ML20149L922
| ML20149L922 | |
| Person / Time | |
|---|---|
| Issue date: | 02/18/1988 |
| From: | Vissing G Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| PROJECT-675A NUDOCS 8802250150 | |
| Download: ML20149L922 (71) | |
Text
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February 18, 1988 Project No. 675 t'EMORANDUM FOR:
The record FROM:
Guy S. Vissing, Project Manager Standardization and Non-Power Reactor Project Directorate Office of Nuclear Reactor Regulation
SUBJECT:
SUMMARY
OF MEETING WITH COMBUSTION ENGINEERING TO DISCUSS THE BASE LINE PROBABILISTIC RISK ASSESSMENT FOR THE SYSTEM 80 DESIGN, JANUARY 26, 1988 INTRODUCTION A meeting of the staff with representatives of Combustion Engineering (CE) was held at the offices of Combustion Engineering in Bethesda, Maryland, on January 26, 1988. The purpose of the meeting was to discuss the Base Line PRA for System 80 design. Enclosure 1 provides a list of those in attendance. provides the viewgraphs which CE used in their discussions.
DISCUSSION The meeting was primarily to obtain the NRC staff response to the CE PRA effort. CE will be submitting their Final PRA for the System 80+ design in June 1989. Comnents on the Base Line PRA were requested in the Spring of 1988. The Base Line PRA will be used as a design tool to assess alternatives in designs, i.e., alternatives within systems.
The Base Line PRA will consider internal events, fire, seismic, and flood.
Sabotage will be treated qualitatively since PRA for sabotage is not advanced yet.
Concern was expressed for considering the RHR system in shutdown mode.
CE indicated that the PRA would address this.
The Base Line PRA will not use fault trees in the analyses but would use event trees. The PRA will use typical models for the Balance of Plant. The requirements of the BCP will be defined in the functional descriptions. One of the computt r codes that CE intended to use is the IRRAS Code, a PC base code developed for NRC by INEL. The staff requested feedback on their experien c.? with this code and the results on floppy disks.
The staff indicated that CE should look at internal events in more detail. Also, the staff expressed a concern for including in the PRA long term cooling following a transient from 100% power original signed by Guy 5. Vissing, Project fianager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects
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February 18, 1988 Project No. 675 MEMORANDUM FOR:
The record FROM:
Guy S. Vissing, Project Manager Standardization and Non-Power Reactor Project Directorate Office of Nuclear Reactor Regulation
SUBJECT:
SUMMARY
OF MEETING WITH COMBUSTION ENGINEERING TO DISCUSS THE BASE LINE PROBABILISTIC RISK ASSESSMENT FOR THE SYSTEM 80 DESIGN, JANUARY 26, 1988 INTRODUCTION i
A meeting of the staff with representatives of Combustion Engineering (CE) was held at the offices of Combustion Engineering in Bethesda, Maryland, on January 26, 1988. The purpose of the meeting was to discuss the Base Line PRA 1
for System 80 design. provides a list of those in attendance. provides the viewgraphs which CE used in their discussions.
DISCUSSION The meeting was primarily to obtain the NRC staff response to the CE PRA effort. CE will be submitting their Final PRA for the System 80+ design in 1
June 1989. Comments on the Base Line PRA were requested in the Spring of 1988. The Base Line PRA will be used as a design tool to assess alternatives in designs, i.e., alternatives within systems. The Base Line PRA will l
consider internal events, fire, seismic, and flood. Sabotage will be treated I
qualitatively since PRA for sabotage is not advanced yet. Concern was expressed for considering the RHR system in shutdown mode.
CE indicated that the PRA would address this. The Base Line PRA will not use fault trees in the analyses but would use event trees. The PRA will use typical models for the Balance of Piant.
The requirements of the BOP will be defined in the functional descriptions. One of the computer codes that CE intended to use is the IRRAS Code, a PC base code developed for NRC by INEL. The staff requested i
feedback on their experience with this code and the results on floppy disks.
The staff indicated that CE should look at internal events in more detail.
