ML20138H950

Summary of Meeting Held at NRC in Rockville on 961216 W/ Westinghouse & WOG to Discuss WCAP-11992, Joint Westinghouse Owners Group/Westinghouse Program:Atws Rule Administration Process. List of Attendees & Handout Encl

ML20138H950
Person / Time
Issue date:	12/24/1996
From:	Craig C NRC (Affiliation Not Assigned)
To:	Matthews D NRC (Affiliation Not Assigned)
References
WCAP-11992, NUDOCS 9701070110
Download: ML20138H950 (29)

Text

._ _ - - __ . _ . .-

[puagk UNITED STATES

< E NUCLEAR REGULATORY COMMISSION E

If WASHINGTON, D.C. 20566 4001 d

4 . . , * ,o Desnber 24,1996 l MEMORANDUM T0: David B. Matthews, Chief Generic Issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation

. FROM:

Claudia M. Craig, Senior Project Manager g'g Generic Issues and Environmental f

, Projects Branch di(

Division of Reactor Program Management (

Office of Nuclear Reactor Regulation

SUMMARY

OF MEETING WITH WESTINGHOUSE AND THE WESTINGHOUSE OWNERS GROUP (WOG) TO DISCUSS WCAP-11992, ATWS P.ULE ADMINISTRATION PROCESS The subject meeting was held at the Nuclear Regulatory Commission (NRC) offices in Rockville, Maryland on December 16, 1996, between representatives of Westinghouse, the WOG, and the NRC staff. The purpose of the meeting was for Westinghouse and the WOG to discuss WCAP-11992, " Joint Westinghouse Owners Group / Westinghouse Program: ATWS Rule Administration Process." Attachment 1 is the list of meeting participants, Attachment 2 is a copy of the non-proprietary presentation material provided at the meeting.

Representatives from the WOG outlined the background and history of the Anticipated Transient Without Scram (ATWS) rule and why WCAP-Il992 was developed. The ATWS rule required Westinghouse plants to install ATWS Mitigating System Actuating Circuitry (AMSAC) in order to meet the ATWS rule risk goals, as discusse,d in SECY-83-293, which included an ATWS core damage frequency risk of lx10' per year and an ASME service level C stress limit of 3200 psig for Westinghouse plants. Prior to the ATWS rule, a generic ATWS analysis basis was performed for Westinghouse plants. The analysis included sensitivity analyses and assumed a base case of -8 pcm/F moderator i temperature coefficient (MTC) with all pressurizer PORVs available and full

' auxiliary feedwater capability. This base case met the 3200 psig pressure limit. WCAP-11992 was developed in the 1987/1988 timeframe in response to NRC concerns with changes in plant's licensed limits on MTC. The methodology was submitted for information purposes in 1989, then submitted for review and approval in 1995 in support of a plant specific licensing action. The methodology provides an assessment of changes in plant conditions on ATWS risk and demonstration of continued compliance with the basis for the ATWS rule.

' Westinghouse believes the methodology maintains the conservatism of the ATWS i

rule basis of 3200 psig and demonstrates continued compliance with the ATWS rule for plants with oositive MTC at part-power conditions.

Representatives from Westinghouse outlined the deterministic analysis considered in the development of the WCAP. The deterministic analysis was ormed in 197 and assumed a reactivity feedback model using a -8 pcm/F pp MIC BE CEEE6 CON mm w ' )s MN8W nant C PDR %4 me

~ . _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ . _ . _ . . ._

l i

2

! I I

MTC which corresponded to a value for which 95% of the cycle the MTC value i would be less limiting. An MTC value was also provided corresponding to 99%

of the fuel cycle as part of the sensitivity analysis. The analysis confirmed I that the ASME service level C stress limits were not exceeded. These analyses '

results formed the bases for the deterministic analyses in support of WCAP- l 11992. Critical power trajectories curves reflect the heatup/ shutdown  !

characteristics which yield the peak pressure of 3200 psig and are used to l determine the Unfavorable Exposure Times (UET) values. The UET is the time during the cycle when the reactivity feedback is not sufficient to prevent the reactor coolant system pressure from exceeding 3200 psig.

The probabilistic analysis was also discussed. The probability of adequate relief capacity is based on a number of factors. These two analyses are combined to show the effect of failure of systems on ATWS risk and the effect of equipment unavailability on ATWS risk. Westinghouse believes this integrated approach provides a better basis for measuring acceptability of parameters affecting ATWS risk than that of the base case from the deterministic model only.

