ML20245K749

Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 4.5.3 Re Reactor Trip Sys Reliability for All Domestic Operating Reactors

ML20245K749
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Site:	Millstone
Issue date:	06/21/1989
From:	Office of Nuclear Reactor Regulation
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References
GL-83-28, NUDOCS 8907050242
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6  % Uft!ITED STATES

[ c i NUCLEAR REGULATORY COMMISSION 7js C E W ASHING TON, D. C. 20555 K,.&r.M' /j' SArETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TO FACILITY OPEPATING LICENSE NO. DPR-65 i

NORTHEAST NUCLEAR ENERGY COMPANY, ET AL.

MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 DOCKET NO. 50-336 GENERIC LETTER 83-28, ITEM 4.5.3 REACTOR TRIP SYSTEM RELIABILITY FOR ALL DOMESTIC OPERATING REACTORS

1.0 INTRODUCTION

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On February 25, 1983, both of the scram circuit breakers at Unit 1 of the I Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal

, from the reactor protection system (RPS). This incident was terminated l manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Prior ts this incident, I on February 22, 1983, at Unit 1 of the Salem Nuclear. Power Plant, an automatic '

l trip signal was generated based on steam generator low-low level.during plant startup. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip.

Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO), directed the staff to investigate and report on the

generic implications of these occurrences at Unit 1 of the Salem Muclear Power Plant. The results of the staff's inquiry into the generic implications of the Salem Unit 1 ircidents are reported in NUREG-1000, " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant". As a result of this investigation, the Commission (NRC) requested (by Generic letter 83-28 dated July 8,1983) all licensees' of operating reactors, applicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of these two ATWS events.

The licensees were required by Generic Letter 83-28, Item 4.5.3 to confirm tht on-line functional testing of the reactor trip system (RTS), including l independent testing of the diverse trip features, was being performed at all l plants. l Existing intervals for on-line functional testing reouired by Technical Specifications were to be reviewed to determine if the test intervals were adequate for achieving high RTS availability when accounting for considerations such as: (1) uncertainties in component failure rates; (2) uncertainties in common mode failure rates; (3) reduced redundancy during testing; (4) operator error during testing; and (5) component " wear-cut" caused by the testing.

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4 2.0 DISCUSSION The NRC's contractor, Idaho National Engineering Laboratory (INEL), reviewed the licensee Owners Group availability analyses and evaluated the adequacy of the existing test intervals, with a consideration of the above five items, for all plants. The results of this review are reported in detail in EGG-NTA-8341, "A Review of Reactor Trip System Availability Analyses for Generic Letter i 83-28, Item 4.5.3 Resolution," dated March 1989 and summarized in this report.

The results of our evaluation of Item 4.5.3 and our review of EGG-NTA-8341 are presented below. 4 The Babcock & Wilcox (B&W), Combustion Engineering (CE), General Electric j (GE), and Westinghoum (W) Owners Groups have submitted topical reports either l in response to GL 83-23, Item 4.5.3 or to provide a basis for requesting i Technical Specification changes to extend RTS surveillance. test intervals ]

(STI). The owners groups' analyses addressed the adequacy of the existing (

intervals for on-line functional testina of the RTS, with the considerations l required by Item 4.5.3, by quantitatively estimating the unavailability of the RTS. These analyses found that the RTS was very reliable and that the I unavailability was dominated by common cause failure and human error.

The ability to accurately estimate unavailability for very. reliable systems was considered extensively in NUREG-0460, " Anticipated Transients Without Scram for Light Water Reactors", and the ATWS rulemaking. The uncertainties of such estimates are large, because the systems are highly reliable, very little experience exists to support the estimates, and common cause failure probabilities are difficult to estimate. Therefore we believe that the RTS unavailability estimates in these studies, while ustful0 for evaluating test intervals, must be used with caution.

NUREG-0460 also states that for systems with-low failure probability, such as the RTS, common mode failures tend to predominate, and, for a number of reasons, additional testing will not appreciably lower RTS unavailability.

First, testing more frequently than weekly is generally impractical, end even so the increased testing could at best lower the failure probability by less than a factor of four compared to monthly testing. Secondly, increased testing could possibly increase the probability of a common mode failure through increased stress on.the system. Finally, not all potential failures are detectable by testino. In summary, NUREG-0460 provides additional justification to demonstrate that the current monthly test intervals are adequate to maintain high RTS availability.

3.0 CONCLUSION

All four vendors' topical reports have shown the currently configured RTS to be highly reliable with the current monthly test intervals. Our contractor has reviewed these analyses and performed independent estimates of their own which

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conclude that the current test intervals provide high reliability. In addition, the analyses in NUREG-0460 have shown that for a number of reasons, more frecuent testing than monthly will not appreciably lower the estimates' of failure pretability.

- E.ased on our review of the Owners Group topical reports, our contractor's independent analysis, and the findings noted in NUREG-0460, we conclude that  !

the existing intervals, as recommended in the topical reports, for on-line  !

functional testing-are consistent with achieving high RTS availability at all operating ' reactors , <

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ENCLOSURE 3 EGG-NTA-8341 Marcn 19E9 TECHNICAL EVALUATION REPORT w

Idaho A REVIEW OF REACTOR TRIP SYSTEM AVAILABIL:Ty Nationa/ AN: LYSE 5 FOR GENERIC LETTER 53-28. ITEM 4.5.3, Engineering RE50LuTIou Laboratory

[ David P. Mackowiak fyi'ij2 Jonn A. Senroecer c ' E m e ';.

