.e., did not f ully penetrate the reactor coolant pressure boundary. The se results were concurred in by CIE-RII in the memorandum, Murphy to Thornburg, dated April 22, 1980, subject. "RII Feport No 50-327/B0-12 Concerning Inspection Per formed to Evaluate Repair of 9equoyah Unit 1 Pressurizer Felief Line."

The minority considers that meaningful metallurgical conclusions cannot and should not be made f rom Xerox reproductions of the in situ metallography, which have been made ava;lable. Given the carbon content (.052/.059%) of the pressurizer relief pipe, ihe minority finds it anomalous that the weld heat af fected tones did not show some sensiti:stion, since then it is infarred that the ba se metal at any distance f rom the molten weld metal essentially did not experience some time in the 800*T to 1500*r sensitization range during Iest weld cooling.

The matter of the sensitization of austenitic st a inle ss steels is cuvoloped in controversy. ArcJuments are made that the weld drawbead repsir wells are no different than adjoining full penetrated in s'.allation weld s. In the ab sen:e of identical metallurgical histories, ho ever, this argament is tenuous. The minority notes the sa f ety implication involved, viz. , that failure of the repaired piping cannot be isolated, which as a consequence, could lead to an uncontrolled blowdovn of the reactor coolant system. The minority is of the opinion that this matter be examined to a much more definitive and conclusive end. It should also be kept in mind tha t the environment , which will be experienced in service by the repair, will be a calculated 0.2 ppm maxi um oxygen bearing steam rather than reactor coolant water conta ining a residual oxygen concentration during posar operationn of 0.005 pps. SWR pipe crack experience and the lack of corrosion data on the performance of sensitized i austenitic stainless stoel weldments in 0.2 ppm oxygen bearing stcam would suggest caution in acceptance of the Sequoyah weld drawbead repair of the pressurizer relief pipe. The a rg ument that P4R service experience has not identified a problem with pressurizer relief pipes is tenuous, because it is unknown how many, if any, operating plants include pressuriter relief pipes which have been repaired as has Sequoyah's. Given thi s uncerta inty, which the minority feels is related to the in situ metallo^ graphy perf ormed, the more definitive laboratory examination and corrosion testing of boat samples parted f rom the weld drawbead repaired Sequoyah prer,urizer relief pipe is proposed for consideration.

l With respect to the finding tha t the weld repair did not full penetr ate the l reactor coolant boundary, it is the minority opinion that it has not been demon'strated that the radiographic technique used has the capability to develop this conclusion. While evidence that the 2T hole in an AS"M No. 12 penetrareter .was visible to TVA Level III film interpreter s and OIE-RII .

personnel may demonstrate that def ects are not present, these criteria may not necessarily demonstrate the capability of the technique to discriminate in a

S. S. Pawlicki _

ra'iograph between sound .ald natal and sound wrmught base netal un,lerlying l the weld metal. The technique e iat be able to provide for thi s di sti,ction in order to confirm whether or not the weld has fully penetrated the re.etor coolant boundary. The capability of the technique could be confireed or denied by radiographing a known f ully penetrated weld and a known partially ,

penetrated weld in the same material and observing if a di= tinction can be (

made in film density diff erences in the weld root area between weld metal and l wrought base meta.

Given the controversy which sometimes attends the interpretation of

. examination results, inspe: tion by third party is desirabla. Attentien is called to an NRC position stated in the memorandum, Raben st e in to Pa r r i s ,

dated September 12, 1979, subj ect , " Qualification of Inspectors, Inspection l Specialists, and Inspection Agencies for Sequoyah." The Rabenstein memorandum I states the NRO position that TVA institute third part inspe= tion for the Sequoyah nuclear plant. The Rubenstein memorandum is provided as an a tt ac hment to this memorandum. The minority opinion conclu ies that third party inspection is required and should be irple-ented in the estter of the  !

acceptance of the weld drawboad repair of the Sequoyah Unit One pressurizer relief pipe.

n.

J. Iblapa t Ma teria l s Eng inee. - Franch civisien of Engineering Office of Huclear Acactor Tegulation

Enclosure:

As stated cc: V. S. Noonan R. L. Tedesco

. A. S:hwencer C. E. Murphy, OIE-RII l l

A. R. He rd t , CIE-RII R. M. Gamble C. Stahle P. K. Van Doorn, CIE-RII MTE3 Reading Tile m

.d

DISTR: EUTION

.- ggp 2113 9 Docket Files bec: NSIC NRC POR T!C Local PDR' A~RS (16)

TERA R. Mattson L'AR-4 File D. Eisenhut Docket Hos.: 50-327/328 D. Vassallo J. P. Knight

- S. Varga L. Shao F. Williams S. Pawlicki Mr. H. G. Parris -L. Rubenstein V. Noonan

, l'ana cr of Power C. Stahle R. Gatole Tennessee Valley Authority M. Service H. Conrad 5'OA Chestnut Street Towr II ELD - C. Woodhead C. Y. Cheng Chattanooga, Tennessee 37401 IE (3) 5. J. Shatt J. M. Grant M.-1al apa t::

Dear Mr. Parris:

F 8. Litton M. Hum C. D. Sellers M. L. 5:v'e SUSJECT: QUALIFICATICM OF lt:SPECTORS, It'SPECTION SPECIALISTS, Af40 INSPECTI0t AGENCIES FOR SEQUOYAH In Arendment No. 61 to the Sequoyah FSAR, you stated that you will provide your own independent review of the Section XI program of the ASME Boiler and Pressure Vessel Code through the ~VA central office staff in Chattanooga, Tennessee. It is T/A's policy to povide its own inspection services en the basis that TVA is a Federal agency and it is not subject to State or other non-Federal inspectors.

It is our position that T/A is not e.xerpt from any of the require ents of 10 CFR Part 50, Section 50.55a(g)(c). Therefore, w recuire that T/A institute the third party inspection system of the Sequoyah nuclear powr plant.

A letter o'f corpliance is requested. ,

Sincarely, M;;i:21 siped til L. S. Rubenstein, Acting Chief Light Water Reactors Sranch No. 4 Division of Project Management cc: See next page .

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'.., .* SEP 211375 Docket Nos.: 50-327/328 Mr. H. G. Parris Manager of Power Tennessee Valley Authority 500A Chestnut Street Tower !!

Chattanooga, Tennessee 3N01

. 1

Dear Mr. Parris:

I

SUBJECT:

QUALIFICAT10ti 0F INSPECTORS, l'!SPECTION SPECIALISTS, AND INSPECTION AGENCIES FOR SEQUOYAH in Amendment No. 61 to the Sequoyah FSAR, you stated that you will provice your own independent review of the Section XI program of the ASM.E hiler and Pressure Vessel Code through the TVA central office staff in Chattance;a.

Tennessee, i t is TVA's policy to provide its own inspection services on :ne basis that TVA is a Federal agency and it is not subject to State or other non-Federal inspectors.

It is our position that TVA is not exempt from any of the re:;uirements of 10 CFR Part 50, Section 50.55a(g)(a).. Therefore, we require that TVA institute the third party inspection system of the Sequoyah nuclear p:wer plant.

A letter of compliance is requested.

Sincerely,

. - b,*u.$ 1 s. u .

L. S. Fl enstein, Acting Chief Light ..ater Reactors Sranch No. 4 Division of Project Management i l

cc: See next page ,

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Tenr:essee Valley Authority' .

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. lle r t e r S. Sanger, J r. Esq.

General Counsel ~

Tennessee Valley Authority

• 400 Co e erce Avenue T- l E11833 -

I;noxvi11e, Tennessee 37902 Mr. E. G. Beasley Tennessee Valley Authority 400 Co m erce Avenue W10Cl31 C ~

r.noxville, Tennessee 37902  :; -

.s, fir. Michael Harding '

Westinghouse Electric Corporation P. O. Box 3.55 Pittsburgh, Pennsylvania 15230 l

1 Mr. David Lambert ..

1 Tennessee Valley Authority 400 Chestnut Street Tower 11 'd,  !

Chattanooga, Tennessee 37401 6

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APPENDIX VI SEQUOYAH 1 AND 2: OPEN SAFETY ITEMS FULL POWER fiDit-TM1 ISSUES ON SEQUOYAll UNIT NO.1 COMPLETE 8 INCOMPLETE 5 IOTAL 13

. FULL POWER TMl ISSUES Ot1 SEQUOYAH UtilT NO. 1 COMPLETE 15 DATED ITEMS 13 N0tt APPLICABLE ITEMS 1

, RULEM A K [ flG ITEMS 1 1/E FUNCTlotJS 1 NOT COMPLETE 9 IOTAL 40 .

0 jQ-$

/

INCOTLETE NON-TMI ISSUES ON SEQUOYXI UNIT N0. I

1. SEISMIC AUDIT PER ACRS LETTER /"v>~ph /< 8. ATWS - REVIEW AND APPR0vE

2. POSITION REQUIRED REGARD 1NG OPERATING PROCEDURES AM'k FOUDATION MONITORING ON SETTLEMENT 9. COMPLIANCE OF IE BULLETIN

3. POSITION REQUIRED ON CONTAINMENT 79-27, LOSS OF UON-CLASS IE SUMP DEBRIS NSTRUMENTATION 1 CMTROL b SYSTEM 3URING GPERATION

4. ECCS EVALUATION M DEL CONCERNING FUEL CLAD $ WELLING 1. DIESEL GENERATOR RELIABILITY -

COMPLIANCE WITH R.G. 1.193

5. POSITION REQUIRED REGARDING PROCESS k CONTROL PROGRAM AND NUREG/CR-0560

) 11. IOPICAL REPORTS WCAP-3226, hi u m/ >d

5. EQurP. QUALIFICATIONS COMPLY WITH 9230 AND 9236 RELATED TO N THE GUIDELINES OF 'lVREG-0538 b- #_' Malm STEAM & FEEDLINE BREAK

7. PAD 3-3 PERFORMANCE CODE - COMPLETE ACCIDENTS EVALUATION REGARDING RESTRICTION IN

12. 9-llST COMPLETE REVIEW OF THE USE OF THIS CODE 9-llST" REQUIREMENTS J. COMPLIANCE OF ole BULLETIN 'f'CMM4k 80-05 RELATED TO Sv-PASS.

OVERRIDE, RESET CIRCUtTS

_ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ .-_ 't _ r __ _ __ m___m_ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

FULL POWER TMI ISSUES ON SEQUOYAH UNIT NO 1 ITEM TASK NO. ISSUE STATUS 1 1.A.1.1 SHIFT IECH. ADVISOR D.1, - 1/81 2 i.A.2.1 IMMED. UPGRADE OF SRO & R0 QUAL, D.l. - 8/80 3 1.A.2.3 ADMIN. uF IRAINING PROGRAM FOR LICENSING UPERATORS D.1, - 8/80 l

4 1.A.3.1 REV. SCOPE & CRITERIA FOR NORMAL L1 CENSING EXAMS D.I, - 8/80

• 5 1.A.3.1 gXAMSEv. SCOPE & CRITERIA FOR SIMULATOR COMPLETE

'6 I.B.2.2 REACTOR INSPECTOR AT OP. REACTORS COMPLETE 7 I.C.1 SHORTIERMACC. ANALYSIS S PROC. l KEVISION SER - 7/31/80 1 3 1.C 6 PROC. FOR VERIF. OF CORRECT PERF.

OF OP. ACTIVITIES

• 9 1.C 7 NSSS VENDOR REVIEW OF PROC. COMPLETE

• 10 1.C 0 PILOT fiONITORING OF SELECTED EMERG.

PROC. FOR NTOL APP. COMPLETE 11 1.D.1 CONTROL ROOM DESIGN REVIEW

• 12 1.G 1 LOW POWER IESTING IRAINING COMPLETE 13 II.B.1 REACTOR COOLANT _ SYSTEM VENTS -

DESIGN REVIEW 2R - 7/13/80 14 II.B.1 REACTOR C0QLANT SYSTEM VENTS -

COMPL. OF INSTALL. 3.l. - 1/01

• 15 II.B.2 PLANT SHIELDING - DESIGN Review COMPLETE 16 ll.B.2 PLANT SHIELDING - COMP;.ETION OF l NODIF. D.l. - 1/81 j

• 17 II.B.3 POST-ACCIDENT SAMPLING - DESIGN REVIEW COMPLETE l 13 II.B.3 POST-ACCIDENT SAMPLING - COMPL.

OF INSTALL. 3.1 - 1/81 19 11.3.4 TRAINING FOR MrTIGATING CORE DAMAGE SAR - 7/31/Y l SER - S/ I/20

/W'M l

1

FULL' POWER TMl ISSUES

{

ON SEQUOYAH U'llT NO. 1 (CONTINUED)

TASK NO. 133gE STATUS ITE.M ll.B.7 1

• 20 ANALYSIS OF HYDROGEN CONTROL COMPLETE 21 11.B.0 DEGRADED CORE - RULEMAKING 22 II.D.1 RELIEF & SAFETY VALVE IEST REQ. D.I. - 6/81

• 23 II.E.1.1 AFW RELIABILITY EVALUATION COMPLETE 24 II.E.1.2 AFW INITIATION 8 INDICATION D,l. - 1/81

• 25 ll.E.3.1 :MERG. POWER FOR PRESSURIZER lEATERS COMPLETE

• 20 ll.E.4.1 CONTAINMENT DEDICATED PENETRATION N/A

• 27 II.B.4.2 CONTAINMENT ISOLATION DEPENDABILITY COMPLETE 20 II.F.1 ADD, ACC. MONITORING INSTRUMENTATION D.1, - 1/81 29 II.F.2 NSTRUMENTS FOR INADEQUATE CORE

{00 LING D.}, - 1/31

'30 ll.K.3 FINAL RECOM. OF B&O TASK FORCE COMPLETE 31 III A.1.1 UPGRADE EMERG. PREPAREDNESS SAR - 7/18/S0 32 lll.A.1.2 UPGRADE EMERG. SUPPORT FACILITIES D , I , -1/81

'33 Ill.A.3.1 DEFINE NRC EMERGENCY ROLE COMPLETE

• 34 Ill.A.3.3 COMMUNICATIONS COMPLETE

'35 III.B.2 IMPL. OF NRC & FEMA RESPONS, COMPLETE 30 lli.D.1.1 PRIMARY COOLANT SOURCES OUTSIDE SutTs '

CONTAINMENT IESTRkEQ-//I8/S PROC. l 37 Ill.D.2.4 0FFSITE DOSE MEASUREMENTS 33 III.D.3.3 IN-PLANT RADIATION MONITORING D.l. - 1/81 33 Ill.D.3.4 CONTROL ROOM HABITABILITY 40 IV.F.1 POWER-ASCENSION IEST 1/E FUNCTION l .. . . . - .= = = * * *

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I, fECESSaRf C0f0lTIOf6 FOR STRESS CORROSI0f! CRACKING II. EVALlMTION OF REPAIR HELD III. CONCLUSI006 #0 LICB61NG ACTION i

l l

1. fECESSARY CONDITIONS FOR STFESS CORROSION CRACKitE A, STPESS HIGH STFESS NEAR YlELD fECESSARY (PESIDUAL STPESS USLALLY FItMffD B. S961TIZED t%iERIAL C. UtFAVORABLE BWIRCNEIT h~8d

1

11. EVALUATION OF WEB REPAIR A, STRESS AT REPAIR WELD IS ASSLKD NO DIFFEPBE FROM FULL ENEBATION WELDS (HIGH ENOUGH TO BE AN ACTIVE CONTRIBLKOR TO STRESS CORPOSION CPACKING)

B. ALL WELDS IN STAINLESS STEEL PIPING (INCLUDItE PEPAIR) AT SEQUOYAH ARE SENSITIZED TO SCTE DEGPEE REPAIR WELD CCMPLETED USING SVE EASIC PROCEDURES USED TO f%KE FULL PBEIPAT10f1 WELDS REPAIR WELD IS WITHIN SNE POPULATION AS FULL P9ETPAT10N WELDS C. BNIR00fBH SEiNICE EXFERIENCE INDICATES NO CPACKING HAS OCCURPED IN NR PRESSURIZER LINE WELDS FWilFACTURED TO SIMILAR PROCEDUPES 4- s7

, aw i

III. CONCUJSI0fE AND LICBEING ACT10t6 l l

PBAIR WELD FABRICATED USING SAT PASIC PROEDUPES ALLOWED FOR .

l FULL PBETPATION WELDS WELD MY BE SENSITIZED AfD IS INCUJDED IN S#E POPULATION WITH FULL PBETPATION WELDS SERVIE EXPERIENE INDICATES THAT SSEITIZED FULL PBEl?ATION WELDS If1 PWR PRESSURIZER LITES DO NOT MVE HIGH POTBfflAL FOR CPACKING to DEFECTS MVE BEB1 FOUfD IN PFAIR WELD IffEGRITY OF PEPAIR VELD IS AT LEAST EQUAL TO FULL PSETPATION WP.as REPAIR WELD INCLUDED lil All AU6EfiE) ItEERVIE ItGECTI0fl PROGP#1 (INCLUDING THIRD PART( ItEPECT10fD l

1 I

APPENDIX VIII

,' SEQUOYAH 1 AND 2: UNDER-CLAD CRACKING OF RPy N0ZZLES REACTOR VESSEL N0ZZLE UNDERCLAD CRACKING I

l BACKGROUND WESTINGHOUSE FRENCH LICENSEE DETECTED CRACKING:

IN BASE MATERIAL OF REACTOR VESSEL N0 m "S IN BROAD AREA 0F N0m c BORE - MORE FREVALENT IN THICKER SECTION ,

CONFINEDTOHAZOFSECONDLAYEROFCLADDING ORIENTED PERPENDICULAR TO CLADDING DIRECTION

- 1.0INCHINLENGTH,0.28INCHINDEFTH BY DESTRUCTIVE AND NON-DESTRUCTIVE OJD EX#41 NATIONS s

CRACKING BELIEVED TO BE:

HYDROGEN-INDUCED RESULT OF WE!. DING PROCESS / HEAT TREAT 74ENT USED IN CUDDING I

'[

l l

l EUROPEANS HAVE INSPECTED = 80 N0ZZLES

- MOST INSPECTIONS IN THE SHOP

- NO FIELD INSPECTIONS OF OPERATING PLANTS l

W HAS INSPECTED c 3S N0ZZLES l

- MOST INSPECTION IN THE FIELD

- INSPECTIONS OF OPERATING PLANT SCHEDULED FOR 1980

- SEQUOYAH INSPECTION CONSTITUTES A BASE LINE ,

AND WILL BE REPEATED 1

CHRONOLOGY OF EVENTS RELATED TO' REACTOR VESSEL N0ZZLE UNDERCLAD CRACKXNG Early Octobe' r 1979

- NRC.'and 'lorthern States Power Company (NSPCo) advised of cracking found by French licensee and that Prairie. Island Units 1 and 2 (operating plants) have French-manufactured reactor vessels October 26, 19,79

- W/N5?Ce Meeting November 26', 1979

- NRC/W/NSPCo Meeting

- -W presented status of ongoing efforts:

e survey of vessel manufacturers e examination of French-manufactured nozzles / blat samples e Prairie Island fracture mechanics analyses e ' development of UT technique

- NSPCo comitted to do 70* UT ISI of nozzles:

e Unit 1 - July 1980 outage o Unit 2 - February 1981 outage

- NRC saw no ir:enediate concern related to continued operation of Prairie Island Units and concluded that W proceeding in an appropriate manner Decemoer 13, 1979

- W transmitted letter to NRC:

e documenting infonnation presented at November 26 meeting e indicating that Rotterdam-manufactured vessels (Sequoyan Unit 1, Watts Bar Units 1 and 2,'McGuire Unit 2, Catawba Unit 1) under investigation and that cladding processes / heat treatment used by CE, B&W, CS&I should preclude

.c'acking Late December 1979

- All customers advised of survey results/W efforts

- Decision made to inspect Watts Bar Unit 2 Early January 1980

- Watts Bar Unit 2 nozzics inspected

{

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l

~ January 31, 1980-

- 'W transmitted letter to NRC s' documenting results of Watts Bar Unit 2 inspection - no underclad cracking Early February 1980

- Decision made to inspect one Sequoyah Unit'l nozzle

- Sequoyah Unit i nozzle. inspected - reheat cracking found February 22, 1980

- NRC/W/TVA Meeting

- Results of Watts Bar Unit 2' nozzles end Sequoyah Unit i nozzle inspections presented

- NRC required inspection of other Sequcyah Unit i nozzles

- NRC stated that all Rotterdam-manufactured nozzles should be inspected

- NRC concern related to satisfying ASME Code Section XI acceptance criteria Late February 1980

- Other Sequoyah Unit i nozzles inspected - underclad cracking found

- Acceptability of all indications in terms of Section XI criteria demonstrated

- NRC granted Sequoyah Unit 1 5% Operating License (February 28,1980) g March 1980

- NRC requested detailed infomation about cladding process / heat treatment used in fabrication of North Anna Unit 2 nozzles in order to perform an independent evaluation (NOTE: North Anna Unit 2 venel manufactured by Rotterdam, nozzles clad by Sulzer.)

- Virginia Electric and Power Company comitted to inspect North Anna Unit 2

- NRC inquired about condition of Salem Unit 2 nozzles (NOTE: Salem Unit 2 vessel manufactured by CE.)

- Public Service Electric & Gas Company comitted to inspect Salem Unit 2

~b

SEQUOYAH I AND RE IABILITY STUDIES l

TVA RELIABILITY STUDIES A) SYSTEMS INTERACTION METHODOLOGY APPLICATIONS PROGRAM B) REACTOR SAFETY STUDY METHODOLOGY APPLICATIONS PROGRAM C) AUXILIARY FEEDWATER SYSTEM RELIABILITY EVALUATION D) PLANT FULL SCALE SAFETY AND AVAILABILITY ANALYSIS

~

g-c.3 _

e SYSTEMS INTERACTION METHODOLOGY APPLICATIONS PROGRAM

SUMMARY

AN OBJECTIVE WAS TO DEVELOP A METHODOLOGY INDEPENDENT OF THE STANDARD REVIEW PLAN (SRP)

FOR IDENTIFYING AND EVALUATING SYSTEMS INTER-ACTIONS IN LIGHT WATER REACTOR COMMERCIAL POWER PLANTS WATTS BAR NUCLEAR PLANT (WBNP) WAS CHOSEN AS THE EXEMPLARY FACILITY FOR DEMONSTRATING THE METHODOLOGY ALTHOUGH IT WAS NOT THE PURPOSE OF THIS STUDY TO JUDGE WBNP,IT WAS CONCLUDED THAT THE FACILITY IS GENERALLY WELL PROTECTED AGAINST INTERACTIONS CONSIDERED WITHIN THE SCOPE OF THIS STUDY

SYSTEMS INTERACTION METHODOLOGY APPLICATIONS PROGRAM OVERVIEW .

OBJECTIVE DEMONSTRATION OF METHODOLOGY .

METHOD IDENTIFICATION OF COMMONALITIES EXISTING AT WBNP THROUGH EXAMINATION OF FAULT TREES DETERMINATION OF POTENTIALLY INTERACTIVE CUT SETS WITH 3 OR LESS INDEPENDENT FAILURES REVIEW AND ASSESSMENT OF POTENTIAL INTERACTIONS LIMITATIONS RCPB MlTIGATING SYSTEMS WERE NOT MODELED FAULT TREES WERE DEVELOPED FOR ANSI N18.2 CONDITION I AND 11 OCCURRENCES ONLY FUNCTIONS RELATING TO THE CONSEQUENCES OF RELEASE OF RADIOACTIVITY WERE NOT MODELED FIRE, EARTHQUAKE, HURRICANES, TORNADOES, FLOOD, SABOTAGE EXCLUDED

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m SPECIFIC ANALYSIS DATA OllTAINEI) Oil ALL COMPONENTS \;

WillCil APPEAR IN CUT SETS, LINKlHG_CilARACIf.RISI1CS

. AC POWER - TRAINS A AllD B

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e DC POWER - TRAINS A AND B g e ACTUATION - INPUTS AND OUTPUTS TO AUI0MATIC CONTROL CIRCulTS I

n . LUBRICATION - INTERNAL AND EXTERNAL D e COOLING .

~

• llYDRAULIC e COMPRESSED AIR

-

• LOCAT10ll - ROOMS, PIPE CilASES, GENERAL AREAS e

1 l

i I l l REACTOR SAFETY STUDY METHOD > OLOGY APPLICATIONS PROGRAM OVERVIEW  ;

i l

OBJECTIVE DETEP.MINATION OF DOMINANT ACCIDENT SEQUENCES l

METHOD SYSTEM EVENT TREES CONSTRUCTED FOR WASH-1400 INITIATING EVENTS SIMPLIFIED FAULT TREES DEVELOPED FOR MITIGATING SYSTEMS ,

RESULTS ICE CONDENSER PLANTS HAVE DIFFERENT DOMINANT ACCIDENT SEQUENCES RISK IS SIMIL/.R TO LARGER DRY CONTAINMENT PLANTS

i

'l.

J t I

SEQUOYAH AUXILIARY FEEDWATER SYSTEM 3 RELIABILITY EVALUATION

SUMMARY

Xaman Sciences Corporation was contracted by the Tennessee Valley Authority to conduct a reliability evaluation of the Sequoyah Unit #1 Nuclear Power Plant Auxiliary Feedwater System (AFS). Kaman employed the G0 canputerized event tree methodology to perform the analyses.

• Results indicate that the probability of successfully starting the auxiliary feedwater system upon demand and providing adequate water flow and pressure to at least two out of four steam generators' is 0.99999 whert the initiatirg event is both feedwater pumps tripped. In event of loss of offsite power (blackout) with diesel generators and battery back-up available the AFS start-up success probability is 0.99997. Other

~

excursions were also evaluated.

The analysis revealed that there are no first order faults in the Sequoyah AFS for the initiating event both feedwater pumps tripped. A total of 116 second order faults were identified for this case. The largest contribution of unavailability resulting from a pair of faults is 10~7 Most second order fault sets contribute to start-up unavailability on the order of 10-10, mW n

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SEQUOYAH NUCLEAR PLANT FULL SCALE SAFETY AND A VAILABILITY ANALYSIS l

l OBJECTIVE: To develop two plant models, one to assess plant safety and one to evaluate plant availability 1

I METHOD GO methodology developed by Kaman Sciences Corporation with funding from EPRI 1

MANPOWER KSC = 80 man-months .

TVA = 30 man-moi 5ths SCHEDULE Phase 1 July 1,19.80 - Dec. 31,1980 Phase 2 Jan.1,1981 - Dec. 31,1981 SCOPE Phase 1 Simplified plant model Detailed plant models of selected systems (Electrical Power, Central Air, Reactor Protection, Safety injection, Main Steam, Main Feedwater)  ;

Preliminary safety and availability assessments i Phase 2 Expansion of simplified model Data collection Final safety and availability assessments incorporation of operator, test, and mainte-nance acti6ns l Determination of critical components Investigation of abnormal scenarios

& 21

i p g s i Ano NoxocenconTaot

SUMMARY

- HYDROGEN STUDIED ABOUT NINE MONTHS  ;

~

- SEQUQYAH CAN WITHSTAND SUSSTANTIAL AMOUNTS OF HYDROGEN ABOVE DESIGN BASIS , )

1

- SIGNIFICANT MODIFICATIONS HAVE BEEN OR ARE BEING INCLUDED TO REDUCE POTENTIAL FOR DEGRADING EVENTS

- LIMITED RISK ASSESSMENT SHOWS SEQUOYAH COMPAR-

ABLE TO THE WASH 1400 STUDY REFERENCE PLANT

-- PROPOSED CONCEPTS FOR RESOLUTION OF HYDROGEN

. ISSUE EVALUATED .

-INTERIM DISTRIBUTED 1GNITION SYSTEM CHOSEN FOR IMPLEMENTATION AT SEQUOYAH. DEVELOPMENT WORK

.ON CONTROLLED IGNITION IS PROCEEDING FOR FINAL IMPLEMENTATION AT SEQUOYAH. H ALO N SUPPRESSION IS ALSO BEING STUDIED.

