ML20128E865

Summary of 930121 Meeting W/Ge & Other Stated Companies in San Jose,Ca Re Status of ABWR Review.List of Attendees Encl

ML20128E865
Person / Time
Site:	05200001
Issue date:	02/04/1993
From:	Joshua Wilson Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9302110115
Download: ML20128E865 (47)

Text

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=4 , pa neg'o{ UNITED 8TATES

,, g NUCLEAR REGULATORY COMMISSION f t, :s WA$HINGTON, D. C,20665 - L

• • • February 4, 1993 l

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Docket No.52-001 l l

APPL.lCANT: GE Nucle.- Energy (GE)

PROJECT: Advanced Boiling Water Reactor (ABWR). i

SUMMARY

OF MEETING W1fH GE ON JANUARY 21, 1993 j A senior management meeting was held on January 21, 1993, betweer the U.S. -'

Nuclear Regulatory Comrrission.'(NRC) staff and GE representatives at GE's 1 office in San Jose, California. The purpose of the meeting was:to discuss the  !

status of the ABWR review. A list of attendees is provided in Enclosure 1 and o the meeting agenda is previd(.d in Enclosure 2. ] '

The meeting began wit' presentation by GE on the accomplishments of.the team-that was formed to re\ sw GE's inspections, tests, analyses, and acceptance

• criteria (ITAAC) for the ABWR.- A copy'of the slides used in this presentation U is provided in Enclosure 3, and a list of the;1TAAC review team' members is-included in Enclosure 3. GE concluded that, based upon the resolution of .

10 system ITAACLand 7 generic' issues. acceptable.ITAAC,can be develo)ed, i However, the process requires a multi-disciplinary team approach witi ';

deteiled, prenriptive-guidance for the preparers and reviewers.-  ;

Based upon the success of the ITAAC review team,-GE requested that:the team stay together and complete 1the remaining ABWR_ITAAC. The staff stated that j  ;

they could not continue their participation in the review team because the a resource impact on the NRC. associated with the'. development of .the 10 system-  ;

liAAC could not be sustained for the remaining ABWR-lTAAC. The staff 1also- ' 1 stated that the approved ITAAC and generic issues should be used by.GE as a- t model to complete development of the remaining lTAAC. The staff wil17 prepare -

ITAAC review guidance based u)on the lessons learned by the review team and j will meet with GE in early-Fearuary to discuss GE's ITAAC preparation- 4 guidance. The staff also plans to participate in another ITAAC review team in early March-1993 that wil1J be-formed to evaluate-a; seismic Class 1: building, n  :

and the turbine building ITAAC.; After GE submits their completed-ITAAC' fori -

).

the ABWR, the staff will perform an acceptance review. . If the ITAAC submittal-is com)1ete and of high-quality, then the staff will initiate its review.-

~

d When tie ITAAr review nears completion and has only a small number of -

significant problems,'then the, staff may reconvehe an -ITAAC review team to'

-h

resolve the remaining ITAAC open issues. -

3

.The meeting continued with;a discussion on the estimated schedule for.a i i de; '-ion on issuance of an~ final design approval for the ABWR design.- GE - .

stated that, based upon lessons learned from the ITAAC. team review,1they would,  :

reassess their schedule for resolution of the remaining open issues,- y

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- 2- february 4, 1993 j completion of the revised safety analysis report, and completion of ITAAC. GE predicted that they would send their revised schedule to the NRC by the end of January. The staff stated that they would use the schedule information from GE to prepare a paper for the Commission on the ABWR review. i The next item un the agenda was the status of the probabilistic risk assessment (PRA) review (Enclosure 4). The approach to identifying insights from the PRA was discussed with GE. Once agreement is reached between the staff and GE on the insights, then the staff will use the insights in their review of the safety analysis report and the ITAAC.

The staff made a presentation on the status of the open items identified in the draft-final safety evaluation report (DFSER) and the progress towards  !

resolution of the open items for each technical branch (Enclosure 5). GE provided their assessment for four technical branches. GE agreed to provide.

their open item status to the staff each Tuesday for NRG management review. ,

lhe remainder of the meeting dealt with three open issues. The first was resolution of station blackout for the ABWR. GE stated that the combustion turbine will be in the certified scope of the design, and it will be verified -;

by ITAAC. The staff agreed to review whether the combustion turbine will-have to be covered in the technical specifications or the reliability assuranco program.

The next open issue involved the diversity of the instrumentation. and control system. G[ made the presentation shown in Enclosure 6. GE and tho staff agreed that in addition to the three analyses that were already performed by GE, three additional analyses will be completed by GE. These analyses will-address:

(1) steamline break (2) RHR suction line break (3) bottom drain line break The new analyses will be completed using the basic format and approach of-the previously. completed ana' lyses. Only the control rod drive hydraulic system-will be modeled in the analyses as a source of vessel makeup. Operator actions will.be defined based upon the ABWR symptom based emergency operating procedures, ,

The final open issue involved human factors operating experience. ' A copy of the handout is provided in Enclosure 7. In order to resolve this issue, GE agreed to include the words--in Enclosure 7 in the next update to the ABWR-standard safety analysis report' (SSAR). GE also agreed to include the appropriate input guidelino documents that were referenced in draft NUREG-5908 '

(fourth quarter of 1992)-in appendix Table-18E.2 of the SSAR under 18.E.2.1 VI and 18.E.2.1 VilI~ The staff will provide a copy of the draft NUREG-5908 to GE by January 27,-1993, and GE will complete their actions by the end of-January.