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U.S. NUCLEAR REGULATORY COMMISSION a o , ., o-.e nr. ,

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a NOTICE Ths report was prepared as an account of work sporpored by a:4 agency of

_the Uruted States Govemment. Neither the Uruted Sates Government nor any agency theteef, not any of their employvts, makes any warmnty, expressed or imphed, or assumes any lega]liabibty or responsibility for any tturd party's use, or the results of such use of any information, apparatus, product or proc.

ess disciowd in this report, or represenu that its use by such third party would not infnnge pnsately owned nghts.

EGG-NTA-83cI l

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l TE:mN::AL EVALUAT:0N REPORT: A REVIEW OF REACTOR TRIP SYSTEM AVAI LL E: LITY ANALYSES FOR GENERIC LETTEF 83-28, ITEM 4.5.3, RESOLUTION Davic P. Mack 0wiak Jonn A. Schroeder EG&G Icaho, Inc.

Icaho Falls, I:aho 83415 FIN 05:01: Evaluation of Conformance to Generic Letter 83-28 for ors (Project 2)

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) ABSTRACT The ::aho. National Engineering Laboratory (INEL) conducted a j technical review of the commercial nuclear reactor licensees' responses te the re:;irements cf tne Nuclear Regulatory Commission's (NRC's)

Gereri: Letter 53-25 (GL S3-28), Item 4.5.3. The results of this review, i' all pla-ts are shown to be covered by an adequate analysis, will i

evice the NRC staff with a basis to close out this issue with no fu*ther review. 'The licensees, as the four vendors' Owners' Groups, su:mitte: analyses to the NRC either cirectly in response to GL E3-28, Item 4.5.2, er to provice a basis for requesting changes to the Technical Specifi:ations (TS) that would extend the Reactor Protection System'(RPS) su veillance test intervals (ST!s). To conduct the review, the INEL te#ine: in-ee crite-ia te cetermine the adequacy, plant applicability, and a:Ceptability cf the results. The INEL examined the Owners Groups' reports to cetermine if the analyses and results. met the established criteria. Fort St. Vrain's responses to Item 4.5.3 were also reviewed.

The :NEL review results show that all licensees of currently operating

m e- ial na: lear react:rs have adecuately demonstrated that their current cr-line RFS test intervals meet the requirements of GL 83-28, Item 4.5.3.

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SUMMARY

ine tw: anticipated transient without scram (ATWS) events at the 56 ee Nuclear Power Plant in February of 1983, focused the attention of the Nu: lear Reculatory Commission (NRO) on the generic implications of ATWS events. The NR then published Geretic Letter 83-28 (GL 83-28) which listed the acticns the NRC required of all licensees holding cperating licenses and eine s with respect to assuring the reliability of tre Rea:ter Prctection System (RPS). GL 83-28, Item 4.5.3, required licensees 1: demonstrate by review that the current on-line functional testing intervals a e cersistent with achieving high reactor trip system (RTS) availability. The licensees respended to the GL 83-28. Item 4.5.3, re:uireme ts as Caners Groups with reports either in direct response to Item 4.5.3, er with a technical basis for requesting extensions to the sa veian:e test inte vals (STIs) that generally included the Item 4.5.3 re:uirec eviews.

Tne.NR:'s Inst u entat er a .d Control Systems Branch (ICSB), Cffice i

of Nu: lear Rea: Or Regulation (NRR), requested the Idaho National E*gineering Laterat:ry (INEL) to review the licensee availability analyses anc evaluate tre overall adequacy of the existing test intervals. INEL review results shewing general compliance with Item l 4.5.3 will p-cvide the NRC with a basis to close out Item 4.5.3 without furthe- review.

4 For tne review, ne INEL defined three acceptance criteria, reviewed the 1":ensees :::ical reports, centractor review reports, and NRC safety l evaluations, and determined the adequacy of the analyses anc the RTS l availability estimates with regard to the review criteria.

The INEL revie criteria to determine the licensees' Item 4.5.3 3

cem;11ance we e, (I) the five areas of concern of Item 4.5.3, (2) the analyses' plant a:pli: ability, and (3) the NRC's RTS electrical i unavailability base case estimates from the ATWS Rulemaking Paper, l SECY-!3-293. I iii i

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Each Owners Groups' reports were reviewed to enture that all five areas cf concern from. Item 4.5.3 were either included in the analyses or q sne n net to be'significant with regard to.RTS availability. The INEL~

review also ensurec that'the individual plants' differences from the ar.alysis' redels were taken'into account.and their' effects were shown'not 1: significantly affect RTS unavailability. The Fort St.'Vrain responses l'

to Item 4.b 3 were also reviewed.

l Tne 0 ers Groups' RTS unavailability' estimates'were compared to the NR 's ATWS Rulemakir.g generic RTS unavailability estimates to determine

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the acceptability of the.0wners' Groups' conclusions that high RT5' availability was demonstrated in the analyses. I The resLits of the INEL review showed that all licensees of cu-rertly ccerating comme *cial nuclear reactors have. adequately cemenstrated that their current on-line surveillance test intervals are censiste ; .th acFieving high RTS availability.

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ACRONYMS A~WS Arti:d: ate: Transient W4 thout Scram Eli Eat:::L & Wiltex EN; E ::kraven National Lateratory i

CE Cc-0.stion Engineering SE Ge e al Electric E1;- ee perature Gas .ocle, n. neactor

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5E :nstra e.tation anc Control Systems Branch

NE.  ::an: National Engineering Laboratory LW; L':-: -nater Rea:ter

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N5: N.:'.ea- Facility Sa'ety Com.ittee NR: N.: lea- Re;ulatery Commissien )

N;; C " :e e' N.: lear Rea: 0- Regulation rusa art .re-a:1cns u neview Committee FSC ~.011: Se . ice Com:any of Coloraco i PWR F-essuri:e: Water Rea::cr R55"i? Rea::c- Sa'ety Stu:, vetne:01cgy Applications Program  !