/}-72

sose 7/ '

l CAPABILITY OF THE SEQUOYAH CONTAINMENT l

- MINIMUM CONTAINMENT PRESSURE CAPABILITY l Dis tGd- / 2. PS t G YlELD - 33 PSIG .

ULTIMATE - 42.5 PSIG

- VOLUME - 1.2 X 106 FT3 l

- CONTAINMENT CAPABILITY TO WITHSTAND HYDROGEN  :

COMBUSTION ASSUMPTIONS:

- BURN IS INSTANTANEOUS AND COMPLETE l

- BURN IS ADI ABATIC

- NO RADIATIVE TRANSFER RESULT:

- SEQUOYAH CAN WITHSTAND A HYDROGEN BURN EQUIVALENT TO APPROXIMATELY 25 PERCENT l METAL-WATER REACTION (USING ULTIMATE l STRENGTH OF MATERIALS)

$- )3 .

s suoc .

t t.

CONCEPTS STUDIED FOR MITIGATION, CONTROL, OR PREVENTION OF CONSEQUENCES FROM HYDROGEN

- MITIGATE THE CONSEQUENCES OF HYDROGEN BURNING .

t VENTED CONTAINMENT: .

1. F

.ILTERED .

2. ADDITIONAL  !

. I

3. COUPLED  !

- CONTROL COMBUSTION i

CONTROLLED (GNITION SOURCES  !

. l

- PREVENT COMBUSTION f

1. INERT CONTAINMENT,WITH N!TROGEN ~

j l

2.

. SUPPRESS COMDUSTION W!TH HALON

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CONCEPTS - ASSESSMENT

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VENTED CONTAINMENT FILTERED s i

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i. tvu i certu i s v c run n/ i L'

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2. ESTIMATED DOSE IN LOW POPULATION ZONE iS (N l EXCESS OF 900 REM -

!' l i;

3. SCME ESSENTIAL FEAT'JRES NOT DEMONSTRATED

'. l 1 i i;

4. POTENTIAL FOR UNNECESSARY BYPASS OF h

CONTA!NMENT . 2

't

5. HIGH INITIAL COST, MODERATE O/M COST '. j j.

ADD!TIONAL CONTAINMENT

1. NOT EFFECTIVE FOR RA?iD PRESSURE TRANSIENTS 1

'd

2. MINIM! ZED RADIATION RELEASE TO THE PUBLIC k 0

(VESSEL LEAKA'GE ONLY)

D

3. VERY HIGH INITI A L COST, LOW O/M COST ij, -

. l-II c

~~ fb - - - . _ _ _ - -

. . _ _ _ _ . . . . . ____..____._____.__....._......_..._._~.-..-

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CONCEPTS - ASSESSMENT (CONT.)

COUPLED CONTAINMENT L NOT EFFECTIVE FOR RAPID PRESSURE TRANSIENTS 4

2. PCTENTI AL FOR DEGRADING SAFETY OF SECOND UNIT

3. LARGE OPERATIONAL PENALTY FOR SECOND UNIT

4. MINIMIZED RADIATION RELEASE TO THE PUBLIC

=e

u < ae g CONCEPTS - ASSESSMENT (CONT.)

- CONTROL COMBUSTION .

IGNITION SOURCES .>

1.HIGH POTENTI AL FOR EFFECTIVENESS DURING MOST ACCIDENTS LE ADING TO CLAD OXIDATION

2. NO EFFECT ON PLANT OPERATION .

3. TECHNICAL DEVELOPMENT REQUIRED

4. REQU RE LOCAL' HYDROGEN MONI.TORING

5. MODERATE INITIAL COST, LOW O/M COST 8 6 e

e i

. It.

. suas 7 i l

. . i CONCEPTS - A3SESSMENT (CONT.) .

l

' I

- CONCEPTS WHICH PREVENT COMSUSTION I  :

, l N!TROGEN !NERT!NG l

1. EFFECTIVE IN PREVENTING l-!YDROGEN COMEUST!ON , l i: 1

': l

2. LARGELY A PASS!VE SYSTEM .

e

. n  !

3. DiFP! CULT BACKFiT TO ICE CONDENSER i, 1

4. OPERATIONALLY PROHIBITIVE BECAUSE OF I FREQUENT MAINTENANCE NEEDED ON ICE P l l

CONDENSER AND OTHER CONTAINMENT SYSTEMS -

b

5. SIGNIFICANT POTENT!AL POR DEGRADED SAFETY (

THROUGH REDUCED MAINTENANCE OF EQUlPMENT [

6. INCREASED LOSS OF ICE

7. HIGH INITIAL COST, EXTREMELY HIGH O/M COST gene ,gmag eo g m 9 ee p penEUS$ m 41an Mg gghd dgg*

N # N SOSW W' N p .g

IH suser 1

CONCEPTS - ASSESSMENT (CONT.)

-IALCN SUPPRESSANT g n.um^t p , ,. i_ i n _ L v. c,_ tc , "ts "r." /ncaI Nm .C .D. m V _t P ^.l' l i'mio HYDROGEN COMSUSTION .

. NO OPERATIONAL Ei~FECTS WITH NORMAL  :

PRECAUTIONS

. UiODERATE HAZARD TO PERSONNEL, i

. TECHNICAL FEAS!SILITY NOT DEMONSTRATED -

i i

i. DECOMPOSITION PRODUCTS MAY PRODUCE SEVERE j CONSEQUENCES  !

l

)

i

!. ACTIVE POST ACCIDENT WITH S!iORT BUT REASONABL!: .

TIME TO MANUALLY ACTIVATE i i

I

. HIGH INITIAL COST, LOW O/M COST I

f 9

9-77

" " ^ " ' " "

sa.

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RESULTS AND CONCLUSIONS 1

- MOST FROMIS!NG CONCEPTS FOR HYDROGEN CCNTROL i er:_ Lmc C r. r:

m Jvr u n ,a, ,a...o r Gnon o . u :as D c u.e ca v . :m v . ..n.mc.. i r ri c ,o civv c .ou-o u n,n i ,n o c .. i o_ .

1. IGNITION SOURCES -

2. HALON SUPPRESS ON .

- S!GN!FICANT IMPROVEMENT [N PHYS! CAL MODELS AND COMPUTER CODES ARE NEEDED

- FILTERED. VENTED CONTAINMENT IS UNACCEPTABLE .

FROM RELEASED DOSE

- !NERTING IS NOT FEASIBLE FOR AN ICE CONDENSER CONTAINMENT

- R!SK AT SEQUOYAH COMPARASLE TO WASH 1400 REFERENCE PLANT 1

l l

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PROGRAM FOR DEALING WITil DEGRADED CORE CONDITIONS 3

IIE IIAVE ORGANIZED AN EIGHT-MAN FULL TIME TASK FORCE FOR.DESI6N AND DEVELOPME DEGRADED CORE ACCIDENTS.

4 WE ARE IMPLEMENTING IMMEDIATELY Tile DESIGN AND INSTALLATION OF AN INTERIM DISTRIi IGNITION SYSTEM (PHASE 1) TO BE OPERATIONAL WITilIN TWO TO TilREE MONTHS. .

WE ARE IMPLEMENTING IMMEDIATELY DEVELOPMENT WORK TO UPGRADE Tile INTERIM DISTRI IGNITION SYSTEM (PilASE 2) AS IMPROVED ASPECTS OF Tile SYSTEM CAN BE DEVELOPED. '

WE WILL COMPLETE A LONG-TERM STUDY AND DEVELOPMENT EFFECT FOR CONTROLLED IGN  ?

SYSTEMS WilICll WILL LEAD TO BACKFITTING Tile PilASE 1 & 2 SYSTEMS, IF NEEDED. (PilASE 3)

, Tile LENGTil 0F Tile STUDY Sil00LD BE APPR0XIMATELY Til0 YEARS.

WE ARE IMPLEMENTING IMMEDIATELY A DEVELOPMENT EFFORT TO UNDERSTAND T ASPECTS OF llALON AS A IlYDROGEN BURN SUPPRESSION.

l

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Ma.ior Ta sk s c

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1.

Controlled Ignition .

2.- halon -

. p . / .

.' 3 Risk Assessment -

,- . .. .- . ~

b 4 . Core Behavior, Hydrogen Generation and "'ransport

. . , . o. . . : .- . , .

.. 5. Hydrocen Eurning and contaiment Responses

^

6.~ Containnent Integrity c, . , e-. . .

"- 7 Equipment Environmental Qualifications

. 8. had.iation Dose Code '

9. ' Hydride converter, Fogging, and others .

~

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%. . 10.' P. ale nakinG .-arad State cf the Art -

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APPENDIX XI 5EQUOYAH 1 & 2: NRC STAFF PLANNED REV OF PROPOSED HYDROGEN MITIGATION SYSTEMS l

RAf1ED STAFF EVIEW FOR TE pig 0fD DISTRIBUTED 161ITIG1 SEEM R E EE: EVALUATE POSITIW MD IEATIVE t.ttCTS TO ASSUE SAFETY MARGIfE AE ItHOVED.

SCDEE: 1. EXERIENTAL TASiG 1.1 ASSESS ELIABILITY T 191ITIm IVIES 1.2 ASSESS CAPABILITY CF It6TRMES TO EASUE hYDF0GS1 COGEPATIG6 1.3 EVALUATE 0780STIG1 PROESS Ill VARIOUS TEM MIXTWES

2. #RYTICAL TAS'S 2.1 EWLOP 191ITIW STRATEGIES FOR VARIOUS ACCIIENT S034ARIOS 2.2 ASSESS entCTIVE?ESS T TE Pfusiu SW EM l SGEDULE: THEE fUiTHS

~

USER'S EQUEST FDRA SAFETY ESEARW FPDGPM PUfP H : EVALUATE SYST95 FOR MITIMTIm T EGPAED COE/CDE ELT ACCIENTS OM)ROGEN QMEU l l

SCTE: EM10P INFDPMTIm m MITIGATIm SYSIDE FOR ALL LWR CGTAltifNTS FOR USE IN UPCTIfE RJLEVKIfiG PROCEDIf4G.

SCEDULE: SHORT TEIN PPASE: 6-12 K N ES LGE TEIM fHASE: 24 KHTHS 1

l

l

! l l

i l

SNORT TEFM R4ASE SCCEE EVAUJATE SM POR MITIGATIG1 CF EGPAID COE ACCIIETTS Ill IE CEER #0 MMK 111 C3ffAlffEffS l IaSsi

1. EVAUJATE h%ROGEN 8EPATlai PATES FOR A PNE CF E-GPAID COE ACCIIBlT SCSIARIOS.

2. ETE!NIfE WAlffETT PESSUE #0 TBfERAlljE ESPG4SE WITH #0 WITN MITIGAT101,

3. EVAUJATE HYDRXEl CNROL SM.

SOEULE IE CEER MAltt9fiS: 6 - 9 MHS MMK lll/BWR NAl?tBITS: 12fUlTHS 1

l I

l h' ?%~

l l

I 1

L0lG TEFM PHAE SCE EVALUATE SYST96 FOR MITIGATIQ17 EGPKED COE/CDE ELT ACCIIBfTS IN AU. LWR CChTAlttEITS.

FEATUES TO E EVALL%TFD 1

'1. WING SYSIDE 1.1 FILTEED-M'NTED CCNTAltfETTS 1.2 CCCPLit!G CF TWD-lHIT SITES

2. HYDPRI C0fTf0L SYSTHE 2.1 SlFFESS4fTS/SCAVEEFS 2.1.1 HALIN SYSIB E 2.1.2 ItETINGSYSTES 2.1.3 WATER FOG SYSTEMS 2.1.4 CATALYTIC ECatIERS (l.AED 2.L5 OXYEN SCAVBEIfE SYSTDE 2.2 C0fff0U.ED BJfNITE 2.2.1 DISTRIBJTED 19tITION KURES 2.2.2 TEit%L ECOSINERS (LARE) I sa m te l

PREPM IS DFB,ltu TO EXTEND DE A TWYEAR ERIOD.  ;

• l

{

D . .-. - -- -

PAGES A87 thru 163 has been deleted as:

p 6 tf u r u ci i Mushkbh le #

)

l l

1 l

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[pMaug*'s UNITE ) sTATfis

.?' ,.

r NUCLEAR REGULATORY COMMISSION 5s 5 ADVtsORY COMMITTEE ON REACTOR SAFEGUARDS W ASHINGT ON. D. C. 20555 k h.;

p .m. .

[J f

( ,' July 2, 1980 i APPENDIX XVI I

TM1-2: BACKGROUND MATERIAL FOR VENTIitG 0F CONTAINMENT MEMORANDUM FOR: ACRS Members j

I j FROM: E

SUBJECT:

TMI-2 VENTING OF CONTAINMENT ATMOSPHERE TO RELEASE Xr-l Attached are four PN0s which describe the reactor building The reactor purgeisstatus.

building being j Purging began on Saturday, June 28, 1980. -

purged through the Modified Hydrogen Control System (MHC), a filtered system The rate with a variable speed fan having a maximum capacity of 1000 cfm.

of purging can be varied as meteorological conditions pemit.

At four minutes into the purge, the system was shut down due to high radiation alarms from the particulate radiation monitor on the plant ventilation stack.

Analysis indicated that the particulate detectors were responding to the Kr-85 concentration in the system. The monitor was modified to eliminate the inter-

[- ference due to the Kr-85. Purging was resumed later in the day under ditions -100 cfm to further evaluate the system until 10:00 p.m.

test con-Purging re-sumed on June 29th with the system and associated monitor response as expected.

Additional instrumentation was installed in an attempt to increase the sensi-Since 2:00 p.m. on June 29, tivity of the stack monitor for particulates.

1980, the system has been operating as meteorological conditions permit with momentary shutdown to change filters and halts for miscellanous maintenance.

Additional details are given in the attached PN0s. ,

1 L' I Attachments:

1. PNG-TMI-80-37

2. PNO-TMI-80-38

3. PNO-TMI-80-38A

4. PNO-TMI-80-383

5. Other PN0s as available before folder deadline v

A-/6

> - ~*

-!: ATE: June 20, l'JL PRELIM!::AkY -fiO1111CA110n Of f Vilf! OR u::uf.uhl OCCt!MCFCf--PfiO-TM. h0-3/

events of POE!!ctf safety or

%is preliminary not ification f.un.,ti tutes L ARl Y notir.e of c Ec i nter est

~

~~

i UyTii s unrf"~ ThP i n f ortr.a tion pr,ese geo i s 5's nit 1sily,r6ceivec (i)thou, I

< 1 f i ca t i on or eva l ua ti on b od ' i s ba s i ca l l y a l l ,tna t i s J_.r,iown _b,y,, NR,C s t.a f t usi fli),.5 d o t t.,

Facility: Hetropolitan Edison Company

~

i

~~ Three Mile Island, Unit 2 Middletown. Pennsylvania Docket Number 50-320

Subject:

REACTOR BUILDING PURGL STATUS .

At appruximately 8:00 a.m. un Jann 25, 1980, purging of the IMI-2 reactor buf1 ding )

at.ms;:here commenced with an initial flow rate of apuruximai,wly 100 clm (.0b Ci/sec in the Modified Hydrogen Control System (MHC). Approximately 4 rainutes inLu the purging the system was shutdown due to particulate radiation nonitor high alanns us, the MHC system (HP-R-223) and the plant ventilat on stack (llP-Rul9). Subsequent i

licenscc analysis ut the monitor sample system filters revealed no particulate er.Livity.

It wes then concluded that the particulate celecturn were respdndino to the nsbl#. gas

-(Kr-Eb) concentraL ion in the sy> Lem. iiPA cnd NRC indepernlent ortAlyhs rusf firmes this cunclusion. Uutween b:00 p.m. sud 10:00 p.m. on June 20, 1930, the licemsev resuined the purgu under test conditiur.3 tu furuier evalu6te systez and ass cialud monitor responsu with a very slow approach fri achieving various I4i? syste flow rates. Prior to this test, Lhe

• 19 / monitur was inuu si led to clin.irm the inter ference due to noble gas. . During the test, additional tiller w&les we ; Laken with subsequent analysis conducted to rcaffirm that no particulate activity conenntrations

-tre present. All analyses by the NRC were in agreement with tne licensce's results.

a result of this lusting, the litanswa ir, c.nnsidering malhuds Lu increase 1.hu senr.;ivity

.Lhc slack monllor systwm tur particulatus with Lhu installation of additional particulate assur11tur ing equ iptr.ent . Resumption ut testing started at 2:00 p.m. on June 29, 1980, to furthur evaluata system and associated inunitur response.

Additional reports on reactor building purging will be updated on a daily cas ts during the initial stages of purging. Thc'NRC Region I mobile laboratory will continue to cm used to verify the licenswe's analytical results.

Mxdia interest has ucturrad because at public sensil,ivity to this evolution and TMI related events. The Commonwealth of Pennsylvania has bean informed. NRC has responded to inquiries. The NRC/TMI Program Office Staff is tr.anitoring events as they occur un a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> basis.

This preliminary notification is issued for information only, and tne information is current as of 2-00 p.m. '-)s , , ', .

. .5:, y :, L.c-d L.s : ~f.Vrl%

Contact:

n. J. Conte 590-3950 - A. Fa.,ano b90-?9b0 .J. I. Collins :90/Jn5 p- C/R DIST. CHM,CMRS.PE.GC.CA.

/  ; i) F,.~.E Ed[Y. .. SECY. HIST. REC 005.

b 02 Niw .

Uni 3 3 '-l y PRELIMINARY NOTITICATION

. . A-/L5" \

l i . .-

. l

\

, ,..a......,.u ~ . . . . - . . . '

i OAT 7.: June 30, 1930 1

t:*r.nti::rJ:r I:0T ir l cA ll C:: Of CVLf4T OR U::U5UAL OCCURin.yCL --Pl;0-TM1 38 Thi s r,rel imina ry notification constit.utmS 1 ARLY. notice of events of POSSISt.C safety or p.;:d Te interest sicnificance. G ht' intermation uresanteo 157Ts. ).n111,a l l y recc1 ved wi tt:c w-i{ci t icri,,c].~al ua t wn g od is Jfe.s 1.ca l_)y a ll tha t} i s .xnown bv 'tlRC :sta f f on tnis date facili}y: He tropu l i t an E d i son ,Co npany Three Mile Isldnd', Unit 2.

Miceletown, Penrisylvania Docket Number.50-320 subject: REACTOR $UILDI.'.G PURGE STATUS l

F.cs., eticn of the reactor building atmosphere purge started at 2:00 p.m. on l Jane 29,193C, with system and associatcc :nonitor r.esponse as e.x pec t ed . Meanwhile, in I an atice;; to increase tne sensitivity of the stack monitor (?l9) for carticulates, c.ic ! t ict.a l instr tentatior .was installed and was made operat iut41 for preliminary.dai:a '

eval atier.. .

Since 2:00 :.m. June 29,.19S0, t.he sys tem was cperated as .:neteorological conditions

,.

r i r t ed . There was one system Shutdown between appr oximately 2:00 a.m. to ?.:00 a.m.

on June .lC,1980, for ressens other than meteorological. :The MHC exhaust f an motor tripped on overload..

Tr.tal calculatec ct, ries reicased as of 12:00 noon was ;2,092. Remaining curies in.the cacter builcing cased ces the last building sample was analy::ed at .95 Ji/cc

.53,SM Ci, tuta l) .

4:diticaal reports on reactar building purging will be updated on :a daily basis curing tra initial stages of purging.' The.NF.C Region 1. mobile laboratory will cantinue to be usec to verify ne licensee's analytical results.

Wi?.. interest has uccurred because of public sensitivity to this evolution and TMI reined events; The Co:rinor wealth of Pennsylvania has monitored these events. NRC nas re sper.Jed to incu tries.

The tiRC/TMI Program Office Staff nas monitored events as tthey occur en a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> casis.

Inis ir:fonnation is current as of 12:00 noon.

-s - R y . ;..cL;,, .

R . J . C on te ',&

Mi.[:cL ",W

/WWs 4 /

a. p

."en t ac t : 590-3950

'A. F ssano 590-3950 lJ,.,T. Collins-590-3955 jvr -

d c, C/R DIST. CHM.CMRS.PE.GC CA-SE;y . HIST. RE COCS.

-1 i...,,

A . .n &

' FELiti1 NARY N011f1CNiION Of FVFN1 0F UNU5UAL CONCURTsCHCL--PNO-ir4i 2CA Thiz acciiminarv nu'.ifica t ion cons ti tutes. i. AP.! Y not ice ut en L3 ut r'estT.;LL tafe.v or l

~

7 a<. - 1 ni t : a l i v et.ce. i vec ,1.nvut

$ 1.c. i~nt. e r f.s. .t5 m n.i.f'.i.ca n c e .

Ih'c inform :1 on .

u'rvsenten i t i i f103 ti on u r ev e . ac t i on ann 1.s._. :n_s_i c.: : l v e t t ine.1 is known uv M t., stail on t 11 do;c

~

Locilitv: Metropolitan Edison Company l Threv Mile. Tsinnc. Unit 2  !

Middletown, Pennsylverija Docket Numt>,-- 50-320 (

Sabject: REAC10R BUILDING PURCF STATUS

~

1 Tur'ging of tna reactor building atmosphere utiliring the audi fied hycengen ::ur:Lrul U.UIC:

-tystem continued al var'ious LyStem flew rater. basec uti t'icL00rclogical concition. Fi/4c6tB N shutdown:'also occurren due tn stack moniLur filter changeouts.

"exir::um purge syLLem flow rate was approximately i n et:n wRile ::ack flow rate average a: proximately 100.000 cfm in the past 74 hours8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br />. Tota l ca lcul4 Leit raainac ivity relenco u of 7:00 :.m. was 5t.09 ti based on stack flew rc.tv erid manual concentrution. P.>.ma i ni n g

. concentration in the rucctor building b4?ed on the last bui! ding dample was analy M cl-1.01 uCi/cc (G7,202 Ci . LoLal).

Fvcluat. ion of the additional particulate munit.ur on the plant e. tack continues.

~

GrbD taraule, continue to vurity lil. tic or ne particulate. activity rulvened.

Additfonal repor s on reactor building purging will be updated un e J.aily batit during th:;

initial Sla9c5 of'Durging. The NRC Reginn ] tuubile laboratory wil! continue to be i.ter.

Lo verify the licensee's analytical results, dia intervst has occurred because of punlic sensitivity to thic. evolutiot; and IMI relateJ 7venn . The Corranonwwullh of Pennsylvania nas monitored thesc eventt, MC nas responded in inquiriec. The NRC TM1 Program Of fice staff hat monitnrec events as they occur un r 24 hu b!$it.

Tnis infopmation 1* [is&current as of_}.7~

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SECY. HIST.RECCDS.

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July 2, 1980

.W.Jti;&.Y ltil'iUil j ott Of TVENT OF UNU5UAL CONCURP.ENCT .PNO-TMI 80-288 s gy i imi t a r v no t i f,i cInc

.it; i n'.e re t t sicaiticonce. a t informa.#on i on,, c o,n presentec t t i,t u te is t,,,f,AR L,,Y_no t i ce ,o f even t_s ds in1tially

'vf?i:WpT 6v}iIiaQo{anc .1 s nasi cTij;y a l Ptno t i s r.nown oy NRC s tafreceived witncut f on tnis da te 2 racility': M.t:tropolitan Edison Company l'nree Mile Island. Uni t 2 Hiddletown, Pennsylvania Docket Number 50-320 Subj ec,t,; RCACTOR SUILDING PURGE STATUS

. veging of tne reactor building atmcconero utilizing the modified hydrogen cont ci (MiC) tystem continued at various system flow rates ba cd en meteorological condition. Momentary shut: towns also occurred due. to stack monitor filter changeouts. Also a scheduled shutdown lasting approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> occurred for computer maintenance.

Yextr.,zm a;orexir,a prge system flow tely 100,000 cfm in rate thewascastapproximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Total 540 cfm while stack flow rate averaged calculated radioactivity released as of 7:00 a.m. was 9676 Ci based en stack flow rate and measured stack concentration.

er.ainingat analyzed concentration

.29 uCi/cc (50,409in the reactor building based on the last building sample was C1, total).

MC evaluotion of the additional particulate monitor on the plant stack is in progress.

Grot samoles continue to verify little or no particulate activity released and are in9cendently verified by tne NRC Region I mobile lobcratory.

k initial .:icnal reports s ages on reactor building purging will be updated on a daily basis during the of curging.

to verify the licensee's analytical The NRCresul::.Region I mobile laboratory will continue to be used Mecia interest has occurred because of public sensitivity to this evolution and TMI related events. The Cot-,onwealth of Pennsylvania has monitored these events. NRC has responded to inaviries.

basis. The NRC TMI Program Office staff has monitored events as they occur on a 2a hour a

This GLL in'a matic/' is curr nt as ofyasmJ n

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.PRiuMINARY NOTIFICATION OF EVEN! OR UNUSUAL CONCURRCNCL--PNO-TMT-80-38C Th,_is ureliminary notification, constitutes FARLY notice ut events of ;0$$LE .afe.ty or pubilt interer.t sicnificance. 'the iniormat1on Dresentec 15 at initiaTiv received vntnntt yy_ritiG Tiun or 'cv'elEatibii and 1s Basically,ait tna t"i's' known' 0'v NB s ti f f on tni s da t el.-

Ality: Metropolitan Edison Company Three Mile Island, Unit 2 Middletown, Pennsylvania Docket Number 50-3?0

Subject:

REACTOR Buti.01Nti PURGL STATUS Purging of the reactor building atnotphere utili7ing the modified hydrogen control (MHC)

~

system continued at various system flow ratet. Liased on meteorological condition. Momentary shutdowns also occurred due to stack monitor filter changenuts. An unschecuted shutdown at 5: 46 a.m. occurred due to the loss of tne sample pump for the eft iuent nonitor (HP-R-2't'if.) .

A replacement pump has been ins talled as of 8:3 red.m. and purging resumed at tha t ti.me.

Maximum purge system finw rate was approximately b.10 ctm while .tnck flow rate averaged appror.imately 100,000 cfm in the pauf. 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;. Tota l calcula ted radioactivity relea:.cd as of 7:00 a.m. was 14,346 Ci based on stack finw rate and ineesurec stack concentration.

Remaining concentration in the reactor building nasec on the last building cample was analyzed at 0.72 uti/cc (40.781 Ci, total).

An additional monitor, consisting of a 3X3 inch sodium iodide crystal detector installed in the plant stack sample line. is providing signals to a multi channel analy2er where f.r 'Ji gaer.a signals are distinguished f rom other potentici isotope, in particular cestium i3/.

This new monitor is being used as backup to the plant ef fluent monitor (HP R-?l%) witn associated filter saJUple anllyseh. The samf.le Analysis rEQuirCs the removal of HP-R 'ij9 particulate filter. Once per day, for spectral analysis to identify any particulate activii.)

'ication. All releases were made in accordance with the Cocantasion order the lechnics1 b..:ifications and the licensee':, procedures.

Additional reports on resctor building purging will be updated on a daily 'uasis during the initial stages of purging. Ihc NRC Region 1 mobile laboratory will cc.ntinue to 'ac used to verify the licenswe':, analytical re.sul ts. ,

1 Media interest has occurred because af public sensitivity to this evolutidn and i.N relatad events. The Commonwealth of Pennsylvania has monitored those eventz. NRC has respondec to 1 inquiries. The NRC iM1 Program Office f.t.4fi has monitored event.s as they occur on a ?a r.aur i i

oss i:5.

i This infonnation is curren / l

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July 4 1980 C . .'OA t E* P l;Ai10N O! E.EN~ Ok UntisgAL CON:Uuthcg __psg.TMi-80-350 4: cisnicarv notificatiun constitu_tes EARLY notice of events of POSilBLE safety or The information Dresentec is as initially received'witnoti h:.:

. i n.t.e r.e s t s i en i f i c_a n..c-.e u.etion - c r eva lua ti on 7.nu 15 casically all tnat is known by NRC staf f on this cate.

. r:

22:ility:~ Metropelltan Edison Company _

~

Tncee Mile Islanc, Unit '2 --

Middletown, Pennsylvania ._. 3

g; Docket Numoer 50-320 .

s 2

s O

Subject:

. REA:TCR BUILDING PURGE STATUS ce Neginy cf the reacter builoing atmosphere utilizing the modified hydrogen control (%C) sfc.e cuntinvec at various system flow rates based on meteorological condition. Momentary sn.;cewn: aise occurred due to stack monitor filter changeouts.

