ML20043A663

Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 4.5.3, Reactor Trip Sys Reliability,On-Line Testing of Reactor Trip Sys

ML20043A663
Person / Time
Site:	Quad Cities
Issue date:	05/17/1990
From:	Office of Nuclear Reactor Regulation
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ML20043A659	List: ML20043A658 ML20043A663
References
GL-83-28, NUDOCS 9005220374
Download: ML20043A663 (2)

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SAFETY EVALVATION REPORT l

GENERIC LETTER ~13-20, ITEM 4.5.3 REACTOR TRIP SYSTER RELIABILITY FOR ALL DURESTIC OPEFATING REACTORS i

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1.0 INTRODUCTION

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

L 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 to this incident, on February 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, en automatic trip signel was generated based on steam generator low-low level during plant startup.

In this case, the reactor was tripped I

manually by the operator almost coincidentally with the automatic trip.

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following these incidents, on February 28, 1983, the NRC Executive Director forOperations(EDO),directedthestafftoinvestigateandreportonthe generic implications of these occurrences at Unit 1 of the Salem Nuclear i

Power Plant. The results of the staff's inquiry into the generic implications i

of the Salem Unit 1 incidents 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 (HRC) 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 ATWS events.

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The licensees were required by Generic Letter 83-28, item 4.5.3 to confirm that on-line functional testing of the reactor trip system (RTS), including independent testing of the diverse trip features, was being performed at all plants.

Existing intervals for on-line functional testing required 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 ratest (2) uncertainties in common mode failure rates; (3) reduced redundancy during testing; (4) operator error during testing; and.(5) component " wear 4ut" caused by the testing.

2.0 DISCUSSION t

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 9005220374 900517 PDR ADOCN 03000254 P

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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 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.

The Babcock & Wilcox (B&W), Combustion Engineering (CE), General Electric (GE), and Westinghouse (W) Owners Groups have submitted topical reports either in response to GL 83-28, Jtem 4.5.3 or to provide a basis for requesting 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 testing of the RTS, with the considerations 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 unavailability was dominated y common cause failure and human error.

The ability to accurately estimate unavailability for very reliable systems was considered extensively in NUREG-0460,lemaking.

" Anticipated Transients Without Scram fr Light Water Reactors", and the ATWS ru 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 useful 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 low RTS unavailability.

First, testing more frequently than weekly is generally impractical, and 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 of a common mode failure through increased stress on the system.

Finally, not all potential failures are detectable by testing, 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 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 frequent testing than monthly will not appreciably lower the estimates of failure probability.

Based 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 that all operating reactors.

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4 EGG.NTA.8341 e

March 1989 l

{e TECHNICAL EVALUAil0N MEPORT 1

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/daho Nat/onal A Review cr REAcioR TRIP $YSTEM AVAILABILITY l

Engineering ANALYSES FOR GENERIC LETTER 83 28. ITEM 4.5.3 RESOLUT10N Laboratory l

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Prepared for the U S. NUCLEAR REGULATORY COMMISSiOtv 4 on wro-e <w Oct Cytset

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NOTICE i

TMs report was prepared as an account of work sponsored Dy an agency of the Usuise State Govemment. Neither the Uruted Sete Govemment not any asetwy thereof, nor any of ther employees. maan any warteaty, empressed i

of imphed, or assumes any legal habihty or responmMay for any third pasty's use of the resulu of such use, of any informauen. apparatus, product or proc..

est dacletee it. Afus report. or represenu that its une Dy such third party would not ininnec prnately owned nghts.

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TECHNICAL EVALUATION REPORT: A REVIEW 0F REACTOR TRIP SYSTEM AVAILABILITY ANALYSES FOR GENERIC LETTER 83-28, ITEM 4.5.3, RESOLUTION l

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FIN D6001: Eval ation of Confor an:e to Generic Letter 83-29 for CRs (Project 2)

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t ASSTRACT The Idaho National Engineering Laboratory (INEL) conducted a L

technical review of the commercial nuclear reactor licensees' responses to the recuirements of tne 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 te covered by an adecuate analysis, will crevice the NRC staff with a basis to close out this' issue with no further review.