RF5 Rea:: - Prote::icn System RTS Eea::Or Trip Syster SER Safety Evaivation Report STI Surveillance Test Interval TER Te:hnical Evaluation Report W Westir:neuse Y

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CONTENTS .

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4.3 GE Piar.ts .... .... . ..... ....................... 9 4.4 kestir; house Plants '

...................... .. 10 4.5 C artitative Review of Vendors' RTS Unavailabilities . .. . 11

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1. Comparison of Vender a..d NR" RTS Unavailability

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TECHNICAL EVALUATION REPORT: A REVIEW OF REACTOR TRIP SYSTEM .

AVAILABILITY ANALYSES FOR GENERIC LETTER 83-28, ITEM 4.5.3 RESOLUTION t

1. INTRODUCTION 'j l

1.1 Historical Backeround In Fe:rua*y of '933,

. two events o::urred at the Salem. Nuclear j Generating Station that focused Nuclear Regulatory Commission (NRC) )

attention on the generic implications of anticipated transient without scrae (ATWS) events. .)

First, on February 22, during startup of Unit I an automatic trip ,

signal generated as a. result of a steam generator low-low level failed to-cause a rea: tor scram. The reactor was tripped manually by an operator. j alrest coin:icentally witn the automatic trip signal, so the ffet that the 1 automati trip had failed to cause a scram went unnoticed.

inree days later on February 25, both of-the scram breakers at Unit 1

' ailed to open on an automatic reactor protection system (RPS) scram si;*al Tre ::eraters took action to control this se:ond ATWS and  !

su::ee:ed in terminating the incident in about 30 seconds. Subsequent investigatier related the failure of the Unit 1 RPS to cause a scram to sticking of the uncervoltage trip attachment in the scram circuit breakers 1

As a result of these ever.ts the NRC Executive Director for Operations 1 directed the staff to undertake three related activities: (1) an j .

evaluation of when and uncer what conditions the Salem plants would be (

allo-e: to restart; (2) a fact finding report of the events at Sa:a 1 and I the circumstances leading to them; and (3) a report on the generic  !

it:14:ations o# these events. i i

To accress (3) above an interoffice, interdisciplinary group was f:rred inclue'n; rembers from the Office of Nuclear Reactor Regulation's I

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e (NRR's) Divisicn af L censing, Division of Systems Integration, Division of human Factors Safety, sivision of Engineering, Division of Safety '

Te:nnelegy, the Office of Insue: fon and Enforcement, the Office for Analysis and Evaluation of Operational Data, and NRC's Region I Office.

Tr.is grou: published NUREG-IDDD as a result of their efforts to resolve the follcaing Ouestions: (1) is there a need for prompt actions to address similar e:vipment in ether facilities; (2) are the NRC and its licensees learning the safety management lessons; and (3) how should the priority and c:ntent of the ATW5 Rule be adjusted.

As a result of the NUREG-1000 findings, the NRC issued Generic Letter E2-2E2 (GL S3-28). The actions described in GL 83-28 address issues relate: to rea: tor -ip system (RTS) reliability. The actions coverec fall i to the following four areas: (I) Post-Trip Review, (2)

E:.i;rert Classifi:ation and Vencer Interface, (3) Post-Maintenance Testing, an: (4) Rea:ter Trip System Reliability Improvements.

Item 4, adeve, is aimed at assuring that vendor-recommenced reactor t*ip breaker modifications and associated reactor protection system changes are Or;1eted in pressuri:ed water reactors (PWRs), that a comprehensive

-am of preventive maintenance and surveillance testing is implemented for the reactor trip breakers in PWRs, that the shunt trip attachment a

• tivates automatically in all PWRs that use circuit breakers in their rea

  tor trip systems, anc to ensure that on-line functional testing of the rea:: r trip system is performed on all light water reactors (LWRs).

The spe:ific re;uirements Of GL 83-28, Item 4.5.3, are that existing intervals for on-line functional testing required by Technical

$;etifications shall be reviewed to determine if the intervals are

ensistent with achieving high RTS availability when accounting for considerations such as: (1) uncertainties in component failure rates; (2) unce-tainties in common mede f ailure rates; (3) reduced redundancy during testing; (?) cre-ate errors during testing; and (5) component " wear-cut" caused by testing.

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. Tne Eab:::k & Wilcox (B&W), Combustion Engineering (CE), General .

Eie:t-i (GE), anc Westinghouse (W) Owners Groups have submitted topical re: Orts either in rescense to GL 83-28, Item 4.5.3,3,4 or to provide a tesis fer reauesting RTS surveillance test interval (STI) e n e n s i c e s . 5 '6,7 ' E ' C~'10'11 In general, the owners groups' analyses were ret c:ne en a plant specific basis. Instead, the analyses addressed a carticular class of reatter tr's system and.then discussed the a:plicatility of the analysis to specific product lines. The NRC reviewed-trese repcrts fer, amon.g cther things, their applicability to GL 63-28, Item 4.5.3 an: summa-1:ed their findings ir, Safety Evaluation

'I Re: Orts (SERs).