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_ - _ ~ - + , , y ,

4 february 4.-1993 1he meeting concluded with Mr. Berglund of GE thanking the staf f for its effort in the ITAAC review. Dr. Murley concluded that, based upon the experience with the ADWR review and the results of the ITAAC review team, Part 52'can work to resolve issues for design certification.

i (Original signed by)

Jerry N. Wilson, Section Chief Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor _ Regulation

Enclosures:

I As stated

• cc w/ enclosures:

See next page 7 DllTR11UT10N w/ enclosures:

Docket File PDST R/F PShea PDR CPoslusny JNWilson RBorchardt DISTRIBUJJON w/o enclolum: .

1Murley/FMiraglia DCrutchfield GKelly, 10E4 JNWilson GGrant, EDO EJordan, MNBe3701- ACRS (11) LShao, RES J0'Brien, RES BHardin, RES JNWilson W1 Russe 11 Alhadani,,8E2 CHcCracken, 801- JRichardson DTerao, 7H15

- WBurton, 801 G1homas, 8E23 MChiramal, 8117 DThatcher, 7E4 CGoodman, 10D24 GKelly, 10E4 JThompson MJanus 1 WHuffman , 1 Moore, 15B18

. OFC: LA:PDST:ADAR SC:PDST: .

NAME: PShea')0 JNWils /I DATE: 02/// j 9[3'302/4 /93

-OfflCIAL RECORD COPY:ABWRHTG3.JNW 4 ,

.r  :

i y y v- v w r . y , - - - - - - -

V

GE Nuclear Energy Dotket No.52-001 cc: Mr. Patrick W. Marriott. Manager Mr. Joseph Quirk Licensing & Consulting Services GE tiuclear Energy GE Nucicar Energy General Electric Company 175 Curtner Avenue 175 Curtner Avenue, Mail Code 782 San Jose, California 95125 San Jose, California 95125 Mr. Robert Mitchell General Electric Company 175 Curtner Avenue San Jose, California 95125 Mr. L. Gifford, Program Manager Regulatory Programs GE Nuclear Energy 12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852 Director, Criteria & Standards Division Office of Radiation Programs V. S. Environmental Protection Agency 401 M Street, S.W.

Washington, D.C. 20460

_ Mr. Daniel F. Giessing U. S. Department of Energy NE-42 Washington, D.C. 20585 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.

Room 8002 Washington, D.C. 20503 Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874 Mr. Raymond Ng 1776 Eye Street, N.W.

Suite 300 Washington, D.C. 20006 Marcus A. Rowden, Esq.

Fried, Frank. Harris, Shriver & Jacobson 1001 Pennsylvania Avenue, N.W.

Suite 800 Washington, D.C. 20004 Jay M. Gutierrez, Esq.

Newman & Holtzinger, P.C. )

1615 L Street, N.W. l Suite 1000 Washington, D.C. 20036

. MEETING ATTENDEES JANUARY 21, 1993 11 Nil afilLIAllD!i  ;

J. N. Wilson NRC D. Crutchfield HRC

.W. T. Russell NRC .

- 1. Murley NRC A. Thadant NRC -l*

C. McCracken NRC.

l J. Richardson NRC D. Terao NRC i G. Thomas NRC M. Chiramal NRC D. Thatcher NRC-t C. Goodman NRC G. Kelly NRC J. Thompson NRC '

W. Durton NRC 4 S. Franks DOE D. McGoff 000 N fletcher DOE /ALWR R. Burke EPRl/ALWR

- D. Wilson EPRI/ Niagara Mohawk  ;

~

T. Gabarain EPRl/UNESA

~

J. Berger EPRl/EDT .;

C. Brinkman ADH-CE A. Ueymer NUMARC .;

1. Fernandez Yankee Atomic Electric ~Co. .-

~'

J. Waldron C'eveElect-(GE)

A. James GE-B. Berglund GE R. Louison GE ,

C. Sawyer GE S. Hucik GE  ;

f. Cooke GE J. Sawabo GE N. Hackford GE J. Duncan GE J. Chambers .GE

- R. Buchholz GE .,

D.-Wilkins GE J. Quirk GE P. _ Marriot_t GE Ehclosure1- ,

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AGENDA GEINRR SENIOR MANAGEMENT MEETING JANUARY 21, 1992 SAN JOSE, CALIFORNIA e introduction e ITAAC Review Team Progress (GE, Staff) e FDA Schedule (Staff) e PRA and Severe Accident Closure Status (GE, Staff) e DFSER Unresolved items (Staff)

Status of Open and Confirmatory items Status of Resolution Meetings and Telecons

}

e Other issues / Unresolved l&C Diver sty Status (GE, Staff)

Statlo;. Blackout Resolution for ABWR (GE, Staff) e Conclusion Enclosure 2

," ABWR DE1LGN CERTIFICATION 1

1/21/93_BANAG.EMENT MEETING  ;

MED.A o SNAPSHOT

SUMMARY

4 o BACKGROUND f o OBJECTIVES OF REVIEW

( o ISSUES REVIEWED AND

SUMMARY

OF RESULTS o EXAMPLES OF AGREED-T0 MATERIAL NUCLEAR BOILER SYSTEM

- REACTOR SERVICE WATER .