The licensees, as the four vencers' Owners'_ Groups, submittee analyses to the NRC either cirectly in response to GL 83-28,

tem 4.5.3, or to provide a casis for requesting changes to.the Technical Specifications (TS) that would extend the Reactor Protection System (RPS) survei'1ance test intervals (ST!s).

To conduct the review, the INEL ce'ine: three crite da to dete*mine the Adecuacy, plant applicability, and acceptacility of the results, The INEL examined the Owners Groues' recorts to cetermine if the analyses and results met the established criteria.

Fort St. Vrain's responses to Item 4.5.3 were also reviewed.

Tre !NEL review results show that all licensees of currently ecerating

om t cial nsclear *eactors have acecuately demonstratec that their current cr-line RPS test intervals meet the requirements of GL 83-28,

tem 4.5.3.

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SUMMARY

The two anticipated tran.sient without scram (ATVS) events at the Salem Nuclear Power Plant in February of 1983, focused the attention of the Nuclear Regulatory Commission (NRC) on the generic implications of ATWS events.

The NRC then published Generic Letter 83-28 (GL 83-28) which listed the actions the NRC requir6d of all licensees holding operating licenses and others with resp 9ct to assuring the reliability of the Reactor Protection System (RPS). GL 83-28 Item 4.5.3, required licensees to demonstrate by review that the current on-line functional testing intervals are consistent with achieving high reactor trip system (RTS) availability. The licensees responded to the GL 83-28 Item 4.5.3, recuirements as C-ners. Groups with reports either in direct response to i

Item A.5.3, er with a technical basis for requesting extensions to the I

sarveillar.ce test intervals (STIs) that generally included the Item 4.5.3 recuired reviews.

The NRC's Instrumentation and Control Systems Branch (ICSB), Cffice of Nuclear Reactor Regulation (NRR), requested the Idaho National Enginee ing Laterattfy (INEL) to review the licensee availability analyses anc evaluate the overall adequacy of the existing test.

I intervals.

INEL *eview results showing general compliance with Item 4.5.3 wili :rovide the NRC with a basis to close out Item 4.5.3 without furtre- *eview.

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For tne 'eview, the INEL defined three acceptance criteria', reviewed tre itcensees tocical reports, centractor review reports,'and NRC safety evaluations, and determined the adequacy of the analyses and the RTS.

availability estimates with regard to the review criter'ia.

j The INEL review criteria to determine the licensees' Item 4.5.3 c:m:11ance were, (1) the five areas of concern of Item 4.5.3, (2) the l

aa.alyses' olant applicability, and (3) the NRC's RTS electrical unavailatility base case estimates from the ATVS Rulemaking Paper, S E C'f-E 3-3 93.

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Each Owners Grokos' reports core revieced to ensure that all five areas of concern from Item 4.5.3 were either included in the analyses or snewn not to be significant withLregard to RTS availability, The INEL review also ensured that the' individual plants' differences from the analysis' models were taken into account and their ef fects we*e shcwn not to significantly affect RTS unavailability. The Fort St. Vrain responses q

to Item 4.5.3 were also reviewed.

The C ners Groups' RTS unavailability estimates wera compared to the NRC's ATWS Rulemaking generic RT5 unavailability estimates to determine the acceptability of the Owners Groups' conclusions that high RTS availability was demonstrated in the analyses.

The results of the INEL review showed that all licensees of currently operating commercial nuclear reactors have adequately

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comonstrated that their current on-line surveillance test intervals are.