1.2 Review Purcose This re;;rt cc: aments a review of the Owners Groups' topical reports, tne NRC SERs, anc ethe- analyses done at the Idaho National Engineering Lateratery (INEL) by pers:nnel in the NRC Risk Analysis Unit of EG&G Idaho, Inc. The INEL cencucted the review at the request of the U.S. Nuclear Regulatery Come ssion, Office of Nuclear Reactor Regulation,

instrumentation and Contrel Systems Branch (ICSB). The review was pe-fermed to cete-mine if the Owners Groups' analyses demonstrated high RTS availability fer the current test intervals, if the analyses included the five areas of concern from GL 63-28, and if all of the plants were covered by tne analyses. The results of the review, if all plants are shown to be

covered by an adecuate analysis, would provide the NRC with a basis for cicsing cut GL S3-28, Item 4.5.3, for all U.S. comeercial n'uclear reactors w4tneut 'urther review.

Tne body of this ree rt presents the review and its findings with regarc to the stated 0:je:tives. .Section 2 describes the criteria used in the review to cetermine the adequacy of the analyses. The review metn:cclogy is cis:vssed in Section 3. Section 4 presents the review results. Tne review ::n:1csions are given in Section 5.

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2. REVIEW CRITERIA To con:Let a review, one must have criteria, or standards, on which a

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; ent or decisions may be based. In this section, the INEL availability analyses review criteria are presented.

GL E3-2B established the three criteria used in the INEL review.

GL E3-25 statec that: (1) all licensees et al., (2) must demonstrate high RTS availability for the current test intervals by documented review when (3) accounting for su:h considerations as the five areas of concern listed ir Section 1.1. Wnile GL 83-28 established all three criteria, it only cefined tw:- of them- who had to do a review and what the review had to take into ace:unt. The third and most subje:tive criterion, "high i

avaiiacility", was not defined.

To esta:lish a cefinition of high availability, the INEL used the ele:trical unavailability base case estimates presented in Table A-1 of A::encix A to SECY-53-293.I# Unavailability is defined as 1.0 minus availability. A low unavailability is equivalent to a high availability.

M:st analyses calculate a system unavailability rather than an availability. Therefore, our criteria for a "high availability" will be ex;ressed in terms of low unavailability for compatibility. These RTS unavailability estimates from Reference 14 were used for two reasons.

First, they were used because they were developed by the NRC's ATWS Task Fo-:e as a reevaluation of the bases for the RTS unavailabilities used in ATW5 rule value-impact evaluations. Second, as stated in Reference 14, this NRC analysis

. bases the RTS unavailabilities on worldwide experience to cate. It is believec that this gives a reasonable estimate of RTS unavailability that includes the common cause contributions that are believed to dominate. The experience based values are dist-ibuted across the four vendor designs based on a com;a ative reliability analysis that evaluates the major ci "erences amon; the designs."

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s Tne estimates from sne NRC ATWS analysis provide a framework with ,

whicn to consicer the to: ital report analyses estimates. The numerical )

estimates in tne SECY-53-293 fer the fcur vendors ccmcined with the five areas c' centern frer GL E3-25, Item 4.5.3, form the criteria used for this review 10 cetermine if the venders' analyses and estimates met the re:;irements of iter 4.5.3.

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3. REVIEW METHODOLOGY P

The INEL conducted this review by examining the vendors' topical e:: ts (References 3, 4, 5, 6, 7, 8, 9, 10, and 11), the technical evaluation reports} ,16,17,18 (TERs) done as a part of the NRC topical re: Ort revie process, the NRC's SERs (References 12 and 13), and NUREG/CR-5197, Evaluation of Generic Issue 115, " Enhancement of Westinghouse Solid State Protection System."IC "

This was done for three

-easons. F1 st, tne re;0rts were examined to find out whether or not the verd:rs' ana;fses addressed the areas of concern from Item 4.5.3 and reflected a high RTS availability. Second, they were examined to determine wnat ;1 arts were ::vered by the venders' analyses. Third, the Generic Issue 1:5 ee:0-t p* viced an independent, updated estimate of the availability O' t he h' s 01i 0 state RTS for comparison to the review criteria.

0 0- the pla,ts covere: Dy the venders' analyses or the NUREG/CR-5197 analy sis, the a::re:riate analysis and availability were compared to the review criteria establisne: in Section 2. If the analysis adequately a: dressed the a-eas of ::ncern and demonstrated a high RTS availability, tre :lart was a::erted as having met the requirements of GL E3-28, Item 4.5.3. Tne results of the com;arisons for plants covered by a vencer analysis are giver ey vencer in Section 4.

Fer plants n:t directly :: vere: by a vendor's analysis, an acceptable means was foun: to extenc tne analyses to cover the plants. This was d:ne fo- tw: plants: Clinton 1 (GE) anc Maine Yankee (CE). The means by which the analyses we"e extende: to ::ve" these two plants are also discussed by vender in Se: tion 4 Cne plant, Fort St. Vrain, a hign temperature, gas-cooled reactor (HIGR), was not covered by any of the four vendors' analyses and required special censide*ation. The INEL examined the responses from Fort St. Vrain l re:vire: by GL E3-28, Item 4.5.3 to determine if the responses demonstrated an a::e;tably nigh RTS availability. The review of the Fort St. Vrain i responses is given in Section 4.6.

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4 REVIEW RESULTS .

Inis section summari:es the results of the INEL review of the vendors' a alyses oitn rega-: to the five area of concern and plant applicability.