After the formal presentation, a number of other topics were discussed. The staff stated that guidance, in the form of a regulatory guide (RG), is currently being developed on the use of PRA. The guidance will outline when risk can be increased, how to quantify risk, the level of detail needed for a PRA, and other areas.

The stair ~ informed the participants that it will not approve the methodology described in the WCAP because the methodology could allow licensees to use the risk goals established when the ATWS rule was developed for plant specific decisions and the PRA methodology may not be sufficiently robust to meet the PRA criteria as currently outlined in the draft RG. The staff stated that the number identified in the ATWS rule should not be used as a " speed limit" for dealing with risk, rather it should be a goal. Plants should not manipulate operation of their plant to reach the limit, but rather should strive to operate below that limit. The staff noted the methodology submitted appears to be an acceptable model to depict an accident progression of anticipated

. transient events that could lead to an ATWS condition. The methodology may be used for informational purposes for managing risk within the existing bounds of technical specifications at plants. The methodology may be used as a tool or " guide," not used as the sole basis for regulatory decisions. The staff stated that plants which may want to use this methodology as a basis for future licensing activities involving increases to their MTC, should begin early dialogue with the NRC, as such discussions would be beneficial to everyone.

The staff also stated that a formal letter to the WOG would be written which would summarize the results of the meeting and provide the results of the  ;

staff's review. i Attachments: As stated cc w/atts: See next page

, -l

5 b *,

2 4-

!* MTC which corresponded to,a value for which 95% of the cycle the MTC value l would be less limiting. .An MTC value was also provided corresponding to 99% l of the fuel cycle as part.of the sensitivity analysis. The analysis confirmed l

that the ASME service'1evel C stress limits were.not exceeded. These analyses .

results formed the bases for,the deterministic analyses in support of WCAP- l 11992. Critical power: trajectories" curves reflect the heatup/ shutdown l 1 characteristics which yield the' peak pressure of 3200 psig and are used to

~

I . determine the Unfavorable ~ Exposure Times'(UET) values. The UET is'the time.

during the. cycle when the reactivity feedback is not sufficient to prevent the .)

j reactor coolant. system pressure from' exceeding 3200 psig.  ;

4

. . . - n .

i The probabilistic analysis was also discussed _ The probability of adequate relief capacity is based on a number of factors. These two analyses are  ;

combined to show the effect.of failure of systems on ATWS risk and the effect..

i of equipment unavailability on ATWS risk; ' Westinghouse believes this )

L integrated approach provides a better basis for measuring acceptability of 1

parameters affecting.ATWS riskethan that of the base case from the '

deterministic model only. -- t.

i L -

After the formal ^ presentation, a number. of 6ther topics were discussed. The ,

i staff stated that guidance, in the form' of a regulatory guide (RG), is .

i

! currently being developed,on the use of PRA. The guidance will outline when  ;

-risk can be increased, how to quantify risk, the level of detail needed for a.

PRA, and other areas. ,

i . .  !

The staff informed the-participants that it will not approve the methodology i described in the WCAP because the methodology could allow licensees to use the l

risk goals established when the ATWS rule was developed for plant specific j decisions and the PRA methodology may not be sufficiently robust to meet the '

1 PRA criteria as currently outlined in the draft RG. The staff stated that-the i number identified in the ATWS rule should not be used as a " speed limit" for dealing with risk, rather it should be a goal. Plants should not manipulate operation of their plant to reach the limit, but rather should strive to operate below that limit. The staff noted the methodology submitted appears

.to be an acceptable model to depict an accident progression of anticipated transient events.that could lead to an ATWS condition. The methodology may be used for. informational. purposes for managing risk within the existing bounds of technical specifications at plants. The methodology may be used as a tool or. " guide," not used as the sole basis for regulatory decisions. The staff stated that plants which may want to use this methodology as a basis for future licensing activities involving increases to their MTC, should begin early dialogue with the NRC, as such discussions would be beneficial to everyone.

' The staff also stated that a formal letter to the WOG would be written which would summarize the results.of the meeting and provide the results of the staff's review.

Attachments: As stated cc w/atts: See next page DISTRIBUTION: ,

See attached page DOCUMENT NAME: G:\ CMC 1\l2 16. MIN  !