o L4PORTANT LESSONS LEARNED o WHAT NEXT?

o

SUMMARY

AND CONCLUSIONS AJJ-1 1/21/93 Enclosure.3

EE/ERC_ITAAC REYlEd5 SNAPSH0T

SUMMARY

o INTENSIVE NRC/GE SESSIONS FOR 10 DAYS STRAIGHT

- CANDID, INTERACTIVE, DECISIVE o AGREEMENTS REACHED ON:

10 SYSTEM ITAAC 7 GENERIC ISSUES 1 DAC SUPPORTING SAR REVISIONS OR TECHNICAL APPROACH o CONCLUSION: HUTUALLY ACCEPTABLE ITAAC CAN BE DEVELOPED SUCCESSFUL EXERCISE AJJ-2 1 1/21/93

+

GE/NRC ITAAC REVIEWS BACKGR011ND o CONTINUOUS SESSIONS 1/11 - 1/20 o OPEN MEETINGS WITH REPRESENTATION FROM: -

NRC GE NUMARC EPRI W

ABB-CE AEP TVA NIAGARA MOHAWK

( DOE CEI SOUTHERN CO.

PROCESS:

GROUP REVIP4-0F GE ITAAC AND GENERIC ISSUES -

GE ENGINEER REVISE GE/ ASSIGNED NRC ENGINEER REVIEW -

GROUP RE-REVIEW AND CONCURRENCE PREPARE CONTROL COPY MARKUP WITH PUNCH LIST ITEMS AJJ-3

-1/21/93

.. GE/NRC ITAAC REVIEWS l

1 l

ATTENDEES  !

l l

J. LYONs NRC S. FRANKS DOE i H. WALKEn NRC N. FLETCHER DOE  !

l W. BURTON NRC S. FRANTz NEWMAN & HOLTz!NGER (GE)

G. THOMAS NRC A. P. NEYMEn NUMARC I D. TERA 0 NRC D. Wits 0N NIAGARA MOHAWK /EPRI-T. SULLIVAN NRC J. REC ABB-CE R. LI NRC E. WHITAKEn TVA/EPRI 1

D. THATCHEn NRC W. L. ZIMMERMANN AEP ,

J. STEWAnt NRC D. ANTOLOVICH .W ,

l

N. CHIMAnMAL NRC J. WALDn0N CEI/SBWR T. POLICH NRC J. WHELEss .SOUTHEnN CO..

T. BOYCE NRC A.-J. JAMES GE J. WILs0N NRC R. L0uzSON GE" R. GRAMM NRC' N. HACKrono GE-S. MALun NRC J. CHAMsEns GE

~

W-russell NRC J. F. 0uInK- GE M.FINKELsTEIN NRC .

C. MCCRACKEN NRC l PLUs SUPPORTING GE TECHNICAL PERs0NNEL As NEEDED.

-AJJ-4A-21/21/93 .

. -. . . _ = __ _ . _ .. . - _ . . _. - -- m . - _ _

.GE/NRC ITAAC REVIEWS S_UPPORTING GE TECHNICAL PERSONNEL SISIEMllSSE ELERGIEEB151 LEAK OETECTION AND ISOLATION SYSTEM H. G. TOTAH REACTOR WATER CLEANUP $YSTEM E. V. NAZARENO NUCLEAR BOILER SYSTEM J. K. SAWABE F. E. COOKE STANDBY LIQUID CONTROL SYSTEM P. F. BILLIG REACTOR COOLING WATER SYSTEM G. E. MILLER

- REACTOR SERVICE WATER SYSTEM G. E. MILLER k RESIDUAL HEAT REMOVAL SYSTEM W. E. TAFT TURBINE MAIN STEAM SYSTEM J. C. BLACK CONDENSER J. C. BLACK CONTROL ROOM HABITABILITY AREA M. MUNSON HVAC SYSTEM CONTROL BUILDING SAFETY-RELATE 0 H. MUNSON EQUIPMENT AREA HVAC SYSTEM WELDING L. FINNEY D. SANDUSKY AJJ-4B l 1/21/93 1

s  !

.- GE/NRC ITAAC REVIEWS.  !

SEPPORTING _GE TECHRLCALEERS_0NNEL 5.15IEM/ISSME GE ENGINEER (S) ,

l EQUIPMENT 00ALIFICATION (50.49) N. G. LURIA I D. C. RENNELS SEISMIC /0YNAMIC QUALIFICATION N. G. LURIA D. C. RENNELS j i

MOV (89-10 ISSUES) B. GENETTI i G. L. MOORE.