Tne vendors' estimates of RTS availability are compared to the review availability crite-ia. Also, some insights concerning RTS availability, gained from an examination of RTS importance measures from selected PRAs, a-e examinec.

4.1 B&W Plants Tne issues of GL E3-22, Item 4.5.3, were acdressed by the B&W Owners G-cup anc the results were submitted to the NRC by the individual utilities ir their responses to GL 83-28. Topical. Report BAW-10167 (Reference 5) was s.cmittec to tne NR to provice a technical basis for increasing the on-line S':s arc allowed outage times (A0Ts) for B&W RTS instrument st-ings. The analysis p esented in BAW-10167 was built upon the previous analysis cone to accress the GL 83-28, Item 4.5.3 issues. However, some infc-mation that was resolved in the generic letter analysis was not repeatec in the subsequent Topical Report because it was not relevant to tre proposec Tecnnical Specification changes. To make BAW-10167 applicable to both GL E3-22, Item 4.5.3 and STI/ACT issues, the Owners Group submitted k EAW-;0:67, Suc lement 1 (Reference 6), to the NRC. Supplement I completed the B&W analysis by acdressing all remaining Item 4.5.3 issues. The EAW -10167 anc Supplement I analyses included the implementation of the autematic snunt trip on the reactor trip circuit breakers as recuired by GL E3-28, Item 4.3.

Ine INEL has previously reviewed the BAW-10167 and Supplement I analyses anc documentec the review in a TER, EGG-REQ-7718 (Reference 15).

For the TER sensitivity stucies which included all of the Item 4.5.3 areas of concern were conducted on the RTS mocels. The sensitivity study results showed the 0cels to be insensitive to variations in the failure rates associatec witn the Item 4.5.3 areas of concern.

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The INEL reviewec BAW-10167, BAW-10167, Supplement 1, and the TER and, _

ceterminec that the B&W analyses ade::uately covered all five areas of concern anc that all currently operating B&W reactors are included.

4.2 CE Plants Licensees with CE reactors responded to the requirements of GL 83-28, l

Item 4.5.3, as the CE Dwners Group by submitting CE NPSD-277 (Reference 3)' I to ne NRC. The NPSD-277 RTS availability analysis specifically included  ;

all five areas of concern and all currently operating CE reactors except Wate-ford 3, which was not in commercial operation until September 1985.

The CE 0.ners Group also submitted CEN-327 (Reference 7) to provide licensees with a basis for requesting RTS STI extensions. This later analysis expanced on the simplified mocels of NPSD-277 to incluce all RTS input carameters. All currently operating CE plants except Maine Yankee were covered in the CEN-327 analysis. The CEN-327 STI analysis specifically included the NPSD-277 analyses of the Item 4.5.3 areas of concern except component " wear-cut" during testing. The CEN-327 analysis showed that the major contributors to RTS unavailability for the four plant classes are common cause failures of the trip circuit breakers which are tested on a monthly basis.

In both NPSD-277 and CEN-327, the CE RPS designs are grouped into four classes by signal processing and trip device differences, otherwise the logic and physical layouts of the RTS are the same for all RTS plant classes. In NPSD-277, Maine Yankee is included in RPS Plant Class 2. In CEN-327, Waterford 3 is incluced in RPS Plant f. lass 3. Between NPSD-277

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arc ~EN-327, all of the CE plants are included in plant classes analyzed in i

CEN-327. This review considers the analysis and results in CEN-327 adecuate for Item 4.5.3 resolution for all classes of CE plants.

The INEL has previously reviewed CEN-327 with regard to STI extension effects anc cocumented the review in a TER, EGG-REQ-7768 (Reference 16).

The results of sensitivity studies cone for the TER show the models to be insensitive to an creer of magritace increase in the com;ocent independent 8 '

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failure rates. The insensitivity to increased component failure rates .

3 alcng with the CE analysis results showing trip circuit breaker common I

ause f ailures to be the major contributor to RTS unavailabi'ity provides a a basis for this rev4e* to conclude that RTS test-induced component wear-cut is not an issue at CE reactors.

The INEL reviewed CEN-327 and the TER and determined that the CE '

aralyses have ade;uately covered all five areas of concern or they have been shown net to contribute to RTS unavailability and that all currently '

cperating CE rea:ters are included.

j 4.3 GE Plants I

Licensees with GE rea:ters responded to the GL 83-28, Item 4.5.3 re:virements is the EWR Owners' Group by submitting NECD-30844 '

(Referer:e 4) to the NRC. The RTS availability analysis specifically incluced the five areas cf concern and covered both generic relay and s lid-state RIS designs which includes all currently operating BWRs. GE stated that the relay RPS configurations for BWR plants have the same

rimary cesign features. Therefore, the generic relay RTS models used.in NECC-3CS44 de not ciffer significantly from the specific BWR plants. GE use

the Clinten 1 crawings fer the solid-state RTS models. Since Clinton 2 is currently the only GE plant with a solid state RTS, no plant unique analysis is necessary.

The EWR Dwners' Group also submitted NECD-30851P (Reference 8) to the NRC. The analysis in this second report used the base case results from NECD-30844 te establish a basis for requesting revisions to the current Te:hnical Specifications fer the RTS. The INEL had previously reviewed l NFCD-30544 and NECD-30351P with regard to botn Item 4.5.3 and STI extension acce:tability and cetumented the-review in a TER, EGG-EA-7105 (Reference 17). Due to insuf ficient information, the .INEL review could net ce S ete the sciid-state RTS review and accepted only the relay RU analysis results.. The NRC reviewed the topical reports and the TER and i

9

issuec an SER (Reference 12). The NRC accepted the analysis results as a .

reference fcr TS changes related to the RTS and as resolution to GL 83-28, 1:er a.5.3, for GE re'.ay plants only. The INEL later completed the solid state RTS analysis review and issuec Rev 1 to the TER (Reference 16), thus a::ertir; tne analyses for all classes of GE plants.

