ML20141B249

Summary of 970415 Meeting W/W in Rockville,Md to Discuss Issues Associated w/AP600 Pra.List of Attendees & Handouts Encl

ML20141B249
Person / Time
Site:	05200003
Issue date:	05/13/1997
From:	Joseph Sebrosky NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9705150204
Download: ML20141B249 (35)

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! }t _ NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2006H001

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+,***** May 13, 1997 f ..

APPLICANT: Westinghouse Electric Corporation.

{ FACILITY: AP600

SUBJECT:

SUPMARY OF APRIL- 15, 1997, MEETING WITH WESTINGHOUSE TO DISCUSS ISSUES ASSOCIATED WITH THE AP600 PROBABILISTIC RISK ASSESSMENT (PRA) 4 1

. The subject-meeting was held on April 15, 1997, in the Rockville, Maryland, -1

.; offices of the Nuclear Regulatory Commission (NRC) between representatives of <

! Westinghouse, and the NRC staff. Attachment 1 is a list of meeting attendees.  !

' Attachment 2 and 3 are the handouts provided during the meeting by the staff '

and Westinghouse respectively._  ;

E ..  !

Prior to the meeting the staff sent Westinghouse a letter dated April 3,1997, which detailed three areas of the PRA in which the staff had concerns. The *

, areas were the focused PRA, steam generator tube rupture (SGTR) accident sequence modeling, and " safety insights" developed from the PRA.

~

Focused PRA i

.The focused PRA, which is a sensitivity study of the baseline PRA, was recognized as being important by both Westinghouse and the staff in resolving the issue of Regulatory Treatment of Non-Safety Systems (RTNSS). .The staff was 'especially concerned with the im)act that revisions to risk-important failure data had on the results of tie focused PRA. In addition, the staff  ;

was concerned about the loss of AC power assumption that Westinghouse used in ,

the focused PRA.

The risk-important failure data that were discussed during the meeting 2 included the data for the following components

In-containment refueling water storage tank injection check valves, squib valves, and Reactor Coolant .

Pump (RCP) trip breakers. The staff believed additional justification is j needed to support the: failure data for the above components. (Westinghouse 1 4

presented the data in attachment 4 during this discussion). The staff agreed 1

. to do a final review of the previously. submitted failure data for the check ,

valves, squib valves,-and RCP trip breakers. Westinghouse agreed to assemble the pertinent data on the above components from their PRA into a single.  :

. ' package and to resubmit this information to assist the staff in reviewing the .h,'1 3 i material.,: Following the review, the staff committed to forward their findings \)

to Westinghouse.-

.The loss of AC power assumption used in the focused PRA was then discu'ssed. {

In_the focused PRA,. Westinghouse assumed that, after the' initiating event  ;

h

occurred, all:non-safety systems (including AC power) failed. However.

-special, consideration was given to the Anticipated Transient Without Scram events where main AC power was assumed available to the rod control system. ,

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The staff believed that the loss of AC power assumption eliminated the likelihood of some high risk-important failures such as the failure of one or more reactor coolant pumps to trip. Westinghouse believed that the modeling of the loss of AC power was consistent with the agreements reached and documented in the SECY papers concerning RTNSS. Westinghouse further stated ,

that they believed the focused PRA was never intended to combine the conservatisms associated with the loss of certain non-safety systems with the

conservatises (e.g. adverse systems interactions) associated with other non-l safety systems being available. . The staff agreed to evaluate the loss of AC ,

3 power assumption and how it could be treated in the " Adverse Systems Interac- i tions" portion of the RTNSS review to capture events like the reactor coolant pumps failing to trip, within the context of the focused scope PRA. l SGTR accident secuence modelina i

The staff was concerned with the lack of thermal-hydraulic analyses supporting  ;

t the assumption that the AP600 design can use non-safety related systems only

as the first line of defense to mitigate SGTR accidents. Westinghouse  :

believes that the AP600 has unique features that make the AP600 SGTR response

, different from other pressurized water reactors. Additionally, Westinghouse stated that they were answering similar questions about SGTR events from the i Reactor Systems Branch. The staff agreed to have the Reactor Systems Branch involved in the review of the thermal-hydraulic analyses that Westinghouse  ;

will be providing in response to concerns regarding the PRA SGTR modeling. l Safety Insichts l

The staff was concerned with a delay in Westinghouse's submission of the  !

safety insights for external events. The staff was also concerned that during 1 a review of the safety insights for the internal events, several important PRA L assumptions may have been overlooked and not included in the submittal.