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l ATWS Anticipated Transient Without Scram

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B&W Babcock & Wilcox BNL Brookhaven National Laboratory f

CE Comoustion Engineering l

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GE General Electric l

Hi3R' High-Temstrature Gas-Cooled Reactor ICSB Instrumentation and Control Systems Branch

!NEL Icaho National Engineering Laboratory LhR

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NFSC Nuclear Facility Safety Committee NRC Nuclear Regulatory Commission NRR Cffice of Nuclear Reactor Regulation PORC Plant Operations Review Committee PSC D blic Service Company of Coloraco u

l rwR Pressuri:ee Water Reac er RSSMAP Reactor Safety Stucy Metnoc:!cgy Applications Program RPS Reactor Pretection System N

RTS Reactor Trip System SER Safety Evaluation Re:c-t STI Surveillance Test Interval TER Technical Evaluation Report L

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o CONTENTS i

ABSTRACT.............................................................

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$UMMARY...............................................................

iii ACRONYMS.............................................................

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INTRCOUCT:CN.....................................................-

1 1.1 Historical Background......................................

1 1.2 Review Purpose.............................................

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REVIEW CRIIERIA...................................................

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REV:EW METHOCOLOGY...............................................

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REVIEW RESULIS...................................................

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4.1 S&W Plants.................................................

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4.2 CE Plants..................................................

7 4.3 GE Plants...................................................

9 4.4 WtStinghoWse Plants.......................................,

10 4.5 Cwantitative Review of Venders' RTS Unavailabilities......

11 4.6 cort St. Vrain.............................................

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REVIEW CONCLUS!ONS..............,.......................,........

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TABLES 1,

Comparison of Vendor and NRC RTS Unavailability Estimates......................................................

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9ECHNICAL EVALUA?f0N R[p0RT: A REVIEW OF REACTOR TRIP $YSTEM i

AVAILABILITY ANALYSES FOR GENERIC LETTER 83-28.

ITEM A.5.3 RESOLUTION I

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INTR 00VCi!ON' t

1.1 Historical Backcreund i-In Feorwary of 1983, two events occurred at the $alem Nuclear i

Generating $tation that focuse: Nuclear Regulatory Commission (NRC)

I attention on the generic implications of anticipated transient without i

scram ( ATW$) events, l

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First, on February 22, during startup of Unit 1 an automatic trip t

signal generated as a result of a steam generator low-low-level iailed to l

aLse a rea::or s: ram, The reactor was tripped manually by an operator almost coin:icentally witn the automatic trip signal ~, so the fact that the i

aut:mati trip had failed to cause a scram went unnoticed.

Three days later on February 25, both of the scram breakers at Unit 1 failed to cpen on an automatic r& actor protection system (RPS) scram sig*al, TFe operators took action to control-this se:end ATW$ and su::eecec in te*minating the inc! dent in about 30 seconds. Subsecuent investigation related the failure of: the Unit 1 RPS to cause a scram to sti: Ling of the uncervoltage trip attachment in the scram circuit breakers, As a result of these events the NRC Executive Director for Operations.

directec the staff to undertake three related activities: (1) an evaluation of when and uncer what conditions the $alem plants would be al':.ec to restart; (2) a fact fincing report of the events at $alem 1 anc:

the circumstances leacing to them; anc (3)~a report on the generic i :14:ations o' these events.

To ac:ress (3) ab:ve an interoffice, interdisciplinary group was

' r ec 1*.:'w:'n; e.oers f*:m the Office of Nuclear Reactor Regulation's

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CAR's) Division of Licensing, Division of syste2s Integration, Division'of Nmv hetors_ $4fety, Division of Engineering, Division of Safety

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Technology, the Office of Inspection and Enforcement, the Office for Analysis and Evaluation of Operational Data, and NRC's Region I office.

This group published NUREG 10001 as a result of their efforts to resolve the follcwing cuestions:

(1) is there a need for prompt actions' to accress similar ecuipment in other facilities; (2) are the NRC and its licensees learning the safety management lessons; and (3) how should the priority and content of the ATWS Rule be adjusted.

As a result of the NUREG-1000 findings, the NRC issued Generic Letter 83-282 (GL 83-28). The actions described in GL 83 28 address issues related to reactor trip system (RTS) reliacility. The actions coverec fall into the following four areas: (1) Post-Trip Review. (2)

E:uipment Classification and Vencer !nterface, (3) Post-Maintenance Testing, anc (4) Reactor Trip System Reliability Improvements.