Tnis revie examined both GE analyses and the Rev I TER and deternined that all five a-eas cf :entern are included in the analyses and that all cc rently c;erating GE rea:ters are included.

4.4 Westinghouse Plants Ld:ensees with Westingn0use reactors did not respond directly to the ee:uirements of GL 83-28, Item 4.5.3. Prior to the Salem ATWS, they had su:-itte: W:AP-10271 (Reference 9) to the NR; to provide a basis for re:uesting :nanges tc the Technical Specifications regarding the RTS. The Westingneuse mein:cclegy attempted to balance safety and operability and was appliec to a typical Westinghouse four loop reactor plant with a solid state RTS in WCAP-10271. The methodology was extended to cover RTSs for two, three, and four loop plants with either relay or solid state logic in WCAD-10271, Supolement 1 (Reference 10).

Tne NRC reviewec the Westinghouse topical reports with the assistance of Erockhaven National Laboratory (BNL) and issued an SER (Reference 13) limiting "Eir acceptance to changes to only the anclog channel ST!s at West'ngn:use plants.

Tne W methodology used fault trees to model the RTS. The models incluced the following five major contributors to RTS trip unavailability:

1. Unavailability of components due to random failures

2. Unavailability of components due to test j

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. 3. U-availability cf components due to unscheduled maintenance l

4. Unavailability of components due to human error )

5. Ur.a.ailability of components due to common cause failure.  ;

While the y analysis dic not directly include any sensitivity studies l ccnce-ning these five areas, the compenent unavailabilities were increased as :ne test interval length increasec. The STI analysis results showed a fact:r of 3 to 5 increase in the RTS unavailability estimates for the icnger test interval. Two conservatism exist in the models that are .

5 l relevant: f i s ; , r.0 credit was taken for early failures that would be i cetectec anc, secenc, no credit was taken for the diversity inherent in the p RTS cesign. These two conservatism, had they been included in the mcce', wcuic cause the increase in the RTS unavailability estimates to be 5 a'*er inan :ne ccse vec fac:crs. 1 1 i l

l Test-inducec component wear-cut was not addressed in any manner in the j l

3 RTS analysis. however, the RTS analyses done by the other vendors, I 1

Referer.ces 3, a and 6, specifically investigated the effects of this issue l cr RTS unavailability. Despite the differences among the other vendors' RTS cesige.s, they all founc the effects of_ test induced component wear-out on RTS unavailability to be insignificant. Based on the other vendors' ,

analyses, the INEL concluced that the effects of test-induced' component wear-cut on y RTS unavailability would also be insignificant. Therefore, the INEL consicers all'E plants to be coverec by adequate analyses.

a.S Caa* itative Review of Venders' RTS Availabilities Sc far, only tne adecuacy of the vendors' analyses has been cisc ssed. No determination has been made of the' acceptability of the nu erical estimates from the various RTS availability analyses. In'this ucti:n, the INEL review censiders the four Owners Groups' RTS availability estimates tc cetermire if iney are inceed indicative of "high availability."

In Table 1, the four vendors' RTS unavailability estimates are -

compared to the review estimates of low unavailability as defined in Secticn 2. The B&W and GE vendors' estimates are given as an overall RTS unavailability per cemand by plant model and RTS type, respectively. The CE and W vent:*s' estimates are given on a similar basis with an additional consideration tnat was not necessary for the B&W and GE analyses. In the CE and W analyses, RTS unavailability was estimated for all input parameters.

For the CE and W unavailability estimates in Table 1, the INEL used the unavailability estimates for high pressurizer pressure, the parameter analy:ec in Reference 19 as the limiting parameter for an ATWS in terms of the number of input channels and diversity of trip signal.

Tne differences in the relative values of the three PWR vendors' RTS unavailability estimates can be attributed to design differences among the RTSs. E&W anc CE RT55 have four analog channel inputs fcr each monitored parameter with four -ip logic channels while W RTSs have three or four analog cnarnel inputs for each paraceter with only two trip logic channels. The 2 of 4 analog channels for the B&W and CE RTS designs are inherently mere reliable than the 2 of 3 analog channels for some parameters in the W design. Also the 2 of 4 trip logic in the B&W and CE RTSs is mere reliable than the W 1 of 2 trip logic. The combination of these two design differences make the W RTS unreliability somewhat higher than the other venders' RTS unavailabilities.

The comparisen shows the B&W, CE, and GE RTS unavailability estimates are lower than the NRC's estimates while the W estimates are the same as the NRC's. The INEL review recognizes the Vendors' estimates and the NRC's estimates are influenced by a number of factors. These factors include, (1) the data uncertainties for both the NRC and Vendors analyses, (2) the scarcity of actual RTS failures world wide, (3) the modeling assumptions and sim;11fications used by both the NRC and the Vendors, and (4) the differing levels of model development between the NRC analysis and the Venders' analyses and between different Venders' analyses. These factors 12

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TAELE 1. COMPARISON OF VENDOR AND NRC RTS UNAVAILABILITY ESTIMATES Vender RTS NRC RTS b

Unavailability Estimates- Unavailability Estimates Verder ( Fa il ure s/Dema nd) (Failures / Demand)