1 EMI/RFI/SWC B. H. SIMON

. I&C ENVIRONMENTAL QUALIFICATION B. H. SIMON q

I INSTRUMENT SETPOINTS A. J. JAMES SECTION 1 0F THE CERTIFIED DESIGN A. J. JAMES DOCUMENT S. FRANTZ (LEGAL)

PIPING DESIGN ACCEPTANCE CRITERIA J. B. KNEPP M. HERZOG E. O. SWAIN ELECTRICAL SEPARATION C. F. CHRISTENSEN AJJ-4C 1/21/93

GEZNRC ITAAC-REVIEWS i

DEJECTIVES_.0F REVIEW FOR A SET OF ABWR ITAAC AND GENERIC ITEMS, DEVELOP AGREED-T0 MATERIAL FOR THE FOLLOWING:

1. DESIGN DESCRIPTION TEXT

2. DESIGN DESCRIPTION FIGURES

3. ITAAC TABLE ENTRIES

( 4. SUPPORTING TIER 2 (SAR) MATERIAL OR WRITTEN TECHNICAL APPROACH 4 1

MD_fR0 DEI MARKED-UP CONTROL COPIES PUNCH LIST OF CLEAN-UP ITEMS AJJ-5

-1/21/93 i

__ _ _ _ . . . _ . . _ _ - . . . . a

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n -SUM 4ARY OF RESULTS

~ AGREED-TO PUNCH MATERIAL IN PLACE: LIST jf SYSTEN OR GENERIC TOPIC D(! ITAAC TIER 2 ITEMS COMMENTS ,

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1. . LEAK DETECTION AND Y Y Y 4 -FAIL-SAFE LOGIC TEST ISOLATION SYSTEN .

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2. : REACTOR' WATER' Y Y Y 3 -

!' CLEANUP SYSTEN ,

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-TIER 2 ANALYSIS REQUIRED -

3. NUCLEAR BOILER Y Y Y- 7 -ADDRESS CURRENT BE5t LEVEL ISSUE SYSTEN -ADD SENSORS 4.. 5TANDBY LIQUID Y Y Y 2 -PREOP INTERFACE i

.. CONTROL SYSTEN  ;

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5. REACTOR C00 LING 1 'Y Y; 'Y 7 -SSAR NEAT LOADS  ;

WATER SYS1!EN- -SAFETY /NON-SAFETY ISOLATION

[:-- ,

i I 6... REACTOR SERVICE .Y .Y Y- 2 -UPDATE SSAR

WATER'SYSTEN; -PICK UP FLOOD INTERFACE AJJ-6 1/21/93- j l

- . - . - - - . . . ~ -_ _ , . _ _ . _ _ ___

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

, ~. .

SUl#lARY10F ~RESULTS-t ' AGREED-TO-- PUNCH MATERIAL IN PLACE:- ' LIST SYSTEM OR GENDaC TOPIC DQ IIAAC TIER 2 ITEMS. C0fetEMTS i: 7.. RESIDUAL HEAT. .Y. .Y Y S l -INTERLOCK LOGIC REMOVAL SYSTEM i

8. TURBINE MAIN 'Y Y Y S -INCLUDE AUXILIARY STEAM VALVE

-STEAM SYSTEM i

V:-

9. CONDENSER. Y. Y 'Y 1 -

1

10. CONTROL ROOM -Y- Y N 7

-REQUIRES REWDRK (E.s., TOKIC  !

HABITABILITY i

. GAS Als SAR DESIGN ISSUES)

AREA'dVAC SYSTEM '

'  ?

II. CONTROLLBUILDING' NOT~ COMPLETED; i

{: SAFETF-RELATED EQUIP- SAR DESIGN ISSUES U MENT AREA HVAC SYSTEM i

y

'12..-PRESSURE BOUNDARY- Y

• Y- .9- -

WELDING-r P

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L1/21/93:

L , . -. .. . . . _ . . . . . ... ... _ . _ . . . _ - _ _ _ _ _ _ . . . _ _ . _ . . _ . . _ _ _ . . _ . _ _ _ _ _ _ .

SUPMARY OF RESULTS. -

AGREED-TO PUNCH MATERIAL IN PLACE: LIST SYSTEN OR GENERIC TOPIC DR ITAAC TIER 2 ITEMS COMMENTS

13. EQUIPMENT GUALI- Y

• Y Z -

NON-PROPRIETARY TIER 2 COMPLETE FICATION-(50.49)

14. SEISMIC / DYNAMIC Y

• Y 0 -

ALSO ADDRESSES INSTRUMENTATION QUALIFICATION AND CONTROL

-l

15. MOV (89/10 ISSUES) Y Y/* Y 0

16. EMI/RFI/SWC' y N N Y- 3

17. IK ENVIR00 MENTAL QUALIFICATI'ON 18.-INSTRUMENT SETPOINTS N N Y 3- - NON-PROPRIETARY TIER 2,'DD AND ITAAC TO BE DEVELOPED

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- AJJ 1/21/93-

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SU R RY OF RESULTS

~

e AGREED-TO PUNCH F MATERIAL-IN PLACE: LIST SYSTEN OR GENERIC TOPIC ILD ITAACf TIER 2 ITEMS COMMENTS l

'19. SECTION 1 0F Y NA NA NA -TECHNICAL AGREEMENT; LEGAL l-l - THE CERTIFIED REVIEW PEWING i

[ DESIGN DOCLMENT i

i. '

i.