Also, the staff expressed a concern for including in the PRA long term cooling following a transient from 100% power Guy S. Vissing, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects
Enclosures:
As stated cc:
See next page I
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t ENCLOSURE 1 ATTENDANCE LIST FOR MEETING WITH CE CONCERNING-SYSTEM 80+ Base Line PRA January 26, 1988 i
NAME ORGANIZATION Guy S. Vissing NRC/NRR/PDSNP i
Tom Kenyon NRC/NRR/PDSNP t
Harold Vandermolen NRS/RES/PRAB Bryan Dolan Duke Power / Design /PRA George A. Davis Combustion Engineering i
Stanley Ritterbusch Combustion Engineering David J. Finnicum Combustion Engineering Brad Hardin NRC/RES/ARGIB Glenn Kelly NRC/NRR/PRAB t
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(WITHOUT (WITH OF j
INITIATING EVENT RECOVERY)
__REE0VERY)
__ TOTAL
,1 t
Sas11 LOCA i
j Vessel Rupture l
Interfacing System LOCA
}
5 team Generator Tube Rupture j
Large Secondary Side Break Transients Loss of Offsite Power Station Blackout ATW5 Baron Ollution TOTAL i
i
(.ORI DAMAGI IRIQUtNCY CONIRIBUTIONS IUR DOMINANT ACCIDfMI Sil)UINCts l'IRCINI Of CORf DAMAGI IUIAI (DRt iRfQUtNCY DAMAGI 5tQUENCE
[ONIRIBiliION I klpleCY i
i (tnis of Of f site Power) (Reac tor Trip) (f ailure to Deliver o
Ausillary feedwater) (failure to Deliver Alternate Icedwater) i (Small 100A) (Reactor Trip) (Iailure of HPSI Inject ion) o j
(failure to Depressurire for LPSI Injection) l
)
l l
(Transient) (Reactor Irip) (Auxiliary f eedwater and Steam o
Removal Successful in short Term) (f ailure to Inter Shutdown Cooling) (Failure to Maintain tong Teen $ccondary Heat Removal)
(Steam Generator lute Rupture) (Reac tor Irip) (Failure of flPSI o
)
Injection) (Failun to Depressurite for LPSI Injection)
I l
(Transient) (Reactor Trip) (failure to Deliver feedwater) e i
l l
(Fallure to Establish Alternate feed Flow)
(Anticipated Transient) (Methanical f ailure of Rods to insert) e 1
(MIC Overpressure) l 6
I l
1 l
I
CORI DAMACI TRIQUENCY CONTRIBUTIONS FOR DONINANT ACflDINT SEQUEKf 5 l'I RCINT Of CORI IIAMAM TOIAL CORI IREQUENCY DAMAM SIQtMNCE CONTRIBUTION TREQUINCY o
(Medium LOCA) (Sif and HPSI Injection Successful) (IIPSI Recirculation Successful) (Hot and Cold leg injection Successful)
(Fallure of Recirculation Cooling) o
($mell LOCA) (Reactor Trip) (IFSI Injection successful)
(Ammillary feedwater and 5 team Removal Successful) (failure of HPSI Recirculation) (Failure to Establish LPSI Recirculation)
(Station Blackout) (Reactor Irip) (Aunillary feedwater -
o Turbine and Steam Removal Successful) (failure to Restore Power litthin 3 Hours)
O
i l
i I
t f0RI DAMAGI TRfQttNCY CONTRIBUTIONS ICR DOMINAMI ACCIDfMI StINNMCIS l
]
PI R(INT DI CORI DAMAGE IDIAI (DRI fRIQtliMCT DAMAGE 5tQigNCE CONTRIBUFION IRIQtKNCY g
(5 tease Generator Tube Rupture) (Reactor Trip) (11 PSI Successful) o
)
(Failure to Deliver Ausillary Feedwater) (failure to Deliver Alternate Feedwater) i i
o (Transient) (Reactor Trip) (Ausillary Icedwater Successful in I
i short Term) (Failure of Turbine Bypass and Atmospheric Dump l