Westinghouse responded that the safety insights for the external events had ,

been transmitted April 11, 1997. Westinghouse believes that the safety )

insights list they developed for internal events was an inclusive list of ,

insights. Westinghouse propcsed that the staff's set of insights resulting l

from their review of the PRA be shared with Westinghouse. Unless Westinghouse determined that there was technically incorrect information in the staff's list there would be no new meetings or information. transfer. The staff's

insights would be added to the AP600 Design Control Document. The staff l agreed to provide feedback to Westinghouse on its proposal.

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_May 13, 1997 A draft of this meeting sumary was provided to Westinghouse to allow them the opportunity to ensure that the representations of their coments and discus-sions were correct.

original signed by:

Joseph M. Sebrosky, Project Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003 Attachments: As stated cc w/ attachments:

See next page DISTRIBUTION w/ attachment:

Docket File PDST R/F TKenyon PUBLIC WHuffman DJackson JSebrosky Alevin, 0-8 E23 GHsii, 0-8 E23 JLyons. 0-8 D1 DISTRIBUTION: w/o attachment:

SCollins/FMiraglia, 0-12 G18 RZimmerman, 0-12 G18 TMartin, 0-12 G18 MSlosson TQuay ACRS (11)

JMoore, 0-15 818 WDean, 0-17 G21 JFlack, 0-10 E4 NSaltes, 0-10 E4 MCunningham, T-10 E50 BPalla, 0-8 H7 GHolahan, 0-8 E2 DOCUMENT NAME: A:FCSD PRA. SUM (6F AP600 DISK) *See previous concurrence n . wy .e m. t. m m m. wm c . coey m.ui .n. chm.ne.new.w. r . copy wwn .n.ch no.noo. n . Nocopy 0FFICE PM:PDST:DRPM _ SPSB:DSSA D:PDST:DRPM l NAME JSebrosky:sg AN) JF1ack* TRQuay 412a DATE -

05/rlf97 i/ 05/05/97 05/13/97 OFFICIAL RECORD COPY

, . l j

l Westinghouse Electric Corporation Docket No.52-003  !

l cc: Mr. Nicholas J. Liparulo, Manager Mr. Frank A. Ross l Nuclear Safety and Regulatory Analysis U.S. Department of Energy, NE-42 Nuclear and Advanced Technology Division Office of LWR Safety and Technology Westinghouse Electric Corporation 19901 Germantown Road P.O. Box 355 Germantown, MD 20874 Pittsburgh, PA 15230 Mr. Ronald Simard, Director Mr. B. A. McIntyre Advanced Reactor Program Advanced Plant Safety & Licensing Nuclear Energy Institute l Westinghouse Electric Corporation 1776 Eye Street, N.W. <

Energy Systems Business Unit Suite 300 ,

Box 355 Washington, DC 20006-3706 l Pittsburgh, PA 15230 i Ms. Lynn Connor I Ms. Cindy L. Haag Doc-Search Associates ,  !

Advanced Plant Safety & Licensing Post Office Box 34 i Westinghouse Electric Corporation Cabin John, MD 20818  !

Energy Systems Business Unit j Box 355 Mr. James E. Quinn, Projects Manager i Pittsburgh, PA 15230 LMR and SBWR Programs GE Nuclear Energy Mr. M. D. Beaumont 175 Curtner Avenue, M/C 165 Nuclear and Aavanced Technology Division San Jose, CA 95125 l Westinghouse Electric Corporation  !

One Montrose Metro Mr. Robert H. Buchholz '

11921 Rockville Pike GE Nuclear Energy Suite 350 175 Curtner Avenue, MC-781 Rockville, MD 20852 San Jose, CA 95125 Mr. Sterling Franks Barton Z. Cowan, Esq.