'%1mur parge syster. flow rate was aporoximately 520 cf:., while stack flow rate averagec m e,xi.ataly 100,000 cfm in the past 24 nours. Total calculated radioactivity released as of 10:00 a.m. was 19,256 Ci based on stack flow rate and measured stack concentration.

L.ex. air ing concentratio 1 in the reactor building based on the last building sample was analy:ed at 0.51 uCi/cc (34,550 Ci, total).

Ine multi-channel analyzer continues to be in operation and indicates no particulate activity being emitted.

acitionel reports on reactor building purging will be updated on a daily basis during t~ initial stages of purging. The NRC , Region I mobile laboratory will continue to be n te verify the licensee's analytical results. .

W:ia interest has occurred because of public sensitivity to this evolution and TMI .

related events. The Commonwealth of Pennsylvania.has monitored these events. NRC has espen1cd to inquiries. The t'RC TM1 Program Office staff has monitored events as they c: cur cn a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> basis.

~his inferr,a+@n is current as of 10:0 a.m.

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PRELIMINArd NOTIFICATION Wuly /, 67QV

. . .W.'Y hCl TAT I M i.  :'!LN1 OR UNUbuAL C0tiCUARLNCL--PNO-TMI- 80.MC a valitainary not ificillnw .00sti etos CARLY . - r.otice of events of ?0$51BLE e.a fety er

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r.:.e res '. 3..va.n i.' i c 3 ne n ,

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. .c i ' Q.i. .'te t rapol , t an L<ii s on f.cgan;.

Tnree Mile islant.!, Unit e Miiioletown . Pennsylvanid L)ocket Number 50 .3?0

(.:pj er.t : %ACT0K BUTI.DiNC PURLiE Sl ATUS Purgias of the red.Ctor bu'1c.ing atmosphere utili::ing the medit'ied hydrogen contrni (MME) system Covitinusd at verious system finw rates ba'. icd on mereprological condition. Momenta ry enutowns al$e occurred due to stack monitor

• iller changeouts.

MaMw purge system flew rate was app ox.in:ately 520 cini, wh'ia s*.ack flow retc averaced upro3imately 100,000 cfm in the pac.t 72. hours , total calculated radioactivity released a ef 7:00 a.m. Was 27.814 Cl based on stack finw rate hne measured stack coricentraticn.

Semining Concentration in t.he rect.t,or buildinc based on inc last builcing sample was analyzed at 0.37 att/cc (2I,070 Cl, Lutel).

De malti-chane.ci anaiyzer centinues tu be in operation and incicates no particulate )

Activi*g baing emitted.

[Silizaf;no6 of the modified reactor builcing purge syste't (MPS) for a faster purga rate is

'oscted by mid Waak.

Additiona.1 'repcrtS on reactne bui h. '..,j MQir.g will be updsled on a daily basis during the initMI Stages of uurijin9 ~ha NId RHVi on 1 'iiobile I Aboratory will Continue to be usec to verify the 1icent.ee 's aridlyciCal rer.ul ts.

Media interest h.>.5 occumd because of nonlir. sensit.ivity to t .is a.valut. ion and TMI reinted events. Inw Comictiweal th of Penti'syl vanie ha'. moni ttsred those event?.. NRC nas responded to inquiries. The N ir. -"J 'h ocram Of# ice *: Laf f has monitorec events &$ they eccur on n 24 hear be$ iS.

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                                                                                                                     .g FnEl.lMINARY NOTIFtcAT10N 01 LYLNT OR UNUSUAL CCNCURRENCC--T.'No-TMi BU-38I

, This.Orelsninary noti fication consti Lutes F,Af.I Y, n,o,tice vi ev e n t'., o f PO'i.

• S L ", ,r.a f A r y o r

  ;:..n t i c. i n e.re n. e.1 n4i. t.i c.a nc e . Tne infonamen cresente:; it at int.tativ reca ne: >:1t:c n rication .or evaluation ena is nar.1,cally cli .nar is kn,sn ny NRC statt en W., ya;i, Fasility: Metropolitan Edison Conipany                                                                                                    j Three Mile Island, Unit 2 Middletown, Pennsylvania Ducket Number 50-320                                                                                                  l l

S.u. abie.c t : REACTOR BUILDING PURCF STATUS Purging of the reoctor building atmor.pnere utilizing the modified hydrogen con rol (M-:.~.) systa.T. Continued at various sy>Lom flow rates bar.Hd on muteorological condit. ion. Momanta r) shutdowns also occurred due tn si.cck monitor filter changcouts. In addit ion , at approx ima. ely . f.) : 20 :.m. the system was shu'.down due to a high alar n or the purge syste.T :criter ( i l F'. ' . ' M l I nob'le gas channel. The reason for tnis e tann is under review. Maximum purge system flow rate was approximately 540 cfm, while stack flow rate aver aged opprci.imately 105,000 cfm in the past, 24 hourt.. Total calculated radioactivity releesea at of 7:00 a.m. was 29,677 Ci cased on s tack flow rate and measured s tack concantration.  ; hmaining concentration in the reactor building bas 0d on the laSt Duilding sanple Wu analyzed at 0.26 uti/cc (14,896 Ci, total). l 1he multi-channel analy:er conLinues to he in operation and indicates no part.itulat.+ activity being emitted. Utili:ction of the modified reactor building purge system (MPS) for a faster purge ratc is be operational later today. Avaitional reports on reactor building puroing will be ur. dated on a daily basis during tne initial stages of purgine. The NRC Region 1 mobile laboratory will continue t.u be used to verify the licensee's analytical resultr.. , i Media interest har. occurred because of public seny.itivit.y to this evolutlun and TH: r*1 steIf  ! even'.s. Ihc CormMnwealth of Pennsylvania ht.'. monitnrad these Cvents. NRi nas reipended tr. inquiries. The NRC Tlil Program Office bla f f has monitored events as they occur un c N noir ba r. i s . This information is current as of 9:00 a.m.

                                                                     -f?l' 3,c --.p f-. M g

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Nf4'3/iC a: 8M l l l tE:c <a 1 June 11,1980 *

                                                                , '" 16    A',' 10 0 7 4.W:. .  . erc.r3,.~

MEMORANDUM FOR: B.H.Grier, Director,RegionIE,'Sh'T. ., J. P. O'Reilly, Director, Region II'W'1.I!5..IN ' ' ' " J. G. Keppler, Director, Region III K. V. Seyfrit, Director, Region IV R. H. Engelken, Director, Region V FROM: Norman C. Moseley, Director, Division of Reactor Operations Inspection, Office of Inspection and Enforcement

SUBJECT:

IE BULLETIN NC. 80 DEGRADATION OF SCRAM DISCHARGE VOLUME CAPABILITY The subject IE Bulletin should be dispatched for action June 12,1980 to all BWR's with an operating license. The bulletin should be sent for information to BWR facilities with a construction permit. The text of the Bulletin and draft letter to licensees are enclosed for this purpose. It is not anticipated that a TI will be issued for this Bulletin. NRR has the lead on evaluation of responses and IE has the lead on inspection of licensee ~. commitments. Inspect in accordance with Inspection Procedure 92703. Norman C. Moseley, Director Division of Reactor Operations Inspection Office of Inspection and Enforcement

Enclosure:

1. Oraft Transmittal Letter

2. IE Bulletin No. 80-14 u

                          >$f;h.O            bn w,, u CONTACT:   W. R. Mil'.   )W                                       SQ                                   l 49-28180                         , g } % , j[\                     tv l

~ WPU:SM ROI:IE,cf.n ROI:IEFj ROI XOC6 , 6/10/30 WRMills ELJordfn o NCMoseley DThe an 9 Job E 6//0/80 6//0/80 6/ /80 ' 6// C A 6-3 1 M 'fAc'n' M 5 N W l k-17 $ l

(Draft letter to GE BWR power reactor facilities with an operating license) IE Bulletin No. 80-14 Addressee: Enclosed is IE Bulletin No. 80-14 which requires action by you with regard to s your power reactor facility (ies) with an operating license. In order to' assist tne NRC in evaluating the value/ impact of each Bulletin on licensees, it would be helpful'if you would provide an estimate of the manpower expended in conduct of the review and preparation of the report (s) recuired by the Bulletin. .Please estimate separately the manpower associated with correc-tive actions necessary following identification of problems through the Bulletin. Should you have any questions regarding tnis Bulletin or tne actions reouired by you, please contact this office. Sincerely, Signature (Regional Director)

Enclosures:

1. IE Bulletin No. 80-14

2. List of Recently Issued IE Bulletins

                                                       ,9f-i<bPt)

        .(Draft letter to all GE power reactor facilities with a construction permit.)

IE Sulletin No. 80-14

        > Addressee:

The enclosed IE Bullet'in No. 80-14 is forwarded to you for information. No written response is required. If you desire additional information regarding this matter, please contact this office. , i Sincerely, Signature (Regional Director)

Enclosures:

1. IE Bulletin No. 80-14

2. List of Recently Issued IE Bulletins l

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1 i SSINS No.: 6820 Accession No.: UNITED STATES 8005050056 NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 June 12, 1980 IE Bulletin No. 80-14 OEGRADATION OF BWR SCRAM 01GCHARGE VOLUME CAPASILITY During our review of BWR operating experience, two events have raised concern on operations related to the control red drive system scram discharge volume (50V). 0 scription of Circumstances: At Hatch Unit 1,'on June 13, 1979, while performing surveillance to func-tionally_ test 50V high level switches, two switches (C11-N013A, 8) were found I to be inoperaole. Redundant switches (C11-N013 C, 0) were operable. The l reactor was in the refuel mode and these switches had been modified-prior to I this occurrence. Inspection of the inoperable level switches revealed that I the float rod was bent.and binding against the side of the float chamber on both switches. The licensee believes that the float rods were bent during or i prior to initial installation and that metal particles from the modification caused binding of the float. (LER 79-038) Brunswick Unit 1 reported that slow closure of the 50V drain valve during a reactor scram on October 19, 1979 apparently caused a water hammer event which  ! damaged several pipe supports on tne 50V drain line. Drain valve closure time  ! was approximately five minutes due to a faulty solenoid controlling air supply to the valve. The damaged pipe sucports were repaired but repair parts for the faulty. solenoid were not available. To prevent possible damage from a scram, the unit started up with the 50V vent and drain valves closed except for periodic draining. During this mode of operation the reactor scrammed , l from high level in the 50V, without prior actuation of either the high level l alarm or rod block switch. Subsequent inspection revealed that the float ball on the rod block switch was crushed and the float ball stem on the high level l alarm switch was bent such that the switches would not operate. The water hammer event discusscd above was the reported cause of failure of these two switch assemblies. (LER 79-74) As a result of these events and related anticipated transients without scram (ATWS) studies, concern arises that the 50V function may be degraded by the I undetected presence of fluid in the 50V. The second event is significant in tnat it indicates the potential for a common cause failure (faulty solenoid) to result in operation of the 50V in a manner which could defeat botn the level switch function and the SOV draining function. The ATWS generic studies (NUREG'0460) have led the staff to propose, among other requirements, improve-ments in the 50V designs to reduce susceptibility to common cause failures.

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2, separate correspondence, the staff will provide example Tecnnical Specifica-tions related to the action items discussed below.

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IE Bulletin No. 80-14 June 12, 1980 Page 2 of 2 A. GE BWR's With an Ooeratine License The following actions are to be taken by licensees of GE designed BWP. facilities with an operating license:

1. Review plant records for instances of degradation of any 50V level switch which was or may have been caused by a damaged or bent float assembly.

Identify the cause and corrective action for each instance.

2. Review plant records for instances of degradation of 50V vent and drain valve operability. Provide the clesure times recuired and typically observed for these valves and the basis for the required closing times.

Identify the cause and corrective action for each instance of degradation.

3. By procedures, require that the SDV vent and drain valves be normally operable, open and periodically tested. If these valves are not operable or are closed for more than 1 hour in any 24 hour period during operation, the reason shall be logged and the NRC notified within 24 hours (Prompt Notification).

4 Review instances in which water hammer or damage which may have been caused by water hammer has occurred in SOV related piping. Identify the cause and corrective action for each instance.

5. Review surveillance procedures to ensure that degradation of any 50V level switch due to a damaged float or other cause would be detected and that inoperability from any cause would be reported to the NRC.

6. If no functional test or inspection which would detect degracation of each SOV level switch has been performed during the past 3 months, make provisions to perform an inspection and functional test of all SDV level switch assemblies at the next reactor shutdown of greater than 48 hours duration.

B. Reoortine Recuirements The action taken in response to the items in Part A shall be completed and a written report on the results submitted to the NRC witnin 45 days from the date of this Bulletin. This report should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the NRC Office of Inspection and

 ,          Enforcement, Division of Reactor Operations Inspection, Washington, D.C.          -

20555. Approved by GAO, B180225 (R0072); clearance expires 7-31-80. Aoproval was given under a blanket clearance specifically for identified generic problems. B-/ P 3

l 1 i IE Bulletin No. 80-14 Enclosure June 12, 1980 RECENTLY ISSUE 0 IE BULLETINS Bulletin Subject Date Issued Issued To  ; No, ' 80-13 Cracking In Core Spray 5/12/80 All SWR's with an Spargers OL 80-12 Decay Heat Removal. System 5/9/80 Each PWR with an OL Operability

 '80-11           Masonry Wall Design             5/8/80     All power reactor facilities with an OL, except Trojan 80-10           Contamination of                5/6/80     All power reactor Nonradioactive System and                  facilities with an Resulting Potential for                    OL or CP Unmonitored, Uncontrolled Release to Environment 80-09           Hydramotor Actuator             4/17/80    All power reactor Deficiencies                               operating facilities and       i holders of power reactor construction permits           )

80-08 Examination of Containment 4/7/80 All power reactors witn j Liner Penetration Welds a CP and/or OL no later l than April 7, 1980 80-07 BWR Jet Pump Assembly 4/4/80 All GE BWR-3 and Failure BWR-4 facilities with an OL-l 80-06 Engineered Safety Feature 3/13/80 All power reactor (ESF) Reset Controls, facilities with an OL 80-05 Vacuum Condition Resulting 3/10/80 All PWR power reactor In Damage To Chemical Volume facilities holding Control System (CVCS) Holdup OLs and to those with Tanks a CP 79-01B Environmental Qualification 2/29/80 All power reactor of Class IE Equipment facilities witn an OL . 80-04 Analysis of a PWR Main 2/8/80 All PWR reactor facilities Steam Line Break With holding OLs and to th'se o Continued Feedwater nearing licensing Addition

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June 23,1980 l Gary G. Zech, Tecnnical Assistant ' Technical Support Granch . Planning & Program Analysis Staff Office of Nuclear Reactor Regulations i 1 Attached are reports discussing events that led to the degradation of the Brunswick scram discharge instrument volume (SDIV). Also attached is an LER describing the discovery of the disabling of a set of level switches I in the Hatch Unit i SDIV. l The ACRS would appreciate a brief discussion of the implications of the above incidents, particularly the Brunswick event. It is suggested that the following items be addressed:

1. Detaf' led description of SDIV/SOV.

2. Potential for common-mode failure of the SDIV level instrumentation  !

(via water-hamer or other events).

3. Consequence of filling the SDIV versus margin of safety'for successful scram function.

Apparent failure to verify operability of SDIV level switches 4. following Brunswick water-hammer event.

5. Consequences of allowing plant operation with SDIV drain and vent valves closed.

6. Lack of apparent "f ail-safe" mode for level valves.

This topic has been tentatively scheduled for discussion at the July ACRS Meeting (July 10-12, 1980) if the NRC Staff will be prepared to discuss the implications of this matter.  ; 1 R. F. Fraley , Executive Director Attachments: as stated cc: D. G. Eisenhut, 00L

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e s.r= sch =c, Ji= %:?.es= (canli=a Tenre.: in I.ight ce=any) , the in=:.swin Steam [ nee:rie 71 ant (3.!?) e y=h:it.s tha fo11 cad g infer::stic: e :5e thi: k.1 scram dischs:sc voltane fleet rwitch fail:::t on Tarvenime 14, 1979. g s [, Durir:; a re.ac:e: se:::= ec Oc:ote: 19, 1979, the scrum dirch:;33 =cimut d:sh g valve did not class b :.5s requd;td time., ca=a1=g a wats.: ha=mut eTT::: which 11 i de.r. aged severa.1 pipe s-c,.c .s oc the sc.rea dise.ht ge Tclusw. disChs:ss F 7 i: 35 k (LII.:-7r-71). D.e vaiva was posit 12;.12g it.self in t.be "c. lase

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                                                                                                                                                 @ -=: t es.: per-                3 d:M r valves closed te ::evec: petsible d.maspe free a se;>=.                                                                                                            i-ser. 31 vsrc to sce.its; t.be serra dis.t.h. tere vel ==ue by periodita.*ly deale. ins, as Latrd 4:                [

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A repNet seleceL4. located at the Eatch Fla=: to cptir the shrum disthargs vols== drain vain5 vs.s( The valve was tasted sat /iafactori.ly sod the scram P scha;gs voluss instdied. vent ed drais system was retu.:ned to a so:inal lise=p. A periodic =ast hu [g e_ Lee of bee:= esemb"-w to tese tbe bish Irrsi s2. ara and tbe red blod 5 y tbese sad.ic.he:s. Alse, a N1::w=.ance I=.strueties h.as bec: prepared := perfers .{.. a W_. odic fametion.a1 test cf the scras M A se volume vs:t a=d d=21: v f Ec travr tha: this i=fe =atic.. natf.sfies tte ccroesr:s a:creweed ce the fsilure E of rhm swirches. If you have ser addia* ' questiocs oc this mat. tat, ;1tSFC Ie contact a =e4*- cf my staf f er an U 1 Very truly 7:== s, ..-- ) m t g

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PS,Wi tswe. Cou*Castw-  ; A ""t* 8'.-"".-e g essge. 3mpecewN atwwfN* NP* M many s Ag=' ,;m g n e A t CpePLANT MCTWCO asOURS A*"eha6"*La Su 6 Caw wg. gypsy ta l . TANJ'g:"e" 04isiil@ A - . L,cJ@ l o I d 'I d. .I c l@ I4l@ lxl@ Ieioi5 i5 q l CAUSE DESC*irTICM AND cDARECT1VE ACT10Ms 27

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5. . .a s . .-_ e . . . es ..ye-e se - ,3er.nisatelv six heu s --ier to 'the scrz '

il'iI* v m . ii2j 1 < -- -<- - e . e .- s- e e . th: dre tsis eeried. it is believed' : bat the 811:e; cake i l I gI a - - -- ., 4- . .w. edi... .. .-. yeth ee,stsie af entrat::er.: due to s*ste- leak. ate,

                                                              -,s..r.--.                . . -             t . . . .:        .s.       vv5 sv,te- ses eres surt:ed and the (Ceci' d '.

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Pt 4&cPrat t 6 ,uussa't3 - . w6vs t. e ts.w. enow ' M } 1 inII__6lololhl a 9 16 12 S: 6ess os o. 04 met to aacurv '7 t / L n kl@l evet :tsc= enc =. . u .72OO W\ s: 4 9 13 N 0*8 U NRC U$( QNt,Y t

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!nj issvc:hl lv J  ?--.i -.dde. televisten. and teus: ster I IIIIIIIIllIII*

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p oug: 918-457-95 1 i u22.g.,,,,g,ing,_ A. c. To111sen. Jr. '

                                                                                                                                        /4 -/A3

LIR COS !h~.'ATION ~ R06 1-79-74: 7acili:y: ,3STJ 1: #1 Eve:: Date: 10-19-79

                           !"DC OISORI?"!OS A'TD PROSA2*.2 CONSIOL*INCIS (Con:'d)
                                        $11th: Increase                      No Chnnee                     .
                                                                                                         .            e
                                        ?cver (17D.)                         Cote A F Vessel ?ressure' Vessel Level Cere nov                             "3" Teed Tlev "A" Teed ncy                         Thro :le Pressure Total S:ein Tlev                     Gross Magava::s Turbise Vibra:1ce                    .

Ives:s =o:ed during a:d a.fte che se:a= vere as follows:

1. Il Diesel Geerator did =c: scar: c the lev level (-3S") 1:i:14:1o:

sig a1 at the cime of the scram.

2. While tryi=g to go 6:o to::s cooli:g and later, shu:dev: ecol 1=g, c:

                                  "3" REI heat excha=ger, de A T could be developed across the hea:

excha ger. * .

3. Af ter bei .g used i=:e==1: et.17 throdshou: the day for vessel level ese:rol, the 101C tuttime could ct be started on .he af te.: noon of the scra=.

S' g

4. Tassel chc=is: 7 exceeded 2 u=ho for app::ximately 31.5 hours, res a .ax1=u= value of 6.63 v=ho.

5. :The CTO sers= discharge vo1E:e drai: 11:e was fou:d to have several pipe ruppe::s da= aged f:= ar. appare: vater ha=me ev e:.

N. Ter$-4c.a?, Specifica:1ces 6.9.1.9b

                                                                     .                                                  l
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4 e Y

l

l. 3 LII. CCC WA-~CN - KC# 1-79-74 i

Zven: Oa:e: 10-19-79 facill:y: 3SI? U:10 # 1. CAUSI DESCRI?-*0F AF3 CDU.I-~VI AC""!ONS (Cont' d) 503 pu=p d-as star:ed. Approxi a:ely cue =in=:e later, the reacter scra=ned. Issh 1=jec:ed isto the vessel caused a cha ge 6 hcociant viscoshy is imbalas=e and surf ace in vessel tension resulti=g 1= flew and pressure disturbances. De  ! che=1st:7 results 6 a release of N-16 v'..ich carries over with the staa=." g E-16 high e=arry ga=a caused the scra=. , vas due to a split mecanic bellows ne on failure the lubeofoil thediverths il Dieselvalve Ce:erate: actua:c , causing a less of eo ::ci ai:. De diverting valve ac: a:o vas replaced vi:h a new vender modified design as all diesel geers:c:s. Diesel ge: era c cc:::ci at pressure has been to be checkad once ,, added to the auxiliary opera:::s Daily Surve111a=:e Repor per shift. . vrt he4: excha:ge 1sle: valve 70473 an oved S e anti-rotatie: device on the the ste= te rotate vi:heu: moving de valve disk, nis n e is anti-rotatics c==sidered device a isolated was adjus:ed a=d de valve tested satisfactorily. eve =:.

                                                                   ~

Disa.sse=bly of the 3.CIC stes= exhaus: check valve shoved da: the stud and nu: c= the back of tha disk had brekas a=d de disk had separated fro = the h and had lodged 2 :he valve. inlet. i=s:alled upes a::ival. I:siseert=g has bes: asked to de:er-%e the cause of the fa11=e. . Yessel condue:1vity 6:: case to 6.63 W vas dv.e to the rash hje::ien andne vesse flux.

                -         res1= treakden due :c ta=perature and neu::::                     Operati=g ?:ocedu:e 14, W OO, to less thas 2 u=ho is app: xi=a:e'ly 31.5 hours.to the conde=ser after precoa:ing has been revised :o direct the GOU efflues:

or as extended shutdev u=:11 a che=ical analysis is :.:n don the afflues:. Also, the revish requi:es tha: ~

                         .de=1:aralizar be used when goi=g.:e '*Ecid/S:ar:", (2) always backvash                  than thirty an p:ecoa: the filters when EC; has been dep;essurized for 3:ea:e:'os opened until
                           =1:u:es, and (3) T041, de GCU 1mj e:: ten valve, vill no the syste= is cc-=ple:ely pressurized and stable.

y nebyC23 sera = discharge vol=e drafh if=a had several va:e: ha=e: duri:g de sera =. nis water ha=e is believed valve. of its supp to have been caused by the e.xcessively loss closhg ti=a of the drais li=e isolatie:of the solesc his proble= has been cor ee:ed by replace =es:n e da= aged supports were aise replaced is required. opera:es the dra h and ve:: valves,is underway to deter =ine if addi:1onal supper and a evaluatie:

                            ?T o

inspections vin be condue:ed c= all velds on this line.the o 4

                                                                                          ,M-                           .

                                                                                                                =

_ p.. . .

• ...i.'* 1.ib:*4TI.:. "137-150T101 3YSTE9 (RIOS) 30C.CATE: 60/05/20 '07A4! ZED: 40 00CKET a

   *i33 I: .       .43 : ic9523Jd!*                                                .

05000321 1**.:s,-32: !:.ia : . -e : *.welea 81amt, Unit 1, Geeacia * .ee C 4- . .; q ast -s Art:Llaitci

• Ge:a,ia *c.e* Co.

K E . . . . a .* E di 181E:.? AFFILIAT! N de-iea 2, attaata, Office of tre Directee Li;ECT: LEE 74-43A/d33*1: Om T90213,3C'em CiSCN8PQe VCl Miqa leveI s.it:mes inc=e ra:1 e dvaine surve ill amee test.Cause: my bent dlcat rec. Recs steticmtemec & smitches reasse==le=, C081ES RECEIVED:LTR ./ ENCL /_. SIZE: ! II57 Ih.T:"N CODE: AC025 TITLi: Incicent 8eecrts ,

  *CTES: _

CORIES RECIPIENT COSIES R E C I

• 1 E:.T LTTE E 4CL ID CODE /dadE LTTR E t.C L 10 C O D E / >. a d E 4 4 f!O it D5BC Of7 g 4 g TER4aL: 41 REG FILE 1 1 C2 NPC PDA* 1 1 2 11 Mpa 3 3 cc ISE 2 1: TA/500 1 1 15 NOVAK/KNIEL 1 1 16 EE3 1 1 17 A0 FOR ENGR 1 1 14 *LANT SYS SR 1 1 19 I&C SYS 98 1 1 20 40 *La'4T SYS 1 1 22 REAC SaFT OR 1 1 23 ENGR BR 1 1 2a K4EGER 1 1 25 PaK Sf3 64 1 1 26 A0/ SITE Ar4AL i 1 27 OPEMA LIC 54 1 1 28 ACDENT ANLYS 1 1 29 Au% SYS 45 1 1 A0/OR8 00A 1 1 AECO 10 10 SERLINGER,C. 3 3 M A:4 A V E R ,5 1 OCJ:
• 1 ' *18 A i 1

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1 Q M S S. D 1 04 NSIC 1 1 TE4NALt 03 L80R 1 1 29 AC25 16 16 4 4 T *, T a t .u 1Es or COPIES CE3utRED: LTTR 61 ENCL 61

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u ..,3,. . n l m m. s. 1 F.ay 20, 19SO  ! TM-SC-435 r FI /JC I. I. EATCE . 1.1cessee E.ven: Repor: Docket No. 50-321 . Uci:ed S:a:es ucles: Ragula:ory Co ission Office =f it.spection and Enforce =an: ' Regie I~ Sci:e 3100 - 101 :tarie::a 5 :eet Atlan:2, Georgia 30303 . X:. James ?. O'Reilly d7. 101:: I Technica.1 Specifiestions,

             ?ursuas: to Section 6.9.1.9.a of Ha :h Uni:                              50-311/1979-38. Rev.1.

please fi:d a:tached Ecportable occurrence Report No.

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                                                                                       -          j Y.' !!.acry Plast Ma. nager )

[ , R 5/ la xc: J . E. .t11er , J r. R. J. Kelly V. A. Vidser C. L. Coggis S R. D. Baker l OD* Con: ol Room ' S TLle

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7) LICENSEE EVENT REFOAT Tot.Lo'.' UP REPORT TO PREVIO"5 REPORT SUEMIMID 7/2/79, a:

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                                                                                                                                              ,.              ..        ..:.          u Event ossemiatics Ano *=cs Aats c

• 5EDVENCES h ee funeen uiin : ee I El k"hile eerf e -i .T su-veillance precedure WY'-1-3004 discharge volu=e hi-level switches IC11-N013A and 3. the switches l

gI sers: l T"") l vere feu=d :o be incoerable. n e reacter was in*:he refuel mode end the l j 31 i redundas: svitches ' 1C11-N013C and D vere coerable. n ere was ne effe:r ie 1 I c oublic hee 1:h er saf erv. n is is not a rece:1:fve ocecrfence.