Valves) (Failure to Maintain long Term Secondary IIcat Removal) f i
TOTAL L
e n
i l
I l
4 1
l i
i i
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T R
V EEOW I E RGD EN f
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i l
A SME I
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T 5
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Y A
RR e
S EE 3
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f EE ART G
PS N
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IMPACT OF SYSTEM 80+ DESIGN FREATURES ON CORE DAMAGE FREQUENCY CONTRIBUTORS DESIGN ENilANCEMENT 4 TRAIN 4 TRAIN LARGER IN CONTAINMENT SAFETY I
INITIATING EVENT ECCS EFW PRESSURIZER RHT DEPRESSURIZATION l
+
+
+
+
Steam Generator Tube Rupture
+
+
+
Small LOCA
+
+
+
AlWS
+
+
+
Mediu1 LOCA
+
+
l iloron Dilution l
1 Large LOCA
+
+
j Station Blackout
+
Large Secondary Side Break i
j N
- 1 l
7 ADDITIONAL CONSIDERATICNS FOR SYSTEM 80+ CORE DAMAGE FREQUENCY REDUCTION ADDITIONAL CONSIDERATION INIIIAIING IVINI LST PLMP ILICIRICAL IlillM0-PROCIDURI S ROOM POW (R SIAIION In0RAULIC COOLING DI51RI80fl0N 8AlI[RII5 ANALYSIS Seill LOCA e
e e
less of Offsite Power e
e ATW5 e
Medium LOCA e
e Transients e
e e
l Steam Generator lobe Rupts.re e
e e
e Boron Dilution e
l 1arge LOCA 4
e j
stati.n stack t e
targe Secondary Side Breaks l
s
+
Vessel Rapture f
l Interfacing Systes 10CA
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l
- l SYSTEM 80+ PRA PROCESS i
e GENERIC SYSTEM 80 Pl:A USEL AS STARTING POINT MODIFY FAULT TREES TO REFLECT SYSTEM DESIGN CHANGES MODIFY EVENT TREES TO REFLECT CHANGES
=
o system design o
plant response EVENT SEQUENCES EVALUATED USING LINKED FAULT TREES J
J i
e TREATMENT OF B0P/SUPPDRi SYSTE!'.3 l
STANDARD FUNCTIONAL DESCRIPTIONS AND DIAGDAMS DEVELOPE 0 PRA MODELING A5SUMPTi0riS INCLUDED AS PART OF STANDARD FUNCTIONAi_ DESCRIPTIONS i
SENSITIVITY ANAi.YSES ON POTENTI.4L CESIGN OPTIONS esammesTeess)N1meses 1
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OBJECTIVE:
SM0OTH THE WAY FOR REVIEW 0F SYSTEM 80+ PRA ADDRESS NRC CONCERNS UP FRONT o
WHAT:
INFORMAL REVIEW OF FF
. NE PRA TO IDENTIFY CONCERNS e
Process e
Modelling Approach, Assumptions, and Data e
Sufficient Coverage of Issues FEEDBACK ON NRC REVIEW (SPRING '88)
FOLLOW-ON MEETINGS ON MAJOR AREAS POTENTIALLY e
Support System Treatment (Late Spring / Summer '08) e External Events (Summer '88) e Containment Analyses (late Fall '88) e Level III Analyses (Winter '88/89) e Other (As Needed) stemmep5TeesshEIOGOBIEER08iiG
d CLOSING COM1ENTS S. E. RITTERBUSCH e
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C-E IS DEDICATED TO ACHIEVING DESIGN CERTIFICATION BY 1991 0
C-E WILL MEET THE SEVERE ACCIDENT POLICY REQUIREMENT FOR A PROBABILISTIC RISK ASSESSMENT EXTEND THE SYSTEM 80 BASELINE (LEVEL 1) PRA TO A LEVEL 3 PRA FOR THE SYSTEM 80+ DESIGN SUBMIT PRA APPENDIX TO CESSAR-DC MEET WITH NRC AND RESPOND TO MlC QUESTIONS AND CO M NTS
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l IE GENERIC SYSTEM 80 PRA TREATNENT OF BALANCE OF PLANI l
e BALANCE OF PLANT (BOP) SYSTEMS ADDRESSED IN PRA USING MODULA!?