Item 4, above, is aimed at assuring that vendor-recommenced reactor i

t*ip treaker modifications and associated reactor protection system changes I

are c:mpleted in pressurized water reactors (PWRs), that a comprehensive program of preventive maintenance and surveillance testing is implemented for the reactor trip breakers in PWRs that the shunt trip attachment activates automatically in all PWRs th6t use circuit breakers in their reactor trip systems, anc to ensure that on-line' functional testing of the reactor trip system is performee on all light water reactors (LWRs).

4 The specific requirement.s of GL 83-28 Item 4.5.3, are that existing intervals for on-line functional testing recuired by Technical Specifications shall be reviewed to cetermine if the intervals are consistent with achieving high RTS availatility when accounting for consicerations such as:

(1) uncertainties in component failure rates; (2) 3 uncertainties in common mode failure rates; (3) reduced reduncancy during testi ;; (4) ccerator treers curing testing; and ($) component " wear-cut" cause0 by t.esting.

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_m7 The Babcock & Wilcox (B&W), Combustion Engineering (CE), General Electric (GE), and Westinghouse (W) Owners Groups have submitted topical i

reports either in response to GL 83-28, Item 4.5.3'3 or to provide a basis for requesting RTS surveillance test interval (STI)

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extensions.5,6,7,8,9,10.11 in general, the owners groups' analyses were not cone on a plant specific basis.

Instead, the analyses add,ressed a particular class of reactor trip system and then discussed the applicability of the analysis to specific product lines. The NRC reviewed these reports for, among other things, their applicability to GL 63-28,

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Item 4.5.3 and summari:ed their fincings in Safety Evaluation ReportsI' (SERs).

i 1.2 Review Durcose This report cocuments a review of the Owners Groups' topical reports, the NRC SEDs, and other analyses done at the Idaho National Enginee-ing Lateratory (INEL) by personnel in the NRC Risk Analysis Unit of EG&G Icaho, Inc.

The INEL concucted the review at the request of the U.S. Nuclear l

Regulatory Commission, Office of Nuclear Reactor Regulation, Instrumentation and Control Systems Branch (IC$B).

The review was performed to cetermine if the Owners Groups' analyses demonstrated high RTS i

availacility for the current test intervals, if the analyses included the five areas of concern from GL 83-28, and if all of the plants were coverec by the analyses, The.results of the review, if all p1' ants are shown to be coverec ey an acecuate analysis, would provide the NRC~with a basis for

-i clos 9; out GL S3-28, Item 4.5.3, for all U.S. commercial nuclear reactors witneut furtner review.

ine body of this report presents the. review and its findings with

• egarc t,o the stated oojectives.

Section 2 describes the criteria usec in 1

the review to cetermine the aceouacy of the analyses. The review retmocology is ciscussed in Section 3.

Section 4 presents the review results.

The review conclusions are given in Section 5.

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REVIEW CRITERIA To conduct a review, one must have criteria, or standards, on which a jud; tent or decisions may be' based.

In this section, the INEL availability I

analyses review criteria are presented.

GL 63-2B established the three criteria used in the INEL review.

GL B3-23 stated that: (1) all licensees et al., (2) must cemenstrate high RTS avai'aoitity for the current test intervals by documented review when (3) acc:unting for such considerations as the five areas of concern listed in Section 1.1.

While GL 83-28 established all three criteria, it only defined two of them-who had to do a review and what the review had to take into account.

The tnird and most subjective criterion, "high ava4', ability", was not defined.

To establish a definition of high availability, the INEL usec sne electrical unavailability base ' case estimates presented in Table A-1 of Accendix A to 5ECY-83-293.I' Unavailability is defined as 1.0 minus availability. A low unavailability is equivalent to a high availability.

Most analyses calculate a system unavailability rather than an availability.

Therefore, our criteria for a "high availability" will be I

i exo'essed in terms of icw 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'ce as a reevaluation of the bases for the RTS unava11 abilities used in ATs$ rule value-impact evaluations.