B&W c d Davis Bessie Pocel IE-10 3E-5 C d 0:enee Class Mocel IE-6 3E-5 CE 8

Flant Class 1 2E-7 2E-5 Plant Class 2 3E-6' 2E-5 Plant Class 3 3E-6' 2E-5 Flar: Class 4 2E-6' 2E-5 GE i Relay Plants 3E-6 2E-5 f

Solid-state Plants 3E-6 2E-5 W

Relay Plants d SE-59 SE-5 d

Solid-state Plants SE-59 5E-5 i

a. All estimates are rounded off to one significant digit.

b. From Reference 14, Table A-1, base case RTS electrical unavailability estimates,

c. From Reference 5, base case.

d. Includes automatic shunt trip on the reactor trip circuit breakers.

e. From Refererce 7, Tables 4.1-1, 4.2-2, 4.1-3, and 4.1-4, respectively; base case test interval, high pressurizer pressure unavailability estimate.

f. From Reference 4

g. From Reference 19, solid state RTS base case. Applied to relay plants basec on similarity cf cesign (see Reference 11, Section 3.2.2 and 3.2.3).  !

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9 help explain the differences between the Vendors' and the NRC's point , ,

estimates of RTS availability.

4.6 Fort St. Vrain Fcrt St. Vrain responded to GL S3-28, Item 4.5.3 in a letter to Eisenhut cated November 4, 1983 20 , stating:

" Existing intervals for on-line functional testing required by the Technical Specifications are curt 3ntly under review by Public Service Company of Colorado (PSC) and the Nuclear Regulatory Commission Region IV staff. The current testine frecueacy at Fort St. Vrain has been dictated by the hu leae Re;ulat: y Commission staff." (Uncerline accec)

In resconse to a re:uest for information from the NRC concerning the Fert St. Vrain resconses to GL 63-28 previously sent, PSC sent the

1. licair; *e:ly to the NRC in a letter to Johnson, dated June 12, 198521;

" Existing intervals for the on-line testing required by the

'e nrical Spe:ifications were reviewed by Public Service Company of Colorado. A Technical Specification change to Limiting Conditions for Operation 4.4.1 (Plant Protective System) and its associated surveillance requirements (SR 5.4.1) are currently being reviewec by tne Plant Operations Review Committee (PORC).

This Technical Specifi:ation change is expected to be approved by the PORC and the Nuclear Facility Safety Committee (NSFC) by June 30, 1985.. As part of the development process for these proposed changes to the Technical Specifications, on-line functional testing requirements were reviewed based on past experience.

Possiele changes to the testing intervals in certain cases where available test data may support such changes has (sic) been discussed at length with the Nuclear Regulatory Commission staff. The Nuclear Regulatory Commission staff has informed Public Service Company of Colorado that no such changes would be at:eptable at this time."

The INEL review interpreted these responses from Fort St. Vrain to m e a a.

the NR; has established Fort St. Vrain's RTS current test intervals, the current test intervals have been evaluated by PSC, and the NRC will not allow enanges to the test intervals at this time.

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I i Free these resp;nses, the INEL concluded that Fort St. Vrain has i

.d.;;ted ne review recuired by GL S3-28, Item 4.5.3, and that the NRC

rsicers the PSC and NRC reviews adequate to meet the Item 4.5.3 ce:vireme'ts.

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5 REVIEW CONCLUSIONS .

All fcur LWR verdo s have submitted topical reports either in response t: GL E2-25, Item 4.5.3, or to provide a basis for RTS STI extensions, or teth. Fcr tne most part, these reports have addressed all cf :,.e issues in Item 4.5.3. Litersees not covered by the topical reports have submitted individual responses to Item 4.5.3.

The analyses in the topical report have shown the currently configured RT55 to be highly reliable with the current test intervals and prior to ie:lementing some of the re uirements of GL 83-28. Implementation of these a::itional requirements will reduce the ATWS risk even further.

The hEL has reviewed the relevant topical reports, TERs, SERs, a::'de a' analyses, and the individual licensee submittals with regard to GL E3-25, Iter 4.5.3, requirements and the review criteria. Based on that eview, the NEL cen:1uces that all licensees of currently operating ccmmercial nuclea power plants have adequately demonstrated that their current RTS test intervals are consistent with achieving high RTS availability.

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, 6. REFERENCES

)

.l U.S. Nuclea* Regulatory Commission, Generic leolications of ATWS Events l l at tne Salen Nuclear Power Plant, NUREG-1000, April 1983.

2. U.S. Nuciear Regulatory Commission Letter, D. G. Eisenhut to All Licensees et al., Recuired Actions Based on Generic Implications of Salem ATWS Events, Generic Letter 83-28, July 8,1983.

3. Comoustien Engineering, Reactor Protection System Test Interval Evaluatier, T.ask 456, CE NPSD-277, Decemoer 1984 4 S.'Visweswaran et al., BWR Owners' Group Response to NRC Generic Letter 83-28, Item 4.5.3, NECD-30844, January 1985.

l

5. R. S. Enzinna et al. , Justification for Increasing the Reactor Trio  !

Syste- Cr-line Test Inte-val, BAW-10167, May 1986.

6. R. S. Enzinna et al., Justification for Increasing the Reactor Trio q Syste- 0--line Test Interval, Succlement Number 1, BAW-10167, l Suppiement Numoer 1, Feoruary 1968. i

7. Cc-oustion Engt eering, RDS/ESFAS Extended Test Interval Evaluation, CEN-327, May 1956.

i S. W. P, Sullivan et al . , Technical Specification Improvement Analyses for l SWE Rea:ter *ctection System, NECD-30851P, May 1985.