20. PIPING' DESIGN. Y Y Y 3 -ASME AM NNS ACCEPTANCE CRITERIA I

j h 21.. ELECTRICAL Y Y Y 0 -APPLIES TO POWER AND

- INDEPENDENCE: INSTRUMENTATION AM CONTROL l i ,

j

22. FUEL POOL NOT REVIEWED: SIMILAR TO COOLING;. SYSTEM - OTHER FLUID SYSTEMS  :

F a' i'

0

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AJJ-9 -:

I/21/93 i

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.G E / N R C I T A A C_- R E V I E W.S EXAMPLES OF_C.0MELETED ITAAC i

l

1. NUCLEAR BOILER SYSTEM KEY HECHANICAL SAFETY-RELATED SYSTEM ILLUSTRATES HOST OF THE AGREED-T0 APPROACHES e

2. REACTOR SERVICE WATER

(

ILLUSTRATES THE TREATMENT OF INTERFACES (PER PART 52)

AJJ-10 1/21/93

l l

GE IMPORIANT LE110NS LEARRED A. PROCESS

1. IT ISN'T EASY; SCRUTINIZING JUST ABOUT EVERY WORD IN WHAT l WILL EVENTUALLY BE A 1000-PAGE DOCUMENT.

l l

2. INTENSIVE GROUP REVIEWS LIKE THIS ARE ESSENTIAL FOR TIMELY RESOLUTION OF ISSUES.

3. A MULTI-DISCIPLINARY CORE REVIEW TEAM OF PART 52-KNOWLEDGEABLE PERSONNEL SHOULD BE IN PLACE FOR THE

( DURATION.

, INTENDED GE INTERNAL APPROACH

4. IN ADDITION TO THE GROUP REVIEW PROCESS, PREPARATION AND REVIEW OF THE REMAINING 80% WILL REQUIRE:

AGREED-T0 EXAMPLES  : DONE STYLE GUIDE  : GE TO UPDATE EXISTING DOCUMENT REVIEW GUIDANCE  : NRC TO PREPARE AN ITAAC "SRP" AJJ-11 1/21/93

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GE IME0RIANT LES10RS_LEARHEQ B. TECHNICAL / CONTENT

1. DETAILED, PRESCRIPTIVE GUIDANCE IS IMPORTANT A. STANDARDIZED WORDS AND FIGURES (BOILERPLATE)

B. SEQUENCING OF SUBJECTS C. GUIDANCE ON SELECTING APPROPRIATE ISSUES FOR TIER 1 TREATHENT

(

D. GRAMMAR, SYNTAX, USAGE

2. TIER 1 SELECTION PROCESS INVOLVES JUDGHENTS BY KNOWLEDGEABLE ENGINEERS

3. ENGINEERS HUST UNDERSTAND T'IE LEGAL SIGNIFICANCE OF CERTIFICATION AJJ-12 1/21/93

GE/NRC_ITAAC REVIEd WHAT NEXI?

EE WORK OFF PUNCH LIST ITEMS RE-WRITE OF STYLE GUIDE TO REFLECT LESSONS LEARNED INTERACT WITH STAFF POLISH AGREED-T0 KATERIAL SEQUENCING OF SUBJECTS EDITORIAL / USAGE ITEMS SUBHIT PACKAGE

(

NRC/GE/NUMARC CONDUCT JOINT BUILDING REVIEW (3/93)

- AFTER STRUCTURAL AUDIT

- SIMILAR PROCESS PREPARE AND SUBMIT REMAINING ABWR KATERIAL FOR STAFF REVIEW

- INTERNAL GE REVIEW NOT CERTIFIED FOR APPENDIX B COMPLIANCE SCHEDULES FOR THESE ITEMS: TO BE ESTABLISHED AJJ-13 1/21/93

S1JMMARY AND_C.0NCLUSIONS_- GE o MUTUALLY ACCEPTABLE ITAAC CAN BE DEVELOPED o AN AGREED-T0 SAMPLE SET OF SYSTEMS AND GENERICS IS NOW IN PLACE o MULTI-DISCIPLINARY TEAM APPROACH SIGNIFICANTLY EXPEDITES THE PROCESS (AND WILL BE RETAINED AT LEAST FOR GE'S ITAAC ACTIVITIES)

(

o VERY DETAILED, EXPLICIT, PRESCRIPTIVE GUIDANCE FOR PREPARERS AND REVIEWERS IS ESSENTIAL TO SUCCESS SUCCESSFUL EXERCISE I

AJJ-14 1 1/21/93

i s.[ 95 AB.WR PRA IS_ SUE STATUS

• CONFIRMATORY ISSUES - Aff 48

• OPEN ISSUES - # 14 THE STAFF AND GE HAVE AGREED ON THE STATUS OF THESE ISSUES.

~

GE AND STAFF DISCUSS ISSUES SEVERAL TIMES PER WEEK.

SEVERAL OPEN ISSUES READY TO BECOME CONFIRMATORY.

OPEN ISSUES IN TIE DF ER BECAUSE WE BROKE SOME ISSUES INTO SUBTASKS A4D BECAUSE WE TRACKED SOME ISSUES ON OUR LISTS THAT WERE NOT IN THE DFSER.

Enclosure 4

~

ADT OPEN ITEM STATUS -

ABWR=SER l'

~l a m m

@ m -

_ m g mh _

r_

m m - - .