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DATE 12/ N 96 12/#/% 'R/ v3&c, C D 12/Jf/96

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DISTRIBUTION w/ attachments: Summary of December- 16, 1996, with Westinghouse dated December 24. 1996  ;

Central File PUBLIC i PGEB i l RArchitzel l

CCraig E-Mail FMiraglia/AThadani BSheron

RZimmerman l '

TMartin l GHolahan MVirgilio TCollins EWeiss l Slee JFlack AAttard I i

RJones h

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l WESTINGHOUSE /WOG/NRC MEETING ATWS RULE ADMINISTRATION PROCESS 1 DECEMBER 16, 1996 AT ROCKVILLE, MD l

i MEETING PARTICIPANTS 4

l H8ME ORGANIZATION ,

l l Claudia Craig NRC/NRR/PGEB i 1 Samuel Lee NRC/NRR/SPSB i Eric Weiss NRC/NRR/SRXB Baard Johansen Westinghouse /CNFD-Core Engineering Roger Newton WEPCo/WOG Vance VanderBurg AEP/WOG

! Gary Ament Westinghouse Transient Analysis j Barry Sloane Westinghouse Reliability and Risk

Analysis I Ken Vavrek Westinghouse /WOG l Brian Sheron NRC/NRR/ADT Marty Virgilio NRC/NRR/DSSA John J. Flack NRC/NRR/SPSB Timothy E. Collins NRC/NRR/SRXB Robert C. Jones NRC/NRR/SPSB Anthony Attard NRC/NRR/SRXB i

ATTACHMENT 1

t NRC/WOG MEETING REVIEW OF WCAP-11992:

ATWS RULE ADMINISTRATION PROCESS DECEMBER 16,1996 MEETING AGENDA 4

2:00 INTRODUCTIONS ALL 4

2:10 BACKGROUND / HISTORICAL LOU LIBERATO.RI

, REVIEW (CON ED) i i

4 2:40 WCAP-11992 DETERMINISTIC BASIS GARY AMENT (W) I l

~

3:00 WCAP-11992 PROBABILISTIC BARRY SLOANE (W)

APPROACil 3:45

SUMMARY

LOU LIBERATORI i 4

4:00 OPEN DISCUSSION ALL 1

1 s

l Attachment 2

4 3..

i = ATWS RULE - 10CFR50.62 FOR WESTINGIIOUSE PWRS - INSTALL AMSAC l

= OBJECTIVE OF ATWS RULE j

i j

• REDUCE PREDICTED ATWS RISK l

= ATWS RULE RISK GOAL - SECY-83-293 REDUCE ATWS RISK TO IX10-8 PER REACTOR l YEAR OR LESS

! PUBLIC RISK CONSERVATIVELY EQUATED WITII 1

CORE DAAIAGE FREQUENCY  !

l

• l l CORE DAhIAGE CONSERVATIVELY EQUATED MTITI j EXCEEDING ASME SERVICE LEVEL C STRESS

! LIMIT l

!

• ASME SERVICE LEVEL C STRESS l LIMIT = 3200 PSIG FOR W PWRS '

!

• AMSAC REDUCED ATWS RISK FOR W PWRS l FROM 3.7X10 " TO 5.8X10 4 i

5 i

s

= AMSAC'S FUNCTION SENSE LIMITING RCS PRESSURE TRANSIENT ATWS EVENT (LOSS OF FEEDWATER)

• TRIP TURBINE TO LESSEN AIAGNITUDE OF SECONDARY-SIDE INVENTORY LOSSES

• INITIATE AFW TO REPLENISH SECONDARY-SIDE  !

INVENTORY  ;

TURBINE TRIP INCREASES RCS HEATUP WIHCII LIMITS CORE POWER DUE TO NEGATIVE REACTIVITY FEEDBACK EFFECTS ,

= W PWR ATWS ANALYSIS BASIS

• '79 ATWS SUBMITTAL - NS-TMA-2182 GENERIC ANALYSES REPRESENTING W PWRS INCLUDED SENSITIVITY ANALYSES FOR NUMEROUS PARAMETERS ESTABLISHED THAT REACTIVITY FEEDBACK, RCS PRESSURE RELIEF, AND AFW SYSTEM PERFORMANCE WERE AMONG THE MOST SIGNIFICANT PARAMETERS WITII RESPECT TO MAXIMUM RCS PRESSURES A'ITAINED BASE CASE ASSUMED MTC = -8 PCM/ F, ALL PRESSURIZER PORVS AVAILABLE, FULL AFW CAPABILITY BASE CASE MET 3200 PSIG PRESSURE LIMIT MODELING AMSAC ACTUATION

i

= WIIY WCAP-11992 l "ATWS RULE ADMINISTRATION 4

PROCESS" FOR W PWRS ?