SUPPORT SYSTEM MODELS CONSTRUCT SYSTEM DEPENDENCY MATRIX FOR FRONT LINE AND BOP SYSTEMS DEFINE TYPICAL BOP SYSTEM DESIGNS OR DESIGNS MEETING FUNCTIONAL SUPPORT REQUIREMENTS DEVELOP FAULT TREE MODELS FOR SUPPORT SYSTEMS FRONT LINE SYSTEM FAULT TREES INCLUDE SMALL SYSTEM / TRAIN LEVEL DEPENDENCY MODELS BASED ON SUPPORT SYSTEM MODELS QUANTIFY USING GENERIC INDUSTRY DATA BOP SYSTEM ASSUMPTIONS WILL BE INCLUDED IN SYSTEM 80+ STANDARDIZED e
FUNCTIONAL REQUIREMENTS ceasessisesshmannanassananas F
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MODULAR SUPPORT SYSTEM MODELS e
' CONSTRUCT FAULT TREE MODEL(s) FOR SUPPORT SYSTEM TRAIN (s) AS FOR '
FRONT LINE SYSTEM IDENTIFY ALL COMMON CAUSE ELEMENTS, SHARED ELEMENT 3 AND INTERFACES e
WITH OTHER SUPPORT SYSTEMS e
RESTRUCTURE MODEL S0 IT IS DIVISIBLE INTO A "COMMON ELEMENTS" PORTION AND A "RANDOM HARDWARE FAILURE" PORTION e
REPLACE "RANDOM HARDWARE FAILURES" PORTION WITH UNDEVELOPED EVENT INCORPORATE REDUCED MODEL AS MODULE IN APPROPRIATE FRONT LINE SYSTEM l
e l
HODELS l
e QUANTIFY "RANDOM HARDWARE FAILURE" PORTION TO OBTAIN FAILURE PROBA8ILITY FOR THE UNDEVELOPED EVENT FOR FUTURE QUANTIFICATION OF FRDNT L NE SYSTEM MODELS coassessYsees EgesaseEEmeses b
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L FIGURE 2.3-2 SAMPLE SAFETY SYSTEM FAULT TREE WITH SIMPLIFIED CONMON CAUSE/COMON ELEMENT SUPPORT SYSTEM MODEL HPSI SYSTEM FAILS.
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HPSIPUMP A F AILS TO RUN 1
l
' MECH ANIC AL COMMON NO CCW TO FAILURE OF CAUSE F AILURE gg, PUMP PUMP OF PUMP o
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' FAILURE OF COMMON FAILURE OF SERYlCE GGG CAUSE CCW-A WATER FAILURE COMPONENTS f-0F CCW 7
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PSA PROCEDURES GUIDE NUREG/CR-2815 EPRI NP-2230 e
NREP GENERIC DATA BASE e
WASH-1400 e
IEEE STD 500/1984 e
NUREG REPORTS ON COMMON CAUSE FAILURES:
/CR-2771
/CR-2098 e
CESSAR-F e
PALO VERDE / SONGS /WATERFORD/FSARs e
C-E STANDARD TECHNICAL SPECIFICATIONS l
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$b ACCIDENT SE0VENCE OUANTIFICATION QUANTIFY SYSTEM FAULT TREES TO GET SYSTEM UNAVAILABILITIES o
CONSTRUCT EVENT SEQUENCE MODELS/ EQUATIONS e
QUANTIFY EVENT MODELS/ EQUATIONS USING SYSTEM MODELS/UNAVAILABILITIES e
e FOR DOMINANT SEQUENCES EXAMINE CUTSETS TO IDENTIFY POTENTIAL REC 0VERY ACTIONS QUANTIFY REC 0VERY ACTION PROBABILITIES REQUANTIFY EVENT SEQUENCES WITH RECOVERY INCLUDED CNN MEN i
I i
I i
MMmMM1
CONDITIDRED SYSTEM FAILURE PROBABILITIES 3
FOR SEQUENCE EQUATION SOLUTION (FOR BASELINE PRA ONLY) e GIVEN CORE DAMAGE SEQUENCE EQUATION CD = IE
- SYSA
- SYSB WHERE SYSA AND SYSB HAVE SHARED ELEMENTS E, FOR EACH SHARED ELEMENT E, CALCULATE P(E /sYSA) e g
g e
QUANTIFY SYSB USING P(E /SYSA) IN PLACE OF P(E )
g g
