ML18040A222

Summary of 940222 Meeting W/Niagara Mohawk Power Corp in Rockville,Md to Discuss 10CFR50,App J,Re Issues & Design Basis Reconstitution Program at Util

ML18040A222
Person / Time
Site:	Nine Mile Point
Issue date:	03/14/1994
From:	Brinkman D Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9403210010
Download: ML18040A222 (95)

Text

Docket No. 50-220 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&0001 March 14, 1994 LICENSEE:

Niagara Mohawk Power Corporation FACILITY:

Nine Mile Point Nuclear Station Unit No.

1

SUMMARY

OF FEBRUARY 22,

1994, MEETING TO DISCUSS 10 CFR PART 50, APPENDIX J, ISSUES AND THE DESIGN BASIS RECONSTITUTION PROGRAM AT NINE MILE POINT NUCLEAR STATION UNIT NO.

1 (NHP-1)

A meeting was held in the NRC One Mhite Flint North Office in Rockville, Maryland, with Niagara Mohawk Power Corporation (NMPC) and NRC staff representatives to discuss 10 CFR Part 50, Appendix J, issues and the Design Basis Reconstitution (DBR) program at NHP-1.

NHPC had requested this meeting.

Enclosure 1 is a list of meeting attendees.

Enclosures 2 and 3 are copies of the handout materials provided by NHPC.

During this meeting, NHPC representatives briefed the NRC staff representatives on NHPC's updated plans for satisfying 10 CFR Part 50, Appendix J, requirements for the:

(1) emergency condenser

system, (2) shutdown cooling system, and (3) containment spray system.

NHPC representatives stated during the next refueling outage (refuel outage 13, currently scheduled for spring 1995),

NHPC will replace/refurbish the emergency condenser containment isolation valves such that these valves will then meet the Appendix J criteria; This pr'oposed action is consistent with the revised schedular exemption issued to NHPC on July 24, 1992.

The NRC staff had no further concerns regarding the emergency condenser containment isolation valves.

proposed approach appears gag It:gX tI,e>TESFP>

@1'1 V'

.i 0 L.a.v.ll

'P4032iOOiO 940314 PDR ADQCK 05000220 P

PDR The July 24,

1992, schedular exemption, was also applicable to the shutdown cooling containment isolation valves.

This exemption had been issued to permit plant operations,to continue until refuel outage 13.

NHPC had committed to replace/add valves to provide containment isolation and Appendix J testing capabilities during refuel outage 13.

During the meeting, NHPC proposed that rather than installing the previously planned modification, they would satisfy the requirements of Appendix J by installing a waterseal pressurizing system between the inboard and outboard isolation valves.

NHPC stated that the proposed modification will meet the requirements of Section III.C.3(a)(b) of Appendix J except when the unit is being cooled from approximately 300 'F to less than 215 'F during which time isolation would be provided by only a single valve.

NMPC stated that utilization of the proposed waterseal would reduce occupational exposures, avoid a 2-23'-week extension of refuel outage 13, and save approximately 2 million dollars required to install new valves.

The NRC staff agreed to consider this proposed change and stated that we would advise NHPC of our decision in the near future.

Subsequent to the meeting, the NRC staff concluded that this

c.

1

~ gM

'I

March ]4, 1994 feasible.

Therefore, NHPC was requested to submit a proposed license amendment if NHPC desires to implement the proposed approach.

The containment spray system isolation valves currently utilize a waterseal.

NHPC had previously committed to replace/add valves to provide containment isolation and Appendix J testing capabilities in lieu of the waterseal.

However, during the meeting, NHPC proposed to utilize the existing waterseal to meet Appendix J on a long-term basis.

The containment spray system waterseal is maintained as long as the containment spray pumps continue to

operate, but this waterseal may not be maintained if the pumps are de-energized.

The NHP-1 emergency operating procedures provide for de-energizing the containment spray pumps during a design basis accident when the containment pressure decreases to below 3.5 psig.

NHPC evaluated the potential radiological consequences of interrupting this waterseal and concluded that the 10 CFR Part 100 and General Design Criterion 19 guidelines would not be exceeded.

NHPC concluded that continued utilization of this waterseal would:

(1) reduce testing of the isolation valves, (2) avoid a

2-4-week extension of refuel outage 13, (3) result in decreased occupational exposures, (4) result in a savings of approximately 4.8 million dollars, and (5) that there would be no overall safety advantage for installing the valves.

The NRC staff agreed to also consider this proposed change and stated that we would advise NHPC of our decision in the near future.

Subsequent to the

meeting, the NRC staff concluded that this proposed change also appears feasible.

The NRC staff also concluded that interruption of the waterseal requires an exemption pursuant to 10 CFR 50. 12 since it is the NRC staff's position that Section III.C.3 of Appendix J requires a continuous waterseal for at least 30 days at a pressure of 1.10 Pa.

Therefore, NHPC was requested to submit a proposed exemption justifying interruption of the waterseal.

NHPC described the DBR program accomplishments.

NHPC reported that 21 system design basis documents and 25 design criteria documents have been prepared in 4 years.

The DBR program is expected to be terminated in mid-1994.

The DBR

1.j March 14, 1994 process is being incorporated into the NMPC design engineering process and will be used on an as-needed basis in the future if any remaining system should require reconstitution.

Enclosures:

1.

Attendance List 2.

NMP-1 Appendix J 3.