U.S. Department of Energy Eckert Seamans Cherin & Mellott NE-50 600 Grant Street 42nd Floor ,

19901 Germantown Road Pittsburgh, PA 15219 l Germantown, MD 20874 1 Mr. Ed Rodwell, Manager j Mr. S. M. Modro PWR Design Certification i Nuclear Systems Analysis Technologies Electric Power Research Institute i Lockheed Idaho Technologies Company 3412 Hillview Avenue  ;

Post Office Box 1625 Palo Alto, CA 94303 l Idaho Falls, ID 83415 Mr. Charles Thompson, Nuclear Engineer AP600 Certification NE-50 19901 Germantown Road Germantown, MD 20874

. =

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WESTINGH0USE AP600 PRA MEETING ATTEM EES April 15, 1997 N8E ORGANIZATION CINDY HAAG WESTINGHOUSE BRIAN MCINTYTE WESTINGHOUSE JIM FREELAND WESTINGHOUSE BARRY SLOANE WESTINGHOUSE TERRY SCHULZ WESTINGHOUSE SELIM SANCAKTAR WESTINGHOUSE STEVE F0WLER WESTINGHOUSE TIM BUETER WESTINGHOUSE JIM FULFORD LIS/NUS/SCIENTECH J0E SEBROSKY NRR/DRPM/PDST l BILL HUFFMAN NRR/DRPM/PDST JOHN FLACK NRR/DSSA/SPSB  :

NICK SALTOS NRR/DSSA/SPSB i MARK CUNNINGHAM NRR/DSSA/SPSB BOB PALLA NRR/DSSA/SCSB TED QUAY (PART TIME) NRR/DRPM/PDST GARY HOLAHAN (PART TIME) NRR/DSSA Attachment I  ;

1, . ,

4 AP600 DESIGN PRA REVIEW l

i

! MAJOR ISSUES i

a 4

i I

NRC/NRR/DSSA/SPSB a

d i NRC/ Westinghouse meeting l April 15,1997 l

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4 Attachment 2

i e

MAJOR PRA REVIEW ISSUES 1

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1. SAFETY INSIGHTS i

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2. SGTR MODELING ISSUES 4

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. 3. FOCUSED PRA ISSUES l

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SAFETV INSIGHTS Identification of design and operational i i requirements based on PRA insights and l

! assumptions l

. 1

• Partial submittal i l

l Insights for external events, due  ;

l 2-28-97, not received l e Incomplete submittal Several important PRA assumptions l were not included in the submittal Uses wording from the SSAR rather than from the PRA Does not include " integrated" insights e Impact on FSER schedule

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SAFETY INSIGHTS (continued)

Several important PRA assumptions were not included in Westinghouse's submittal i

Examples:  :

1 i

j e Separate hardware and application l

! software modules for the Rx trip and  !

) ESF functions of the PMS, with the l l exception of a limited set of software modules which control fundamental

computer operations. -

l e Operators will be trained to recognize the need for and take action to manually trip the reactor in time to avoid an ATWS when automatic trip

fails following a LOCA or transient event (skill-based assumption in HRAD.

l l

e The reliability of the RCP trip breakers  ;

(to open on demand) is important :D-RAP:. /

\-

! l y . _. . _. . _.

l i

SAFETY INSIGHTS (continued ll e insights need to reflect that adequate net positive suction head (NPSH;l  ;

exists for the NRHR pumps to prevent l pump cavitation and failure even  !

l when the IRWST and surnp inventory '

is saturated. ,

l e Insights need to reflect that NRHR interlocks, administrative controls,

operator actions, procedures, etc. are necessary.

e ADS is actuated automatically when l AC power to 4160 V busses is unavailable for 15 continuous hours.

In the PRA it is assumed that the operator will act to override automatic depressurization in certain cases.

Insights need to identify cases and reflect the need for procedures and i training.

r D

4.-

SAFETY INSIGHTS (continued?

l j

Several important insights are based on

! statements from the SSAR and do not

reflect accurately the PRA assumptions i Examples: '

L e Sof tware features, such as " fail safe" l l capability and " functional diversity," l

! are assumed in the PRA. " Insights" submittal just mentions a V&V l program in accordance with IEEE j standards as stated in SSAR.

l l e No I&C failures are modeled in the

l. PRA due to loss of HVAC. Submittal l states that "PMS equipment is

! designed to accommodate a loss of l

HVAC." However, such statement

should be tied to mission times and i l refer to features and analyses which

! prove that this statement is true.

4

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i l SGTR MODELING ISSUES l

e Lack of T-H analyses supporting the.