9. R. L. Jansen et al. Evaluation of Surveillance Frequencies and Out of ,

Seevice Times fer the Rea:ter Protection Instrumentation System,  !

WCA;-10271, Janua y 1983.

10. R. L. Jansen et al . , Evaluation of Surveillance Frequencies anc Out of Se vice Times for the Rea: tor Protection Instrumentation System, Su::le ent 1, WCAP-10271, Supplement 1, July 1983.

11. R. L. Jansen et al. , Evaluation ef Surveillance Frequencies and Out of Servi:e Times for the Rea: tor Protection Instrumentation System, Su: clement 1-0-A, WCAP-10271, Supplement 1-P-A, May 1986.

12. U.S. Nuclear Regulatory Commission Memorandum, G. C. Lainas to E. J.

But:her, Ac:ectance for Refe-encine of General Electric Company (GE)

Tecical Repe-ts NECO-30844, "EWR Owners' Group Response to NRC Generic Letter 83-28," anc NECD-30251P, " Technical Specification Improvement Ana'yses for SWR Reactor Pretection System,"' April 28, 1986.

13. U.S. Nuclear Regulatory Commission Letter, C. O Thomas to J. J.

Sheopaed, Acce:tance for Referentine of Licensing Toeical Recort WCAD-10271. " Evaluation of Surveillance Frequencies anc Out of Service Times for tee Reactor Protection Instrumentatier Systems," February 21, 1955.

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U.S. ' Nuclear Regulatory ' Commission, Amendments to- 10 CFR 50 Related to Anticipated 19, 1953. Transients Without Scram (ATWS) Events, SECY-83-293, July 15.

J. P. Poloski' and S. D. Matthews, Review of B&W Owner's Grouc Analyses for Ircreasing The Reactor Trip System On-line Test Interval, EGG-REQ-7728, Septemoer 1988.

16. D. P. Mackowiak and B. L. Collins, A Review of the Combustion:

Engineering Evaluation For Extending the RPS and ESFAS Test Intervals, EGG-REQ-7768, Septemoer 1988.

17. R. E. Wrignt and B. L. Collins, A Review of the BWR Owners' Group Technical Specification Improvement Analyses for the BwR Reactor Protection System, EGG-EA-7105, January 1986.

15. :R. E. Wright and B. L. Collins, A Review of the BWR Owners' Grouc Technical 5: edification 1mocovement-Analyses for the BWR Reactee Se:te:: en System, EGG-EA-7105, Rev 1, March 1987.

19. D. A. Reny et al., Evaluation of Generic Issue 115, Enhancement of the Reliability c' Westinghouse Solic State Protection Systems, huREG/CR-5197,-January 1959.

20. Public Service Company of Colorado Letter, O. R. Lee to D. G.

Eisennut, Reseense to Generic Letter 83-28, November 4,1983.

21. Public Service Com:any of Colorado Letter, J. W. Gham to'E. H.

Johnson, Reseense to Generic Letter A3-28, June 12, 1985.

)

I 18 4

. *s c . . ui a a .ca*a aam'o*' coaaas oa *"c"'**a"'*******"~*** ~

f,*. ,,

" .y BIBUOGRAPHIC DATA SHEET EGG-NTA-8341 in i ..e e e . ...on TEC C CAL EVALUATION REPORT: A REVIEW OF REACTOR TRIP SYSTEM AVAILABILITY ANALYSES FOR GENERIC LETTER 83-28, ITEM 4.5.2, RESOLUTION ~ ;* "* " " '* " "

Mareh 1989 David P. Mackowiak * :* ' ": " " "

John A. Schroecer -o '- "-

l March 1989

. . . u . . . u . a . - 4. . ; . . . . . c . . . c . u u. -- . .a , c . ..ac.u- ~ ac.. -a-Reculatory and Technical Assistance n.

9 G. 6.. *s. st.

P. O. Box 1625 Idaho Falls, ID 83415 D6001

.c  :.~.:.c... w .o .... ;.... c..u. u .- -<.c- ".**c""c" Instrumentation anc Cor. trol Systems Branch Technical Evaluation Report Division of Engineering and System Technology Office of Nuclear Reactor Regulation * * " ' ' * " ' " " ~ ~ ~

U.S. Nuclear Regulatory Commission Was*in; ten. DC 2SEEE 2..........:,n

. u...a a- .

The Idaho National Engineering Laboratory (INEL) conducted a technical review of the co :mercial nuclear reactor licensees' responses to the requirements of the Nuclear Regulatory Commission's (NRC's) Generic Letter 83-28 (GL 83-28), Item 4.5.3. The results of this review, if all plants are shown to be covered by an adequate analysis, will provide the NRC staff with a basis to close out this issue with no further review.

Tne licensees, as the four vendors' Owners' Groups, submitted analyses to the NRC either directly in response to GL 83-28, Item 4.5.3 or to provide a basis for requesting changes to the Technical Specifications (T5s) that would extend the Reactor Protection System (RPS) surveillance test intervals (STIs). To conduct the review, the INEL defined three criteria to determine the adequacy, the plant applicability, and the acceptability of the results. Tne INEL examined the Owners Groups' reports to determine if the analyses and results met the established criteria. Fort St. Vrain's responses to Item 4.5.3 were also reviewed. The INEL review results show that all licensees of currently opera-ting commercial nuclear reactors have adequately demonstrated that their current on-line RPS test intervals meet the requirements of GL 83-28, Item 4.5.3.

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