NMP-I Design Basis Reconstitution Donald S.

Brinkman, Senior Project Manager Project Directorate I-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation yacc w/enclosures:

See next page

l

Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station Unit No.

1 CC:

Hark J. Wetterhahn, Esquire Winston 5 Strawn 1400 L Street, NW Washington, DC 20005-3502 Supervisor Town of Scriba Route 8, Box 382

Oswego, New York 13126 Vice President - Nuclear Generation Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station P.O.

Box 32

Lycoming, New York 13093 Resident Inspector U.S. Nuclear Regulatory Commission P.O.

Box 126

Lycoming, New York 13093 Gary D. Wilson, Esquire Niagara Mohawk Power Corporation 300 Erie Boulevard West

Syracuse, New York 13202 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Hs.

Donna Ross New York State Energy Office 2 Empire State Plaza 16th Floor

Albany, New York 12223 Mr. Richard B. Abbott Unit 1 Plant Manager Nine Mile Point Nuclear Station P.O.

Box 32

Lycoming, New York 13093 Hr. David K. Greene Manager Licensing Niagara Mohawk Power Corporation 301 Plainfield Road

Syracuse, New York 13212 Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, New York 10271 Hr. Paul D. Eddy State of New York Department of Public Service Power Division, System Operations 3 Empire, State Plaza

Albany, New York 12223 Mr. Martin J.

McCormick, Jr.

Vice President Nuclear Safety Assessment and Support Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station P.O.

Box 63

Lycoming, New York 13093

'r.

B. Ralph Sylvia Executive Vice President, Nuclear Niagara Mohawk Power Corporation 301 Plainfield Road

Syracuse, New York 13212

~,

l Hr. B. Ralph Sylvia March 21, 1994 This exemption is required for the current condition and will be required if NHPC decides to pursue the proposed change in commitment.

Please feel free to contact me on (301) 504-1409 if you have any questions regarding these matters.

Sincerely, Original signedby:

~t

Enclosure:

February 22, 1994 Meeting Summary w/o enclosures cc w/enclosure:

See next page DISTRIBUTION:

'Docket File NRC 5 Local PDRs PDI-1 Reading SVarga JCalvo RACapra CVogan DBrinkman OGC ACRS (10)

RBarrett, 8/H/7 JPulsipher, 8/H/7 RLobel, 8/H/7 CCowgill, RGN-I Donald S. Brinkman, Senior Project Manager Project Directorate I-1 Division of Reactor Projects I/II

'Office of Nuclear Reactor Regulation OFFICE PDI-1:LA PDI-1: PH BC:

PDI-1:D IIAHE CVo an ~

DBrink a RBa t

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RACa ra+

DATE 94 X j 94, 7 Z/ 94 9 2/94 OFFICIAL RECORD COPY FILENAME: G: iNHPliNHPIAPPJ. LTR

P

Enclosure 1

Attendance List Februar 22 1994 Meetin To Discuss NMP-1 10 CFR Part 50 A

endix J Issues and NMP-1 DBR Pro ra

~Na e

Donald S.

Brinkman Robert A. Capra Gene Imbro Richard Lobel Jim Pulsipher Rick Plasse Lee Klosowski Car l Terry David K. Greene Robert F. Oleck Timothy Lee George Inch Michael G. Hosier Gary Wilson David Baker Norm Rademacher George K. Wierzbowsky Robert Cazzolli Robert Kirchner Hark Wetterhahn Kevin Graney Position Senior Project Manager

Director, PDI-1

Chief, Spec Insp Br Section Chief, SCSB Containment Sys.

Branch, NRR NRC Resident Inspector, NHP Gen Supv. - Unit, 1 Design VP Nuc.

Eng.

Hgr. - Licensing Program Director - DBR Supv.

Mech Design Ul Engineer Ul Plant Eval Licensing Engineer Managing Counsel Licensing Program Director Operations Manager Lead Systems Engr.

Supervisor - Chemistry 8

RP Sup Engineer NMPC Nucl Technology Attorney Serch Licensing/Bechtel Power NRC/NRR/PD I-1 NRC/NRR/PDI-1 NRC/NRR NRC/NRR/SCSB NRC/NRR/SCSB NRC/Region I NHPC NHPC NHPC NHPC NHPC NHPC NHPC NMPC NMPC NMPC NHPC NHPC NMPC Winston 8 Strawn Bechtel

(

P

Enclosure 2

NIAGARAMOHAWKPOWER CORPORATION NINE MILEPOINT NUCLEAR STATION UNIT 1 February 22, 1&4 V NIAGARA 4 MOHAWK

I I

I

Niagara Mohawk Power Corporation Nine Mile Point - Unit 1 Appendix 1 - NRC Presentation AGENDA I

INTRODUCTION & PUIU'OSE L. A. KLOSOWSKI II EMERGENCY CONDENSER T. D. LEE III SM3TDOWN COOLING T. D. LEE IV CONTAIMNENTSPRAY

~

DESIGN

~

WQ)IOLOGICALASSESSMENT

~

CONTAIÃHENTLEAKAGEANALYSIS T. D. LEE R. J.. CAZZOLLI G. B. INCH

~

PRA REVIEW R. F. KIRCEXNER V

S Y & CONCLUSIONS C. D. TERRY

I

INTRODUCTION

~

10CFR50 APPENDIX J PROMULGATED 1973 NMP1 REQUIRED NUMEROUS PROCEDURE/H&U3WARE CHANGES TO COMPLY STATUS

~

NMP1 SATISFIES APPENDIX J REQUIREMENTS FOR ALLCONTAINMENTPENETRATIONS AND ISOLATION VALVES EXCEPT:

EMERGENCY CONDENSER SYSTEM (SCHEDULE EXEMPTION TO RFO 13 - 1995)

~

SHUTDOWN COOLING SYSTEM (SCHEDULE EXEMPTION TO RFO 13 - 1995)

~

NMPC'S PREVIOUS PLAN TO INSTALL"TYPE C" TESTABLE VALVECONFIGURATIONS TO COMPLY

~

NMPC PLAN TO REPLACE EXISTING CONTAINMENT SPRAY WATERSEAL WITH "TYPE C" TESTABLE VALVECONFIGURATIONS

I J

PURPOSE

~

PROVIDE NMPC'S UPDATED PLANS FOR SATISFYING APPENDIX J REQUIREMENTS FOR:

~

EMERGENCY CONDENSER SYSTEM

~

SMJTDOWN COOLING SYSTEM CONTAINMENTSPRAY SYSTEM

~

CEGWGES FROM PREVIOUS PLANS FOR SHUTDOWN COOLING AND CONTAINMENTSPRAY RESULT FROM REVIEWS OF:

SHUTDOWN SAFETY

~

OPERATIONAL SAFETY

~

PERSONNEL tuQ3IATION EXPOSURE OUTAGE DURATION

~

COST

~

APPENDIX J REQUIREMENTS FOR CONTAINMENTSPRAY SYSTEM WILLBE MET BY CONTINUED USE OF A WATERSEAL

~

APPENDIX J REQUIREMENTS FOR SEG3TDOWN COOLING SYSTEM WILLBE MET BY INSTALLATIONOF WATERSEAL INSTEAD OF VALVEMODIFICATIONS

~

APPENDIX J REQUIREMENTS FOR EMERGENCY CONDENSER SYSTEM WILLBE MET BY INSTALLING "TYPE C" TESTABLE CONFIG&&TIONS

) J I

J

APPENDIX J - COMPLIANCE UNIT 1 PLANS TO MEET APPENDIX J FOR:

~

EMERGENCY CONDENSER

~

SHUTDOWN COOLING

~

CONTAINMENTSPRAY PRESENTATION:

~

PREVIOUS PLANS

~

CURRENT PLANS

~

PURPOSE FOR CHANGING THE WAYWE COMPLY WITH APPENDIX J

4 P

r I I

EMER EN NDE E

~

PREVIOUS PLAN

~

REPLACE/REFURBISH VALVES SUCH THAT THEY MEET APPENDIX J CRITERIA

~

CURRENT PLAN MEET CURRENT APPENDIX J COMMITMENT

~

REPLACE AIR ACTUATORS ON OUTSIDE ISOLATION VALVES REPLACE'INSIDE ISOLATION CHECK VALVES

~

TEST ISOLATIONVALVES PER APPENDIX J CRITERIA

~

COMPLETE DURING REFUEL OUTAGE 13 (CURRENTLY SCHEDULED FOR SPRING 1995)

~

J

~

PREVIOUS PLAN

~

REPLACE/ADD VALVES TO PROVIDE CONTAINMENT ISOLATIONAND APPENDIX J TESTING CAPABILITIES

~

CURRENT PLAN WILLSATISFY APPENDIX J BY WATERSEAL PRESSURIZING BETWEEN INBO&U3AND OUTBOP3U) IV'S

~

NMPC PERFORMED A CONCEPTUAL DESIGN EVALUATION WITH RESPECT TO NMP1 DESIGN BASIS, TECHNICAL SPECIFICATIONS, 50.59, SAFETY, APPENDIX J AND EOP'S.

NMPC WILLPROVIDE A QUALIFIED CONTINUOUS WATERSEAL SOURCE TO MEET THE FOLLOWING SPECIFICATIONS.

+

DESIGN FOR A LOCA, LOOP, SINGLE ACTIVEFAILURE

+

WILLMEET APPENDIX J SECTION III.C.3(a)(b)

~

CURRENTLY WATER TEST THE SDC IV'S IN THE ACCIDENT DIRECTION; WILLCONTINUE TO

'ERFORM THESE TESTS IN ACCORDANCE WITH TECHNICALSPECIFICATIONS.

~

CAPABLE OF PROVIDING A WATER SEAL FOR 30 DAYS AT 1.1 P,

+,.

EOP ACTIONS WILLNOT AFFECT THE WATERSEAL FOR DESIGN BASIS EVENTS

+

WATERSEAL SUPPLY SYSTEM WILLBE A SAFETY RELATED DESIGN WITH POST-ACCIDENT FUNCTION

+

DESIGNED FOR SEISMIC LOADS

+

SEPAIMTE EMERGENCY DIESEL SUPPLIES FOR REDUNDA1A'OMPONENTS

r I'

~

SDC WATERSEAL IS DESIGNED FOR SINGLE ACTIVE FAILURES DURING REACTOR OPERATION EXCEPT UNDER THE CONDITIONS WHERE THE SDC SYSTEM IS PLACED IN OPERATION DURING A SHUTDOWN, IE., REACTOR PRESSURE BELOW 120 PSI