Second, as stated in Reference-14, v

inis NRC analysis

. bases the RTS unav411 abilities on worldwide experience to i

cate.

It is believed 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 cistributed across the four vencor designs based on a comoarative reliability analysis that evaluates the major cif#e'ences among the designs."

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The estimaSes from the NRC ATUS analysis provide a framework with which to consider the topical report analyses estimates. The numerical estimates in the SECY-83-293 for the four vendors combined with the five areas'of concern from GL 83-28.. Item a.S.3, form the criteria used for this review to determine if the vencers' analyses'and estir.ates met the recuiretents of Item 4.6.3.

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REVIEW METHODOLOGY i

f The INEL conducted this review by examining the vendors' topical f

reports (References 3, 4, 5, 6, 7, 8, 9. 10 and 11), the technica) evaluation reports 15,16,17,18 (TERs) done as a part of the NRC topical report review process, the NRC's SERs (References 12 and 13), and i

NUREG/CR-5197, Evaluation of Generic Issue 115. " Enhancement of Vestinghouse Solid State Protection System."

This was done for three First, t'he reports were examined to find out whether.or not the easons.

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vencers' ana'yses addressed the areas of concern from Item 4.5.3 and reflected a high Ri$ availability.

Second, they were examined to determine what plants were covered by the venders' analyses. Third, the Generic 1

Issue 115 report provided an ince:endent, upcated estimate of the l

availability of the W solic state RTS for comparison to the review criteria.

• c tne plants covered by the vencers' analyses or the NUREG/CR-$197

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l analysis, the a:propriate analysis and availability were compared to the review criteria established in Section 2.

If the analysis acequately addressed the areas of concern and demonstrated a hig5 RTS availability, the plant was accepted as having met the requirements of GL 83-28, i

Item a.5.3.

The results of the comparisons for plants covered by a vencor analysis are given by vencer in Section 4 For plants not directly coverec by a vencor's analysis, an acce: table means was founc to extene the analyses to cover the p1snts.

This was cone fer two piants: Clinton 1 (GE) anc Maine Yankee (CE). The means by which the analyses were extended to cover these two plants are also discussed by vencer in Section 4 i

One plant, Fort St. Vrain, a high temperature, gas-cooled reactor (HTGR), was not covered by any of the four vendors' analyses and required-s:ecial consideration.

The INEL examined the responses from Fert St. Vrain i

recuired'by GL 83-28, Item 4.5.3 to determine if the responses demonstratec i

en acce:tably righ RTS availaDiiity.

The review of'the Fort St. Vrain i

responses

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REVIEW RESULTS i

This section summarizes the results of the INEL review of the vendors'-

l' analyses with. regard to the five areas of concern and plant applicability.

The vendors' estimates of RTS availability are compared to the review availability criteria. Also, some. insights concerning RTS availability, gained from an examination of RTS importance measures from selected PRAs, a'e examined.

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L 4.1 B&W Diants The issues of GL S3-28, Item 4.5.3, were aedressed by the B&W Owners

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Group and the results were submitted to the NRC by the individual utilities f

in their responses to GL 83-28. Topical Report BAW-10167 (Reference 5) was l

suomitted to the NRC. to provice a technical basis for increasing the j

on-lire STIs and allowed outage times (A0Ts) for B&W RTS. instrument strings. The analysis presented in BAW-10167 was built upon the previous analysis done to acdress the GL 83-28 Item 4.5.3 issues. However, some information that was resolved in the generic-letter analysis was not t

repeated in the subsecuent Topical Report because it was not relevant to ne proposec Technical Specification changes. To make BAW-10167 applicaele to both GL 83-28, Item 4.5.3 and STI/A07 issues, the Owners Group submitted I

l EAW-10.57, Supplement 1 (Reference 6), to the NRC.

Supplement I completed t

tre B&w analysis my acdressing all remaining Item 4.5.3 issues. The BAW -10167 and Supplement I analyses included-the implementation of the s

aut:matic s unt trip on the reactor trip circuit. breakers as required by GL a

33-28. Item 4.3.