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a u .i 42 2 o CALtSE CE5*miPTION AND comett:T1vt ACTIONS h I i;eij ' Level svi:ches 1C11-K013A and 3 vere ocened f e i=seec:ien. n e fles: red ne reds .were straighte=ed the I i,ijj was f ou=d to be be:: en both svi:ches. switches rea.sse= bled and successfully fune:icnally :ested, ne vender's (

   ;77,i review did not dete-tine :he exae: cause. It is believed : hat the floar i
- 77] l                                                                                                                                                                                                  '

_J reds vere be.s: du:i=g er prier to initial installa:1cn. gl 30 f 8 9

• MCWee e* OtsCQvtRY DisgmtaTions 32 p A:tew Ottstm sTATVS et:0 EVE #v ST.TV5

                                       *.PC't. t .                                                                               Surveillance                                                        }

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Rg($ A11: =8 AI'.E ASI AMCss. r :A ActivtTY  !

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Pt a t C'.*.t k (IPeg,s g3 sett a Tvet 1 lti I l 10 lo 10 GL:_J@tts.a 9i 1 12 PTieN -"A 90 7 8 9 P1m$0N%8 L seou aits , ou..u . stua.rtiev ..! NA u. l t - l 10 l o 10 lhl t . e si ir

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NATJ:A~1'.~e TI?OTC ' Cec; sis Po.*er C: pny Pisa: T. 1. H a*.:h 32xley, cec:sta 31513 50-311/1979-38, Rev. 1., Reportable Occurrence Report No. a

o ' fune:ionally tes:

k* nile perferr.ing su:veillece procedure HXP-1-3004, scran discharge volu e hi-level svi:ches IC11-N013A and B, the svitches were f ound :o be inoperable. The reac:or was in the refuel mode and the Level svi:ches 1C11-redundet svf ches, IC11-N013C and D, were operable. i l-N013A and 3 vere opened for inspec:icn. The tofloat bindred was f :he agains: ound to be side of-bent c: both svi:ches and causing the floaThe ficat rods were straigh:enod, the svitches reassemb the floa: chambe r. and functiona.11y tesced. ~ The modification These switches had been modified prior to this occurrence.1:s purpose was to per=it added a new ve:: perr to the floac chamber. t ! total isela:1es of :he level switch f rom the scram disc.harge header during f calib ra:1:n. After the modification 1: vas found : hat the flest was s:uck The top of the float chamber was removed and the flos I and wouldn't move. Wear marks on the inside of the flos: chamber ste: vas found to be bcut. had been rubbing on the tank for so:e time; hoveer, indica:e tha: :he fica: It is believed . the sv'.tch had never f ailed to pass surveillance tes:ing.

ha: the floa: rod was originally ben: prior to installation o: during 1: is This caused lish: centact between the float and tank.

insta11a:i:n. setal par:icles from the modification performed on the chs=ber believed :ha: fre: movin g.' Normal quarterly esused the binding which s:cpped the flos: surveillance has been increa. sed to monthly on these No f ur:her' switches action f or one quarter. is planned. No f ailures have been noted.co date. e l l 1 e 4-/7

SNUNSWICK 1/ HATCH 1: BULLE11Nd Abdutu REGARDING P.0TENTIAL LOSS OE SCRAM - DISCHARGE VOLUME CAPABlLITY l l I I TWO-PART PRESENTATION ON RECENT OPERATING EXPERIENCE RELATED TO THE BWR SCRAM FUNCTION

                         --IE BULLETIN 80-14 -- DEGRADATION OF SCRAM DISCHARGE
                           ' VOLUME CAPABILITY, ISSUED JUNE 12, 1980 e MULTIPLE FAILURES OF SDV LEVEL SWITCHES AT HATCH AND BRUNSWICK
                         --IE BULLETIN 80 -FAILURE OF CONTROL RODS TO INSERT DURING A SCRAM AT A BWR, ISSUED JULY 3,1980 e FAILURE OF 76 0F 185 CONTROL RODS TO FULLY ,                          ,

i INSERT DURING SCRAM AT BROWNS FERRY 3 ON JUNE 28,1980 < I l l l i l l

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i l l I l l l l l TYPICAL SDV LEVEL SWITCH ASSEMBLY I

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l l l l BULLETIN 80-14 DEGRADATION OF SCRAM DISCHARGE VOLUME CAPABILITY THE CONCERNS WHICH LED TO ISSUANCE

    --TWO EVENTS INVOLVING MULTIPLE SDIV LEVEL SWITCH FAILURES RAISED CONCERN THAT A COMMON CAUSE OF FAILURE EXISTED.

HATCH 1, JUNE 13,1979 FOUND iWO INOPERABLE HIGH LEVEL SCRAM SWITCHES--CAUSE WAS BENT STEM ON FLOAT ASSEMBLIES BRUNS,'ICK 1, NOVEMBER 1979 FOUND INOPERABLE ALARM l l AND ROD BLOCK. SWITCHES--CAUSE WAS DAMAGED FLOAT l ASS EMBLIES l

   --REACTOR OPERATION WITH SDV VENT AND DRAIN VALVES CLOSED BRUNSWICK 1, NOVEMBER 1979 REACTOR STARTUP WITH CLOSED SDV VENT AND DRAIN VALVES DUE TO UNAVAILABILITY OF REPAIR PARTS
                                   ~ }e K

l l BULLETIN 80-14 1 l DEGRADATION OF SCRAM DISCHARGE VOLUME CAPASILITY l 1 OBJECTIVES OF BULLETIN 80-14 l

   --REQUIRE OPERASLE SDV VENT AND DRAIN VALVES
   --REQUIRE OPEN SDV VENT AND DRAIN VALVES DURING OPERATION
   --REQUIRE PERIODIC TESTING OF SDV VENT AND DRAIN VALVES
   --REQUIRE OPERABLE ROD BLOCK AND ALARM SWITCHES
   --REQUIRE PERIODIC TESTING OF ROD SLOCK AND ALARM SWITCHES
   --CBTAIN FAILURE DATA TO EVALUATE POTENilAL FOR LEVEL SWITCH MALFUNCTION, ESPECIALLY FROM COMMON CAUSE 0

1 _ z.o a

1 l l l l i SULLETIN 80-14 DEGPADATION OF SCPAM ;ISCHARGE VOLUME CAPABILITY FINDINGS TO DATE , J

    --IfEEDI ATE SUPVEY OF PLANTS DETERMINED VENTS AND DRAINS OPEN ON ALL PLANTS
   --45-DAY REPORTS DUE JULY 27, 1980
   --REPORT OF INOPEPABLE (STICKY) ROD SLOCK AND ALARM SWITCHES AT BR0h"'S FERRY l

l l

                           - 2.A

I i m, G. > 1 l L., t i 1 1 Pages A ,fco' thru 4 -J o 1 has been NOTED: deleted as I l l s. 9 4 il ,P.# ;f I'. ( '. l '

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PRELIMINARY NOTITICATION June 30, 1980 Page 1 of 2 PKILIMINA?.T NOTif1 CATION 07 EVENT OR UNUSUAL OCCURFINCE PNO-II-60-119 This,p,reliminarvnotificatinnconstitutesEARLYnoticeofan'ev'estofP655!3;E l sa fety or public interes t,_ significance. The ,io t o rma ti o n p r e,s en,t ed i s of ini t1211'/ receaved witmou,t, veri fi ca tion or eval uation and,is basi cally all that is known by II staff _as et this date f , ,., y -- no FACII.iTY : Tennessee Valley Authority F- ~ ~ 25 - IU Browns Terry Unit 3

                                                                                      .       E'     ca       CD            c Dock-t Mo. 50-29e

• c' Attens, Alabama SITRJ E C".*
• FAILURE OF C0h" TROL ROOS TO INSERT DtIRTNG A SCFM At '1:30 a.m. on June 28, seventy six cont rol rods failed to fully insert during a routine shutdown. The reactor was manually scrammed from about 30% power to effect repair to the feedwater system. Following the manual scram, 76 of Lhe 185 control rods failed to fully insert. The partially inserted rods were all on the east side of the core. Reactor power level sax indicated to be leas than one percent in the east si'le of the core. The west side of the core was suberitt-cal.

A seceed manual scram was iniLiated four minutes inter and all partially inserted rods were observed to drive inward but 59 remained partially withdrawn. A third manual scram was made one minut.e later and 47 rods remained partially withdrawn. Core coolant finw, temperature and pressure remained normal for plant conditions. Six minutes later an automatic scram occurred from a scram discharge volume tank high water level signal. All the rods full inserted at this time. The unit remaina shutdow= and the cause is still under investizilion. No definite cause has been determined at this time but the problem appears to be hydraulic in sature. The scram discharge volume, which :eceives scram discharge water from the east control rod drive units, is postulated to have been partially filled before the event; however, no level annunciation was received prior te shutdown. A Confirnation of Action Letter is being issued, requiring the licensee to keep the plant shutdown pending comp 12 tion of their investigation of the fa Llures sud  : enmpie!1on of corrective actions. A restart will not occur without NRC concur- l rence. ' A Region II inspector was dispatched to the site to assist the resident inspectors on June 2S. Two NRO resident inwpeeLors are onsite and a Region 11 supervisor is enrnute. Media interes: has occurred. A nevh r*lesse was made by the licensee. The NRC ) plans oc news release at this time. The State of Alabama has been informed. i i Tte NEC Emergency Response Center received notification of this_2cturrence by

u. telephone froc the. licensee at 2:25 a.m. CDI on June 30. This information is current as et 11:30 a.m. on June 30.

C/R 0:57. CHM. CMP.S. PE.GC.CA. Con t.n e t : C. Julian, 811 242-5533; x. C. Lewis, SECY.H.. ui. 3:....a. vo. . M

                                                                     =-
                                                  ! PRELIMINARY MTT.TPATION July 3 1980 PRT.LIMINARY NOTIFICATION OF EVF_NT OR UNUSUAL OCCURRENCE PNO-II-80-119A 7: .'         e This    preliminary     notification saf'ety or public interest constitutes,  EARLY     notice  of an'_eTent_of     POSSIBLE significance.      The inf ormation precerrted is'as-ing ially received vitbout ve ri_ fica tion or evalua ti,on and is basi cally - a].1" thate-is knowa by IE' s ta f f a s of thi s da t e                                              iG'         -

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FACILITL Tennessee Valley' Authority Browns Ferry Unit No. 3 ' M[r E $

                                                                                            % . C. u 2.             7
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Docket No.' 50-296

                            ! Athens, Alabama                                               Spf'2 g N               M           '

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SUBJECT:

TAILUPd 0F CONTROL RODS TO. INSERT.DURING A SCRAM This PN updates PNO-Il-80-119. The cause of the seventy-six control rods net to fully insert following a routine shutdown on June 28 has not been identified. i l The two inch diameter piping which serves Lo drain the east ceram discharge  ! volume manifold was disassembled. No blockage was found and the piping did not contain excessive foreign material. . This. piping is being restored. i 1 Additional flow tests are planned in order to compare the drain capabilities of the east and vest scram discharge volume manifolds. Future efforts include examination provides of vent check valves in the Clean Radioactive Waste systatn, which the vent control rods. Thirty-five for the scram discharge volume, and scram time testinS of all of the affected rods have been friction and scran 1 tested, witl1out any abnormalities being identified. l

                                                                                                                                \

The licensee bas' verified by unitrasonic testing that t.he scram discharge wiuces i l of ot Units 1 and 2 are emptv and capable of receiving discharge water in the c. vent I a reactor scrac. Un2ts I and 2 are operating, j NRC has a ten man investigation team onsite that is headed by the Regien TI Director. i I I A news release was made by the licensee. The NRC has not made a news release. The State of Alabsma has been informed. This infomation is current as of 10:30 a.m. on July 3. Centa c,t : C. Julian, RII, 242-5533; R. C. Lewis, RII, 242-5535 I Du t ribu tion: Transmit.ted H. St. l i Chairman Ahearne Commissioner Hendrie l Comissioner Gilicsky S. J. Chilk, SECY i Commissioner Bradford C. C. Kaccerer , CA Cocmissioner Kennedy. ACRS  ! (For Distribution) Transmitted : MNBB P. Bldg. V. J. Dircks , TD0 II:XOOS (II:HQ Dist.) i' H. R. Denton, NRR C. Mi.helson, AEOD R. H. Vollmer, NRR l J. J. Toucha rd, PA Landow (6 min /pagc) R . J . Ma ttson , N7J1 J. J. Cunnings, CIA N. M. Haller, MPA D. F. Ross, NRR ! i R. G. Ryan, SP D. Eisenhut, NRR MAIL H. K. Shapar, EI.D S. H. Hanauer, NRR R. Minogue, SD IE:XOOS Willste_ Document Mgt. Br. (For PDR/LPDi DIST. CHM.CMRS.PE.GC.CA. J c. Davzs, NMSs SECY . HIST . RE CODS . E' J' 8"d"It ' EE8 PPILTNTNARY NOTIFICATION h ~2 h f

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NRC StafiBacks 24 Test Shutdowns  ; 31111 stone Unit 1. Waterford, Conn - ! ums'*a "" Further tests and studies wiu be Nine Mlle Point Unit 1. Oswege. N.Y i Tne staff of tne Nuclear Be;ulatory co'1 ducted on the Alabama reactor Comn11ssion has proposed that the thJs weekend in an effort to detri. Oyster Creet. Lacey Townslup. N.J.. } Peach Bottoni Units 2 and 3. Pe6ei; i NT:C order test sbt.tdowns of two mine the cause of the malfunetton. Bottom Township Pa.; Pilgrim. P'ym dozec nu: lear pow er plants in en ef- Ste!!o said. On Monday, the NRC staff will tele, outh. 31 ass; Vermont Yankec. Ver fort to find out why an Alabama plant non. VL; Browns Ferry Units I and 1 , exncrienced a mysteriou3 malfunctien phone the operators of the 24 ott er ' dutta; a rouune snutdown la.st sees. reactors around the nataon and tell Decatur. Ala.:. Brunswick Unb I and

2. Southport. N.C.

end. them nbether or not to test the:r Eden 1. Hatch Units 1 and 1 Bas-Victor Stello. tnt N.lC's director of snutdown mechanisms. ley. Ca.: Big Rock Point. B:; Rock inspection ard enforcement, told the Stello said the staff tuay order both Point. Mich.; Dresden Units 1. 2 and 2. commission Thursday that Brown 5 manual and automatic shutdowns alorns. 111.; Duane Arnold. Palo, Ferry Un:t 3 was brought to near zero from low. power levels to make sure Jowa; La crosse. Genoa, Wis.; Mont. power generation despite the malfune. the systems are working proper!y and tion ar.d there was no ris); of an acei' to gather detailed infonnation whien C?ll o. Mot.ticello. Minn.; Guad Cities dent. mignt help explain the Browits Terry Lnits 1 a.nd 2, Browtville, Neo.Cordova,111.; Cooper, B .it tbc une.s pected and un ex- failure. plained shutdom failure concerned Such tests, he said, would take ; the NBC because of the vital impor- each reactor out of operation for j about 2% days, tance of the mechanisms that control Tne 24 plants involved in the pro. a reactor's power production and are relied upon to shut it down instantly posed tests are: in an emer;cac; Jatnes A. FitzPatrick. Scriba. N.Y.. l A-t/O

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 .. y ' ,~                                                                    BROWNS' FERRY 3:     FAILURE OF ALL RODS TO                ,

(.'.~ INSERT FULLY

                                                                                             .FOLLOWING A ...

SCRAM . SEQUENCE OF EVENTS BROWNS FERRY 3 r'j - FAILURE T0' COMPLETE SCR M 6/28/80 - l Time Event 01:31 Manual Scram from 400 MW (s30% Power). All rods on west side fully insert - on east side 13 rods travel full in, 5 rods were already fully inserted -- 18 rods on east. side' fully inserted,.76 rods partially inserted after scram 01:36 Reset Reactor Protection System (RPS) and initiated manual scram - rods on east move 12 inches average - 34 rods fully inserted. 01:37 Reset RPS and initiated manual scram - rods on east move 7 inches aherage - 56 rods fully inserted 01:43 Reset RPS and mofe Scram gDischar' e Volume (SDV) switch to

                                    " Normal" - received auto scram on high discharge volume.

All east rods fully inserted. NOTE: SDV vents and drains opened between scrams but drain times were not sufficient to completely drain ' system. b

                                                          $9- s-/ /

j' 00 SOUTH NORTH 59 2 1 3 4 2 1 3 55 3 4 2 1 3 4 2 1 3 I 51 1 3 4 2 1 3 4 2 1 3 4 47 1 3 4 2 1 3 4 2 1 3 4 2 1 i 43 2 1 3 4 2 1 3 4 7 1 3 4 2 1 3 39 4 -2 1 3 4 2 1 3 4 2 1 3 4 2 1 35 4 2 1 3 4 2 1 3 4 2 1 3 4 2 1 31 2 1 3 4 2 1 3 4 2 1 3 4 2 1 3 27 2 1 3 4 2 1 3 4 2 1 3 4 2 i 3 23 4 2 1 3 4 2 1 3 4 2 1 3 4 2 1 19 4 2 1 3 4 2 1 3 4 2 1 3 4 2 1 15 1 3 4 2 1 3 4 2 1 .3 4 2 1 11 3 4 2 1 3 4 2 1 3 4 2 I 3 07 3 4 2 1 3 4 2 1 03 4 2 1 3 4 2 1 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 1800 l ., FIGUR E 7.1-3 Control Rod Scram Group Assignment

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1. Hycrauli: Control Valve Alienment Verified

2. Eas: Bank Vent Valve Verified Operable

3. Friction Tested 35 Rods

4. Veri #iec Calibration of 3-Gallon, 25-Gallen, anc 50 ^alion LeYel Swittnes en Instrument Volume

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c. Verified tha: Off Gas Radia:icn Levels Were Normai

10. Completed V'sual and Me:hanical Insc. ettiens cf Vents an: Orains Ii in Scram Discharce Volume l

11. Verified that No Maintenance or Mccification Perfcrmed that Wcui:

l Aff ect Cor. trol Rod Drives l

12. Reviewec Scram History for Previcus Failures l

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1 l I i l 1 l l l I f ' I z t i 1

BROWNS FERRY - ALL UNIT 5 NEAR TERM ITEMS IMPLEMENTED

1. UT check of scram discharge volume piping for water after each scram.

2. Instruct all shift crews how to respond to an event of this type.

3. Perform surveillance of each scram discharge instrument volume level switches at least once/ month.

4 Visually check the CRD valves at least once/ shift.

5. Unit 3 to remain shut down until investigation is completed and NRC concurs in restart, l

1

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1. For 5WR's That Are Operatine Wi-hin 3 days of bulietin perform presCrice: surveillance ests on the 5: ram DisCharoe Volume System.

Witnin 20 days, unless otherwise Cirectec, De-f Cr one automeCic 2. (' and one manual scram at normal operatinC tem:erature and Cressure ( 1 with more than 50% of the rods fully withCrawn. e.e

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4. F.eview emergenC.v C;eracinc. ;rCCeCures\ 'incluCe th0sc DrepareC by tne .

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3. Develop surveillance procedures to moniCor the SCrE~ Cischarge VClume fCr 'ater w aCCumulECion.

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. ci-,u_q e 0 . c ,, 3. / : c Snutcown Facili y 5:atus Date Resta-: Estimate Commer:

Fitzpatrick 5/D 5/5/80 7/19/SD T:rus M :. 9+ Milistone 1 oper. 94 Nine Mile Point 1 coer. Oyster Creek S/D 1/5/8D 7/5/3D Refuel 1 e Peach Ec::o 2 S/D 3/21/50 7/24/50 Refuel Peach 30: tom 3 oper. Pilgrim oper. 94 Vernont Yankee oper. 1 crowns rerry i* oper. I Browns Ferry 2 oper. t Browns Ferry 3 S/D. 7/2/ED RJ Brunswick i S/D 5/25/ED 7/7/5D Refuei 2runswick s e S/D, 2 / i / c,; // : ,, :- rJ

                                                                                                                                         ..e;ue ;

Hatch 1 oper. Hatch 2 oper. ti Eig Rock ? int oper. l 7+ Drescen i S/D 10/1/75 Long r- Chem r i o. ., n . . . X' Dresden 2 oper.

• Dresden 3 oper.

Duane Arnold oper. La Crosse oper. Monticello coer. M4 7T Quad Cities i 3/D 7/2/50 7/5/ED ee: water-

                                                                                                                                           . . ,a_.K    v. .,. l i Quad Cities 2                                      Ocer.

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NOTED: Pages f9.JA 3 thru # ->> v' has been deleted as t [ A ", * ,O. _% Pe r.., i. ,.., 4 h

                                           .! J$

i l[ ),* ': . i k i 4 V

r APPENDIX XXIII ST LUCIE: PRESSURE INSTABILIlY DURING C00LUOWN ON NATURAL CIRCULATION. EVENT 4 0F JUNE 11, 1980 ST. LUCIE TRIP AND COOLDOWN (6/11/80)

1. SITE DESCRIPTION II. EVENT DESCRIPTION III. REACTOR COOLANT PUMP SEAL PERFORMANCE IV. STEAM VOID INDICATIONS V. ANOMALOUS SOLID PLANT INDICATIONS O

m..-

ST. LUCIE UNIT NO. 1 LICENSEE: FLORIDA POWER AND LIGHT COMPANY SITE: TWO NUCLEAR UNITS

1. OPERATING

2. UNDER CONSTRUCTION (36%, FLD 10/82)

LOCATION: 12 MILES SE OF FT. PIEPCE, FLORIDA REACTOR: COMBUSTION ENGINEERING- PWR 2560 MWT 802 MWE INITIAL CRITICALITY: APRIL 22,1976 COMMERCIAL OPERATION: DECEMBER 21, 1976 CURRENT CYCLE: CYCLE #4 (16-MONTH OPERATING CYCLE)

    . CRITICALITY THIS CYCLE:    MAY 7,1980 RESTARTED:   JUNE 30, 1980 AT 100% JULi' 1,1980 CONDENSER COOLING:   ONCE-THROUGH HEAT SINK: ATLANTIC OCEAN l

I 1

                                       ~

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1

1 l EVENT DESCRIPTION l l INITIAL CONDITION: FULL POWER l TIjj{ EVENT / ACTION 0226 HCV-14-6 FAILEC SHUT l SHORTED SOLENCID TERMINAL BOARD LOST CCW FLOW TO ALL RCPS 0233 MANUAL REACTOR TRIP 0235 STOP ALL RCPS 0238 START 181 RCP TO ENCHANCE NATURAL CIRCULATION 1 0239 STOP 151 RCP 0300 START NC C00LDOWN CDR: 60-70F/HR 0350 OPEN HCV-14-6 AIR LINE JUMPER 0600-0630 DEPRESSURIZE 1140 PSIG TO 690 PSIG CHARGE VIA PZR AUX SPRAY LINE 061S PZR LEVEL VARIATIONS NOTED 0630 - 1230 ALTERNATE CHARGING FLOW BETWEEN PZR AUX SPRAY AND LOOPS - II-l

i TIME EVENT /A TION 0700-0730 INDICATIONS OF VOIDING OTHER THAN IN PZR (SUSC00 LING 200-150F) 1051 START LPSI 15 IN SHUTDOWN COOLING MODE 1 1227 START LPSI 1 A IN INJECT!m.' N.'LE TO GO SOLID 1 1230 PZR LEVEL HOT PEGGED HIGH) l COLD STEADY AT 64%) I NOTED ANOMALOVE SOLID PLANT RESPONC-(PRESSURE CONSTANT WHILE CHARGING AT 88 GP.M)

 -1300      SHUT LPSI 18 MINIFLOW ISCLATION VALVE                                      ;

(FOUND CRACKED OPENI ' 1357 NOTED SLIGHT RISE (RAMP) IN RWT LEVEL STOP LPSI 1A SHUT MINIFLOW LINE MOVS SL!GHT RISE IN PZR LEVEL (COLD) AND PZR PRESSURE 1500 DPAW STEAM BUESLE IN PZR, DRAIN TO INDICATING RANGE 1600 NORMAL RESPONSE TO CHARGING AND LETDOWN VARIATIONS 4 II-2 l l l i l l l l

                      - [2.5 c   E.                                                        l 1
                                                                                      )

REACTOR COOLANT PUMP SEAL PERFORMANCE e ERRATIC CONTROLLED BLEEDOFF FLOW INDICATION e N0 VAPCR SEAL LEAK INTO CONTAINMENT e VISUAL INSPECTION RESULTS NO DAMAGE SLIGHT HEAT CHECKING e ALL SEALS REPLACED e SEAL INFORMATION BYRON JACKSON CONTROLLED BLEED 0FF TO VCT 3-STAGE SEAL PLUS VAPOR SEAL e SEAL REPLACEMENT: RCP~lAl APRIL 1977 (LOSS OF CCW) RCP 1A2 APRIL 1977 (LOSS OF CCW) RCP 181 NOVEMBER 1978 (SUSPECTED CAUSE OF MOTOR PROBL EMS ) RCP 182 OCTOBER 1979 (PLANNED MAINTENANCE) l III-1 A - 2;:t

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           /            hg                           UNITED STATES
         !;      ,       j               NUCLEAR REGULATORY COMMISSION
         -                 c          ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 b         #

July 17, 1980

              %e APPENDIX XXV 3UMMARY OF COMMISSIONER HENDRIE's COMMENTS ACRS Members                                2    JN TREATMENT OF CLASS 9 ACCIDENTS ACRS Technical Staff                                                    -

COMMENTS BY COMMISSIONER HENDRIE ON TREATMENT OF CLASS 9 ACCIDENTS Commissioner Hendrie at the Joint Commission /ACRS Meeting on July 11, 1980, in response to a question from Dr. Xerr on treatment of Class 9 accidents in the NRC decision making process, made the following comments:

1. The present design basis concept ought to be retained but ought to be aimed at severe core accident events. Core melt events should be reduced to some prescribed level in some quantitative set of terms. Beyond the prescribed level you should look at reasonable and practical measures to limit the consequences, taking into account probabilities and effort expended.

2. The measures to be used beyond the design basis should be more flexible than those for design basis and should be based on best engineered design calculations. It is impractical to try to draw the design basis for all events since one can always find events not covered.