~

~

" ~

DE - 152 _

DRCH - 19 O - -

99gg _7 DSSA - 144~

O ',"

,. - DRIL - 6 DORS ~- 3 wm -

ADAR -J=

(I) ,. _ _ . . .

n 2n

- TOTAL -343 -

_ I I  ! i 8 I f I I ! I I k

i I T f i I I 3 ! ! $

f I f  ! i armt w so s'te 117 12e

• WEEK STARTING BASED ON FSER INPUTS DATA AS OF 01/15/93

3

g. e t

DFSER ISSUE RESOLUTION MEETING /TELECON STATUS

BRANCH ISSUES CO' CHAPTERS STATUS j SPLB 160 3,6,9,10,11 DESIGN MEETINGS 10 & 12/92 l l TELECON CH. 6 ON 1/12/93 l F ITAAC RESOL. IN PROGRESS L. SRXB- 23 4,5,6,15 TELECOM 1/13/93  ;

F ITAAC RESOL. 'N PROGRESS L

SD RISK. TELECON WEEK OF 1/18/93 SPSB 78 19.1 TELECON 1/11, MEETING 1/20

, ITAAC MEETING NEEDED .

L L SCSB 152 19.22 TELECON 1/17. TELECON' WEEK OF l

p 1/25. ITAAC MEETING NEEDED.

l HICB 43 7 . TELECON/ MEETING BY END OF JAN.

. ITAAC RESOL. IN PROGRESS HHFB '24 18 ' MEETING BY..END OF JAN.

n FOR DESIGN AND'ITAAC ISSUES  :

I 2

INCLUDES lOPEN.AND CONFIRMATORY ITEMS. NOT INCLUDED IN DFSER.

v

~

. . . . . - - - - ~_ .. . __ --.- . -- -- -_.

BRANCH I_SSUES(#)' CHAPTER _S STATU_S EELB 142 8 Nov. & DEc. MEETINGS /TELECONS.

DRAFT SSAR REVISION FROM GE EXPECTED BY MID-JANUARY.

ITAAC MEETING REQUIRED ECGB 2 122 2,3,5,9,14 TELECONS BY JAN. END.

AUDIT IN FEB.

ITAAC MEETING NEEDED.

PEPB 3 13 TELECON IN DECEMBER,SSAR MARKUP PROVIDED. AWAITING ITAAC SUBMITTAL PSGB 1 13 TELECON IN DECEMBER, SSAR MARKUP PROVIDED. AWAITING ITAAC SUBMITTAL.

PRPB 3 2,6,15 MEETING ON 1/25-DES.& ITAAC 8 12 TELECON 1/11.

AWAITING ITAAC SUBMITTALS

' INCLUDES OPEN AND CONFIRMATORY ITEMS.

2 INCLUDES MECHANICAL, STRUCTURAL, GEOSCIENCE, CHEMICAL, AND MATERIALS REVIEW AREAS.

4

. , _ - _ _ __i- m_

7 BRANCH H ISSUES (O' CHAPTERS STATUS

. RPEB- 22 14,17- MEETING ON 1/8/92  :

AWA m MG JAN. SU8MITTAL i AWA m MG ITAAC Suen m AL -

p 0TSB. 3 16- AWA m MG JAN. SUsM m AL DAR 6 14 ITAAC' RESOLUTION IN PROGRESS 8 1,20 TELECOM 1/13 i MEETING IN FEBRUARY '

I i h-'- lINCLUDES OPEN AND CONFIRMATORY-ITEMS. I i

I

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DSFER ISSUE RESOLUTION STATUS (NON-ITAAC)

Confirmatory items COL Action items Tectt Soc items Ooen items Status Sta*us Status Status Total Tota!

NRC Total Tota!

items Closed Rescived items Gipsad Resolved items Closed Resolved Branch items Closed Resolved 54 17 0 5 1 0 53 17 4 52 29 0 SPLB 30 0 0 0 0 3 0 0 74 22 0 1 EELB" 2 0 2 0 0 0 2 0 2 0 0 0 PEPB 8 0 8 0 0 0 0 0 0 0 0 0 PSGB Transmittal of Chapter 8 markup resoiving ali non-ITAAC issues scheduled for 1/25/93 Closed - No further GE action required (NRO to "close")

Resolved - rdarkup of SSAR provided to NRC I

' JNF/ARP1/012193

O GENuclearEnergy -

ABWR-I&CDiversity Presented to U S. NuclearRegulatory Commission

!? M. A. Ross

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January 21,1993 sm

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I&CDiversityissue Issue: "The staffis concerned that the use of digital computer technology in I&C system could result in safety-significant common mode failures"

ABWR Design Privention simple software '

comprehensive software design / validation and verification Mitigation through diversity rapid response (scram and isolation) previded in Main ;

. ControlRoomi Core makeup water capability of continuously operating '

^isedwaterand CRD systems.

longer term actions supported by analog, two divisions of .

Remote'ShutdowiSystemi

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I&C Diversityissue (cont'd) l o

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Status '1 NRC agreed that longer term actions from outside MCR is.

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l; reasonable' and defined " longer term" to be after first hour of l initiating event (5y92) ,

o  :

l GE completed analyses of Chapter 15 events which demonstrated j '

[ that feedwater and CR0 flows alone are sufficient for at least l firsttwo hours (6/92) u NRC completed review of GE analysis (12/92)

NRC Preliminary EvaluationConclusions

. ]

Provide diverse HPCFsystem manualinitiation in MCR

-: Feedwater Control System to be designed and tested to l verify operation during Chapter 15 events j

- LRationale i

-many of the events rely on feedwater system to

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operate formitigation ?