4 assumption that the AP600 design can use nonsafety-related systems only as the first line of defense to

! mitigate SGTR accidents i

e Lack of T-H analyses supporting

success criteria and modeling of sequences involving failure to trip one or more RCPs e The staff asked W for such analyses over a year ago

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j l

l W responded w/o the analyses L

staff followed up Waiting for W's response impact on baseline and focused PRAs .

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FOCUSED PRA ISSUES i 1. Revised risk-important failure data

) Examples:

• IRWST injection check valve failure .

rate reduced from 1 E-6/hr to 2E-7/hr l

CCF is highly risk important Unresolved DSER open item i questioning the 1 E-6/hr rate i

o -

Unresolved DSER open item questioning assumed CCF multiplier

• Explosive valve failure rate reduced from 3E-3/d to 5.8E-4/d without l

justification l l e RCP failure to trip probability reduced l L

from 1.2E-3/d to 4.2E-4/d without

justification  !

-O c .

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FOCUSED PRA ISSUES (continued)

2. No bases for some important l modeling assumptions Example:

• Success criteria.for systems and operator actions in SGTR sequences (important contributors to LRF) l

3. Inappropriate use of software CCFs e The CCF probability which includes l failure of both PMS and PLS is used in several top cut sets instead of the higher CCF probability of just the PMS software i

-g-

FOCUSED PRA ISSUES (continued)

4. Issues with criteria for screening l SSCs, contributing to initiating I

events, for RTNSS Examples:

e On'y one criterion (% contribution) e Criterion relaxed from 1 % to 10%

o Reliability assurance not addressed

5. Issue with assumption about AC power failure at the onset of an accident initiating event LIE?

l e Unrealistic and not consistent with the objectives of the focused PRA e Impact on PRA results and insights e Assess assumption's impact

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AC POWER IN THE FOCUSED PRA e Assumed that AC power is lost at the onset of an accident IE, except to the l Rx trip breakers

• Focused PRA objectives assess CDF and LRF using the

" passive" safety systems only use PRA results & insights to decide on RTNSS

• Assumption is not consistent with focused PRA objectives o

l An "unlikely" event is assumed to always occur and eliminate the likelihood of important failures (CDF and LRF are " artificially" decreased? !

results are impacted and insights masked i

AC POWER IN THE FOCUSED PRA

{ continued?

e assumption is unrealistic and not 3 conservative 3

it is unlikely that AC power will be g unavailable following a non-LOSP IE 1

e inconsistent use of assumption AC power to the Rx trip breakers but not to every thing else .

e assumption eliminates the likelihood of some high risk-i.mportant failures failure of one or more RCPs to trip failure of PRHR actuation and CMT 4 injection due to I&C software and I hardware .CCFs f l r

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_ _ _ _______m_-------

AC POWER IN THE FOCUSED PRA (continued)  !

e Assumption impacts PRA conclusions A preliminary study by W indicates  !

significant impact of this  !

assumption on results and insights l

CDF increases by over 1 F-5/yr LRF increases by over 1 E-6/yr issues with W's study (further CDF

& LRF increases very likely) l l

e Assumption ma.sks important insights Importance of software reliability and DAS mitigation 01 SGTR events when ,

I one or more RCPs fail to trip (T-H .

analyses pending?

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! NRC/ WESTINGHOUSE MEETING i

4 l ON PRA TOPICS i APRIL 15,1997 l

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l e

Attachment 3

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--w--a-s4 a. neM.:n- .o.44.- 1 Ae a M 9- um -A w>-- ake + 4ae sM..

L, SAFETY INSIGHTS -

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" Systematic search to identify additional design and l operational requirements for the certified design, based '

on PRA insights and assumptions" d

b i

e How do we see the status l e A suggestion for closure i

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SAFETY INSIGHTS .

How We See The Status e The NRC SER authors have already developed an extensive " safety insights" list.

e Westinghouse concern is centered around the potential impact of this " insights" list on the future Design Certification Document - i.e. much of l this list will end up in the DCD ?

e if the " safety insights" list is unduly detailed (e.g. -

compared to previous submittals), how does it potentially impact the DCD?

Otherwise, an " insight" is in the eye of the  !

beholder; the NRC SER authors could include what they see as " insights" in their document.

___,,_________m_ ____m_-_---__

SAFETY INSIGHTS ,

1 .