j

• GENERATED BY W/WOG IN 1987/88 IN RESPONSE 4

TO NRC CONCERNS WITH CHANGES IN PLANT'S LICENSED LIMITS ON MTC

PROVIDES STRUCTURED METHOD OF ASSESSING j CHANGES IN PLANT CONDITIONS ON ATWS RISK j AND FOR DEMONSTRATING CONTINUED l COMPLIANCE WITH THE BASIS FOR ATWS RULE l -

PMTC AT PART-POWER CONDITIONS UPRATED POWER CONDITIONS j -

REDUCED RCS PRESSURE RELIEF CAPACITY l -

REDUCED AFW CAPABILITY j

• USES APPROACH THAT PARALLELS THAT USED IN l SECY-83-293 l l -

DETERMINISTIC ACCIDENT ANALYSES THAT l

ESTABLISH CONDITIONS THAT LIMIT 1 MAXIMUM RCS PRESSURE TO 3200 PSIG l -

EQUATES EXCEEDING 3200 PSIG WITH CORE

! DAMAGE l l -

PRA MODEL WITH TARGET ATWS RISK OF IX10 5 PER REACTOR YEAR OR LESS l

i 1

I

e-

=

BENEFITS OF ATWS RULE ADMINISTRATION PROCESS I FOR W PWRS

, l PROVIDES A STRUCTURED BASIS FOR ASSESSING COMPLIANCE WITH A PRESCRIIYI'IVE RULE FOR THE NON-DESIGN BASIS ATWS EVENTS '

l PROVIDES A METHOD OF ASSESSING CHANGES IN

PLANT CONDITIONS IMPORTANT TO THE CONSEQUENCES OF AN ATWS EVENT USING A DETERMINISTIC SAFETY ANALYSES BASIS i

COMBINED WITH PRA MAINTAINS CONSERVATISM OF ATWS RULE BASIS l EQUATING THE RCS PRESSURE LIMIT OF 3200 PSIG WITH CORE DAMAGE i

l PROVIDES A METHOD FOR DEMONSTRATING l CONTINUED COMPLIANCE WITH BASIS FOR ATWS

! RULE FOR PLANTS CURRENTLY LICENSED WITH l PART-POWER PMTC l

CLARIFIES PROBABILISTIC ARGUMENTS IMBEDDED j IN DETERMINISTIC ANALYSIS BASIS FOR ATWS

! RULE WHILE ENSURING THAT THE OVERALL OBJECTIVE OF ATWS RULE IS MAINTAINED.

l ..

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= NRC REVIEW OF WCAP-11992 i

W/WOG ATWS RULE ADMINISTRATION PROCESS (WCAP-11992) USED IN 1988 " ASSESSMENT OF

COMPLIANCE WITH ATWS RULE BASIS FOR

W PWRS" (WCAP-11993)

! THE W/WOG PROGRAM WAS PRESENTED TO THE ACRS ON TWO SEPARATE OCCASIONS FEB.19,1988 SUBCOMMITTEE MEETING ON CORE PERFORMANCE & SCRAM SYSTEM RELIABILITY l -

APRIL 21,1989 INSTRUAENTATION &

! CONTROL SYSTEMS SUBCOMMITTEE ON l ATWS RULE IMPLEhENTATION l

j

• 25 COPES OF WCAP-11993 WERE PROVIDED TO

! NRC FOR INFORMATION PURPOSES ON MARCH 1,

! 1989 l

• WCAP-11992 ATWS RULE ADAUNISTRATION 4 PROCESS PRA MODEL USED AND REFERENCED IN l NUMEROUS PLANT IPES IN RESPONSE TO GL 88-20 i l

1

• WCAP-11992 FORMALLY SUBMITTED TO NRC l FOR REVEW AND APPROVAL BY WOG ON MAY 2, 1995, AFTER BEING INFORhED BY NRC THAT PROCESS WAS NOT RECOGNIZED WHEN

] REFERENCED IN RESPONSE TO PLANT SPECIFIC i

LAR FOR PMTC (BYRON /BRAIDWOOD UNITS) l 2

j ..