3. The safety objective for the design basis for accidents causing severe core damage should be a fairly low risk level. A further safety objective beyond the design basis ought to be expressed in terms of likelihood of exposure to the general public, taking evacuation measures into consideration. Once these objectives are developed a framework within which to fit such things as  !

hydrogen evolution in severe core damage accidents and what to ' do in general with degraded core measures. The objectives would also be helpful in providing a framework to think about implementa- , tion of the emergency planning rule and would provide helpful l background for new siting criteria. Without such a safety princi- l ple or objective the NRC will to' continue to carry en individual efforts and not have any consistency among those efforts. Please let me know if you would like to have a cooy of the transcript of the meeting whicn provides slightly more detail. G. R. Quit schreiber, Chief Project Review 3 ranch No. 2 i cc: D. Bessette U 0. Johnson ! M. Griesmeyer l I. Catton j J. Lee ! W. Stratton ! R. Saale --@ ] i

l N U REG-0699 i l

  .s Comments on the NRC Safety Research Program Budget for Fiscal Year 1982 ateNbhe           u 1980
      .. Advisory Committee on Reactor Safeguards
         . U.S. Nuclear Regulatory Commission
     . Washington, D.C. 20555
              ,p"'em,
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I l l l l 1 l d'ys>*\ REQUg#'o UNITED STATES

                    ~,,
       !"     ,       g                NUCLEAR REGULATORY COMMISSION g               c            ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g                        WASHINGTON, D. C. 20555 O,

July 17, 1960 Honorable John F. Ahearne Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Dr. Ahearne:

The Advisory Committee on Reactor Safeguards submits herwith its comments on the budget for FY 1982 of the Office of Nuclear Regulatory Research. Only that portion of the budget relating to Program Support has been con-7 sidered. The funding levels considered are those allocated by the EDO s~ Staff in its preliminary markup of 2 July 1980 and those requested by RES in its reclama of 9 July 1980. Consnents on personnel requirements and allocations are included in a few instances where particularly appropriate. Sincerely, Milton S. Plesset Chairman

Attachment:

NUREG-0699

s. .-

 . -       - , ~                                         _                                                .                       __         __ __

1 l l 1 1 TABLE OF CONTENTS l i Page 1 PART 1: GENERAL C0MMENTS........................................ 7 TABLE 1................................................. PART II: SPECIFIC C0MMENTS....................................... 11 CHAPTER

1. LOCA AND TRANSIENT RESEARCH................................. 13 1.1 Introduction........................................... 13 1.2 Semiscale.............................................. 13

1. 3 Separate Effects Experiments and Model Development..... 14 1.4 3-D Program............................................ 15

1. 5 Code Improvement and Ma i ntena nce. . . . . . . . . . . . . . . . . . . . . . . 16 1.6 Ccde Assessment and Application........................ 16

1. 7 Fuel Behavior Under Operational Transients............. 16 1.8 Core Damage Beyond L0CA................................ 16 1.9 PBF Operations......................................... 17

2. L0FT........................................................ 19 2.1 Introduction........................................... 19

2. 2 The LOF T Te s t P rogram. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 19 i 2.3 R e c o m me n d a t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3. PLANT OPERATIONAL SAFETY.................................... 21 3.1 Introduction........................................... 21

3. 2 Aan-Machine Interface.................................. 21

3. 3 Inst rumentation and El ectrica l . . . . . . . . . . . . . . . . . . . . . . . . . 22

3. 4 P l a n t Sy s t ems B e h a v i o r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23

3. 6 Aechanical Components Safety........................... 23

3. 6 St ruc t u r a l Sa f e ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24

3. 7 Fracture Mechanics..................................... 26

3. 3 Ope rat i ng Ef fects On Ma t eri al s. . . . . . . . . . . . . . . . . . . . . . . . . 25 3.9 dondestructive Examination............................. 25 4

v

l CHAPTER Page

8. SYSTEMS AND REL I AB I L ITY ANALYS I S. . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 8.1 Introduction........................................... 41

8. 2 Methodology Development................................ 42 1

8. 3 Reliabili ty and Human Error Data Analysi s. . . . . . .. . . . . .. 43 8.4 Systems Analysis....................................... 43

8. 5 Consequences Analysis.................................. 44

9.

SUMMARY

..................................................... 45 B I B L I OGR AP H Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 GL0SSARY........................................................... 49 m vii

I l PART I

GENERAL COMMENT

S f i l i s-

PART I: GENERAL COMtENTS

1. Introduction The FY 1982 safety research budget is being formulated during a partic- l ularly complex era. On the one hand, there exists a large array of re-search needs for operating reactors and for reactors to be constructed that arise directly or indirectly from the implications of the Three Mile Island, Unit 2 (TMI-2) accident. On the other hand, national economic conditions place considerable emphasis on the need to control government expenditures. If those safety resea rch areas which are judged to have potentially the greater impact in protecting the public health and safety are to receive the necessary priority, several steps will need to be taken, including the following:

          .      The NRC will have to provide policy guidance on the major open safety issues.
          .      The user offices will have to reevaluate their approach to formulating requests for research and strive to consider these in some broad framework which takes into account the major issues confronting the agency.
          .      Nuclear Regulatory Research (RES) will have to reevaluate its current and proposed programs in terms of risk-reduction poten-
 ~

tial and major regulatory needs.

          .       The NRC will have to judge whether some research, particularly that which involves large scale component testing or the appli-cation of existing methodology, should be the responsibility of the it dustry rather than of the NRC.
          .       The NRC may have to reduce sharply son. research wnich is merely confi rmatory in nature where there is good reason to believe that the current regulatory requirements provide adequate pro-tection to the public.

We elaborate on some of these points below.

2. TMI-2 Accident Related Research Needs For operating reactors and those under construction, the principal study areas that have come to the forefront following the TMI-2 accident include the following: the accomplishment of highly reliable shutdown heat re-moval; the study o.f anomalous transients and small loss of coolant acci-dents (LOCAs); the improvement of operator capability to understand and 3

respond to transients and accidents;. a reexamination of the overall _ design ~~ adequacy with regard to the possible existence of relatively high prob-ability accident sequences; and measures to deal with or mitigate degraded core and core melt accidents. For reactors yet to be constructed, the additional issues of importance include the following: siting issues; the development of new general design criteria, for example, to deal with i nadequacies in the single-failure criterion and with any new 1RC policy on core melt accidents; L possible major changes in system design approach, . such as a dedicated, bunkered, shutdown heat removal system, vented-filtered contain.nent, or other similar features; and a long-range NRC philosophy toward standard reactors. tiany of these topics require policy guidance from the NRC if an effective and timely NRC safety research _ program is to be implemented.

3. Reevaluation of Priorities for User Needs In NUREG-0603, we recommended that the user offices give early attention to an evaluation of the priorities of their existing research requests in the light of their changed perceptions of safety research priorities.

The NRC has established a procedure which requires that the user offices request or endorse most of the safety research program; for this and other reasons, it is important that these offices devote adequate atten-tion to assessing their current and future safety research needs. The - user offices have provided some comments on the proposed research program. However, we are not satisfied that this matter has received the needed attention, and recommend that the user offices devote the effort needed to develop a cohesive set of research requests fonnulated with .the neces-sary perspective and within some broad framework of regulatory needs.

4. Reevaluation of Research Priorities We and others have recommended that RES apply the methodology of risk assessment to its own program in order to define those areas having 'the greatest potential for improving the protection of the public health and safety. We recommend that RES give priority to such an effort during the next few months, both to provide an improved basis for setting priorities for FY 1982 and for an evaluation of possible changes in priority and ,

funding level for FY 1981. l We recom.nend also that the NRC develop criteria for when safety research should be done by industry.

5. Class 9 Accidents The general subject of Class 9 accidents, including but not limited to  !

the proposed rulemaking on degraded cores and core melts, introduces a l very important and complex research area. During the past several months, there has been developing a considerably expanded effort compared to the limited program previcusly pursued. 4 J

   - - ., , . . . - , .      -,    ,.n..  -..-v     .   --

 .["um However the research program needs to be geared to providing that infor-mation most important to the NRC decision-making process as expeditiously as possible, and the appropriate resources should be assigned, not only        !

in FY 1982 but earlier in FY 1981. It is therefore of overriding impor-tance that policy guidance from the NRC and additional participation by-the user offices be made available to RES at an early date. We believe that the proposed level of effort may fall short of what the NRC needs. For example, rather than a program that consecutively examines the dif-ferent containment ' designs such as the large dry pressurized water reactor (PWR) containment, the ice-condenser containments, and the different boil-ing' water reactor (BWR) contairunents for hydrogen control and core melt. RES should be addressing all of those containment types concurrently, by examining realistic design approaches.

6. Other Areas Requiring Emphasis The NRC research program currently includes major expenditures for re-search on _ the large LOCA and for confirmatory research intended to dem-onstrate that the current regulatory requirements are adequately conserv-ative in areas where this is quite likely to be the case. On the other hand, the current research program, and that proposed for FY 1982, lacks sufficient empnasis in many areas where either there are large uncer-tainties or there is reason to expect that a significant improvement in safety may be achievable. We believe that the FY 1982 program (and the FY '1981 program, as practicable) should be reoriented to provide appro-priate emphasis on topics such as the following:

a) The proposed program includes considerable growth in areas re-lated to operational safety. However, it still lacks signi-ficant, cohesive research on light water reactors (LWR) plant operational behavior as a function of design and control. b) The impact of control systems and other nominally non-safety systems on safety has become a matter of increasing interest. A research program devoted to this matter should be formulated. c) To complement the research program on operator error, a re-search program should be initiated to evaluate the effect of design errors on LWR safety and to provide a basis for the development and application of improved approaches to reduce the impact of design errors on safety. We recommend that such research be initiated in FY 1981 and given strong support in FY 1982. d) The General Design Criteria should be reexamined, using among  : other things, probabilistic methodology, for the purpose of developing improvements in the current criteria.

                                                                                                             ^'

In view of the above . recommendations, we believe that a budget level of ,~ about $265 million for research program support is- required for FY 1982.

                ,             This recommendation is based on the assumption that the needed large shifts in programs and priorities will be made in the program description provided to us by RES during our review of this. subject.

7. Specific Reconrnendations The succeeding numbered chapters of this report contain our recommenda-tions regarding - the programs and . funding levels for each of the eight decision units .of the RES. budget for FY 1982. The funding levels . re-ferred to, and given in Table 1, are those requested by RES and those resulting from the ED0's preliminary markup as of 2 July 1980.

We note with approval that the FY 1982 budget request has been presented in eight decision units rather than the fourteen used for the FY 1981 budget. The proposed regrouping of program subelements is more logical and coherent, more representative of the program objectives, and more amenable to effective management. There is no longer a separate decision unit for Improved Reactor Safety; the several program subelements that were formerly in this category have been distributed to the appropriate decision units of the new format. This change is responsive to and con-sistent with the recommendations in Chapter 15 of our report to the Con-gress on the FY 1981 budget (NUREG-0657). Specific comments on the levels of funding f or each decision unit and, in ~ genera l , for each subelement are given in Part II, which follows. The reconnendations of the ACRS are included in Table 1. ~ 6

           ._ ~          ._ _

TABLE 1 PROGRhi SUPPORT BUDGET FOR FY 1982 (IN MILLIONS) EDO RES ACRS MARK REQUEST RECOMMENDATION 7/2/80 7/9/80 7/12/80

1. LOCA AND TRANSIENT RESEARCH

a. Semiscale 7.5 7. 5 7.5

b. Separate Effects Experiments and Model Development 5. 7 7.8 7.8

c. 3-0 Program S.O 6.0 5.0

d. Code Improvement and Maintenance 4.5 4.5 4.5

e. Code Assessment and Application 7.9 7.9 7.9

f. Fuel Behavior Under Cperational Transients 6.4 6. 4 6.4

g. Core Damage Beyond LOCA 11.1 12.1 11.1

h. P8F Operations 4.6 4.8 4.8 52.9 57.0 55.0

2. LOFT

a. Engineering and Analysis 10.4 10.4 10.4

b. Fuel 4.5 4. 5 4. 5 c, Instrumentation 10.0 10.0 10.0

d. Operations 9.5 9. 5 9. 5

e. Facility Support 13.6 13.6 13.6 48.0 48.0 48.0 l

l I i l

l 1 l l EDO RES ACRS MARK REQUEST RECOMMENDATION 7/2/80 7/9/80 7/12/80

3. PLANT OPERATIONAL SAFETY

a. Man-Machine Interface 4.8 4.8 4.8

b. Instrumentation and Electrical 7. 3 7.3 7. 3

c. Plant Systems Behavior 1.5 1.5 1. 5

d. Mechanical Components Safety 8. 4 9.0 9.0

e. St uctural Safety 5. 5 6. 5 6. 5

f. Fracture Mechanics 4.5 6. 0 6. 0

g. Operating Effects On Materials 7.6 7. 6 7.6

h. Nondestructive Examination 3.4 3.4 3. 4 43.0 46.1 46.1

4. SEVERE ACCIDENT PHENOMENA AND MITIGATION RESEARCH

a. Fuel Melt dehavior 9.0 10.5 *

b. Fission Product Release and Transport 4.3 4.3 ) 18.7

c. Severe Accident Mitigation 3. 9 3.9 /

d. Fast Breeder Reactors 0 8.0 17.5

e. Advanced Converter Reactors 0 2.0 1. 3 17.2 28.7 37.5 8

i l EDO RES ACRS MARK REQUEST RECOMMENDATION 7/2/80 7/9/80 7/12/80

5. SITING AND ENVIRONMENTAL RESEARCH

a. Seismology and Geology 3. 5 5. 3 5.3

b. Meteorology and Hydrology 2.0 2. 0 2. 0

c. Airborne Effluents-Environmental Impacts 2.3 2. 3 2. 3

d. Aquatic Effluents-Environmental Impacts 1.8 1. 8 1. 3

e. Occupational Exposures and Health Effects 3. 6 3. 6 3.6

f. Socioeconomic Impacts 0. 5 0.7 0.7

g. Siting Alternatives 0 0.4 0. 4

h. Emergency Preparedness 0.5 0.5 1. 0 14.2 16.6 17.1

6. WASTE MANAGEMENT

a. High Level Waste 16.3 16.7 16.3

b. Low Level Waste 5. 5 5. 5 5.5

c. Uranium Recovery 3.0 3.0 3. 0 24.8 25.2 24.8 l

l l 9 1

1 1 I EDO RES ACRS MARK REQUEST RECOMMENDATION 7/2/80 7/9/80 7/12/80

7. SAFEGUARDS AND FUEL CYCLE SAFETY

a. Physical Protection 3.1 3.1 3.1

b. Material Control and Accounting 1.4 1. 7 1. 7

c. Threat and Strategy 0.4 0. 4 0. 4

d. Fuel Cycle Facility Safety 1. 3 2. 0 2. 0

e. Decommissioni ng 1.o 1.6 1. 6

f. Transportation 0. 8 0.8 0. 8

g. Effluent Control 1. 2 1.2 1.2

h. Product Safety 0.3 0.3 0. 3

i. Occupational Protection 0.6 0.6 0. 6 10.7 11.7 11.7

8. SYSTEMS AND RELIABILITY ANALYSIS

a. Meteorology Development 5.0 5. 7 5. 7

b. Reliability and Human Error Data Analysi s 2.3 3. 5 J. 5

c. Systems Analysis 10.4 13.1 13.1

d. Consequences Analysis 1.2 2. 5 2.5 18.9 24.8 24.8 I l

TOTAL 229.7 258.1 265.0 l 10

6 m e O ,A>> .s- w l l I 1 l J l PART II SPECIFIC COMMENTS i I

 % d' 11

1. LOCA AND TRANSIENT RESEARCH l 1

l l 1.1 Introduction This item . includes several programs which are directed toward improved understanding of reactor behavior in large break LOCAs and small break LOCAs, and there has been extensive reorientation of the program to empha-size the latter. In the past by far the greatest attention was given to large break LOCA problems. We strongly support the change of emphasis. l Also included here is the improvement and assessment of codes which have j as their objective an analytic description and understanding of LWR tran-sients. The last group of programs in this item are directed toward the understanding of fuel and core behavior under conditions in which the core is inadequately cooled. Comments on these programs follow. 1.2 Semiscale (Item 1.a*) This facility has shown itself to be increasingly useful as an experimen-tal tool for contributing to an understanding of PWR transients. RES has undertaken a serious study of the limitations and scaling questions which arise in translating observations in Semiscale to full scale. We strongly approve and commend this. effort. There are some modifications in Semiscale which should receive comment. One modification, Mod 2A, is already under-way and includes improved simulation of the Westinghouse type PWR. The f acility will have tvn steam generators with correct, full-heignt geometry, a highly desirable modification in view of our strong interest in natural circulation and reflux-boiling heat transfer. The modification will in-clude a new pump, upgraded instrumentation, and a new core. Most essen-tially, a strong effort will be made to improve the thermal insulation in the facility. In view of the improved data that may be obtained from Mod 2A, we view its cost as moderate and support this effort. A second modification of Semiscale which is under consideration is Mod 5 which would have a substantially higher cost of about $10 million. This modification would be directed toward the simulation of a Babcock and Wilcox (B&W) PWR and will involve not only a different core but two once-through steam generators. An integrated control system would be installed, , and a new vessel would be required with vent valves and proper upper head geomet ry. The central question relating to this modification is its cost effectiveness and its potential contribution to code development for the description of transients which are peculiarly characteristic of B&W type plants. We view the Mod 5 program favorably since it will contribute to code assessment. As will be noted below, RES in general tends to under-estimate the needs for code assessment. We support the amount requested by RES with a high priority. These item numbers refer to the decision unit subelements in Table 1. m 13

                                                                                                   .~

1.3 Separate Ef fects Experiments and Model Development (Item 1.b) There are several programs grouped in this item which deserve some sepa-rate discussion. One of these is the Two Loop Test Apparatus (TLTA) which is presumed to do for BWRs what Semiscale does -for PWRs. TLTA, however, is entirely. inadequate for the purpose and should not be. used to relate to licensing problems or to code assessment for BWRs. The facility _ is particularly misleading in applications to small break LOCAs. Si nce results from the present TLTA cannot be used for code _ assessment, we strongly recommend that no further work be done with the present facility and - that it be replaced with a new facility. The cost of a useful,'new TLTA depends very strongly upon the decision whether it should contain one fuel bundle, or several. The power requirements for a facility with more than one bundle are so large that proper steady-state conditions before initiation of a transient may not be attainable. The need for more than one bundle arises from concerns regarding asymmetry effects. Such asymmetries could very possibly be studied in a proposed Japanese facility, ROSA IV, in which case a single bundle TLTA facility would be acceptable. We support the construction of such a facility and urge its early implementation. Another facility related to BWRs is the Steam Sector Test Facility which has the objective of studying BWR core spray behavior. The program is directed toward large break LOCAs and cannot readily be reoriented to small break problems. We recommend that the program be phased out in FY 1982. Other programs in this area are FLECHT-SEASET at Westinghouse and Thermal Hydraulic Test Facility (THTF) at ORNL. FLECHT-SEASET has been reoriented to examine natural circulation. The facility has good steam generator representation and may be useful for code assessment. The facility is, however, limited to low pressure and the program should be phased out in FY 1982. The THTF at ORNL has not been productive of useful data and should be terminated at the earliest possible date. Tests are scheduled to end in FY 1980 and further expenditures should also end at that time. We believe that the present effort in code assessment is inadequate. For-an improved program additional experiments on separate effects are re-qui red. Many of these experiments would not requi re large facilities since the experimentation should be directed toward getting basic physical and engineering bases for the codes. RES should extend its efforts in this direction as soon as possible. The model development program consists largely of relatively small proj-ects in various university laboratories. We strongly endorse this kind of program as being useful, productive, and cost effective. At the same time the program provides a helpful interaction with an important part of the engineering and scientific comunity which should be extended and in-creased. This program suffers from the bureaucratic difficulties of get-ting contracts. The NRC should make a serious effort to resolve this dif-ficulty. 14 i

As regards the budget request, we strongly support, with a high priority, the RES request for $7.8 million. Any further reduction in the original request would seriously jeopardize the program. 1.4. 3-D program (Item 1.c) This program is an international one involving Japan, tne Federal Republic of Germany (FRG), and the United States and was begun when LWR safety re-search was preoccupied with large break LOCAs. It was with this problem in mind that two large facilities were designed and built in Japan. One, the Cylindrical Core Test Facility (CCTF), will be completed shortly. The second facility, the Slab Core Test Facility (SCTF) is under con-struction. Both f acilities are limited to low pressure. Ar, experimental program is planned in CCTF on natural circulation, and two-dimensional effects in core refill following a large LOCA are planned in SCTF. Both facilities, .within -their capabilities, are well constructed but suffer from an insufficient test engineering staff. It would be productive if the NRC would arrange for the assignment of two or three researcn engi-neers from the U.S. to facilitate the effort. There was early appreciation in Japan of the low-pressure limitations of CCTF and SCTF, and they have undertaken the construction of a high-pres-sure facility, ROSA IV. This facility could be of such importance to the U.S. reactor safety program that NRC could readily justify the assignment in Japan of several engineers to participate in its design. The German effort in the international 3-D program will consist primarily in the construction of the Upper Plenum Test Facility. This facility will presumably provide some information relating to special questions regarding large LOCAs. One is the so-called ECL bypass question which ceased to be of concern many years ago. The facility will also make some contributions to the interaction of hot leg injection with steam upflow through a core. This question relates to a special feature of German PWR design. We suggest that the NRC attempt to secure a redirection of the German program. The U.S. contribution consists of two programs. One of these is the sup-ply of experimental measuring devices for these large foreign research facilities. We have for some time urged the development of new and im-proved instrumentation which could be installed in operating power reac-tors and would encourage some contributions from the 3-D program to this end. The second contribution from the U.S. consists in applying a bridge be-tween the various tests in Japan and FRG by means of the TRAC computer code. RES should carefully consider whether this computational effort contributes effectively to the basic requirement of a useful code descrip-tion of nuclear power plant transients. As regards the budget for this subelement, we endorse the EDO funding level of S5 million. 15

1.9 PBF Operations (Item 1.h) PBF is the only reactor in the country that is dedicated to studying the behavior of fuel in operational and short-period transients. It has provided useful licensing information. We continue to believe that its longer-term usefulness will depend on the new roles it may find in the study of fuel behavior under accident conditions. We support the funding for this subelement with the provisions expressed in Sections 1. 7 and 1.8 above concerning the future role of PBF. 1 I 17 l

 . .   -       -.-       -..-     . _-          .-.-       -    .~   . . -.             . .

I 1 l

2. LOFT 2.1 Introduction The_ LOFT facility is the only integral facility which models a PWR. The shortcomings of the. facility are well known and relate for the most part The nuclear core is slightly to deficiencies in vertical dimensions. This reduced height introduces less than half the height of a PWR core.

some uncertainty in translating the early quench observed in the large LOCA test in LOFT to a full-size system. Further, the height relation-ship between the core and the steam generators affects the interpretation of measurements of natural circulation heat transfer. 2.2 The LOFT Test Program LOFT . tests were for some time directed toward a design basis accident involving the instantaneous double-ended cold leg break (DECLB). Tests of this type have contributed to the understanding of this kind of acci-dent and also have contributed to code assessment. In response to e strongly modified view of more immediate needs, the LOFT program wa3 redirected in FY 1980 to the study of reactor transients which were the result of small breaks. The current plans call for further tests of this kind in FY 1981. Both the FY 1980 and the FY 1981 programs as now planned include other. types of transients, i ncludi ng, particularly in FY 1981 tests concerned with anticipated transients without scram. The signifi-cant test proposed for FY 1982 is a DECLB at the higher core power of 16 kw/ft. No further small break tests are scheduled for FY 1982. A test has been proposed for FY 1983 with pressurized fuel. Although we believe that LOFT will essentially complete its NRC mission in FY 1982 with NRC funding phased out at the end of FY 1982, the LOFT System could still be a valuable tool for the nuclear power industry. The LOFT installation could be of fered to the nuclear industry to be operated with industry financial support as a facility wnich would en . hance operational capabilities of the nuclear industry. 2.3 Recommendations LOFT represents the largest single expenditure in the safety research budget so that its program must be considered with special care. We recommend that the tests through FY 1982 be adequately funded and that following the 1982 tests the f acility be decommissioned unless it is taken over by the nuclear industry. The final tests to be run to the completion of the program should be carefully scrutinized and evaluated by RES to obtain the most useful final series. We would also wish to contribute to the choice of these tests. Efficient operation of the facility appears to require the requested level of support and therefore we endorse that level. i 1 1 19 ' 1

3

3. PLANT OPERATIONAL SAFETY 3.1 Introduction-The RES request for FY 1982 is consistent with that' considered necessary and desirable for providing guidance for standards which licensees and applicants need for improvement .of operation and maintenance of reactors in a safe and reliable manner. Funds requested have been increased over FY 1981 but are considered appropriate to provide the programmatic effort for NRC to demonstrate tne leaderchip and guidance for correction of.

deficiencies in reactor operations indicated by lessons learned during the past year. The level of funding requested by RES for 1982 is $46.1 million. We support this level; however, a close review of the functional listing of programmatic items indicates that the allocation of requested funds may be improperly prioritized. Specific comments on the program subelements follow. 3.2 Man-Machine Interface (Item 3.a) The requested level of funding for this subelement is $4.8 million. The work on developing improvements in instrumentation and information display

                    - is expected to have progressed by FY 1982 to a point where firm recom-mendations can be made for status monitoring and diagnostic display re-                    ,

quirements. A systematic study is expected to be completed on instrumen-tation to follow the course of an accident and specific recommendations provided. Initial simulator experiments to study operator behavior should , be completed. Utility response to training requirements should be evaluated in FY 1982. A program of human factors measurements and improved instrumentation and , control displays will be continued through FY 1983/1984. Detailed accoin-plishments for that program are as follows:

                          .        Analyze the responsibilities of plant personnel witn respect to normal and .off-normal operation, inspection, testing, maintenance, and design.
                          .        Relate these responsibilities to associated selection, training and management requirements.
                          .        Analyze accident sequences to identify operator i nformation re-quirements and to identify improvements in emergency response procedures.
                          .        De velopment and feasibility testing of concepts for computerized             i display and diagnostic systems.

21

                 . Conduct experiments to test the effectiveness and reliability of-proposed improvements in display and diagnostic systems.
                 . De velop design requi rements and regulatory review critical for operational aids to reactor operators and other plant personnel.
                 . Assess the net effect of such systems on risk.
                 . Determine the feasible and effective improvements in the capa-bilities and utilization of training simulators.

The man-machine interface programs have been initiated either as a result of NRR user requests, in response to the Congressional request for im-proved reactor research or as a direct result of the TMI action plan. These progrants will provide data and information which will assist NRh in strengthening and revising license requirements to improve safety and re-duce risks. These programs are considered important to plant operational safety and should be continued and expanded within reasonable manpower and equipment c resources. 3.3 Instrumentation and Electrical (Item 3.b) The requested level of funding for the subelement is $7.3 million. Ad- 1 vr.nced two-phase instrumentation to follow the liquid level in nuclear power plants will be tested for possible use to alleviate TMI-type prob-lems. Work on fire protection research concerning fire suppression , systems which has been endorsed by NRR is scheduled for completion in FY ' 1982 and full scale replication tests of actual cable area configurations are scheduled to be in progress. LOCA tests will be completed in the qualification testing program and work will be initiated to address safety concerns from the environment of non-large LOCA accidents. Qu al i-fication testing and postmortem analysis will be perfonned on TMI-2 i n conjunction with DOE sponsored programs during plant decontamination and recovery in FY 1982/1983. A system review of generic safety related instrumentation and electrical equipment to identify the ability to withstand temperature and steam conditions, basic design problems, fabrication problems, wear, ' aging and other reliability problems should be initiated in FY 1982. The fire protection and qualification testing programs are in support of SD's programs to develop regulatory guides and standards. In addition, tasks within these programs have been carried out at the request of NRR to investigate existing plant installations for adequacy. We believe, however, that the expense of fire replication tests is far too great for the infonnation to' be obtained and do not support this particular part of the program. Industry should be encouraged to perfonn more confirmatory

         -testing of fire protection concepts.

22

l l l The initiation of new programs, including problems with safety related instrumentation and electrical equipment , software verification and the  ; study of nuclear plant electrical supply design problems , lead to in- l creased funding needs. These areas have been identified in inspection  ! and licensee event reports (LERs) as significant contributors to plant incidents. Work on these problems offers a potentia 1 for reducing the level of risk f rom accidents, will contribute to improve safe plant operations and should be supported with additional funds or diversion of funds from other programs which we do not support. i 3.4 Plant Systems Behavior (Item 3.c) RES has requested funding for this subelement at the level of $1.5 mil- , lion. The test of a continuous on-line surveillance system to show how pattern recognition can be used to alert plant operators of anomalous conditions . is expected to have completed its demonstration phase at the TVA Sequoyah plant by FY 1982. A significantly increased effort on assessing nuclear plant operational behavior should be initiated in FY 1982. This effort will include assessments of operational transients on system behavior, the safety consequences of shared systems within a plant and facility design requirements for safely coping with accident condi-tions. These programs demonstrate and develop diagnostic tools which will contribute to operator knowledge of plant conditions. Within the small fund allocation, priorities should be carefully reviewed to obtain the maximum benefit expected from each program task.