+

SDC VALVES ARE WATER TESTED TO MEET A MAXIMUMOF 5 GPM LEAFAGE

+

SDC WOULD ISOLATE ON A LOCA SIGNAL

+

SDC OPERATION IS INTERLOCKED TO PREVENT OPERATION ABOVE 120 PSI WHICH IS 10% OF THE DESIGN PRESSURE

+

SDC IS IN OPERATION A VERY LIMITEDAMOUNT OF TIME

+

DUE TO THE ABOVE, THE PROBABILITYOF CORE DAMAGEWHICH WOULD REQUIRE THESE VALVESTO SEAL PRIMARY CONTAINMENTATMOSPHERE IS EXTREMELYLOW; PROBABILITYOF CORE DAMAGE WHjILESDC IS IN OPERATION IS 1.1E-8 PER YEAR

+

DURING THE TIME SDC IS IN SERVICE, THE CORE DECAY HEAT LEVELS ARE A SMALLFRACTION OF OPERATING CONDITIONS

+

THEREFORE, NMPC CONCLUDES THIS HAS EXTREMELYMINIMALSAFETY IMPLICATIONS

I

v

~

~

ADVANTAGES OF WATERSEAL APPROACH SHUTDOWN SAFETY AVOIDEDRAG)IATIONEXPOSURE TO INSTALLAND TEST NEW VALVES

+

55 6 MANREMFOR MODIFICATIONINSTALLATION

+

) 20 5 MANREMFOR LONG-TERM IN SERVICE TESTING AVOIDEDREFUEL OUTAGE 13 EXTENSION OF APPROXIMATELY2 TO 2.5 WEEKS

~

SAVINGS OF APPROXIMATELY$2 MILLIONTO INSTALL NEW VALVES

~

SUMMARY

~

PLAN TO COMPLETE DURING REFUEL OUTAGE 13 (CURRENTLY SCHEDULED FOR SPRING 1995)

REQUIRES AN OUTAGE TO IMPLEMENTTHE WATERSEAL

~

NRC ACTIONS-

+

NMPC WILLSUBMIT A TECHNICALSPECIFICATION CEGQ4GE FOR PRESSURE ISOLATIONVALVES ASSOCIATED WH'H THE WATERSEAL.

W

)

SDC WATER SEAL SOC WATER SEAL SUPPLY SOURC ORAIN TO REACTOR BUILOING EQUIPMENT ORAIN TANK FLOW ~ 28GPM I/2'EACTOR REACTOR PROTECTION SYSTEM FROM SHUTOOWN COOLING 38-13 38-12 38-79 38-78 r

HC I)-

I HO AC

~~REACTOR PROTECTION SYSTEM I

I

~PRESSURE INTERLOCK I'Q~

38-81 38-82 38-73 38-72 TO SHUTOOWN COOLING p-

I 1

NTAINVIENT PRAY

~.

DESIGN A MODIFICATIONPLAN T. LEE

~

IUG3IOLOGICALASSESSMENT R. CAZZOLLI

~

CONTAINMENTLEAIMGEANALYSIS G. INCH

~

PRA REVIEW R. KIRCHNER

I I

NTAINMENT PRAY DESIGN

~

PREVIOUS PLAN

~

REPLACE/ADD VALVES TO PROVIDE CONTAINMENT ISOLATIONAND APPENDIX J TESTING CAPABILITIESIN LIEU OF WATERSEAL

~

CURRENT PLAN UTILIZETHE EXISTING WATERSEAL TO MEET APPENDIX J ON A LONG-TERM BASIS

+

DESIGN BASIS ASSUMES CONTINUOUS SPRAY THROUGHOUT THE DBA

+

EXISTING WATERSEAL UTILIZES THIS DESIGN BASIS

+

ON THIS BASIS NMPC SATISFIES. APPENDIX J FOR CONTAINMENTSPRAY

I I

~

NTAIMCENT PR%'

NMPC'S PREVIOUS PLAN TO REPLACE THE VALVES WAS BASED ON THE COMPLEXITYOF MAINTAININGA WATERSEAL WHEN EOP'S SECURE CONTAINMENTSPRAY NMPC PROCEDUI& N1-OP-14 WHICH IMPLEMENTS THE WATERSEAL HAS ELIMINATEDTHIS COMPLEXITYDURING EOP IMPLEMENTATION

~

EOP IMPLEMENTATION

+

CONDITIONS IN EOPS WHERE THE CONTAINMENT SPRAYS ARE SECURED

+

WATERSEAL IS NOT MAINTAINEDWHEN THE SPRAYS ARE SECURED

+

BY NOT MAINTAININGTHE WATERSEAL WHEN THE SPRAYS ARE SECURED THE COMPLEX OPERATOR ACTIONS WHICH WOULD BE REQUIRED ARE ELIMINATED

+

THESE EOP ACTIONS WHICH INTERRUPT THE WATERSEAL DURING DESIGN BASIS EVENTS HAVE BEEN EVALUATEDFROM A I4Q)IOLOGICALSTAt'G) POINT AIG) DO NOT IMPACTTHE LICENSING BASIS FOR NMP1

+

THE EVALUATIONWAS PERFORMED IN ACCORDANCE WITHTHE NRC SAFETY EVALUATIONWHICH IMPLEMENTED BWROG-EPG, REV. 4

+

CONCLUDED THAT 10CFR100 AND GDC 19 GUIDELINES WOULD NOT BE EXCEEDED

)