=

NRC REVIEW OF WCAP-11992 (CONTINUED)

W/lVOG FORMALLY RESPONDED TO NRC RAIS IN NOV.1995 AND AGAIN IN MARCH 1996 l

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! Deterministic ATWS Analysis Basis in i

! WCAP-11992

) )

= Deterministic analyses of ATWS events j were performed by W in 1979 according l to requirements specified in NUREG-0460.

i l

= The results of these analyses were I provided to the NRC in December 1979

) (letter NS-TMA-2182) and are the

] deterministic analyses supporting the i

i bases for the Final ATWS Rule for W l PWRs.

l l

l = The 1979 ATWS analyses considered the various NSSS & SG configurations of W j

PWRs. For the reference case, the 4-loop j Model 51 SG configuration was used.

i 3

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i j = Results of sensitivity analyses were j included as prescribed in NUREG-0460 i for variations in numerous plant

parameters and system configurations.

f l

= The 1979 ATWS analyses assumed a i

reactivity feedback model using a

-8 pcm/ F MTC which corresponded to a j

value for which 95% of the cycle the j MTC value would be less limiting.

! Results of analyses assuming a 99% MTC i value (-7 pcm/ F) were also provided to document the sensitivity of ATWS to

i. MTC.

l

) = The 1979 ATWS analyses modeled a l

i Turbine Trip at 30 seconds and AFW j actuation at 60 seconds. These mitigating j actuations are those provided by AMSAC.

l l

A

{ ..

= The 1979 ATWS analyses showed that l ATWS events following a loss of main

{ feedwater are RCS pressure limiting l events.

i k

= The 1979 ATWS analyses results for l W PWRs showed that the ASME Service

! Level C stress limits (equivalent to 3200 psig) were not exceeded.

1

= These ATWS analyses also form the bases l for the deterministic analyses supporting l the WCAP-11992 methodology.

l l

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Use of Deterministic ATWS Analyses in the

!l WCAP-11992 PRA Model

}

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= The deterministic ATWS analyses are I used in establishing the values assumed in the WCAP-11992 PRA model and input

via the PR node (RCS Pressure Relief).

i

= The PR node values are based on UET

) values established by examining the reactivity feedback characteristics for a

given core design.

i

= UET values are based on Critical Power Trajectories which reflect the reactivity feedback model used in deterministic ATWS analyses.

l 1

i i- t Critical Power Trajectories l

= Critical Power Trajectory curves represent the reactivity feedback model l used in deterministic analyses of the j limiting ATWS events.

i

= These events are analyzed assuming j

j 1

various Pressure Relief and AFW capacity l configurations and consider Uprated plant l

) conditions.

l

= The Critical Power Trajectory curves l reflect the ATWS analysis heatup /

l shutdown characteristics which yield a

! peak RCS pressure of 3200 psig.

I j = The Critical Power Trajectory curves are

used to determine UET values.

i

)

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~

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ATWS CRITICAL POWER TRAJECTORIES Fraction of 3579 MWt NSSS Power at constant 3200 psig RCS Pressure Loss of Load ATWS Full AFW Capacity I Tin (*F) 2 PORVs 1PORY 0 PORVs 570 0.951 0.946 0.941 i 580 0.884 0.867 0.850 600 0.734 0.697 0.661 620 0.561 0.508 0.456 640 0.360 0.294 0.229 660 0.I20 0.042 -

POWERVS TIN, LOSS OF LOAD ATWS FULL AFW/ PEAK RC.P=3215 PSIA 1.0 g h 3

las g0',  %

= 0,  %

j 0,  %

MN M\

\

NNh ,

h.

0 .

560 500 600 620 640 die COi1EINLET TEMPERATURE, Degree F

= 2 PORVS + 1 PORY o O PORYS

j ATWS CRmCAL POWER TRAJECTORIES Fraction of 3579 MWt NSSS Power at constaat 3200 psig RCS Pressure Loss of Load ATWS Half AFW Capacity I Tin (*F) 2 PORVs 1PORV 0 PORVs 570 0.949 0.944 0.939 580 0.877 0.861 0.845 600 0.720 0.684 0.649 620 0.541 0.490 0.439 640 0.335 0.271 0.207  !

660 0.091 0.014 -

POWERVS TIN, LOSS OFLOAD ATWS HALF APW/ PEAK RC.P=3215 PSIA LG 0.9 05 g"  %

,, g

" b l"  %

I" \\\

0.1 ,

3 0

560 500 600 620 640 660 COREINIEr TEMPERATURE,DegreeF

= 2 PORVS + 1PORY o O PORVS

i UET - Unfavorable Exposure Time i

4 i

j = UET is the time during the cycle when the reactivity feedback is not sufficient to j prevent the RCS pressure from exceeding l 3200 psig (the ASME Service Level C l stress limit).

l i

= UET values are determined by comparing l

j plant specific (or bounding) core heatup / l l shutdown characteristics with the Critical '

Trajectory curves that reflect the ATWS l

analysis.

l l = UET values are calculated for the various

! combinations of Pressure Relief Capacity and AFW Capacity, with and without i

manual operation of the rod control i

l system.

d

a i'

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! PR - Pressure Relief

\

= If an ATWS event occurs, the probability l of adequate pressure relief capacity i depends on a number of variables.