       ~

3.5 Mechanical Comoonents Safety (Item 3.d) The requested level of funding for this subelement is $9 million. The Seismic Safety Margins Research Program (SSMRP) derives its support in large part from this portion of the subelement on Mechanical Compo-nents Safety and from a similar portion of the subelement on Structural Safety. We continue to support the SSMRP and recommend it be funded at the requested level for FY 1982. We reiterate our recommendation made previously in NUREG-0657 that the SSMRP be structured to provide input as early as is feasible into the broad safety policy considerations concern-ing the seismic design bases of nuclear power plants. This should include a timely preliminary evaluation of the seismic contribution to the prob-ability of serious accidents and the principal contributors to uncertainty in such probability estimates. We hope to see significant results per-taining to these matters by the end of calendar year 1980. The goal of the other programs in this subelement is to determine and en- I hance reliability under various accident and operating conditions; however, a successful approach still needs to be developed. It should begin with a . definition of the NRC problems to be solved and the criteria to be used. l Currently, considerable emphasis is being placed on seismic impact on l mechanical components. Clearly, the great majority of potential accidents l l 23 l l l l L . - . - . _ .-.

_ _ _ _ _ _ _ . _ _ . . . _ . _ _ . . _. _ . . _ ~- _ _ __ 1 i and . reliability problems in the life of a reactor do not involve earth-l quakes, and those .that do are covered under another portion of the pro-gram.: A great deal of industrial experience exists with.many of the . I components in' question and the program could profit greatly if this expe- { rience could be utilized. 1 3.6' Structural Safety (Item 3.e)  ; The requested level' of funding for this subelement is $6.5 million. This program is well defined and well balanced among the several identified needs. We support funding at the requested level and offer the following - comments: l

                  .       The program is oriented strongly toward questions relating to                        !

the safety of operating plants.

                  .       Major emphasis is given to seismic-related problems, as is appropriate for structural safety.

The research on flood effects and hazards is long overdue but now appears to be headed in the right direction. Some increase in this effort in both FY 1981 and FY 1982 would be warranted. The program for international cooperation is essential and should provide much useful information at low cost. It is im-portant, however, that most of this effort should be conducted by RES rather than by independent contractors, and suitable allocations of manpower and travel funds should be made to per- ' mit this mode of operation. The division of seismic research programs, including the SSMRP, between the Structural and Mechanical Engineering Branches, re-quires special attention by RES to the interfaces between these programs. i The nature of many of the problems related to structural safety is such that special attention should be given to the question i of whether the needed research should be done by the NRC or by  : the industry.  ! RES should maintain cognizance of the structural research being done by industry and should be in a position to utilize the results of it to the greatest extent practicable. A significant portion of the research in this program is to be done by independent contractors rather than by National Labo-ratories. The results of this action, in terms of cost, effec- l tiventss, and timing, should be evaluated as the program pro- l gresses. ' l 24

i l

      .      The proposed research to determine the effectiveness of QA procedures, especially nondestructive testing methods for concrete, appears at this time to have little research content and to be of dubious value. At the minimum, it requires further evaluation and definition.

3.7 Fracture Mechanics (Item 3.f) The requested = funding for this subelement is $6 million. This is a good , long-range . program that is providing a sound basis for decisions on the -l integrity of pressure vessels. It should continue. The question of l thermal shock in pressurized systems represents an important uncertainty to the integrity of the older reactor pressure vessels. This program has not been supported by NRR but should be actively pursued to provide a , basis for decisions in this area. In the piping area, RES should continue i I to work with NRR to define programs which will provide an acceptable basis for reducing the number of constraints on primary piping systems wnfle f maintaining adequate safety margins. 3.8 Operating Effects on Materials (Item 3.g) The requested level of funding is 57.6 million. The largest uncertainties in assuring the integrity of the primary pressure boundary are contributed by operating environment, radiation and water chemistry. The programs in this area address these issues in a sound, coherent manner. We look for-ward to a continuing definition and deployment of the new program on en-vironmentally assisted cracking. The study of the Surry steam generator will be valuable in providing information on the relation between eddy current indications and actual defects that will aid the NRC in its deci-sion on other operating steam generators. We are still concerned about the merits of the subsequent program and reiterate the comment made in NUREG-0603: "The work should be limited to the correlation between NDE indications and tube integrity until a careful study has indicated the positive contribution to be made by additional work." Substantial in-I dustrial participation in this program is to be encouraged and would aid in defining any future program. l l 3.9 Nondestructive Examination (Item 3.h) The requested level of funding is $3.4 million. Periodic inspection of reactor components are regularly carried out to assure that no dangerous flaws are . present. NRC must be capable of judging how reliable these techniques are and be able to develop criteria for the acceptability of new techniques. Several good programs are planned or in place to enhance this NRC capability. We are less certain that NRC should be funding several_ other programs which involve the development of new techniques to be used for inspections. i 25

l 1 l

4. SEVERE ACCIDENT PHENOMENA AND MITIGATION RESEARCH 4.1 Introduction Activities in this decision unit fall into two distinct categories, Severe Accident Phenomena, and Fast Breeder and Advanced Converter Reactors. The two will be discussed separately.

The work on Severe Accident Phenomena is closely tied to planned NRC rule-makings which will deal with degraded core cooling, power plant siting, and emergency planning. Because the rulemakings explore as yet uncharted regions, and because the NRC has as yet reached only preliminary positions concerning the rulemakings (especially in the area of degraded core cool-ing), it is difficult to judge whether the proposed work is appropriate to the needs. We continue to recommend, as we have done repeatedly in previous reports, that a viable program in Fast Breeder and Advanced Converter Reactors should be continued. 4.2. Fuel Melt Behavior (item 4.a) Fission Product Release and Transport (Item 4.b) Severe Accident Mitigation (item 4.c) These three subelements represent one of the highest priority research areas in the entire research program. The program should be fonnulated, structured, supported and directed in a manner such as to provide tne information needed by the NRC in its planned rulemaking which will dea! with degraded core cooling, in its actions on accident mitigation at hiy population density sites, and its efforts to provide a better understand ing of the course of severe accidents, an understanding which might be important in the unlikely event of a real accident. The research areas involved are many, challenging and complex. RES is to be comended for its efforts to generate a proposed research program in a situation in which a minimum of guidance has been provided by the licensing staff and the NRC. However, we do not believe that the program, as proposed, is likely to provide the information likely to be needed by the NRC for its decision making on these matters during the next few years. We recommend that a high level task force containing appropriate representatives of NRR, RES and S0 be established with the charter of recommending promptly the research program and resources required to meet the NRC needs. We also recommend that the NRC devote the necessary time to provide needed , insight on safety philosophy and objectives which should guide this work. l l We anticipate that the $18.7 million currently requested by RES for work l on subelements 4.a. . b, and c (which covers the LWR portion of this de- ' l cision unit) is likely to be insufficient when the program receives better definition and that the currently defined emphasis is likely to change markedly. l 27 l r y y9 -

1 l l I

                                                                                                                                                       ~

We suggest the following as a possible approach to begin defining the program:-  ;

                                    . Ascartain the major categories of information needeo                               including the following:

1) Indian- Point / Zion / Limerick /other high population density sites.

2) The rulemakings on degraded core cooling, on environmental im-pact statements, on emergency planning and on power plant siting.

3) The interim approach for small or low pressure containments.

4) Policy guidance for near tenn construction pennits.

5) The understanding needed to provide appropriate operator guidance, should a potentially serious accident occur.

                                   .      Define the informational needs for each major category and the rel-evant time scale for the information to be developed.

We foresee that a major i nformational need of the decision making pro-cesses will be for sufficiently detailed conceptual design studies of potential mitigating features for the various reactor / containment com-binations, including their costs, benefits, pros and cons, to enable proper judgments. Such studies should have a high priority and should be carried out concurrently. It is anticipated that high priority short-and long-term research needs will arise as a result of such studies and the overall program should be structured to have the flexibility and re-sources to pursue needed avenues expeditiously, as practical. We antici-pate that significant changes are likely to be desirable in the currently proposed research program as the result of such an approach. We believe that while the proposed funding level of $18.7 million cur-rently may represent a reasonable floor on support for this research program for FY 1982, it would be prudent either to request an additional

                                   $5 to $10 million or to assure that flexibility to pursue needed research is.readily achievable from related decision units.

4.3 Fast Breeder Reactors (Item 4.d) t Congress authorized $13.7 million for Liquid Metal Fast Breeder Reactor (LMFBR) research in FY 1980 and the House Appropriation Subcomnittee has authorized $11.1 million for FY 1981, whereas the NRC and ACRS endorsed a level of $18 million for FY 1981. However, RES nas proposed $8 million for F.Y 1982, while the E00 has proposed that no funds be provided or ex- g pended. It is expected by RES that 507, or more of its existing advanced reactor safety research resources would be redirected to resolution of degraded core cooling problems in LWRs.  ! 1 28

1 At the same time, significant design and developmental efforts in the , LMFBR area are underway outside the NPC. Department of Energy (00E) is performing the conceptual design of a 1,000 MWe LiiFBR plant (Conceptual Design Study) and intends to deliver a report to Congress next spring. 00E would hope to submit a Preliminary Safety Analysis Report (PSAR) on such a plant to NRC within a year of any Congressional approval. Clinch River Breeder Reactor (CRBR) design and procurement is proceeding, and DOE budget authority for FY 1980 is over $170 million. DOE is spending over $140 million in FY 1980 on breeder technology (including $36.5 mil-lion for LMFBR safety) and $76 million for Fast Flux Test Facility (FFTF) which achieved initial criticality this year and will likely achieve beneficial use next year. Other nations are pursuing comercialization of the LMFBR, and the French may market a 1,000 to 1,500 MWe unit by 1985 or so. All of the DOE effort cited above is proceeding with little or no input by NRC even though new safety concepts are under development and new safety precedents are being established. In NUREG-0657 and in other reports, we have consistently supported an NRC LMFBR research program " based on the perception that many of the current safety problems associated with LWRs have resulted from the fact that safety research lagged behind reactor development." We have said also that, if foreign LMFBR technology is likely to be imported in the next 10-20 years, "it is important that the NRC program of safety research on advanced reactors be maintained to ensure an adequate technical basis for U.S. regulatory standards, guides, and criteria." We reiterate our general support of such a program. Further, until a consensus is reached that the U.S. will not utilize LMFBRs, we believe it important that the NRC ensure that a sound, long-range, LMFBR research and licensing activity exists within NRC. We believe that the NRC should have an input to DOE activities such as the Conceptual Design Study and the design of CRBR while they are in progress and that it should have considerable liaison with the DOE technology and FFTF activities. Fu rt her, we believe that the NRC should endeavor to keep aware of the safety cri-teria and design features of foreign plants having commercial potential. Such efforts will require people versed in and active in both licensing and research activities, but no effort is made here to separate one func-tion from the other. However, it is important that these people cooperate closely and perhaps even be interchanged frequently. It is difficult for us to comment on a specific budget level because no one has prepared a budget incorporating the licensing activities sug-gested above. However, we believe that the level recommended for FY 1981 as adjusted for inflation (thus a total of about $17 million) will allow both new and continuing work. We believe that expenditures of this magni-tude are reasonable to complement a U.S. development effort which amounted to over $600 million in FY 1980. We believe that these funds should be set up in a separate account where they will not be utilized for LW3 safety or other work. We do not endorse a decrease to $8.0 million; such 29

a reduction is too drastic and will not support an effective program. We ~' recommend that at least $17.5 million be requested, not only because it is warranted, but also to minimize the possibility that Congress will act in such a manner as to' direct funds from other high-priority work rather than appropriating money specifically for LMFBR work. We continue to believe that greater priority should.be given to accident delineation, accident prevention, and studies of alternate containment systems than has been the case. We believe that the priority given to analytical code development, especially that for SIMMER, should be de-creased. We endorse the priority of the experimental program at SANDIA and the aerosol experimental and analysis programs. The aerosol work appears to need better coordination and focusing than it has received. 4.4 Advanced Converter Reactors -(Item 4.e) Advanced Converter Reactor research is centered almost entirely on gas cooled reactor studies, with about 75% of the effort directed toward Ft. St. Vrain and the remainder at more-generic topics. We support a level 'L of about $1.3 million which will allow for a continuation of the Ft. St. Vrain effort; as with the LMFBR funds, these should be set up in a sepa-rate account. Studies will be directed primarily toward the long-term degradation and strength of the graphite, techniques to measure this strength, helium-air mixing. under emergency cooling conditions, and fre-quency response to power variations in the system. m 30

   .m. ..w,.-.._   . -- ---.--    -.  ._.m-__ .._,~ , . ,,               .v.r    . - - - . . , . , .n _w_.,.n.- ,..m,  ,

_. , ..c., . . . _ , . ,..,r r.. y---

5. SITING AND ENVIRONMENTAL RESEARCH 5.1 - Introduct m This decision unit includes eight subelements ranging from seismology, geology, meteorology, hydrology and the movement of radionuclides through the environment to the assessment and evaluation of occupational exposures, the planning for emergencies, and the evaluation of related socioeconomic impacts.

5.2 Seismology and Geoloqy (Item 5.a) This subelement is devoted primarily to developing a better understanding of the seismic and geologic behavior of several important regions of the U.S., and is responsive to our recommendation of several years ago for such a program. The studies are of considerable importance to the estab-lishment of an improved seismic design basis for future LWRs and to' an assessment of the seismic safety of existing LWRs. The causes of the significant earthquakes which have occurred in historic times east of the Rockies are important to understand if the NRC is to avoid excessive conservatism on the one hand or a significant underestimate of seismic ri:;k on the other . hand. The northwest portion of the U.S. poses similar concerns. We continue to place high pri rity on this research category and urge that the requested RES funding level, of $5.3 million for FY 1982 be provided. 5.3 Meteorology and Hydrology (Item 5.b) We reviewed the meteorological research underway within the Site Safety

          ,Research Branch.       We endorse this work, particularly as it relates to the development of methods for handling the dispersion of airborne effluents over complex terrain and at. greater distances and longer time periods from the point of release. We endorse also the careful' review and eval-uation being conducted by this group of the ARAC system, as well as the in-depth assessment of the available alternatives. It would appear unwise for the NRC to move forward with the implementation of the ARAC system at operating nuclear power plants without first having completed this work.

We also reviewed the hydrological research being conducted by the Site Safety Research Branch. This work pertains to the contamination of ground water, predictions of its movement and the development of methods for interdiction. We endorse this work and urge that it include the development of the basic information necessary for the establishment of acceptable criteria for the hydrological characteristics of nuclear power plant sites. One, project that might be added to these efforts is a care-ful evaluation and assessment of methods that can be effectively applied 31

N I by municipal water purification facility personnel for removing radi o-nuclides from surface and ground waters. that have been contaminated by reactor effluents. The overall funding level requested for this. subele- ) ment is considered satisfactory. ' 5.4 Airborne Effluents - Environmental Imoacts (Item 5.c)-

      . Projects - reviewed within this subelement included those on "Radiciodine Pa thway. Analysis," "Early Ef fects of Inhaled Radionuclides," and " Acute Morbidity and Mortality f rom Nuclear Accidents." Although the first of
      ,these is considered important, we note that it is very similar to a Tech-nical . Assistance Project being conducted by the Radiological Assessment Branch.      For this reason, we recommend that the two projects be combined.

There is no need to do this work twice. Altnough we believe the last two projects would yield data useful for making better estimates of the health effects of . accidental releases from nuclear power plants, we recommend that this work be carefully correlated with similar work underway within

       'other Federal agencies. Such an evaluation may lead to changes that will make this research more productive.

Although we have in past years called for reduced efforts on improving models for the environmental transport 'and behavior of radionuclides under conditions of routi ne plant operations, the recent challenges to U.S. population dose estimates by scientists in Japan and the FRG show We support the the need for a continuing effort in this subject area. requested $2.3 million funding level for FY 1982. 5.5' Aouatic Effluents - Environmental Imoacts (Item 5.d)  ; Although, in general, we endorse the NRC research on the liquid pathway, there ic a need for a shift in its emphasis. To be specific, we recom-mend that this work be modified to place less emphasis on sediments and more on the sediment-biota interface and associated implications in tenns of the resulting population dose. In addition, the work should be di-rected to the assessment of the behavior of specific radionuclides rather than to radioactive materials, in general. The specific nuclides selected should be those of primary public health interest. We endorse the re-quested FY 1982 funding level of $1.3 million. 5.6 Occupational Exposures a'nd Health Effects (Item 6.e) Projects cowed in this subelement include those pertaining to Neutron Dosimetry and Lifects, the - Behavior and Health Ef fects of Ingested and Inhaled Radionuclides, and Epidemiological Studies of Exposed Populations. We endorse the projects on Improved Neutron Dosimetry and Effects Eval-uation, and on "Decorporation Techniques for Radionuclides." However, we believe that the project relating to " Health Ef fects Assessment" is in need of better definition. Similarly, we believe that the project en - titled, "Do simetric Model - ALARA," should be more clearly defined, 32

                   ,     -.,..m..,                                             e+. ~

particularly with respect to the newer types of data to be generated. Until this is done, we do not endorse this study. Overall, the requested funding level of $3.6 million for FY 1982 appears to be appropriate. S.7 Socioeconomic Impacts (Item S.f) We have no comments on the projects within this subelement. 5.8 Sitino Alternatives (Item 5.g) Although we did not review any specific research projects in this category, we believe there are several problems that should be addressed. One would be to develop data and information related to the forthcoming rulemaking on siting. Specific aspects that need attention include the establishment of criteria for determining the acceptability of sites for single- as well as multiple-unit stations. With respect to the latter, we are particularly concerned about the lack of definitive approaches for evaluating their advantages and disadvantages. In order to address these prob le:ns , we recommend that a funding level of at least 30.4 million for FY 1982 be provided within this subelement. S.9 Emergency Preparedness (Item 5.h) We continue to be aware of the need for reliable and accurate instruments for assessing nuclear power plant radiation levels and releases under accident conditions. Since the data generated by such monitors will be used to make major decisions relative to post-accident actions, it is imperative that they yield data of as high quality as possible. Research to achieve these goals should be actively pursued and should include the application of up-to-date technology in the design of such equipment. Where monitors involve the use of portable field equipment, we want to caution that care be taken to assure that the people involved have a clear understanding as to the conditions under wnich such instruments would be used and the types of decisions that would be based on the data they gen-erate. We are pleased to see that a project on " Human Factor of Emergency Re-sponse" has been proposed as a new area for study within this subelement. There are several areas in whicn such research could be helpful in emer-gency preparedness. One would be to study the relative benefits of sheltering versus the use of potassium iodide (KI) pills for reducing radiciodine intakes among population groups. For example, will people l be evacuating so rapidly that they will not take time to take KI pills?  ! What can we expect of population groups during tne initial pnases of a reactor accident? l i 33

1 1

                                                                           ~

Another area for this type of research relates to trade-offs in accepting a given low dose now (witn a probability of one) versus the possibility of a higher dose later (with a probability of less than one). An example of the application of the results of such studies would be controlled versus uncontrolled venting of reactor containments such as at TMI-2. To assure adequate support for these additional studies as well as those already proposed, we recommend that this subelement recei ve a funding level of 31 million for FY 1982. 5.10 Recoaraendations We have observed during this review that the degree of overlap in research projects being conducted by RES and those being conducted as Technical Assistance Projects within other NRC divisions is increasing. In addition, there appears to be a lack of coordination in the researcn efforts between RES and NRR in areas such as emergency response alternatives, radionuclide transport, and environmental monitoring. We reconnnend that these matters be explored to reduce any possible duplication of effort and losses of research efficiency. We were impressed with the extent of the workload of the RES members in-volved in monitoring research projects in this subject area. There is a clear need for at least one additional meteorologist within the Site Safety Research Branch and efforts should be made to employ at least one radiation biologist to provide in-house competence relative to the bio-logical effects of ionizing radiation and associated countermeasure ac-tions. I 34

l

                                                                                                                                                     )

6. WASTE MANAGEMENT l l-6.1 Introduction This decision unit includes research on the safety. problems of handling and ultimate disposal of high and low level radioactive wastes and uranium mill tailings. The safe disposal of all these types of wastes  ;

has been and continues to represent a major public concern in the ex-ploitation of nuclear energy for large scale power generation. 6.2 Hiah Level Waste (Item 6.a) We agree with the NRC staff that research work on high level waste handl-ing and disposal should be vigorously pursued so that the necessary tech-nical information is made available on a timely basis for decisions re-garding licensing ' and regulatory activities. The ultimate safe disposal of these. wastes poses one of the most difficult and complex problems in the nuclear fuel cycle. We have observed continued improvement in RES and NMSS in managing the research work in this program. We believe that the major area in need - of improvement ~is in the decision-making steps for selecting tne re-search work realistically needed and in the setting of priorities. _ These matters assume greater importance when funding is limited. We have ob-served that increased attention and cooperation are being given to these matters by both NMSS and RES. However, the reviews, although frequent and extensive, have --for the most part been made internally by the NRC staff. We recommend augmentation of the NRC reviews by including assist- ' ance and participation of outside qualified people. We suggest that consultants give special attention to the geological exploration needed for site characterization. The increased expertise and broader perspec-tive that can be made available by judicious use of consultants can greatly assist the NRC in deciding now much research work is realistically needed and whether it should be supported by NRC or by other organizations, e.g., 00E. We believe that the NRC should expedite its planned studies on the devel- ' opment of risk assessment methodology for potential early application of this technique to assist in the selection of research work to be under-taken and in setting priorities for it. Based on our review of the research program, we believe that only a mod-erate reduction in funding can be accommodated before the timely comple-

                                                                                                                  ~

tion of required research would become difficult and as a consequence, lead to delays by NRC in the licensing of repositories. We believe that this reduction can .be arranged by a combination of deferrals and reduc- i tions of some of the research areas as determined by application of best judgements on the urgency and amount of information needed to answer specific questions. 35 l l l

l l l We recommend a funding level of $16.3M in FY 1982 for this researcn pro- , gram. 6.3 Low Level Waste (Item 6.b) In NUREG-0657, we emphasized the need of sufficient research work to expedite the licensing and regulation of handling and disposal of low level radioactive wastes. We reiterate that position for FY 1982. We have urged the RES and NHSS Staffs to reexamine the FY 1982 program witn the assistance and participation of outside consultants. Particular at-tention should be given to that work necessary to permit the NRC to make ' licensing decisions rega rding low level waste. The existing situation mandates the selection of new disposal sites within the near future. Re-search related to the development of criteria for judging acceptability of such sites should be expedited. We recommend a funding level of $5.5 million for this program. 6.4 Uranium Recovery (Item 6.c) The disposal of uranium mill tailings which result from the uranium re-covery and concentration operations on uranium ore has long been a public concern in connection with nuclear power generation. We agree with the NRC Staff on the need for research on these problems. We recommend that the requested funding of $3 million be provided in order to deal satisfactorily with the large number of existent uraniuro mill tailings piles and to provide early guidance for tne licensing and regu-lation of new mills so that the public safety problems encountered earlier ~ can be avoided, or at least, ameliorated. 36

7. SAFEGUARDS AND FUEL CYCLE SAFETY l

          .1     Introduction in addition to subelements dealing with Safeguards and Fuel Cycle Facility            ;

Safety, . this decision unit has subelements addressing the radiological problems of handling materials in situations other than those'specifically covered in other decision units. Thus, subelements concerning Decommis-sioning, Transportation, Product Safety, and others, are included in this

       . decision unit.

As a general observation it may be noteu *. hat the situations and materials proposed for study are those associated with the operation of LWRs. In a number of instances the scope of the proposed studies ought to be recon-sidered, and possibly broadened, should the country's present policy con-cerning reprocessing and breeder reactors be changed by the time the FY 1982 Budget is in effect. 7.2 Physical Protection (Item 7.a) A major fraction of the effort in this item will be devoted to applying techniques already developed for use in the licensing and regulatory pro-cess. Some work will be continued, and new work started, on spent fuel storage problems. Potential ~Tonflicts between safety and safeguards re-quirements for operating reactors will also be studied. 7.3 Material Control and Accounting (Item 7.b) Here, also, a major fraction of the effort will be devoted to transferring developed techniques for use in the licensing and regulatory process. In-creased attention will be given to determining the amount of material held up in processing equipment. 7.4 Threat and Strateqy (Item 7.c) This subelement is a small program to develop appropriate responses to threats or appropriate actions in the event of successful sabotage or theft. In our view, the work in this subelement would have a lower pri-ority than the work in subelements 7.a or 7.b. 7.5 Fuel Cycle Facility Safety (Item 7.d) A major research effort in this area is devoted to analyses of accident scenarios for aerosol generation in fuel cycle facilities and to deve- l lopment of realistic models for aerosol transport within such facilities and to atmospheric release points. We agree with the importance of this effort and support the RES plans for it. Another significant research-effort in this program is that directed at the development and application 1 of risk assessment methedology in the fuel cycle. We recommend funding of this subelement at the requested level. 37 l

7.6 Decommissioning (Item 7.e) ., We have previously recommended a larger NRC research program on decommis-sioning. We continue to support' this position. We recommend funding at the' level requested by RES. 7.7 Transportation (Item 7.f) We believe that the research studies related to safety in the transpor-tation of radioactive. materials is generally needed and endorse the re-quested level of funding for it. 7.8 Effluent Control (! tem 7.g) We agree with the RES on the need for this research program which is mainly directed at improving the accuracy in evaluating effluent control. system performances in .PWRs and fuel cycle facilities. In order to help achieve this objective, a major research effort will be made to obtain more accurate radionuclide source term data. We question the value. of. the study on the " Decontamination Effects on Radwaste Systems" and recom-mend that the~ study on an " Advanced PWR Effluent Treatment Model" be either combined with the one on " Source Term Measurements" or deleted. We recom-mend the requested funding level for this program. 7.9 Product Safety (Item 7.5) This subelement - is a new program. Logically, a first step would include developing an i nventory of the products requiring consideration and a scale of relative public risks associated with these products. An ade-quate fraction of the total funds allotted to this subelement should be available for the purpose of a preliminary relative risk survey. 7.10 Occupational Protection (Item 7.1) This subelement covers several projects related to the measurement and control of the buildup of radionuclides within reactor systems and to the , post-accident decontamination of LWR plant sites. These efforts are in ' direct response to our recommendations over the past several years. We endorse these efforts and consider the requested funding levels to be i adequate. We recommend, however, that consideration also be given to the i expansion of related research on the reduction of occupational exposures associated with major maintenance and repair operations such as the re-placement of steam generators. This work, coupled with ongoing research on fuel failures due to pellet-cladding interaction and radionuclide re-leases derived from PBF experiments. should provide the types of infor-mation necessary for making progress in controlling occupational exposures in LWRs. 38

l l l 7.11 Sumary

                                                                                                           ]

1 The RES request for funds in this decision unit was $13.2 million, and i the E00 markup $10.7 million.  ! l Of these funds, $5.2 million has been requested by RES and $4.9 million has been approved by E00 for Safeguards (Items 7.a. b, c). This compares with $4.9 million for these items in FY 1981. There is the familiar dif-ficulty of comparing the priority of work in this problematical field with work aimed at improving the operational safety of reactors; but, in view of the public interest and potential importance of possible acts of , theft or sabotage, we believe that this work should be continued at I

• about the existing level and that the amount requested by RES is in the I low range of acceptability. l 1

For the remaining items RES has requested $6.5 million, and the EDO has l approved $5.8 million. We recommend funding at the level requested by l RES, but suggest that the amount allocated to Occupational Protection (Item 7.i) should be increased sufficiently to support a meaningful study of crud build-up in LWRs. In summary, for this decision unit we recommend funding at the level re- l quested by RES. j l 39

I I l

8. SYSTEMS AND RELIABILITY ANALYSIS i

8.1 Introduction This decision: unit, Systems and Reliability Ana lysis (SARA), includes many but. not all of ~ the programs which-previously were grouped under the former decision unit entitled Risk Assessment. SARA has four subelements:

Methodology Development; Reliability and Huaan. Error Data Analysis; Sys-

                 . tems analysis; and Consequences Analysis.