1 L History

• PRA insights were presented in chapters of the

. AP600 PRA submittal to the NRC

!

• NRC SER authors wrote their own version of

" safety' insights" containing more elements and 4

formatted by system.

o l

• Westinghouse wrote another set of insights similar

to the System 80+ submittal scope and format, as suggested by the NRC reviewers. This was done

! systematically with an expert group. .

e All-Westinghouse insights material has been submitted to the NRC with the external events material transmitted April 11 completing the set.

. e There is nothing more Westinghouse could add to i

the set already compiled between what the NRC  !

l SER authors wrote and what Westinghouse has

[ submitted. )

• NRC SER authors currently have enough  !

! information and have already written a

considerable amount of " safety insights". This

process can.be wrapped up by the NRC SER authors soon without any additional " safety  ;

I insights" input from Westinghouse. . H i

SAFETY INSIGHTS A Suggestion For Closure i

• Closure can be accomplished by the following

path:

i i

! 1 NRC SER authors wrap up their " safety insights" writeup based on their already  ;

j gathered input plus what Westinghouse l submitted.

[

k. .

! 2 The SER writeup is shown to Westinghouse i

for final potential input.

i 3 Westinghouse will examine the vniteup for l suggestions that the SER authors may choose l to incorporate. These suggestion could be of ;

three types: i l A Editorial comments.

B Westinghouse position that an item does 1 not belong to the list.

I  !

l C Westinghouse position that an item is technically incorrect. l L

L

SAFETY INSIGHTS A Suggestion For Clo'sure, Cont.

4 NRC SER authors would consider Westinghouse comments if time is available, and in accordance with the following:

Type A and B comments are optional for the NRC SER author to consider.

Type C comments are expected to be very few, if non-existent. If any exist, they must be addressed mutually.

e Unless type C comments arise, no new meetings

or information transfer is expected.

r l

C

Modeling of SGTR in the AP600 PRA Brief History:

• Reviewer sent RAls in March,1996.

General questions about the success criteria Westinghouse answered questions verbally in subsequent discussions Formal replies submitted in November,1996.

l e in January 1997 meeting, received more detailed l questions from the reviewer.

Westinghouse provided verbal responses and presented T/H analyses at the meeting.

• Received more RAls in March 1997 requiring additional T/H analyses.

l Mainly concern the use of CVS and operator actions.

l These questions can be answered verbally today.

The written responses will be available soon.

l l L

Figure 1 - AP600 SG Tube Rupture Levels of Defense CURRENT PWR AP600 SGTR , SCTR 8

8 I E E SSAR AUTO HH9. AFWS AUTO CVCS. SFws SAFE TY WANUAL SG ISR. LEAK ISOL ATED '

MANUAL SC ISOL. I LE AK ISOLATED 8

CASE RCS COOL /DEPRES dCS COOL /DEPRES I e h 4 AUTO HHSt.

' SSAR AUTO CWf. PRHR HX.

8 CONT SPRAY I LEAK NOTISOL SAFETY CVCS/5FuS ISOL. I LEAK ISOLATED WAN RCS VENT RCS WNTED CASE SC ISOL, PCS qJ l i 4 4 CORE DAWACE AUTO CMT.

PARTIAL A05 > I E AK NOT ISOL WAN RNS INJECT RCS VENTED i NO CONT FLOOD t B

' h AUTO Cuf. FULL ADS IRWST PCS I LEAK NOT ISOL RCS VENTED

' CONT FLOOD e a h

MAN FULL ADS AUTO ACCOW. tRwST. LE AK NOT ISOL

' PCS RCS VENTED

' CONT FLOCD

i s 1 r amust CORE paung

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SC 'R in the AP600 PRA l e PRA model is based upon Understanding the AP600 response to an l SGTR Integral tests  ;

Safety analysis T&H analysis Computer simulations e Uses conservative assumptions relative to the  ;

expected plant performance. (Operator action times, need for ADS, etc.)

l e Multiple, diverse types of responses to a SGTR.

1. active features and operator actions (similar to e

current plants)

2. passive features

a. automatic, and

b. independent of AC power.

e 5

i i

. SGTR IN THE AP600 PRA l

e Active System Response l (This plant response not applicable to the focused )

PRA.) )

l

> Nonsafety-related, active features and operator j actions equalize the RCS and SG pressure and terminate the leak.