The INEL has previously reviewec the BAW-10167 and Supolement 1

. a a'yses and documentee the review in a TER, EGG-REQ-7718 (Reference 15).

r t e TER, sensitivity stucies which ir:luded all of the Item 4.5.3 areas of ::r:ern were conducted on the RTS mocels.

The sensitivity study results 5.ee the moeels so be insensitive to variations in tne fatiure rates i

ass::'atec witn the Item 4.5.3 areas of concern.

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l The INEL reviewed BAW-10167, BAW-10167, Supplement 1, and the TER and determined that the MW analyses adequately covered all'five areas of concern and that all currently operating MW reactors are included.

I 4.2 CE plants I.

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Licensees with CE reactors responded to the recuirements of GL 83-28, Item 4.5.3, as the CE Owners Group by submitting CE NPSD-277:(Reference 3) to the NRC. The NPSD-277 RTS availability analysis specifically'incluced all five areas of concern and all currently operating CE-reactors except Waterford 3, which was not in commercial operation until Septemcer 1985, i

l The CE Owners Group also submitted CEN-327 (Reference 7)-to provide licensees with a basis for requesting RTS ST! extensions. This later l

analysis expanded on the simplified models of NP50-277 to.-include all RTS

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in:wt parameters. All currently operating CE plants except Maine Yankee I

were covered in the CEN-327 analysis. The CEN-327 STI analysis I

s:ecifically included the NPSD-277 analyses of the Item 4.5.3 areas of concern except component " wear-out" 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 j

tested on a monthly basis, In cot.n NPSD-277 and CEN-327..the CE RPS cesigns are grouped into four classes ey signal processing and trip device differences, otherwise the 5

logic and ohysical layouts of the RTS are the same for all RTS plant classes.

In NpS0-277, Maine Yankee is included in RPS Plant Class 2.

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CEN-327 Waterford 3 is included in RPS Plant Class 3.

Between NP50-277 arc *EN-327, all of the CE plants are included in plant classes analy:ed in CEN-327. This review considers the analysis and results in CEN-327 j

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.

I ef fects anc cocu*ented the review in a TER, EGG-REQ-7768 (Reference 16).

The result.s of seasitivity sit.cies cone for the TER show the models to be insensitive to an orcer of magnitude increase in'the com:0nent independe n

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failure rates.

The insensitivity to increased component failure rates along with the CE analysis results shewing trip circuit breaker common cause failures to be the major contributor to RTS unavailability providet a

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a basis for this review 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 analyses have adequately covered all five areas of concern or they have teen shown not to contribute to RTS unavailability and that all currently cperating CE reactors are included.

4.3 GE elants I

i Licensees with GE reactors responded to the GL 83-28, Item 4.5.3 f

recuirements as the BWR Owners' Group by su mitting NECD-30844 (Reference 4) to the NRC.

The RTS availability analysis specifically inciveed the five areas of concern and covered both generic relay and solic-state RTS designs which includes all currently operating BWRs.

GE i

stated that the relay RPS configurations for BWR plants have the same primary cesign features.

Therefore, the generic relay RTS models used in NECO-30844 50 not ciffer significantly from the specific.BWR plants. GE usec the Clinton I crawings for the solid-state ~RTS models. Since Clinton 1 is currently the only GE plant with a solid state RTS, no plant unique analysis is necessary.

The EnR Owners' Group also submitted NECO-30851P (Reference 8) to the MC.

S e analysis in this second report used the base case results from NECO-30844 to establish a basis for recuesting revisions-to the current l

Technical Scecifications for the RTS.

The INEL had previously reviewed NECD-30844 and NECD-30651P with regard to both Item 4.5.3 anc STI extension acceptacility and documentec tre review in a TER, EGG-EA-7105 (Re'erence 17). Due to insufficient information, the INEL review could net ec :'ete the solid-state RTS review and accepted only the relay RTS analysis *esults.