• i i

! These include:

i l

ATWS initiating event Initial power level

)

Number of PORVs available

{

+ Reactivity Feedback l

Auxiliary Feedwater Capacity i

• Operator Intervention I = The PR node values account for the probability of ad. equate pressure relief capacity considering these variables and are based on the 'UETs.

f

1 FEATURES OF THE WCAP-11992 RISK l

MODEL 4

l

= The UET approach for determining

! adequacy of pressure relief (PR) uses the 1979 Westinghouse Deterministic ATWS Analysis submittal am basis for event tree  !

success criteria

= The integrated approach provides a better '

basis for measuring acceptability of parameters affecting ATWS risk than that available from a base case, best-estimate deterministic approach

• Model shows effects of failures of plant systems on ATWS risk

• Model shows effects of unavailability of plant equipment on ATWS risk

MODEL FEATURES (continued)

= PR Node addresses twelve sets of conditions affecting reactivity feedback l needed to prevent RCS pressure from i i exceeding limit

l I l i

i Available AFW Primary Rod Control Capacity PORV Status System 100 % 0 Blocked Manually Actuated 50 % 1 Blocked Not Actuated 2 Blocked

e W

MODEL FEATURES (continued) i 4

WCAP-11992 Event Sequence Model Includes l Conservative Risk Prediction Assumptions l

= No credit for Containment Systems j (assumes Risk = CDF) 1 4

= PR failure is equated to risk:

i j

• No credit for Plant Systems to prevent j l core damage given RCS pressure >

j ASME Service Level C limit (assumes i core damage rather than potentially mitigatable LOCA) '

= Assumes core damage if AMSAC fails

l MODEL FEATURES (continued) l

= For avoidance of ATWS risk, model j requires successful RCS Boration j following PR success i

• For PR, need only limited control rod i insertion for peak RCS pressure reduction benefit (cases with successful

) actuation of rod control system have i

lower UETs than those without l actuation) '

i i

l

• In order to be successful, model j reqmres boration for shutdown, even I for cases where rod control system actuation succeeded

0 j .-

i-l

! Validity of WCAP-11992 Methodology Does Not Depend on Target Risk Values i

= The SECY-83-293 Basis Risk Target was l

used mainly as a point of reference:

i

• No other risk context was available l when the method was developed

• Results for most cases analyzed in the 1

examples provided in WCAP-11992 are substantially lower than the target i

= Example calculations provided in the WCAP are based on conservative assumptions

h l'

l Effort was made to maximize predicted l ATWS core damage frequency rather than

! calculate best estimate: l

! l j = Conservative Reactivity Feedback l assumptions were used i l

= Analysis reflects a limiting plant configuration, per the deterministic

) analysis (# loops, SG type, etc.)

l j = Risk model includes conservatisms (as

noted earlier in this presentation)

a-l

SUMMARY

l i

! WESTINGIIOUSE PWRS ARE IN COMPLIANCE WITH 10 l CFR 50.62 ("ATWS RULE") BY VIRTUE OF HAVING

INSTALLED NRC-APPROVED AMSAC SYSTEMS.

i IN ADDITION TO TIE REQUIREMENTS OF THE RULE, l

THE WCAP-11992 METHODOLOGY PROVIDES A hEANS l FOR EVALUATING PLANT SPECIFIC VARIATIONS AND l CYCLE SPECIFIC CHANGES IN PARAMETERS IMPORTANT l TO TIE ANALYTICAL AND RISK BASES BEHIND TIE

, RULE.

WCAP-11992 UTILIZES DETERMINISTIC ANALYSES IN A PROBABILISTIC FRAMEWORK CONSISTENT WITH SECY-i 83-293 FOR TIIE ATWS RULE.

't 4

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4 o

cc:

Mr. Nicholas J. Liparulo Westinghouse Electric Corporation Mail Stop ECE 4-15 P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Hank A. Sepp Westinghouse Electric Corporation Mail Stop ECE 4-07A P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Andrew P Drake, Project Manaer Westinghouse Owners Group Westinghouse Electric Corporation Mail Stop ECE 5-16 P.O. Box 355 .

Pittsburgh, PA 15230-0355