e P(E /SYSA) = rCS,
+ P(E ) ((P(SYSA) - ICS )/P(SYSA))
g g
g P(SYSA)
= PROBABILITY ASSOCIATED WITH CUTSETS CONTAINING l
I ELEMENT EI b seasmusTseas)saamaanssmaana F
n M M mM l
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CEREC BASICALLY PREP AND KITT WITH SONE ENHANCENENTS 8
MODEL DISTRIBUTIONS e
INCLUDE REPAIR TIME AND OUT-OF-SERVICE TIME USED FOR FAULT TREE QUANTIFICATION FOR BASELINE PRA e
CEDAR UTILITY CODE FOR SYSTEM FAILURE PROBABILITY CONDITIONING I
USED IN BASELINE PRA e
CESAN SAMPLE CODE WITH ADDITIONAL DISTRIBUTIONS USED FOR SEQUENCE EQUATION SOLUTION IN BASELINE PRA l
e SETS /KITT I
USE FOR LINKED FAULT TREE QUANTIFICATION IN FINAL PRA KITT USED FOR CALCULATION e
IRRAS PC BASED CODE DEVELOPED FOR NRC BY INEL
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l CORE DAMAGE FREQUENCY CONTRIBUTIONS 8Y INITIATING EVENT CORE DAMAGE CORE DAMAGE FREQUENCY FREQUENCY PERCENT (WITHOUT (WITH OF INITIATING EVENT RECOVERY)
_ RECOVERY)
TOTAL Large LOCA Medium LOCA Small LOCA Vessel Rupture Interfacing System LOCA Steam Generator Tube Rupture Large Secondary Side Break Transients Loss of Offsite Power Station Blackout ATWS Boron Dilution TOTAL e
b emanmesnen)amenssammes 1
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CORI DAMAGI IREQUENCY CONTRIBUTIONS FOR DOMINANT ACCIDINI SEQUENCES
~
l'1RCINT Of CORf DAMAGE 10lAt CORE IRIQUINCY DAMAGI SEQUENCE CONIRIBUTION
__IRIQU[NCY (Inss of Of f site Power) (Reactor Trip) (Failure to Deliver o
Auxiliary Icedwater) (Failure to Deliver Alternate feedwater)
(Small LOCA) (Reactnr Trip) (Iailure of HPSI Injection) o (failure to Depressurize f nr LPSI Injection)
(Transient) (Reactnr Trip) (Auxiliary Icedwater and Steam o
Removal Successful in Short Term) (f ailure to Enter Shutdown Cooling) (Iailure to Maintain Long Teir Secondary lleat Removal)
(Steam Ocnerator lotse Rupture) (Reactor Trip) (Failure of IIPSI o
Injectiure) (failure to Depressurize inr LPSI Injection) o (Iransient) (Reactor Irip) (f ailure to Deliver feedwater)
(Failure to Estatslish Alternate f eed Flow)
(Anticipated Iransient) (Mechanical f ailure of Rods to insert) o (MIC Overpressure)
N N
v CORI DAMACf TREQUENCY CONTRIBUTIONS FOR DOMINANT ACCIDfMT SEQUENCES PIRCINT Of CORI DAMAGE TOTAL CORE FREQUINCY DAMAGE SEQUfkCE CONTRIBUTION FREQUENCY (Medium LOCA) (SIT and HPSI Injection Successful) (HPSI o
Recirculation Successful) (Hot and Cold Leg Injection Successful)
(failure of Recirculation Cooling)
(Small LOCA) (Reactor Trip) (HPSI Injection Successful) o (Auxiliary Feedwater and Steam Removal Successful) (Failure of HPSI Recirculation) (Failure to Establish LPSI Recirculation)
(Station Blackout) (Reactor Trip) (Auxiliary Feedwater -
o Turbine and Steam Removal Successful) (failure to Restore Power Within 3 llours) e m N mB m