HPCFprovides additionalassurance '

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I&C DiversityIssue (cont'd)

GEResponse

- ABWR Feedwater System enhanced performance

-- two stages vs three stages ofpumps

-- adjustable speedmotordriven pumps

-- vesselpressurization rate (~ 25%) slower

-- triplicated fault tolerant control system

-- control logic to avoid level 8 pump trip

- Bounding analyses completed with only CRD flow

-- results demonstrate that at least two hours capability with only one CRD pump (second pump normally in standby)

-- Isolation transients 1430'F PCT (vs22WF limit)

-- HPCFline break 17WF PCT

-- Feedwaterline break 1860'F PCT (previouslysubmitted)

- Conc!ude that ABWR design is adequate as is

-- no MCR hardwired HPCF manua! initial required

-- no increase in feedwater system design requirements and

[ testing necessary

4 DRAFT 1/20/93 EVENT

15.2.2.2.1.3 GENERATOR LOAD REJECTION WITH FAILURE OF ALL DYPASS VALVES, ,

15.2.2 2.1.3 TURDINE TRIP WITH FAILURE OF ALL BYPASS VALVES These events are postulated to occur coincident with a undiscovered common mode f allure of the Essential Multiplex System (EMUX) in such a manner that all valid and correct EMUX control and rnonitonng data transmissions are lost. The reactor response to these two events are similar, AUTOMATIC ACTIONS Upon a turbine / generator trip, the reactor scrams immediately The scram signals generated by turbine stop valve or centrol valve instruments are hardwired to RPS. The SRVs open on spring setpoint since it - t is assumed that the SRVs cannot be open by its normal relief mode due to the postulated common mode f allure. RPV makeup is performed only by one CRD pump.

EOP ENTRY CONDITIONS:

The following alam1 condition is provided by equipment independent of the EMUX and is an entry condition to the emergency operating procedures: ,

SAR Appendix 18F Reference

1. RPV W ATER LEVEL LOW [ FIXED PCslTION), Column 8, IBF-14.

OPERATOR ACTIONS For this event scenario, the CRD pump is used to inject water into the RPV. It is assumed that because of the failure of the bypass valves, the bypass valves can not be reopen for decay heat removal.

1, Enter EOPs developed from the RPV Control Guideline, upon receiving the RPV Water Level Low

, alarm.

2. Restore and maintain water level (water level signal is hardwired) above Levei 3 using the CRD -

pump.

3. If water level cannot be mp!ntained above Level 3, control water level above top of active fuel.

4. If water levet drops below the Minimum Zero-Inect water level, perform an emergency depressurization (because of the assumed failure of the bypass valves and the control capability of ,

ADS and SRVs, this operation is not possible).

CONTAINMENT RESPONSE Since steam is discharged to the suppression pool, suppression pool temperature will increase. As

- calculated for the Steam Line Break Outside Containment event, the suppression pool temperature rise  ;

due to the integrated decay heat for one hour was estimated to be approximately 40 'F. For one hour, condenser hotwellinventory needed for decay heat removal is conservatively calculated to b4 approximately 23% of the totalinventory in the hotwell 11is concluded that initiation of suppression pool cooling using RHR will not be required for the first hour of this postulated scenario in which all operator actions are limited to those performed in the main control room. The RHR suppression pool cooling function is assumed to be not available because of the EMUX common mode failure Subsequent operator actions at the Remote Shutdopwn System willinclude initiation of the RHR suppression pool cooling function.

I ANALYSIS RESULTS The analysis assumed only the use of one CRD for RPV makeup. The SRVs are assumed to open on the spring relief setpoint and discharge steam to the suppression pool. The turbine bypass valves are assumed to f ail and hence cannot be used 1or pressure control ant:t decay hea'. removal. The analysis results are given in Figures 1 to 3. The peak clad temperature is calculated to be a maximum of1430 T, well below the 2200 'F limit.

SUMMARY

For the postulated event ci turbine trip or generator trip with total failure of the bypass valves coincident with an EMUX common rnode f ailure, sufficient information and controls that are independent of EMUX are available in the main control room to mitigate the event, assuming that all operator actions will be limlted to the main control room for the first hour. Sufficient water inventory is available for decay heat removal during that one hour period. Subsequently, reactor cold shutdown conditions and post accident recovery operations can be initiated using the Remote Shutdown System.

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DRAFT 1/20/93 EVENT: LOCAINSIDE CONTAlNMENT This event is postulated to be a break of HPCF(C) line coincident with a undiscovered common mode f ailure of the Essential Multiplex System (EMUX) in such a manner that all valid and correct EMUX control and monitoring data transmissions are lost.

AUTOMATIC ACTIONS After a HPCF(C) line break inside containment, the reactor would expected to automatically scram on High Drywell Pressure. Because of the assumed EMUX common mode failure,this scram is postulated to f ail. In adddion, all ECCS systems are assumed not be available because of the postulated comrnon mode failure and the break postulated in HPCF(C). The feedwater and condensate systems are assumed to f ail. For tha postulated line break, the RPV pressure will drop rapidly, resulting in closure of the turbine control valves by the pressure regulator and automat <: scram signals (hardwired) f rom the subsequent turbine trip. The MStVs should close when RPV water level drops to Level 1 but because of the assumed common mode f ailure, the MSIVs will remain open. Refer to Figures 1 through 3 for the reactor response to the break .

EOP ENTRY CONDITIONS:

The following alarms are provided by equipment independent of the EMUX. These are the entry conditions for emergency operating procedures expected f or a LOC A inside the primary containment from instruments that are hardwired.