We have previously given strong support 'to this research program. In. NUREG-0657, we placed our highest. research priorities on the FY 1981 de-cision units entitled Improved Reactor Safety and Risk Assessment; we recommended increases in the President's budget requests for these two decision units. The gi owing emphasis during recent months on the use of reliability and risk analyses and the development of quantitative risk criteria supports our recommendation. We support growth in the SARA budget for FY 1982. The extent of growth which is needed depends in part on the extent to which probabilistic methodology is used in other research decision units and by other organi-zational entities within the NRC, particularly NRR and NMSS. The role, scope, priorities, and resources of the. Interim Reliability Evaluation Program (IREP) will trongly influence the SARA budget requirements. Similarly, the role t ,at SARA will play in the evaluation of probabilistic studies performed by industry will influence the SARA resource require-ments for FY 1982. The overall NRC resources in probabilistic and risk analysis must be substantial and the extent to which SARA will provide support to other groups in' the NRC should be factored realistically into the FY 1982 bud-get. In general, we support the work areas planned for FY 1982 under SARA. However, we believe that some aspects have not received sufficient pri-ority and resources. These include the following:

                    . The early development of quality assurance criteria for probabilistic analyses to be used in the regulatory process.
                    . The early development of a changed approach to the single-failure crit eri on.
                    . An examination of possible weaknesses in current application of the single failure criterion.

41 4

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1 4 . e; i The development of a basis for an improved regulatory approach to minimizing significant design errors. The early. development of infonnation needed to detennine the appro-priate. regulatory approach to control systems and to ' infonnation. needs of the reactor operator. An increased level of. resources on the program to develop quanti- > tative risk criteria.

                          . A considerable acceleration of the research program on flood risk to nuclear power plants.
                          . A large increase in emphasis and resources for the task on alternate decay heat removal systems, including consideration of sabotage.

A considerable acceleration in the development of information needed to estimate the likely effect on risk of various potential design changes intended to prevent or mitigate the consequences of accidents leading to severe ccre damage or core melt in LWRs. A program to better define property damage from accidents involving large releases of radioactive materials, including the effect on societal resources. All necessary support -for the proposed NRC rulemaking on accidents involving degraded cores and core melt. Unless there are major resources made available for similar work elsewhere within the NRC, we believe that $24.8 million will be appropriate budget for SARA in FY 1982. In any event, we recommend that the matters listed above be given priority in both FY 1981 and FY 1982, even if it means reducing other useful pro-grams, ongoing and proposed, in this decision unit. We note further that the Probabilistic Analysis Staff which is responsi-ble for the research in this decision unit perfona more scientific work in-house than that of many of the other organizational units. This must be taken into account in manpower allocations. 8.2 Methodology Development (Item 8.a) The RES justification and planned accomplishments are reasonable. However, as discussed in Section 8.1, we believe that priority in this subelement should be given to the most pressing needs of the NRC. These needs include the following: the development of a methodology suitable for early use by the industry and the NRC in system and accident probability evaluation; quality assurance guidance and a peer review technique for probabilistic analyses perfonned by the NRC and the industry; flood risk to LWRs; and quantitative risk criteria. 42

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Also, a methodology for evaluating the regulatory approach to LWR control systems should receive priority. 8.3 Reliability and Human Error Data Analysis (Item 8.b) The propused research program on human error should have the benefit of consideraole interaction with Inspection and Enforcement (IE). Such interaction, if carried on down to include personnel from training and inspection, could be useful in both directions. The large program proposed for LER failure rate analysis should be co-ordinated with the work of the Of fice of Analysis and Evaluation of Oper-ational Data (AE00), as well as similar efforts in NRR. This research program should be responsive to the needs of such groups. , While we agree that work on component f ailure rate and downtime is' worth-while, we recommend that this program, as well as that on Methodology Development, be evaluated to see if the proper priority is being given to systematic and common cause failures of all kinds, including sabotage. 8.4 Systems Analysis (Item 3.c) Part of this subelemont is focused on the IREP program, while part appears to consist of a collection of largely new FY 1981 programs involving the application of probabilistic analysis. Although we forsee a need for an applications program in addition to the IREP program, it is not clear how the currently proposed program interfaces with other NRC staff efforts. For example, the proposed effort involving a review of LERs and a study of operational occurrences should be supportive of AE00, if performed. The proposed effort in standardized relaibility design guidance, which appears to represent a reexamination of the si ngl e-f ai lure criterion, might serve as the vehicle for research on this topic, and warrant greater emphasis in that case. The task on risk-related resident inspection, if pursued, should be closely coordinated with IE. The analysis of plant log data on forced outages requires better defini-tion and coordination with the subelement 8.b if, perforad. As mentioned earlier, the tasks en alternate heat removal systems should receive much greater emphasis. This emphasis should be sufficient to provide a basis for regulatory decision making by the end of FY 82. l 43 4

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8.5 Consequences Analysis (Item 8.d) It is important that the interface between this subelement and that on Systems Analysis, as well as the interface with the subelement on Severe Accident Phenomena and Mitigation, be defined. As a minimum, close co-ordination among these several efforts is required, and a group having an overall perspective on the entire LWR risk picture should be maintained. As outlined in the NRC Staff document providing justification and planned accomplishments, the following should receive priority in this subelenent: resolution of liquid pathways; support of the licensing office in power plant siting; design and emergency planning; and reexamination of nearby and distant effects of a large atmospheric release of radioactive mate-rial on property damage and societal resources. 44

9.

SUMMARY

The recommenoations in this report, if followed, would result in a total RES Budget for Program Support of $265 million. This is somewhat greater than the RES request of July 9,1980, chiefly because of the recommended increases for Fast Breeder and Advanced Converter Reactor Research. This total is also significantly greater than the 5229.7 million recommended by tne EDU Staff on July 2,1980. The accident at TMI-2, the lessons lec ned from it, and the ensuing rulemaking proceedings, all seem to us to mandate the highest prio rity for research relating to the safety of LWRs, both those now operating or under construction and those yet to be designed or constructed. These considerations lead us to conclude that the highest priorities should be assigned to the following areas: That research related to transients and small LUCAs in Decision Units 1 and 2. Research on Plant Operational Safety: Decision Unit 3. That research related to Severe Accident Phenomena and Mitigation in Decision Unit 4. Support of the in-house and contract research related to Systems and Reliability Analysis: Decision Unit 8. 45

l l l l I BIBLIOGRAPHY Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program for Fiscal Year 1981, NUREG-0657, February,1980. Comments on the Nuclear Regulatory Commission Safety Research Program Budget, NUREG-0603, July,1979. NOTE: The above reports are available for purchase from the NRC/GPO Sales Program, U.S. Nuclear Regulatory Comission, Washington, DC 20555, l and the National Technical Information Service, Springfield, VA 22161. l t 1 47 I 1 l l

l' l l GLOSSARY ACRS Advisory Couaittee on Reactor Safeguards AE00 Office for Analysis and Evaluation of Operational Cata ALARA As Low As Reasonably Achievable ARAC Atmospheric Release Advisory Capability BaW Babcock and Wilcox BWR Boiling Water Reactor CCTF Cylindrical Core Test Facility CRBR Clinch River Breeder Reactor DECLB Double-Ended Cold Leg Break 00E Depart;nent of Energy ECC Emergency Core Cooling EDO Office of the Executive Director for Operations ESSOR Multi National Research Reactor Complex at Ispra, Italy FFTF Fast Flux Test Facility FRG Federal Republic of Germany FY Fiscal Year IE Of fice of Inspection and Enforcement IREP Interim Reliability Evaluation Program LER Licensee Event Report LMFBR Liquid Metal Fast Breeder Reactor LOCA Loss-of-Coolant Accident LCFT Loss of Fluid Test LWR Light Water Reactor 49

NDE Nondestructive Examination ' NMSS Office of Nuclear Material Safety and Safeguards i NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NRU Atomic Energy of Canada Ltd., Test Reactor ORNL Oak Ridge National Laboratory PBF Power Burst Facility PSAR Preliminary Safety Analyis Report PWR Pressuirzed Water Reactor QA Quality Assurance RES Office of Nuclear Regulatory Research SARA Systeins and Reliability Analysis SC TF Slab Core Test Facility SO Of fice of Standards Development SSMRP Seismic Safety Margins Research Prograni THTF Thermal Hydraulic Test Facility TLTA Two Loop Test Apparatus TNI-2 Three Mile Island, Unit 2 TRAC Transient Reactor Analysis Code TVA Tennessee Valley Authority l l 50

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• U.S. NUCLE AR r.EEULATOMY COMMl8810N BIBLIOGRAPHIC DATA SHEET NUREG-0699

4. TITLE AND SU8TlTLE (Add Votume Na,if eprenate) 2.lle ** Dim kl Comments on the NRC Safety Research Program Budget 3. REclPIENT'S ACCESSION No.

for F1 scal Year 1982 7 AUTHOR (S) S. D ATE REPORT COMPLE TED M ON TH l YEAR Advisory Committee on Reactor Safeguards wv 1980

9. PERFORMING ORGANIZATION N AME AND MAILING ADORESS (Include lia Code / OATE REPURT ISSUED MoNTM l YEAR Advisory Committee on Reactor Safeguards Jtil y 1990 US Nuclear Regulatory Commission 6. (te*, o<e ns

         'Aichington , DC 20555.

8.(Leme Olv e)

12. SPONSORING ORGANIZ ATION N AME AND MAILING ADORESS (lactuos 2,p Codel 10 PROJECT!T ASK/ WORK UNIT NO.

Same as 9., above. i t. CONTRACT NO.

13. TYPE OF REPORT eE ntoo cove Rf o (loc /upre deres)

15. SUPPLEMENTARY NOTES I 4. (Lews o/m*/

16. A85TR ACT Q00 words or leu)

Recommendations of the Advisory Committee on Reactor Safeguards are presented to the Commissioners for their consideration for FY 82 budget for the NRC sa fe ty research pmgram.

17. AEY WOROS ANO DOCUMENT ANALYSIS 17a OESCRIPToRS 1

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[ W A$HINGTON, D. C. 20555 July 15,1980 APPENDIX XXVII ACRS REPORT ON SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 i The Honorable John F. Ahearne Chairman U. S. Nuclear Regulatory Commission Washington, D. C. 20555

SUBJECT:

REPORT ON THE SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 & 2

Dear Dr. Ahearne:

During its 243rd meeting, July 10-12, 1980, the Advisory Committee on Reactor Safeguards completed its review of the application of the Tennessee Valley Authority (hereinafter referred to as the Applicant) for authorization to operate the Sequoyah Nuclear Plant, Units 1 & 2 at full power. The Committee had considered aspects of the application during its 242nd meeting, June 5-7, 1980; 236th meeting, December 6-8, 1979; 229th meeting, May 10-12, 1979; and 228th meeting, April 5-7, 1979. A tour of the facility was made by members of the Subcommittee on January 24, 1976 and the application was considered at Subcommittee meetings on July 9,1980; June 2,1980; November 5,1979; and March 12, 1979. During its review, the Committee had the benefit of discus-sions with representatives and consultants of the Applicant, the Westinghouse Electric Corporation, and the Nuclear Regulatory Commission (NRC) Staff. The Committee also had the benefit of the documents listed. The Committee reported on interim low power operation of Unit 1 on December 11,1979 and on ' a construction permit for this plant on February 11, 1970. In its letter of December 11, 1979 the Committee addressed the proposed special low power test program. to be carried out on Unit 1, the seismic reevaluation of the Sequoyah plant, actions on recommendations resulting from the review of the accident at the Three Mile Island Station, Unit 2, and actions on various generic problems. These generic problems were further discussed in the Committee's report, " Status of Generic Items Relating to Light-Water Reactors: Report No. 7," dated March 21, 1979. The Committee's j recommendations in ii.s December 11, 1979 letter are also applicable to Unit 2 1 except that the special low power test program will not be repeated on l Unit 2. l The special low power test program has been reviewed by Westinghouse Electric Corporation and by the NRC Staff. The Applicant began these tests on July 11, 1980 and the Applicant, Westinghouse, and the NRC Staff will review the results of these tests. It is expected that the additional operator training and operator experience will prove to be beneficial. l

                                                  . N l

Honorable John F. Ahearne July 15,1980 The Committee has reviewed and reported on NUREG-06c0, "NRC Action Plans Developed as a Result of tne TMI-2 Accident," Draft 3. The status of the Applicant's compliance with the NTOL licensing requirements as well as a number of non-TMI-related items were. reviewed during its 243rd meeting. There - are a number of both non-TMI and TMI-related requirements not fully resolved. Both the NRC Staff and the Applicant expect that the complete resolution of these outstanding items is essentially a procedural or documentary matter which will be completed within a very few weeks. These items should be resolved to the satisfaction of the NRC Staff. The Committee wishes to be kept i nfo rmed. The Committee believes that the implementation of the Action Plan as it will be realized at Sequoyah is adequate to assure the safe operation of this plant. The Committee, in its March 11, 1980 report on the NTOL items, recommended that the licensees develop reliability assessments for thei r pl ants and that design studies of possible hydrogen control and filtered vented contain-ment systems be required. The Applicant has conducted studies of a number of means for hydrogen control, and as an interim measure, has proposed installa-tion of a distributed array of ignition sources which it expects to have in place by the fall of 1980.- The Applicant has concluded that by this means the containment would be able to cope with the pressure resulting from the combustion of hydrogen released by the reaction with water of up to about 70% of the zirconium in the core. This compares with the 25% which the contain-ment could cope with without any additional control measures and the 30 to 50% estimated to have reacted in the accident at TMI. The NRC Staff plans to review the proposed system in detail to assure itself of its efficacy and that all safety aspects have been taken into account. The Committee wishes to be kept informed of the further conclusions reached by the Staff and the Applicant in their continuing consideration of these matters. The Applicant has conducted reliability assessments of some features of the plant and has considered some aspects of the effects of a possible filtered vented contain-ment. Though the work accomplished to date is limited in scope, these studies are definitely responsive to the Committee's recommendations on these points. The Applicant proposes to continue studies of this nature and to extend the range of their application. While these efforts, as well as those concerned with hydrogen control, should be vigorously pursued, in view of the commitments made by the Applicant, it is the opinion of the Committee that

                  ~

their present incomplete status need not delay the issuance of a full power operating license. Early this year a differing professional opinion was advanced by a member . of the NRC Staff concerning the acceptability of a particular weld repair in the piping to a pressurizer relief valve of Sequoyah Unit No.1. All other qualified and responsible members of the NRC Staff, as well as professional personnel on the staff of the Applicant, take the position that the weld should be regarded as acceptable since there is no evident reason why it should not be at least as capable as other (more standard) welds whien would

l Honorable John F. Ahearne J aly 15,1980 l be considered acceptable. The differing opinion is not that the weld is demonstrably less capable than it need be, but 1) tnat the evidence available is inconclusive on this point, and 2) that more specifically relevant infor-mation could be obtained without serious difficulty. This could be done by constructing a mock-up of the weld in question using material and procedures as similar as possible to those which apply in the actual case and subjecting the mock-up to a through-wall metallographic examination. The results of this examination could then (for example) be compared with those from a full penetration weld in the same material, which has been performed in the stan-dard fashion and deemed acceptable based on satisfactory operational experi-ence with which the majority opinion has compared the preser,t weld. This has not been done. The Committee does not consider it to be particularly likely that this weld repair presents a serious hazard; but it does believe the evidence on this point could be improved. The Committee believes that, i n ' the interest of resolving the question that has been raised to the maximum extent readily possible, steps of the nature outlined should be taken. The Committee believes, that if due consideration is given to the items mentioned above, the Sequoyah Nuclear Plant, Units 1 and 2 can be operated at levels up to full power without undue risk to the health and safety of the public. Si ncerely ,  ; Milton S. Plesset Chai rman

References:

                                                                                                     )

1. Tennessee Valley Authority, " Final Safety Analysis Report, Sequoyah l Nuclear Power Plant," Volumes 1-13, and Amendments 1-63.

2. U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Operation of Sequoyah Nuclear Plant Units 1 and 2," NUREG-0011, March 1979.

3. U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Operation of Sequoyah Nuclear Plant Units 1 and 2," Supplement No.1, 1 NUREG-0011, February 1980. l

4. U.S. Nuclear Regulatory Commission, "NRC Action Plan Developed as a Result l of the TMI-2 Accident," NURE3-0660, May 1980.

5. U.S. Nuclear Regulatory Commission, "TMI-Related Requirements for New Operating Licenses," NUREG-0694, June 1980.

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July 15,1980 APPENDIX XXVIII SEQUOYAH 1 AND 2: ICE CONDENSER The Honorable Victor Gilinsky CAPABIW AND HME QN Commissioner U. S. Nuclear Regulatory Commission Washington, DC 20555

Dear Dr. Gilinsky:

This is in response to your request of July 10, 1980 concerning particular aspects of the Sequoyah Nuclear Plant. Fi rst , as mentioned in the Committee's Sequoyah latter of December 11, 1979, the capability of the ice condenser containment design to cope with the steam resulting from a large LOCA was the subject of detailed discussion over a period of years involving the NRC Staff, the vendor, and the ACRS. As a result of this effort it was concluded that this type of design was fully capable of fulfilling the function mentioned. We have no reason to change that conclusion. Second, the matter of the control of large amounts of hydrogen is discussed to some extent in the Committee's Sequoyah letter of tnis date. Although the information available at present is prelimi nary and will require further detailed confirmation both by the Staff and the Applicant, we expect the present general conclusions to be confirmed. The Applicant has committed to proceed quickly with the installation of a distributed ignition system. The Committee does not believe that there is any practical' need to hold up the issuance of an operating license pending completion of the proposed ignition system. Sincerely, Milton S. Plesset Chairman l l l 1

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Dear Dr. Plesset:

In the Committee's review of the TVA reques: for authorization to operate.the Sequoyah plant at full power, I would appreciate the Committee giving particular attention to whether there is reasonable assurance that the ice condenser containment would hold in the event of a serious accident. Specifically, . l I request that the Committee's letter include: (1) an assessment of.whether the ice would adequately . suppress the steam pressure in a large loss of. coolant accident; and (2) the Committee's view on whether additional hydrogen control measures should be required for full power operation to limit the effects of large amounts of hydrogen, such as that generated during the Three Mile Island accident. 1 I understand TVA is proposing to install a distributed , ignition system to mitigate the effects of a hydrogen burn. ) I would appreciate the Committee's assessment of the likely 1 effectiveness of such a system. 1 Sir,cerely, , ie lVictor ION Gilinsky _ m Commissioner cc: Chairman Ahearne Commissioner Hendrie - Commissioner 3radford t l l

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          ***s-                                July 16, 1980 APPENDIX XXIX ACRS COMMENTS ON THE REACTOR COOLANT PUMP TRIP AND HIGH PRESSURE INJECTION TERMINATION CRITERIA Honorable John F. Ahearne Chairman U. S. Nuclear. Regulatory Commi ssion Washington, D.C. 20555

SUBJECT:

ADDITIONAL ACRS COMMENTS ON THE RCP TRIP AND HPI TERMINATION CRITERIA

Dear Dr. Ahearne:

In your letter of April 1,1980, you requested that we clarify our concerns with the present reactor coolant-pump (RCP) trip and the high pressure injection (HPI) termination criterion. You also indicated in a memorandum , to R. Fraley on February 22, 1980 that you would welcome cur comments on NUREG-0623, " Generic Assessment of Delayed Reactor Coolant Pump Trip During i Small Break Loss-of-Coolant Accidents in Pressurized Water Reactors." 1 1 The cresent requirements for RCP trip and HPI termination have developed from ' the lessons learned from the Three Mile Island accident and from the extensive number of small break LOCA calculations subsequently carried out. There are two distinct requirements in the ISE Bulletins issued, as referenced below, which can be considered separately. The first concerns the directive which requires prompt shutdown of all reactor coolant pumps in ?WRs following a depressurization transient which initiates safety injection. The second is the requirement that the safety injection system continue to be operated until a specified degree of subcooling is attained in the primary system. The prompt reactor coolant pump trip mandated by the Bulletins followed analyses by the vendors of nuclear steam supply systems which seemed to show d that leadthere was a "windew" to calculated peakofcladding break sizes and pump trip temperatures indelay times excess ofwhich wou}F the 2200 licensing limit. These same .riethods of analysis indicated that with prompt j pump trip the peak cladding temperatures would remain below 2200 F. l The NRC Staff prepared a useful critique in NUREG-0623 of these vendor calcu-lations and, while this report clearly presented the deficiencies in the analytical methods used, the report agreed witn the vendors' cenclusions. The short-term action by the Staff therefore was the requirement of prompt trip of the reactor coolant pumps; as a long-term action the Staff recommended that licensees propose and submit design changes tnat will assure auto.natic trip of all reactor cooiant pumps. We do not, at this time, disagree entirely with the Staff's requirement of prompt coolant pump trip, but in view of the analytical limitations upon which prompt trip is based we believe that the emphasis on immediacy of the

       ,   trip and on eventual automatic trip may not be desirable. Recent experimen-tal data has put doubt on the existence of the "windew" wnich is the basis l
                                                      -950

Honorable John F. Ahearne July 16,1980 for requiring ' prompt pump trip. Additional experimental data will become available before the end of the year. The prompt trip has been carried out in four transients since the Sulletins have appeared. In none of these was there a LOCA in the primary system; all of these transients arose from disturbances on the secondary side. No significant plant damage ensued in these transients and there was no harm to plant personnel or to the public. There has been complaint, however, that without reactor coolant pump flow the operator loses reactor pressurizer control 'since, in many PWRs, pressuri:er spray flow depends on coolant pump flow. Further, natural circulation must also be established to remove decay heat. It must be said that the Staff's hope to develop a clear distinction between depressurization from a small break on the primary side and depressurization from a secondary side transient seems quite optimistic. We believe that reactor coolant pump trip upon primary depressurization is an acceptable procedure, but we see no urgency at this time for installation of automatic ' pump trip. With regard to primary pressure control, we believe that it is desirable to provide pressuri:er spray flow which is independent of main coolant pump flow. The present set of requirements for HPI termination criteria is based upon achieving a specified degree of subcooling in the prima ry coolant system along with, in some cases, a specified water level in the pressuri:er and steam generators. These requirements are intended to prevent a recurrence of the TMI-2 situation in which HPI flow was terminated while still necessary; these requirements, however, do not address the conditions in which HPI should be tenninated when not requi red. We are concerned that relatively frequent system transients which activate HPI might progress to liquid disenarge through safety valves or PORVs, valve failure under liquid flow, and a resultant small break LOCA. It should also be pointed out that Westinghouse has recently reported a significant deficiency under 10 CFR 50.55(e) for a l numoer of reactors with high head centrifugal charging / safety injection pumos. Failure to stop these pumps promptly when high pressures are reached could result in pump f ailure from low flow - a common mode failure of the redundant HP1 pumps. Changes in operational procedures may also affect the design limits of other components. These interactions need to be carefully reviewed. We note that a numoer of plant transients that have occurred in the past year have been affected by the NRC approved HPI termination and RCP trip criteria. These include events, as referenced below, at North Anna, Unit 1, September 26, 1979; Prairie Island, 'Jnit 1, October 2,1979; 'and ANO, Unit 2, January 29, 1980. Some changes have been made in criteria in response to these events. We believe that continued Staff attention in this area is required. Sincerely, Milton 3. Plesset Chairman

                                                         / pas /

1 Honorable John F. Ahearne July 18,1980 i

References:

1. U.S. Nuclear Regulatory Commission, Office of Inspection and Enforcement, "I&E Bulletin 79-05A," April 5, 1979.

2. U.S. Nuclear Regulatory Commission, Office of Inspection and Enforcement, "I&E Bulletin 79-06A," April 14, 1979.

3. U.S. Nuclear Regulatory Commission, Office of Inspection and Enforcement, "I&E Bulletin 79-06B," April 14, 1979.

4. U. S. Nuclear Regulatory Commission, Office of Inspection and Enforcement, "I&E Bulletin 79-05C and 79-06C," July 26, 1979.

5. NUREG-0623, " Generic Assessment of Delayed Reactor Coolant Pump Trip During Small Break Loss-of-Coolant Accidents in Pressurized Water Reactors," November 1979.

6. Letter, C. M. Stallings, VEPC0, to J. P. O'Reilly, NRC, Submitting Licensee Event Report for September 25, 1979 North Anna Number 1 Cooldewn Incident (October 9, 1979).

7. Letter, L. O. Mayer, NSP, to J. G. Kepolor, NRC, Submitting Licensee Event Report for October 2,1979 Steam Generator Tube Rupture Inci-dent (October 16,1979).

8. U.S. Nuclear Regulatory Commission Preliminary Notification of Event or Unusual Occurrence, PNO-IV-30-05, January 30, 1980.

9. Letter, D. C. Trimble, AP&L, to R. W. Reid, NRC, Submitting Startup Report, Supplement 2 for ANO-Unit 2, Maren 6, 1930.

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                 /                                           April 1, 1980 AIMMAN Jr. Milton S. Plesset Chairman, Advisory Cemittee en                                                                           i Reactor Safeguards                                                                                    i U. S. Nuclear Rec'latory Cemission Washington, D.               . 20555                                                     '

Dear Dr. Pl sset:

4y ph I appreciate receiving your letter of March 11, 1980 which outlined ACRS . coments on the recomendations of the NRC Task Force en Bulletins and

• Orders. One of the cements was that, ,

                    "The NRC Staff should, in cenjunction with the licensees, review the criteria for HPI termination and reactor coolant pump tric to reduce unnecessary challenges to the pressuri:er safety valves and prevent unnecessary trips of the reactor coolant pumos which may increase the difficulty in establishing uninterrupted core cooling."

Does this coment reflect a Cemittee concern that the fundamental philosophy (racid reactor coolant pump trip, HPI termination only after

    ^        certain criteria have been met) currently esocused by the NRC Staff may
       '     be in errer and should be reexamined, or rather a concern that while the fundamental philosophy is correct, the quantitati.ve criteria for trip and termination shculd be reviewed and pernaps revised?

Sinchely, k )

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1 hn (F. Ahearne j cc: Cemissioner Gilinsky ,- Comissioner Kennedy Cemissioner Hendrie Commissioner Bradford SECY OPE EDO - NRR . O e

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APPENDIX XXX l USE OF IN-CORE THERMOCOUPLES FOR . DETECTION OF 0FF-NORMAL CONDITIONS AND I The hanoracle Vic .or 3ilinsky FOR ACCIDENT ANALYSIS Commissioner U.S. Nuclear Regulatory Commission Washington, D.C. 2U555

Dear Dr. Gilinsky:

Your letter of June 4, 1980 requested the Committee's views on tne utility of in-ccre tnermoccuples as an aid in tne detection of off-noraal conditions and for ese in acciaent analysis. Your request was in tne context of instru-r,1entation to follow the course of an accident. As yoa noted, tne ACRS has long been concerned with assuring that qualified instrumentation should be available to follow the course of a serious acci-dent. This concern nas extended to tne use of data from core outlet thenao-couplas in PWs. For example, the ACRS has commented on the desiracility of the rmocou pl e availability and use in its reviews of the Davis-Besse, Oconee, and Indian Point Unit 3 plants. Mo re recently , t he Comt,;i tte a ccanented on the use of thermocou pl es dJring a discussion With the Com-mission immeciately following the accident at TMI (see tne attacned corre-spondence in which aooropriate paragraphs are underlined). Tne Conmittee Delieved then, as it does now, that core outlet thennoccupies can provide a readily available eneck on certain tyces of core conditions and ' serve as an additional means for detection of core benavior ano:,ialies. Tne Conittee also believes that, as practical, such information Poul d be mace readily avail abl e to the operator over the full teaperaturt unge of tne installec cetectors. Tne Committee recommends that, as prac. .ic al , faulty thermocouples should be replacea during refueling intervals. Hew PWRs shoul d be de si gnet' for thermocoupl e repl a c e.le nt . Core sucessenbly exit tnermocouples or otner core outlet tnermocouples have not generally been used in BWRs; however, the ACRS favors a caraf al examination of the feasi-bility of tneir use in BWRs and tne prcs and cans of such use. The ACRS notes that the NRC Staff nas developed proposed criteria f or installation and use of core outlet tnennoccuples as part cf the Lessons Learned item addressing inadequate core cooling. NRC Staff efforts in this area include the requi rement specified in proposed Regulatory Guide 1.97 for core exit temperature measurements. The ACRS supports this Regulatory Guide requirement but suggests tnat the NRC Staff consider PWR and BWR - design dif f erences in its implementation. The Committee believes that i nstrume nt s displaying thermocouple readings should be readily available in pl ant control rooms, consistent with tne philoso;ny underlying ACRS Generic Item 43: " Instrumentation to Follow the j 1 l l l l [ "M ' l

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                                      .Comissioner Bradford                                     l Corntnissioner Ahearne                                   i 4

FRCP.: R. F. Fraley, Executive Director Advisory Corn:nittee on Reactor Safeguards Attached for your infor: nation and use is a copy of the reccomenda-tions of the Advisory Corn: ittee on Reactor Safeguards which were orally presented to atx! discussed with you on April 17, 1979 re-gardirg the recent accident at the Three Mile Island Nuclear Sta-tion Unit 2. V R. F. Fraley Executive Director Attach. ment: Recomendations of the NRC Advisory Comittee on Reactor Safeguards Re. the 3/28/79 Accident at The Tnree Mile Island Nuclear Station Unit 2 ATTACHMENT 1 k~&S

April 17, 1979 RECCWDCATIONS OF WE NUCLEAR RS3ULATORY COM*lSSION ADVISORY CCWlTTEE CN REACTOR SAFEGUARDS RD3ARDING THE MARCH 28, 1979 ACCIDOC AT THE NREE MILE ISUCC NUCLEAR STATION UNIT 2 Presented orally to, and discussed with, the NRC Commissioners durire the ACRS-Commissioners Meeting on April 17, 1979 - Washington, D. C. Natural circulation is an importanc node of reactor coolirg, both as a planned process and as a process that may be used under abrormal circumstances. Se Committee believes that greater understanding of this node of cooling is required and that detailed analyses should be developed by licensees or their suppliers, ne analyses should be supported, as necessary, by experiment. Procedures should be de-veloped for initiating natural circulation in a safe manner and for providing the operator with assurance that circulation has, in fact, beer established. This may require installation of instrumentation to measure or indicate flow at low water velocity. The use of natural circulation for decay heat removal following a loss of offsite power sources requires the maintenance of a suitable over-pressure on the reactor coolant system. This overpressure may be

 ! assured by placing the pressurizer heaters on a qualified onsite power source with a suitable arrangement of heaters and power distri-bution to provide redundant capability.                                                     Presently operating PWR plants should be surveyed expeditiously to determine whether such arrangements can be provided to assure this aspect of natural circula-tion capability.