~

I

~

1 I

~

ADVANTAGESOF WATERSEAL APPROACH

~

REDUCES TESTING REQUIRED

~

AVOIDS REFUEL OUTAGE 13 EXTENSION OF APPROXIMATELY2 TO 4 WEEKS INCREASED IUQ3IATIONEXPOSURE TO INSTALLAND TEST VALVES

~

SAVINGS OF APPROXIMATELY$4.8 MILLION

~

NO OVERALLSAFETY ADVANTAGEFOR INSTALLINGTHE VALVES

J

CONTAINMENT SPRAY SYSTEM CONTAINMENT SPRAY MOOE 88-41 88-15 88-48 00 K

EI0Z0O ILI III IL00 Ka0Z0O ILI III 88-35 88.45

-16

- PRIHARY LOOP PRIMARY LOOP ORWELL 88-11 88-12 BB-IIB 88-32 I

88-31 88-86 BALL JOINTS UL CONTAINHENT SPRAY Ply Ill 88-81 88-21 I

BALL JOINTS LS.

CONTAINMENT SPRAY PUHP 112 88.87 BALL JOINTS CONTAINMENT SPRAY PUMP 121 88.82 88-22 BALL JOINTS 88-27 LAL CONTAINMENl'PRAY PUHP 122

I

)

CONTAINMENTSPRAY RADIOLOGICALASSESSMENT

l

F IIV54ENT PRAY%ATER E L RADI I I AI A E

MENT OBJECTIVE:

PERFORM A RADIOLOGICALASSESSMENT OF INTERRUPTION OF THE CONTAINMENTSPRAY WATERSEAL

~

ESTABLISH A REASONABLE LEAKRATE WITHINTHE LICENSING BASIS FOR FURTHER ENGINEERING EVALUATIONOF POTENTIALLEAKPATHS ACCEPTANCE CRITERIA:

~

COMPLIANCE WITH 10CFR100 AND 10CFR50, APPENDIX A, GDC-19 BASIS:

~

EOP ACTIONS WHICH INTERRUPT WATERSEAL DURING DESIGN BASIS EVENTS ARE VERIFIED TO NOT IMPACTTHE LICENSING BASIS IN ACCORDANCE WITH NRC SAFETY EVALUATIONFOR REV 4 OF BWROG-EPG'S

~

DESIGN BASIS SOURCE TERM (LOCA) IS CONSERVATIVELYCOMBINED WITH WORST-CASE ANTICIPATED INTERRUPTION OF WATER SEAL

~

LEAKAGEIS CONSERVATIVELYASSUMED TO BE AT 3TE MA3OMtMRATE FOR THE FULL30-DAY ANALYSISPERIOD

0

DOSE SUI'PNMARY:

AI P

Y ATREL I L I ALE AL ATI EAB:

LPZ:

THYROID GAMMA BETA THYROID GAMMA BETA DESIGN BASIS DOSE, REM 3.80 0.29 0.15 8.28 0.33 0.26

[1,2]

REGULATORY LIMIT,REM 25 25 LOSS OF WATERSEAL DOSE, REM 7.06 1.03 0.50 41.5 2.49 1.93 TOTAL DOSE, REM 10.86 1.32 0.65 49.78 2.82 2.19

% OF LIMIT 3.6 5.3 16.6 11.3 THYROID CONTROL ROOM:

GAMMA BETA 5.65

'.27 1.81 30 5.0 30 13.7 1.95 14.5 19.35 2.22 16.31 64.5 44.4 54.4

[1]

EAB & LPZ REGULATORY LIMITS PER 10CFR100

[2]

CONTROL ROOM REGULATORY LIMITS PER GDC19

l

F AINIVIENT PRAY IATER EAI.

RADI I I AI A E

MENT CONCLUSION:

~

A LEAKRATE EQUIVALENTTO AN ADDITIONAL11%

PER DAY OF PRIMARY CONTAINMENTLE&AGEHAS BEEN DETERMINED THAT SATISFIES THE FOLLOWING REQUIREMENTS:

~

MAINTAINSOFF-SITE DOSES WELLWITHIN

((25%) 10CFR100 LIMIT MAINTAINSCONTROL ROOM DOSES SIGNIFICANTLYBELOW (< 65%) GDC"19 LIMITS PROVIDES A REASONABLE LEAKRATE FOR ENGINEERING EVALUATIONOF POTENTIAL LEAKPATHS

~

~

CONTAINMENTLEAKAGE ANALYSIS

)

~

~

UTILIZEDESIGN BASIS RECONSTITUTION TO DEVELOP A CONTAINMENTMODEL

~

ANALYZELONG TERM POST DBA LOCA CONTAINMENTRESPONSE ASSUMING EOP REV 4 GUIDELINES FOR OPERATION OF CONTAINMENT SPRAY SYSTEM

~

EVALUATEPOTENTIAL LEAFAGE PATHWAYS FROM PRIMARY CONTAIICVIENT(PC) THROUGH THE CONTAINMENTSPRAY SYSTEM TO THE REACTOR BUILDING

~

DETERMPtE THE LEAIMGECONDITIONS WHICH ARE BOUNDED BY THE RADIOLOGICAL ENVELOPE

~

ANALYSISAPPLIES REALISTIC BUT CONSERVATIVE ASSUMPTIONS TO REVIEW THE CONSEQUENCES OF EOP ACTIONS CONSISTENT WITH THE NMPC EVALUATIONOF THE IMPLEMENTATIONOF REVISION 4 OF BWROG EPG RESULTS USED TO CONFIRIVf EOP ACTIONS DO NOT VIOLATELICENSING BASIS AND WILLNOT BE REFLECTED IN THE RADIOLOGICALEFFECTS OF THE DESIGN BASIS EVENTS