SAR Appendix 18F Reference Column 8,18F-121

1. DRYWELL PRESSURE HIGH [ FIXE 0 POSITION).

OPERATOR ACTIONS PER EOPS The expected principal operator actions are given herein. All control functions and process parameters are provided by equipment independent of EMUX.

Upon entering the EOPs deve. loped f rom the RPV Control Guideline on High Drywell Pressureas an entry condition, and concurrently enter EOPs developed f rom the Primary Containment Control Guideline on Drywell Pressure High alarm as an entry condition, the following sets of actions are executed concurrently:

1. RPV Control

1. Initiate a manual scram if a scram has not been initiated .

2. Maintain RPV water level (water level signal is hardwired) above Level 3 using one CRD pump only. Since the postulated break is located at the HPCF(C) injection line which is below Level 1 but above top of the active fuel, water will spill out the break and into the drywell.

3. 11 RPV water level cannot be maintained above Level 3, maintain RPV water level above top of the active fuel.

4. When RPV water tevel drops below the Minimum Zero-inject water level (below top of fuel),

perform an emergency depressurization (because the reactor pressure is depressurizing due to uncovery of the assumed break, this action is not necessary).

l I

OPERATOR ACTIONS PER EOPS (continued)

11. Primary Containment Control:

1.

Initiate wetwell sprays using the fire protection system and the firewater addition mode of RHR(C) for primary containment pressure control.

2. If necessary, initiate drywell spray using the fire protection system and the firewater addition mode of RHR(C ) tor primary containment pressure control (drywell pressure signe'is hardwired),

CONTAINMENT RESPONSE As calculated for the Main Steam Line Break Outside Containment event, the suppression pool temperature rise due to decay heat over a period of one hour was less than 40 'F. For this event, the suppression pool temperature rise is expected to be less than 40'F due to the wetwell and/or dr)well sprays initiated by the operator.

1.OCA ANALYSIS The complete circumferential break of the HPCF(C) injection line was analyzed. For this case only the control rod drive makeup system was assumed to be available for RPV makeup. The CRD takes suction from the the condensate sysiem (hotwell) or from the CST.

After scram the CRD injects water into the RPV. The CRD continues to inject water into the vessel but it is not enough to overcome the break tiow, The water level continues to drop and the core is uncovered (Figure 3), At approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from the start of the event the peak cladding temperature is only about 1700 'F which is well below the 2200 'F limit.

SUMMARY

For the postulated event of HPCF(C) line break inside the containtnent, sufficient automatic controt f unctions, information, and controls that are independent of EMUX are available in the main Control room to mitigate the event and maintain the fuel clad temperature below its limit, assuming that all operator actions will be limited to the main control room for the first hour. Sufficient water inventory is available for dect , leat removal during that one hour period. Subsequently, reactor cold shutdown conditons and post accident recovery operations can be initiated using the Remote Shutdown System.

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I&C Diversity-#4 Conclusions

• Current text reads, in part "3. The following display capability shall be added:

a. RPVwaterIbvel

b. RPVwaterlevelllevel3) alarm

c. Drywellpressure

d. Drywellpressure(high) alarm

e. CUWisolation valve status

f. RCICsteamline isolation valve status

g. HPCFflow These displays shall be safetygrade. The displays may be analog components or simple, dedicated and diverse software-based digitaldisplays.

4. The remote shutdown station displays shallbe operable during normal operations. This willpermit an operator to assess the status of the displayedparameters without transferring control from the main controlroom."

I&C Diversity '#4 Conclusions (Cont'd) e L

[

• Delete HPCFflowfrom3g(notrequired) v

• Revise item #4 text to read as follows:

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"In the event that instrumentation channels utilized at the Remote Shutdown Station are also used to implement the requirements of.

Item #3 above; those instrument channels shall be configured such that the Remote Shutdown Station displays are operable during i normal operations. This willpermit an operator to assess the statss of the displayedparameters without transferring control from the main controlroom." ,

. This text modification required for clarity of the NRC intent. .

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GENuclearEnergy ABWR - HfE Operating Experience Review Presentedto U.S. Nuclear Regulatory Commissionn

, J l I g M. A. Ross 2 January 21,1993

'S

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1 .

Proposed Resolution - Operating Experience Review Issue l

L l Incorporate the following into the Tier 11 DAC, SSAR Table 18.E.2.1, Article II.2 as newitems "c"and "d".

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l.

The HFE Program Plan shall also establish:

"c. That the nfCR and RSS designs shall be implemented using HSi equipment technologies which are consistent with those defined ,

in Section 18.4.3 of the Standard Safety Analysis Report (SSAR).

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c Proposed Resolution - Operating Experience Review issue (cont'd) e d. That in the event other HSI equipment technologies are alternatively selected for application in the MCR and RSS design

[ implementations:

(i) a review of the industry experience with the operation of those' selected new HSI equipment technologies shall be conducted;

a i-(ii) the Operating Experience Review (GER) of those new HSI

' equipment tecimologies shallinclude both a review of _

literature penaining to the human factors issues related to L similar system applications of those new HSI equipment technologies and interviews with personnel experienced with the operation of those systems;and (iii) any relevant HFEissues/ concerns associated with those .

selected new HSI equipment technologies, identified through ;

i the conduct of the OER, shall be' entered inta the HFIissus '

Tracking System forclosure."

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