The!RC reviewed the topical reports and the TER and 9

6 1

'issuedanSEA(Reference 12). The NRC accepted the analysis results as a t

reference for TS changes related to the RTS and as resolution to GL 83-28, Item 4.5.3, for GE relay plants only.

The INEL later completed the solid state RTS analysis review and issuec Rev 1 to the TER (Reference 18), thus accepting the analyses for all classes of GE plants.

This review examined both GE analyses anc the Rev 1 TER and'cetermined that all five areas of concern are included in the. analyses and that all currently operating GE reactors are included.

4.4 Westincheuse plants Licensees with Westinghouse reactors did not respond directly to the l

recuirements of GL 83-28, Item 4.5.3.

Prior to the Salem ATWS, 'they had suomitted WCAP-10271. (Reference 9) to the NRC to provide a basis for recuesting changes to the Technical Specifications regarding the RTS. The Westingneuse methodology attempted to balance safety and operability and was acclied 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 WCAP-10271, Supplement 1 (Reference 10).

The NRC reviewed the Westinghouse topical reports with'the assistance of Brocinaven National Laboratory (BNL) and issued an SER (Reference 13) limiting their acceptance to changes to only the' analog channel STIs at westingmouse plants.

Ine W methodology used fault trees to model the RTS.

The models included the following five major contributors to RTS trip unavailability:

1.

Unavailability of components due to random failures 2.

Unavailability of components due to test e

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e.

3.

Unavailability of components due to unscheculed ceintenance 1

4 Unavailability of components cue to human error 5.

Unavailaoility of comoonents due to common cause failure.

While the y analysis did not directly include any sensitivity studies conceaning these five areas, the component unavailabilities were increased as the test interval length increased. -The STI analysis results showed a factor of 3 to 5 increase in the RTS unavailability estimates for the i

longer test interval, Two conservatisms exist in the models that are re'evant:

first, no credit was taken for early failures that would be l

cetected and, second, no credit was taken for the diversity inherent in the-W RTS cesign. These two conservatisms, had they been included in the mocel, would cause the increasa in the RTS unavailability estimates to be smal'er than the observed factors.

Test-induced comoonent wear-cut was not-accressed in any manner in the W RTS analysis. However, the RTS analyses done by the other vencors, References 3, 4 anc 6, soecifically investigated the effects of this issue on RTS unavailability. Despite the tifferences among the other vendors' RTS designs, t,ney all found the effects of test induced component wear-cut I

on RTS unavailability to be insignificant. Based on the other vendors' analyses, the INEL conclucec that the effects of test-induced'comoorent wear-cut on y RTS unavailability would also.be insignificant.

Therefore, t e INEL consicers all' y plants to be coverec by adequate analyses.

a.5 0.antitative Review ef Vencers' RTS Ava11 abilities So far, only the adecuacy of the vencors' analyses has been cisc.ssed. No determination has been mace of the' acceptability of the.

l

n. erical estimates from the various RTS availability analyses.

In snis i

section, the INE'. review considers the four Owners broups' RTS availability estimates to cetermine if they are inceed indicative of "r ;n availaoility."

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In Table 1, the four venders' RTS unava11aM11ty estimates are compared to the review estimates of low unavailability as defined in Section 2.

The M W and GE vendors' estimates are given as an overall RTS

{

vnavailability per demanc by plant'mocal and RTS type, respectively.

The CE and y vendors' estimates are given on a similar basis with an additional

{

consiceration that was not necessary for the MW and GE analyses.

In the CE and W analyses, RTS unavailability was estimated for all input

{

FortheCEandyunavailabilityestimatesinTable1,theINEL parameters.

used the unavailability estimates for high pressurizer pressure, the

'I

arameter analyzed in Reference 29 as the limiting parameter for an ATWS in f

terms of the number of input channels and diversity of trip signal.

The differences in the relative values of the three PWR vendors' RTS unavailability estimates can be attributed to design differences among the l

RT$s.

B&W and CE RT$s have four analog-channel inputs for each monitored parameter with four trip logic channels while W RTSs have three or four I

analog channel inputs for each parameter with only two trip logic

{

channels.