1 7
F E 0 R Y
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SYSTEM 80+
RISK REDUCTION FACTORS
~
IMPROVED ECCS IN CONIAINMENT RHI
/ NO SWIICll OVER I
!! TRAIN AFW la TRAIN AFW LARGER PRESSURIZER LARGER STEAM GENERATOR 11 TRAIN AFW LARGER PRESSl!RIZER
' LARGER STEAM GENERATOR FEED AND BLEED e
IMPACT OF SYSTEM 80+ DESIGN FREATURES ON CORE DAMAGE FREQUENCY CONTRIBUTORS DESIGN ENilANCEMENT 4 TRAIN 4 TRAIN LARGER IN CONTAINMENT SAFETY INITIATING EVENT ECCS EFW PRESSURIZER RWT DEPRESSURIZATION Loss of Offsit,. Power
+
+
+
+
Steam Generator Tube Rupture
+
+
+
Small LOCA
+
+
+
+
+
+
Medium LOCA
+
+
+
+
Station Blackout
+
Large Secondary Side Break d
.s
ADDITIONAL CONSIDERATIONS FOR SYSTEM 80+ CORE DAMAGE FREQUENCY REDUCTION j
l ADDITIONAL CONSIDERATION INITIATING LVINI
[SF PUMP It[CIRICAL lilIRMO-PROCIDUiti S ROOM POWER stall 0N llVDRAULIC l
COOLING DISIRIBUIION BATIERilS ANALYSLS Small LOCA Loss of Offsite Power
+
AIWS Medium LOCA Iransients Steam Generator Tube Rupture
+
+
Boron Dilution large LOCA Station Blackout large Secondary Side Breaks e
Vessel Rupture interfacing System LOCA
'""**"*'r e asy
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SYSTEM 80+ PRA
~
PROCESS e
GENERIC SYSTEM 80 PRA USED AS STARTING POINT H0DIFY FAULT TREES TO REFLECT SYSTEM DESIGN CilANGES H0DIFY EVENT TREES TO REFLECT CHANGES o
system design o
plant response EVENT SEQUENCES EVALUATED USING LIKt:ED FAULT TREES e
TREATMENT OF B0P/ SUPPORT SYSTEMS STANDARD FUNCTIONAL DESCRIPTIONS AND DIAGRAMS DEVELOPED PRA MODELING ASSUMPTIONS INCLUDED AS PART OF STANDARD FUNCTIONAL DESCRIPTIONS l
1 SENSITIVITY ANALYSES ON POTENTIAL DESIGN OPTIONS b soammusiseeshE-
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PROPOSED NRC ACTIONS
~
e OBJECTIVE:
SMOOTH THE WAY FOR REVIEW OF SYSTEM 80+ PRA ADDRESS NRC CONCERNS UP FRONT o
WHAT:
INFORMAL REVIEW OF BASELINE PRA TO IDENTIFY CONCERNS e
Process e
Modelling Approach, Assumptions, and Data e
Sufficient Coverage of Issues FEEDBACK ON NRC REVIEW (SPRING '88)
FOLLOW-ON MEETINGS ON MAJ0P. AREAS POTENTIALLY Support System Treatment (Late Spring / Summer '88) e e
External Events (Summer '88) e Containment Analyses (Late Fall '88) e Level III Analyses (Winter '88/89) l e
Other (As Needed) l seassessTHMehEIOG880EERS30G
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CLOSING COMMENTS S. E. RITTERBUSCH
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C-E IS DEDICATED TO ACHIEVING DESIGN CERTIFICATION BY 1991 0
C-E WILL MEET THE SEVERE ACCIDENT POLICY REQUIREMENT FOR A PROBABILISTIC RISK ASSESSMENT EXTEND THE SYSTEM 80 BASELINE (LEVEL 1) PRA TO A LEVEL 3 PRA FOR THE SYSTEM 80+ DESIGN SUBMIT PRA APPENDIX TO CESSAR-DC MEET WITH NRC AND RESPOND TO NRC QUESTIONS AND C0f9 TENTS
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