0 1

I

~

POTENTIAL FL W PATH

~

LEAIMGEPATHS

~

ISOLATION VALVES ASSUMED TO LEAK, PRESSURIZING SYSTEM PIPING

~

MAJOR SYSTEM VALVES ASSUMED TO HAVE POTENTIAL FOR PACKING LEAKAGE

+

SIMPLIFIED PACKING LEAKAGEMODEL DEVELOPED TO ESTIMATE LEVEL OF PACKING DEGRADATION

~

CONTAINMENTSPRAY SYSTEM COMPONENT AND PIPING INTEGRITYASSUMED

~

COMPONENTS (PUMP SEALS, AND VALVES) IN THE CONTAINMENTSPRAY SYSTEM PREDICTED BY THE MODEL TO HAVE ONLY LIQUID LEAKAGEARE ASSUMED WATER SEALED

lA

)

TORUS COOLING MOOE CONTAINMENT SPRAY SYSTEM

'IO 8-41 88 15 88-48 G.OO CCCOZOO WN d.OO K

OZOO UJ 88-35 88-45 88.16 V

PRIHARY LOOP PRIHARY LOP ORYWELL 88.44 88-11 88-12 88-118 88-31 88-86 BALL JOINTS BALL JOINTS CONTAINHENT SPRAY PUHP Ilt 88-81 88-21 LO.

QHTAINHENT 112 7

BALL JOINTS QTAINHENT 121 88-82 BALL JOIN'IS

.88-27 LO.

CONTAINHENT SPRAY PUHP 122

I

)

4 ~

~

UTILIZEGOTHIC 4.0 (QA VERSION) TO PREDICT LONG TERM POST LOCA PC RESPONSE

~

BENCHMAMMDTO THE ANALYSIS COMPLETED UNDER DBR

~

6 LONG TERM DESIGN BASIS LOCA CASE PROFILES ANALYZED

~

EOP ACTIONS MODELED

~

EVALUATEVALVEPACKING LEAIMGEFLOW AREA WHICH SATISFIES RADIOLOGICAL ENVELOPE

~

DETERMINE AVERAGE PACKING LEAK ASSUMING 3 LOOPS EXPOSED TO ATMOSPHERE - 13 VALVES WITHALL 13 VALVEPACKING DEGRADED

~

DETERMINE MAXIMUMSINGLE VALVE PACKING DEGRADATIONWHICH SATISFIES RADIOLOGICALENVELOPE

b

)

~

~

NCLU I N

~

THE MAXIMUMLEAKAGERADIOLOGICAL ENVELOPE BOUNDS SIGNIFICANT PACKING DEGRADATION

~

THE BOUNDED PACKING DEGRADATIONWOULD BE DETECTED DURING NORMALQUARTERLY SURVEILLANCE AVERAGE PACKING DEGRADATION CORRESPONDS TO -3.5 GPM LIQUID LEAFAGE DURING SURVEILLANCETESTING CONDITIONS

~

MAXIMUMSINGLE VALVEPACKING DEGRADATION CORRESPONDS TO -50 GPM DURING SURVEILLANCE,TESTING CONDITIONS

~

THE BOUNDED PACKING DEGRADATIONWOULD BE DETECTED DURING CONTAINMENTILRT

~

RESULTS CONFIRM THAT THE EOP ACTIONS WHICH INTERRUPT THE WATERSEAL DURING DESIGN BASIS EVENTS DO NOT IMPACT THE NMP1 LICENSING BASIS

i

PRA REVIEW

4

~

PR~

ENGINEEMNG EVALUATIONTHAT CALCULATESTHE PROBABILITYOF SEVERE ACCIDENTS BY ENUMERATING POSSIBLE ACCIDENT SCENARIOS.

SCENARIO PROBABILITYIS DEVELOPED BY MULTIPLYINGTHE PROBABILITYOF THE INDIVIDUALFAILURES THAT OCCUR IN THE SCENARIO.

SCENARIO CONSEQUENCE IS DEVELOPED USING THERMAL EGG)RAULIC ANALYSISTHAT SIMULATES PLASM ACCIDENT RESPONSE.

A

~

l

~

~

~

~

t I

PRA REVIEW OF IS UE

~

CHANGE TO INITIATINGEVENT FREQUENCY a

NO CHANGE NO EFFECT ON NORMAL PLANT OPERATION

~

CHANGE TO PLANT ACCIDENT RESPONSE a

NO CHANGE LEAIMGECRITERIA HAS NO EFFECT ON SYSTEM RESPONSE

~

CHANGE TO ACCIDENT CONSEQUENCES

~

YES - INCREASED LEAIMGE/SOMEWHAT LARGER RELEASE TO THE ENVIRONMENT

~

~

I' 4

~

I ~

~

~

4 N

~

NO CHANGE IN ACCIDENT FREQUENCY

~

CORE DAMAGEEVENT CONSEQUENCES ARE CHANGED FOR THOSE SEQUENCES WHERE CONTAINMENTOTHERWISE INTACT

~

FOLLOWING AN ACCIDENT, RADIONVCLIDE RELEASE IS LARGER FOR LESS LEAKTIGHT ISOLATION

0

'I l

~

PJ

INCREASED DOSE CONSEQSkNCES FROM RADIOLOGICALANALYSIS

~

DOSE WITHIN 10 MILES = 0.75 REM

~

DOSE WITHIN50 MILES = 0.11 REM POPULATION WITHIN10 MILES = 50,000 POPULATION WITHIN50 MILES = 1,000,000

~

~

VALUE-IMPA T

~

COST = 1.48ES PREM X $9000 PER PREM

~

FREQUENCY = CDF = 7.2E-7 PER YEAR

~

VALUE = COST X FREQUENCY COST X FREQUENCY = $960 PER YEAR 14 YR CASH FLOW NPV = $7,915 (8%)