The plant operator should be adequately informed at all times con-cerning the conditions of reactor coolant system operation which might affect the capability to place the system in the natural circu-lation mode of operation or to sustain such a node. Of particular importance is that information which might indicate that the reactor  ; coolant system is approachirs the saturation pressure corresynding l to the core exit temperature. 21s impending loss of system over-pressure will signal to the operator a possible loss of natural circulation capability. Such a warning may be derived from pressur-izer pressure instruments and hot leg temperatures in conjunction with l conventional steam tables. A suitable display of this inforTnation should be provided to the plant operator at all times. In addition, consideration should be given to the use of the flev exit tempera-tures f rom the fuel subassemblies, where available, as an additional indication of natural circulation. l l

, . - - .- -. . = _ - - l 1 The exit temperature of coclant frx the core is currently measured by thermocouples in many PWRs to determine core performance. Se Committee recommends that these temperature measurments, as currently available, be used to guide the. operator concerning core status. We range of the information displayed and recorded should include the full capability of the thermocouples. It is also recommeMed that other existing instrumentation be examined for its possible use in assisting operating action durina a transient. Se ACRS recommends that operating power reactors be given priority with regard to the definition and implementation of instrumentation which provides additional information to help diagnose and follow the course of a serious accident. This should include improved sampling procedures under accident conditions aM techniques to help provide improved guidance to offsite authorities, should this be needed. We Committee recommends that a phased implementation approach be em-ployed so that techniques can be adopted shortly af ter they are judged to be appropriate. We ACRS recommnds that a high priority be placed on the development and implementation of safety research on the behavior of light water reactors during anomalous transients. The NRC may find it appropriate to develop a capability to simulate a wide range of postulated tran-sient and accident conditions in order to gain increased insight into measures which can be taken to improve reactor safety. The ACRS wishes to reiterate its previous recommendations that a high priority be given to research to improve reactor safety. Consideration should be given to the desirability of additional equipment status monitoring on various engineered safeguards features and their supporting services to help assure their availability at all times. De ACRS is continuing its review of the implications of this accident and hope to provide further advice as it is developed. l l l , ( l

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                             ~dNITED STATES ATOMIO ENERGY COMMISSION WAsHINGT O N. O.C. 30545 ROY l ' UU
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1 Honorable Dixy Lee Ray Chairman U. S. Ato ic Energy Com ission Washington, D. C. 20545

Subject:

INTEF.IM REPORT ON INDIAN POINT NUCLEAR GENERATING STATION UNIT No. 3

Dear Dr. Ray:

At its 163rd teeting, Nove:ber 8-10, 1973, the Advisory Co :ittee on Reactor Safeguards co:pleted an interi: review of the appli-cation of Consolidated Edison Co:pany of New York, Inc., for authorization Unit No. 3. to operate Indian Point Nuclear Generating Station The ptoject has been previously considered at Sub-co: ittee meetings on July 11, 1973, October 10, 1973 and November 7, 1973. cembers on Nove:berA2,tour of the facility was cade by Co :ittee 1973. In this review, the Co::ittee had the benefit of discussions with representatives and consul-tants of Consolidated Edison, their contractor, and the AEC Regulatory Staff. docu=ents listed. The The Co :ittee also had the benefit of the Com:ittee reported on the applicatier. for construction of Indian Point Unic No. 3 on January 15, 1969. Indian Point Unit No. 3 includes a four-loop Westinghouse nuclear steam supply syste with a design power rating of 3025 MW(t). The design is rating of 2760 MW(t). si=ilar to that of Unit No. 2 which has a power The three-unit Indian Point Nuclear Gene-rating Station is located approx 1:ately 2-1/2 ciles southwest of Peekskill, boundary line.New York, and 24 ciles north of the New York City The Coti=i t t e e 's repert of January 15, 1969, called attention to various catters including the following: consideration of ther=al shock to the pressure vessel in the unlikely event of a loss-of-coolant accident (LOCA); =easures to deal with possible hydrogen concentration buildup in the containment following a LOCA; greater independence in the on-site power system; =ain-coolant-l ATTACHMENT 2 1 - $Q ) _ _ _ . . _ - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ ^ ~ ~ ~ ~ - ~

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l 1 Bonerable Dixy Lee' Ray ty/1453 i l pump flywheels as a poten:Lal source of =issiles; protection (

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against potential effects of a fuel-handling accident; and the possible _ effects of systematic or coccon mede failures. Mos Lof.there No. 3. items are generic, not unique to Indian Point Unit Acceptable measures have been taken on Indian peint Uni: No. 3 vith regard to the on-site power syste=, hydregen concentratien buildup, and postula:ed fuel-handling cccidents. Studies are still underway on the potential for cissile genera: ion fro: gross reactor coolant pu:p overspeed in the event of certain postula:ed LOCAs; this =a::er should be resolved in a manner satisfactcry Oc the. Regulatory Staff. It is believed tha: r e s c i u t i e r, of the *hermal shock catter can avait the development of further Progra: andinfor:ation free the heavy Section Steel Technology other s:udies. With regard to anticipated tran-sien:s without scra=, the Concittee recommends that the recen:1~v a oninnf. u u r;c e d Regulatory Staff position be i=plemented for Indian P ) Unit Nc. 3 in ti ely fashion. l Because there is limited operating experience with very large, high power density reactors, the ACRS believes that initial operation should be limited to power levels no greater than 2760 MW(t) and that fur:her review by the Committee is appro-priate before higher power levels are per:10:ed. The Cc==ittee believes that, in the censideration of the operation of Unit i

       - No.      3 at higher including                  power levels, several fac: ors are pertinent, the folleuing:

and other similar reae:crs;satisfac cry experience in Unit Sc. 3 adequa:e knouledge of fuel perfer- ' mance; extent to which an independent confirmation of LOCA-ICCS l analysis has been =ade by the Regulatory S:aff; further resolution of relevant generic ca::ers; and consideration of the possibility of improvements in ECCF effectiveness. , I i The Committee recognires that re-evaluation of operatins limits

         =ay be necessary as a result. of possible changes in the accep-tance criteria for emergency core cooling systems. The Coc=1 :ee wishes to be kept infor=ed.

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Honorable.Dixy Lee Ray ggy 14 un The Applicant stated that he will apply and utilize suitable squipment to enable periodic tes:ing of the proper positioning of check valves in: ended to isolate low pressure sys:ets con-nected to the primary system. This =a::er should be resolved in a =anner satisfactory to the Regulatory Staff. Studies are underway with regard to the reliability of the service wa:er distribution to the diesel-generators. This matter should be resolved in a =anner satisfactory to the Regulatory Staff. The original turbine design has been found by the Applicant to have the possibility of overspeed somewhat beyond the manu-facturcr's design condition if the turbine should trip a: or near the design power. The Applicant is preparing design odi-fications to eli inata this condition, and will propose appropriate power li=itations until acceptabic :odifications have been made. This matter should be resolved in a manner satisfactory to the Regulatory Staff. The Co::ittee believes that several considerations are appro-priate in the further development of the Technical Specifications, as follows: operating heatup and cooldown pressure-te perature curves as conservative as practical with respect to 10 CFR Part 50, Appendix C; appropriate baseline inspection and pericdic

*n-service inspection of the stea= generator shells; startup of an idle loop at power; acceptable cumula:ive li=1:s on downtime of protec: ion systems and engineered safety fea:ures; and con-tinuing availability of core outle     ther occuples.

The Com:1: tee also believes that further consideration should be given to aug=ented use of :ovable in-core de:ectors, appropriate in-service inspection of noccles,in the primary head of the stea generators, and to the detailed specification of ad=inis:rative controls intended to prevent overpressurization of the reactor vessel below operating te=peratures. Generic problems relating to large water reactors have been identified by the Regulatory Staff and the ACES and discussed in the Co::ittee's report dated December 18, 1972. Those prob-le=s and additional generic proble=s identified in more recent ACRS reports should be dealt with appropriately by the Regulatory Staff and the Applican:. Y

l l Honorable Dixy Lee Ray , Roy I 4 UD The Advisory Connittee on Eeactor Safeguards believes that, if due regard is given to the ite=s nentioned above, and , subject to satisfactory completion of construction and pre-operational testing, there is reasonable assurance that Indian Point Nuclear Generating Station Unit No. 3 can be operated without undue risk to the health and safety of the public. The Con 10:ee believes that operation should be a: Power levcis no greater thaa 2760 MW(t) prior to further Co=nittee review. Sincerely yours,

                                                                      . Gusrav

8. G. Mangelsdebk Chair =an References A:: ached 1

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l Ow? ADVISORY COMMITTEE ON REACTOR SAFEGU ANDS UNITED STATES ATOMIC ENERGY COMMISSION W ash t NGTON. D.C. 20HS September 23, 1970 l . l Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Commission Washington, D. C. 20545

Subject:

rep 0RT ON OCONEE NUCLEAR STATION UNIT NO. 1

Dear Dr. Seaborg:

During its 125th meeting, September 17-19, 1970, the Advisory Committee on Reactor Safeguards completed its review of the application of the Duke Power Company for a license to operate Unit 1 of the Oconee Nuclear Station at power levels up to 2568 5'(t). The Comittee met with the applicant during its 124th meeting, August 13-15, 1970 and Subcocr ittee meetings were held on June 23, 1970, at the site and on July 31, 1970 and Septetber 9, 1970, in Washington, D. C. In the course of the review, the Comittee had the benefit of discussions with representatives and consultants of the applicant, the Babcock and Wilcox Cot pany, the Bechtel Corporation, and the _AEC Regulatory Staff, and of stud of the documents listed. The Oconee Station is located in a rural area of Oconee County, South Carolina. The nearest population centar is Anderson, 21 miles south, with a population of about 41,000. T1.e minimum exclusion distance for .- the completed three-unit pm statiec will be one c:ile and the Low Popu-lation Zone radius will be six miles :entaining about 3,400 people. The water supply for the plant is taken from Lake Keowee which was created by the applicant. The lake and associated recreational facilities are ex-pected to attract a transient population to the area. The application covers Oconee Units 1, 2, and 3, but this report applies only to Unit 1, which will employ the first of the Babcock and Wilcox two-loop, four-pump, pressuriced water reactor, nuclear steam supply systems.- The three units are designed to be nearly identical, but some facilities and services are shared in various arrangements. ne Cocruittee has reviewed the temporary arrangements necessitated by operation of Unit 1 while Units 2 and 3 are still under construction. It is believed that the proposed physical measures and administrative procedures to isolate the operating unit from construction activities are adequate. , ATTACHMENT 3

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ag' Honorable Glenn T. Seaborg September 23, 1970 The Comittee reported to you on the construction permit application for this power station on July 11, 1967. At that time the proposed operating power was to_ have been 2452 W(t); the current proposal for operating at powers as high as 2566 W(t) is justified by the applicant, primarily on the basis of a flatter power distribution. Prior to operation at the higher power level, reactor operation should be reviewed by the Regulatory 1 Staff. l The prestressed concrete containment building is similar to those for the l Palisades and Point Beach plants which have been reviewed recently for operation. 1 The Comittee recommends that the applicant accelerate his studies of means of preventing common failure modes from negating scram action and of design features to make tolerable the consequences of failure to scram when required during anticipated transients. As solutions develop and are evaluated by the Regulatory Staff, appropriate action should be proposed and taken by the applicant on a reasonable time scale. The Coccittee wishes to be kept in-formed. The applicant has proposed using a power-to-flow ratio signal as a diverse means to cause shutdown of the reactor if emergency core cooling action should be initiated. The Comittee believes it is necessary that either the I equipment associated with this signal be demonstrated to be able to survive the accident environment for an adequate time or a dif ferent, diverse trip signal be employed. This matter should be resolved to the satisfaction of the Regulatory Staff. The Comittee suggests that developmental techniques, such as neutron noise analysis and use of accelerometers, be considered as an aid in ascertaining displacements, changes in vibration characteristics, and the presence of loose parts in the primary systems. The Comittee notes the desirability of the continuing use of some thermocouples in the core. The Connittee has commented in previous reports on the development of  ; systems to control the buildup of hydrogen in the containment which might follow in the unlikely event of a loss-of-coolant accident. The' applicant proposes to make use of a purging technique after a suitable time delay sub-sequent to the accident. Relatively high off-site doses possibly could result follocing p'eging of the containment. The Cocnittee reco= mends that l purging systems be incorporated in the plant but that the primary protection j in this regard should utilize a hydrogen control method which keeps the  ! hydrogen concentration within safe limits by means other than purging. The i ( NhJe m" b

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         ,     gy,                                                                      l 6~ d u Honorable Gisnn T. Saaborg                                September 23, 1970 i

i hydrogen control system and provisions for containment aereosphere mixing ( and sampling should have redundancy and instrumentation suitable for an engineered safety feature; these should be made available within the firrt two years of power operation. The Committee wishes to be kept informed of the resolution of this matter. The applicant stated that the amount of radioactivity in liquid wastcs nor- I mally will not be greater than one percent of 10 CFR Part 20 limiting con- I centrations af ter dilution with the minimu= flow (30 efs) below the Keowee dam. Larger flows will have proportionately smaller limiting concentrations. The mean annual discharge from the Keowee dam is expected to be 1,100 cu. f t./ sec. The off-gas system has holding tank and filtering capability and gas re-lease rates are not expected to exceed a few percent of 10 CFR Part 20 limits. In order to protect against the postulated consequences of the accidental dropping of a fuel element, the applicant has stated that either, he will install filters in the fuel pool building exhaust system, or the equivalent control and protection will be assured by another method. This matter should be resolved to the satisfaction of the Regulatory Staff within the first year of power operation. l Improved calculational techniques are being applied to the analysis of the efficacy of the emergency core cooling system in the unlikely event of a loss-of-coolant accident. Interim results appear to be acceptable, but further calculations are needed and some phenomena important to the course of the accident require further study. This matter should be resolved in a manner satisfactory to the Regulatory Staff prior to operation at power. The Com=it-tee wishes to be kept informed. The reactor is calculated to have a positive moderator coefficient of reactiv-ity at power which will become negative as boron is removed from the coolant concurrent with build-up of fission products and fuel burnup. The applicant plans to perform tests to verify that divergent azimuthal xenon oscillations cannot occur in this reactor. The Cocnittee recommends that the Regulatory Staff follow the measurements and analyses related to these tests. A conservative method of defining pressure vessel fracture toughness should be employed that is satisfactory to the Regulatory Staff. Other problems relating to large water reactors which have been identified , by the Regulatory Staff and the ACRS and cited in previous reports to you ' should be dealt with appropriately by the Staff and applicant in the Oconee Unit 1 power plant as suitable approaches are developed. l ( l

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                . (v wu .         u Honorable _Glenn'T. Seaborg.                                         Septseber 23, 1970 The Advisory Committee on Reactor Safeguards believes that, if due
                    . regard is given to the items mentioned above, and subject to satisfac-tory completion of construction and preoperational testing there is reasonable assurance the Oconee Nuclear Plant Unit 1 can be operated:

at pcVer levels up to 2568 KJ(t) without undue risk to the health and safety of the public. Sincerely yours, , i I 1

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Joseph M. Hendrie Chairman l Additional comments by.Dr. W. R. Stratton are presented below:

                                 "The high off-site doses which are stated to accompany the proposed purging operation are based on calculations which include a number of assumptions which I believe to be overly                                            ,

conse rva tive . It is my opinion that the situation, should i it ever 'arise, would be much less severe and that the pro-posed purge system would provide adequate protection for the health and safety of the public in this regard and therefore the additional hydrogen contrel equipment required by this letter is not necessary."

Attachment:

List of References I i e em LlF[jf'f I l l

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UNITED STATES ATOMIC ENERGY COMMISSION wAsWNGroN, D.C. 20545 Augu s t 20, 1970 1 l l 1 Honorable Glenn T. Seaborg Chairma U. S., Atomic Energy Co= mission Washinsten, D. C. 20545

Subject:

REPORT ON DAVIS-3 ESSE NUCLEAR P0 tier SIATION '

Dear Dr. Seaborg:

N At its 124th meeting, August 13-15, 1970, the Advisory CoE$.ittee on Reactor Safe;uards co=pleted its review of the application by the i Toledo Edisen Cc=pany and The Cleveland Electric Illuminating Ccapany for a per=it to construct the Davis-3 esse Nuclear Power Station. A subco==ittee met to review the project on May 2b, 1970, at the site and in Toledo, Ohio, and on August 4,1970, in Washington, D. C. t During its review, the Committee had the benefit of discussions with representatives and consultants of the applicants, the Babcock and. Wilcox Company, the 3echtel Corporation, and the AEC T. ;ulatory Staff. The Co==ittee also had the benefit of the documents listed. i The plant will be located on the southwestern shore of Lake Erie ap-proxi=ately 21 miles east of Toledo, Ohio. The nearest populatien centers are Toledo and Sandusky, Ohio, each about 20 miles from the site, with pcpulations in 1960 of 379,000 and 32,000, respectively. Tbc city of Fremont, Ohio, with a 1960 population of about 15,000, is located 17 miles free the site. The minimum exclusion distance is 2400 feet and the low population zone distance is two miles. Approx-imately 3200 people live within five miles of the site. < Camp Perry, an Ohio National Guard facility, is located on Lake Erie about five miles east of the site. This installation is used during a short period each year for target practice with small arms and with 40-==. anti-aircraft At the Erie Industrialshells Park, armed about only with a scall destruct charge. three to four miles east of the site, Cadillac Gage Company is engaged in testing ordnac,ce equipment firing 120-mm. mortar shells with a taxt:a= range of about two miles. All firing from both locations is directed into restricted areas in Lake Erie. Ite applicants have prcyided studies which demonstrate that none , of the projectiles now being fired from these installations could 1 l l l

                                                                                        . ATTACHMENT 4 l

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Bonercble Clenn T. Seaborg .2- August 20, 1970 l penetrate the heavy reinforced concrete structures provided to protect the essential portions of the plant. The Com=ittee recocoends , how- - ever, that the applicants and the Regulatory Staf f make suitable arrange-ments to be informed of any changes in these activities so t, hat their

possible ef fect on the safety of the plant may be evaluated.

l. An area in Lake Erie about ten miles north of the site is used by air- ' craft from the Scifridge Air Force Base.in Eichigan as an Anti-Submarine Warfare practice area and by the Lockbourne Air Force Base at Columbus, Ohio, as an impact area for automatic weapon firing from aircraf t. The applicants have been given assurance by officials of the Department of l Defense that military aircraf t.enroute to or from this area vill not be l routed closer than ten miles from the site. The Co=mittee believes that  ! this arrangement reduces, to acceptably low levels, the probability of an aircraft strikin; tbc plant, but reco= mends that formal arrangements be made to enabic the applicants and the Regulatory Staf f to caintain I i continuing awareness of the operational patterns of military aircraf t in this area. The Davis-Besse plant will include a two-loop pressurized water reactor similar to those for the Midland units except that the internal vent valves have been eliminated by changes in the elevations of the steam generators to obviate their need. Since the proposed arrangement elim ' inates the possibility of coolant flow bypass through an open vent valve, the Davis-Besse reactor is designed for an initial core power level of I 2633 MWt as compared to 2452 MNt for the Midland units. The applicants stated that it will be pessible to anneal the pressure vessel if this should become necessary at so=a time af ter operation is begun.

          'A suitable preoperational vibration testing program should be e= ployed for the primary sys te=. Also, attention should be given to the devel-opeent and utilization of instrumentation for in-service monitoring for excessive vibration or loose parts in the primary system.

The containment consists of a steel vessel surrounded by a reinforced concrete shield buildi.ig, with the annular space maintained at a slightly negative pressure and the air from this space exhausted through filters. This design is similar to that for the Prairie Island, Kcwaunce, and Butchinson Island plants, except that the free volume of the steel con-t a inme n t is much greater, nearly thre.e million cubic feet. .The Regula- - tory Staff should review the contain=ent design pressure to assure that an adequate margin of conservatism exists. l [

l l Eenorable Clenn T. Scaborg August 20, _>19 7 0 l l l r Detailed criteria re=ain to be formulated by the applicants fer tha de- j sign of the penetrations for the het process pipes which traverse the l annulus between the two containment barriers. I= view of the i=portance of these penetrations, criteria should be reviewed by the Resulatory Staff to assure adequate conservatis=, and the applicants should arrange for an independent review of the actual design. , The Cdt:ittee has co==ented in previous reports on the develop =ent of syste=s to control the buildup of hydrogen in the_ containment which might follow in the unlikely event of a less-ef-coclant accident. The applicants are studying various methods of coping with this proble=, including purging and the use of catalytic rece=biners. The Cc=mittee reco=nends that the primary protection in this regard should utilize a hydrogen control =ethod which keeps the hydrogen concentration within safe 'li=its by means other than purging. The capability for purging should also be provided. The hydrogen control syste= and provisions for centainment at=csphere mixing and sampling shou 1d have redundancy

   ,   and instru=entation suitable for an engineered safety feature.        The Co=-

mittee wishes to be kept infor=ed of the resolution of this matter. The applicants have stated that they vill provide additional evidence ob-tained by i= proved culti-node analytical techniques to assure that the e=ergency core cooling syste= is capable of limiting core te=peratures to acceptably conservative values. They will also =ake app:cpriate plant i changes if further analysis demonstrates that such changes are required. This =atter should be resolved during construction in a manner satisfac-tory to the Regulatory Staff. The Co==ittee wishes to be kept inforced. The Cercittee reco== ends that the applicants accelerate the study of means ' to prevent ce==on f ailure modes f:c= negating scra= action, and of design features to =ake tolerable the consequences of failure to scra= rsring anticipated transients. The applicants stated that the engineering design would maintain flexibility vich regard to relief capacity of the pri=ary syste= and to a diverse means of reducing reactivity. This =atter should be resolved in a manner satisf actory to the Regulatory Staff during con-s t ruc tion. The Co==ittee wishes to be kept informed. The' Co==ittee believes that consideration should be given to the utiliza-tion of instrumentation for prompt detection of gross f ailure of a fuel elenent. Consideratico should be given also to the use of core exit ther=ocouples as an aid to reliable operation and as an additional method ) of detecting behavior ano=alies. - 4 l i

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Eonorable Glenn T. Seaborg 4 August 20, 1970' l l l

     .The applicants propese ba:ch discharge of liquid wastes following treat-ment. Concentrations of radionuclides in the discharge will be kept well below 10 CFR 20 limits with positive dilution being provided from several equip = ant cooling water streams. Plans for operation of waste treatment equipmen: should be such as to mini = ice the quantities of           ,

radioactivity discharged, and provisions should be made to achieve rapid dispersion in the lake. Other proble=s rela:ed to large water reactors have been identified by the Regulatory Staff and the ACRS and cited in previous ACRS reports. The Co==ittee believes that resolution of these items should apply equally to the Davis Besse plant. ~s The Co==ittee believes that the above ite=s can be resolved during con-struction and that, if due consideration is given to these items, the Davis-Besse Nuclear Power Station can be constructed with reasonable assurance that i: can be operated without undue risk to the health and safety of the public. Sincerely yours , ,' W Joseph M. Hendrie Chairman p2 fe r enc e s :

1. Letter fre: Toledo Edison Cc=pany, dated August 1, 1969; License Applica: ion, Volures 1, 2 and 3 of the Preliminary Safety Analysis Report (PSAR) --

2. Volume 4 of the PSAR, dated April 16, 1970

3. A=end=ents 1 through 9 to License Application

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UNITED STATES [60 htGu\ !E { c qff e NUCLEAR REGULATORY COMMISSION $ t,, y yy.. i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 g v%*:/j"ej o,,1

     .....                                                  July 17, 1980 APPENDIX XXXI SEQUOYAH NUCLEAR PLANT DESIGN OF INTAKE STRUCTURE MEMORANDUM FOR:        William J. Dircks, Acting Executive Dirac* r ior                erations FROM; R. F. Frale , uecutive Upctor, ACRS

SUBJECT:

SEQUOYAH NUCLEAR PLANT DESIGN OF INTAKE STRUCTURE During the Committee's review of the full power operating license for the Sequoyah Nuclear Plant, Units 1 and 2 (243rd ACRS meeting), a question was raised regarding the safety of the new coolant intake building which will be used when both units begin operation. The NRC Safety Evaluation Report states that the new intake building has been constructed such that the probability of losing the water supply due to collision of a drifting barge is 4 x 10-8 per year. The intake structure may survive the impact of a drifting barge, but the ACRS would like to know whether the structure is vulnerable to ' collision of a barge at full speed from any credible direction, in-cluding a tow proceeding in the upstream direction, and the proba-bility of such an event. The Committee would like to be informed regarding the ability of the intake structure to withstand the effects of barges carrying flam-mable cargoes including LNG. The Committee wishes to be informed regarding the results of this - study. cc: T. Rehm, E00 G. Zech, NRR R. Savio, ACRS ACRS Members [ 2)[ _ __

l APPENDIX XXXII

 ,$      ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE

1. Proposed rule,10 CFR 60, Technical Criteria for Regulatory Geologic Discosal High-Level Radioactive Waste

2. Update of Chapter V of TMI Action Plan: "NRC Policy, Organi:ation, and Management."

3. Press release by Sen. Gary Hart, Chmn. U. S. Senate Subcommittee on Nuclear Regulation, concerning Special Senate Investigation of the Nuclear Accident at Three Mile Island, dta July 2, 1980.

4 Memo from B. W. Sheron, NRC Staff, to D. F. Ross, NRC Staff, Status of Reactor Coolant Pumo Trip and Effect on Non-LOCA Decressurizing Transients, dtd. Apr. 17, 1980.

5. Letter, G. G. Sherwood, General Electric Co. to M. S. Plesset, Chmn. , ACRS, Anticioated Transients Without Scram - General Electric Comments on ACRS Letter Dated April 16, 1980, dtd. May 30, 1980

6. Letter, M. P. Oncavage to M. W. Carbon, ACRS, Request for ACRS Involvement in Turkey Point steam generator problems, dtd. June 13, 1980.

7. Letter, J. O. Mingle, Kansas State University , to Rep. M. McCornack, U. S.

House of Representatives, on regulatory reform of NRC, dtd June 3,1980. I I t l A- 05 2}}