4

> CVS and SFW are part of this response.

Supported by testing and analysis

> The startup feedwater pumps provide RCS cooling through the intact SG. l l

j > SG overfill is prevented through automatic control on feedwater.

4 4

l > One CVS makeup pump provides sufficient

RCS makeup to prevent RCS voiding.

> Operator actions:

Isolate the faulted SG Cool the RCS using the intact SG Reduce RCS pressure with CVS auxiliary spray or ADS stage 1 (backup) i Analyses show sufficient time to perform these actions This response has been shown to.be effective through ,

analysis and computer simulations. I i )

• Active System Response This response is similar to current plants.

Most notable difference is the high head safety injection in current plants have greater capacity than ,

the AP600 CVS pumps. This higher capacity maintains l the RCS pressure higher during an SGTR relative'to the i SG pressure than the CVS does in the AP600. This makes the leak flow larger and reduces the time

- available for operator actions in current plants.

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S 4

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AP600 Focused PRA l Summary of Actions to Date:

I e September,1993: Completion of initial RTNSS evaluation documented in WCAP-13856 l e June,1995: Focused PRA updated to latest (Revision 2) baseline PRA model e June,1996: Meeting to discuss focused PRA held between NRC reviewers and Westinghouse; ,

request for sensitivity study made by the reviewers '

e September,1996: Focused PRA updated to latest (Revision 7) baseline PRA model  !

o- December,1996: Sensitivity study transmitted to NRC reviewers ,

I e February,1997: Questions received from the NRC reviewers regarding sensitivity study, no further comment on the submitted focused PRA L

I l ,

l l

l L

'\

The AP600 Focused PRA Method i 1

• Consistent with SECY-94-084

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• The scope and frequency of initiating events are maintained in the focused PRA as in the baseline PRA i

• Following the initiating event, all nonsafety system l credit is removed in Level 1 and Level 2 PRA i calculations i 6

i i

l I

4 l

4

l Modeling and Assumptions i 1

l

e Clarified the definition of the " initiating event" to l l be the point at which a demand for reactor trip 1 ii should occur j e Assume failure of all nonsafety systems after l initiating event occurs i i

j

• Special consideration given to the ATWS events:

i l

Initiating event defined to be the point at which the safety-related trip function fails to trip in response to an ATWS precursor 4

i -

After the initiating event occurs, it is assumed main AC power is available to rod control

system (and thus, to the RCPs since they are l on same bus) so as to not change the initiating L event frequency l

l- e This approach adopted or else ATWS l l event would not occur (rods drop on loss  !

of main ac power to rod control system) i e Power to reactor coolant pumps is a j penalty 1

t 5

i

AP600 IRWST injection check valves e simple check valve design

.

• RCS grade water environment e stainless steel construction Previous AP600 PRA Revision URD Value Basis Value used in PRA 1.0E-06/hr extrapolation of 0 failures in data 1.0E-06/hr

= 8.76E-03/ demand (for adverse conditions)

Current AP600 PRA (based upon current design)

URD Value Basis Value used in PRA 2.0E-07/hr 0 failures in data 2.0E-07/hr

= 1.75E-03/ demand (for plant design conditions)

Value in NUREG/CR 2815 and used in IPEs on some current plants 2.0E-07/hr

= 1.75E-03/ demand

Conclusion:

The AP600 check valve value is based upon the current design and the URD value.

Attachment 4

v RCP Breaker Data URD Value Basis 6.0E-04/ demand 1993 generic plant data 1.4E-06/ hour calculation from URD data Plant Specific NPRDS Value Basis 4.8E-07/ hour 1986-1996 Westinghouse plant specific data (1.7E-04/ demand)

Conclusion:

The AP600 RCP breaker value is based upon significant' Westinghouse plant specific data.

l l

l

Squib Valve Data .

URD Value Basis 3E-03/ demand small data population with 0 failures & assumption Sandia Data Basis 2.0E-04/ demand large population of data with 0 failures & analysis 3.2E-04/ demand large population of data &

analysis Value used in the AP600 PRA is a geometric average of the above values, giving each of them an equal weight.

1 The squib valves in the Sandia data are the same or

! similar to the AP600 design.

l

Conclusion:

l The AP600 squib valve value is based upon extensive

! experience with similar equipment.

l l

1-i 1 I

l