~

I t

~

NCLU I

~

~

VALVEREPLACEMENT IS NOT COST JUSTIFIABLE

I l 4

I

Enclosure 3

NIAGARAMOIIAWKPOWER CORPORATION NINE MILEPOINT NUCLEAR STATION UNIT 1 DESIGN BASIS RECONSTITUTION February 22, 1994 V NIAGARA ILMOHAWK

I l

t

Nine Mile Point 41 - Design Basis Reconstitution Update with NRC - February 22, 1994

~ Introduction/Purpose ------- C.D. Terry

~ DBR Accomplishments to-date -- R.F. Qleck

~ Present Schedule/Scope ----- R.F. Oleck

~ DBR Integration to Eng'g --- L.A. Klosowski

~ Conclusions ------------C.D. Terry

\\ I

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't

DBR Accomplishments:

System Design Basis Documents

~

21 SDBD's in 4 years

~

IPE Systems-important to safety

~

Generic Letter issues:

~

Service Water

~

Instrument Air

~

Reactor Vessel Instrumentation

~

Plant LCO mitigation:

~

Lake water temp.-Containment Spray

~

Diesel Load Sequence/Transient

~

Neutron Monitoring Secondary Containment Leak Test

~

Remaining System Design Basis adequate

~

Reconstitute as-needed

~

Safety Assessment

-Senior Eng. Review Team

~ Configuration Controlled Documents

0 r

> I I

9 DBR PROGRAM L

Term~

No.

IPE PROTOTYPE 1

1 CoreS FIRST YR TYPE PRIMARYBASES 1991 1

2 1

2 3

4 1

2 Ah>;. "5 3

4 1

2 1993 3

4 1

2 3

4 4

1 2

3 1

2 3

2 3

4 1

1997 1

2 3

4 1

2 3

4 1

2 3

4 2

3 4

5 6

2 Sew Wtr 3 RBCLC 4 TBCLC 5 EDG 6 DC125V SECOND YR MECH GL 89-13+IPE HACH GLN-13+IPE hKCH GL N-13+IPE ELEC EDSFI+IPE ELEC IPE+EDSFI+SBO+

1 2

3 4

1 2

wit C! gg ~:4teccnatttgtlen eccnatituticn "N Y c.R 4

1 2

3 4

7 8

9 10 11 12 13 14 15 16 17 18 7

8 9

10 11 12 13 14 15 16 19 Cont Spray Primary Cont Emerg Cond ADS Instr Air AC4160V AC120V THIRD YR Contain(Systems)

SD Cool RPS Neutr Mon Feedwtr/HPCI FOURTH YR MECH hKCH HKCH Mri&C MECH ELEC ELEC MECH MECH t&C I&C MECH IPE+Lake temp IPE+Torus IPE+FuelseMOV IPE+FUELS GL 88-14 IPE+EDSFI GL9146 IPE RG1.

IPE IPE+Safety IPE+Safety Fuels+ ReliabiTity IPE+Reliabi

'991 1

2 3

4 1

3 1

2 4

1 2

1 2

3 4

1 2

3 4

1 2

3 4

24 25 RV Internal Proc Com SEVENTH YR 26 20 Uq. Poison 27 Remote Shtd 19 18 RPV Ins.

20 22 RB HVAC 21 17 CRD/ATWS FIFTH YR 22 23 Main Steam 23 21 Condens Tnsfr t&C IPE+RG 1.97 MeCH IPE+Safety urt&c IPE+Safe hKCH Fuels+Reliability MECH IPE+Reliabi'ECH Safety+PLEX t&c Plant MOD MECH IPE+Safety t&c Safe

+RG1.97 1991

~'-accelerate

~ accelerate I~acceierate 2

3 4

1 2

3 4

1997 1

2 3

4 1

2 3

4 1

2 3

4 30 31 DC24V Recirc RWCU CR HVAC OTHER SYSTEMS ELEC Safety+RG1.97 MECH Safe aKcH SafetyKNonSafety MECH Safe 1991

=

1993 32 33 34 35 36 37 38 OffGas Proc Rad Monit ARM SPDS Circ Wtr Spent Fuel Post Acc Sam MECH t&C I&C I&C hKCH hKCH urt&C Safety&NonSafety Safety Safety Safety Rehabiirty Safety Safe

I P

.s

~

DBR Accomplishments:

Design Criteria Documents - DCD's

~ 25 DCD's in 4 years=Topical DBD's

- Supported SDBD Issues/schedule

- Generic Issues:

- Reg.Guide 1.97

- Service Water

- EDSFI

- Computer Software Provided for:

- Electrical Transients-PTI Transient Code

- Cable/Raceway Database-ELECTRAK

- Thermal/Hydraulics-PCTI%(

- Piping Stress Analysis - PIPEMASTER

- Finite Element Analysis - COSNIOS/M

- GOTHIC-Containment Leak rate

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~ I F

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~

PROGRAM YEARS TYPE RESPON 1991 1992 1993 1994 1997 1 DCD-101 2 DCD-109 3 DCD-115 4 DCD-120 5 DCD-208 6 DCDM1 7 DCD-306 8 DCD-311 9 DCLM17 10 DC~

INITIALYEAR Pipe Supports Primary Containment Seismic Analysis External Events Piping Design It Analysis DC E ectrical Dis ribution Electric Cable Design AC Load Electrical Isolation VitalArea Access Struc StrIM Struc Slruc Mech Elec Ekc Elec Sec HP MPR MPR IMPELL IMPELL IMPELL IMPELL GILBERT IMPELL NMPC NMPC 1

2 3

4 1

2

% inc les lip 3

4 1

2 3

4 1

2 3

4

';N( F. <

Schedule slIP 1

2 3

4 1

2 3

4 2

3 4

1 2

3 4

1 2

3 4-11 DCD-102 12 DCD-110 13 DCD-114 14 DCD-204 15 DCD-205 16 DCO418 17 DCD-319 18 DCD401 19 DCD-117 20 DCD-213 21 DCD-305 22 DC~

SECOND YEAR Seismic CassNcation Reinforced Concrete Steel Structures Hydraubc Design 8 Analysis Heat Transfer Design IIIC Setpoint Design Fusing Design Accident Load'HIRD YEAR Fire P otection HVACGeneral Design Nuclear 8 Process Instr.

Source Term Struc Struc Struc Mech Mech Elec Elec Fuels Struc Mech Elec Nuc MPR IMPELL IMPELL IMPELL IMPELL GILBERT GILBERT SLI Pac Nuc BumsIIR GILBERT SYMMET 1

2 3

4 1

2 3

4 1

2 3

4 2

3 4

1 2

3 4

1 2

3 4

1 2

3 4

1 1997 2

3 4

1 2

3 4

FOURTH PHASE 23 DCD-119 Cable Tray II Duct Support Struc 24 DCD-202 Component Functknal Req'mts Mech 25 DCD407 Post Accident Sam 4D Elec StIL TBD OEI 26 DCD-118 27 DCD-203 28 DCD-304 FIFTH PHASE Component Supports Pipe Break Loads Com Control Struc Mech Elec NMPC NMPC NMPC 1

2 3

4 1

2 3

4 1

2 3

4 2

3 1

2 3

4 1

2 3

4 1

2 3

4 1

2 3

1 2

3 4

29 DCD-111 30 DCD-206 31 DCD-309 SIXTH PHASE Masonry Block Structures Installation Design Requirements Control Panels Display Design Struc Mech Elect NMPC NMPC NMPC 1991 1997

(g

~

o t

T

~ I I

~

i~4')

4 t

t DBR Accomplishments:

Configuration INanagement Upgrade

~ Component I evel Database Integration PRIME VAX PRIME Nlod Track'g Database Eqojp.5, 'Q'ist CDS=Control Doc's Database Database

@3'C MOSSE

~ I O

4 C

DBR Integration into Design Engineering

~

DBR Products ( eg: SDBD's, DCD's, Calc's}

~

Reviewed/Approved by cognizant Design Engineer prior to acceptance

~

DBR Products entered into Controlled Document System and Configuration Management System

~

Use of DBR Products integrated into appropriate Division and Engineering department procedures

~

eg: SDBD's, DCD's used to develop design inputs for plant changes

~

DBR Products used frequently for:

~

Mod's

~

Operability determinations

~

System functional inspections (EDSFI, SWSFI, etc.)

gi ~

A (~r

()I ~

S"

~

~

Ph DBR Integration into Design Engineering

~ Ownership of Configuration Management System transfer to Engineering

~ Future DBR - type product development on as-needed basis:

- Major Modifications

- Emerging Regulatory/Safety Issues

- Plant Economics & Reliability

- PLEX

1 Jj e QAll

( r.Z'C'"3-March 14, 1994 process is being incorporated into will be used on an as-needed basis should require reconstitution.

Enclosures:

1.

Attendance List 2.

NMP-1 Appendix J 3.

NMP-1 Design Basis Reconstitution cc w/enclosures:

See next page ig the NMPC design engineering process and in the future if any remaining system J

Ori inal si ned b

g Donald'S.

Brinkman, Senior Project Manager Project Directorate I-1 Division of Reactor Projects - I/II Office, of Nuclear Reactor Regulation Distribution: *w/handouts

• Docket File

• NRC & Local PDRs

• PDI-1 Reading WRussel 1 /FMiraglia, 12/G/18

LReyes, 12/G/18 SVarga

• CCowgill, RGN-I JCalvo RACapra

• DBrinkman CVogan OGC

EJordan, MNBB 3701

• RLobel, 8/H/7

• RPlasse, RGN-I

• JPulsipher, 8/H/7

• GImbro, 9/A/1 ACRS (10)

VMcCree, 17/G/21 LA:PDI-1 CVo an

'$94 PM: PDI-1 DBrinkman. smm 94 BC:S RBr t f9 94 D: PDI-1 RACa ra+

/g 94 OFFICIAL RECORD COPY FILENAME: NM150220.MTS

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