The 2 of a analog channels for the M W and CE RTS designs are inherently meu reliable than the 2 of 3 analog channels for some

{

parameters in the y design. Also the 2 of 4 trip logic in the M W and CE RT$s is more reliable than the y 1 of 2 trip logic.

The combination of j

these two design differences make the y RTS unreliability somewhat higner than the other vendors' RTS unavailabilities.

The comparison shows the MW, CE, and GE RTS unavailability estimates are icwer than the NRC's estimates while the y estimates are the same as l

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) tne data uncertainties for both the NRC and Vendors analyses, (2) the scarcity of actual RTS failures world wide, (3) the modeling assumptions ard simplifications used by both the NRC and the Vendors, and (4) tne

[

ciffering levels of model development between the NRC analysis and the Vencers' analyses and between different Venders' analyses. These factors t.

5 12

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TABLE 1.

COMPARISCN OF VENDOR AND NRC RTS UNAVAILA8!LITY ESTIMATES' b*

Vender RTS NRC RTS b

Unavailability Estimates Unavailability Estimates i

Vender (Failures /Comand)

(Failures /Demane)

B&W i

r Davis Bessie Model IE 10" 3E-5 d Oconee Class Mocal IE-6*

3E-5 d

CE Plant Class 1 2E-7' 2E-5 Plant Class 2 3E-6' 2E-5 l

1 Plant Class 3 3E-6' 2E-5 l

Plant Class 4 2E-6' 2E-5.

i i

GE Relay Plants 3E-6 2E-5 Solid-state Plants 3E-6 2E-5 i

W Relay Plants SE-59 d

SE-5 Solid-state Plants SE-59 d

SE-5 t

All estimates are rounced' of f to one significant ' digit.

a.

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

estimates.

c.

From Reference 5, base case.

d.

Includes autcmatic shunt trip on the reactor trip circuit breakers, b

F om Reference 7 Tables a.1-1, 4.2-2, 4.1-3, and 4.1-4, respectively; l

e.

l base case test interval, high pressuri:er pressure unavailability estimate.

f.

From Reference 4.

g.

-om Reference 19, solid state RTS base case. Acclied to relay-plants

{

j base: on similarity of cesign (see Refe ence 11, Section 3.2.2 anc 3.2.3).

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. t help explain the differences between the Vendors' and the NRC's point estimates of RIS= availability.

4.6 Fort St. Vrain Fort St. VrainrespondedtoGL83-28, Item 4.5.3ina5etterto Eisenhut cated Novem'ber 4,1983 0,,g,ggng

" Existing intervals'for on-line functional testing-required by tne Technical Specifications are currently.under

- i review by Public Service. Company of Colorado (PSC) and the

. Nuclear Regulatory Commission Region !V staff. ' The current 1

testing frequency at Fort St. Vrain has been dictated oy tne l

Nuclear Reculatory Commission staff." (Uncerline accec)

In response to a reavest'for information'from the NRC.concerning the Fort St. Vrain responses to GL 83-28 previously sent, PSC:sent the following reply to the NRC-in a letter-to-Johnson, dated June 12, 19852L;

" Existing intervals for the on-line testing required by th'e Technical Specifications 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-1 being' reviewed by the Plant Operations. Review Committee (PORC).

This Technical'$pecification change.is expected to'be-approvec by 1

the PORC and 'the Nuclear Facility. Safety Committee (NSFC) by June 30, 1985.. As part of the development process for these procesec

'l changes to the Technical Specifications, on line functional-testing recuirements were reviewed based on past experience.

' l Possible 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.R6gulatory Commission staff.

The Nuclear Regulatory Commission staff has informed DuDlic Service Company-of Colorado that nc such. changes would be acceptable at this time."

The INEL review interpreted these responses'from Fort St. Vrain to mean the N$ has establishec Fort St. Vrain's RTS current test intervals, the current test intervals have been evaluated by PSC, and the NRC will not al'ow changes to the test intervals it this time.

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From these responses, the INEL concluded that Fort St. Vrain has conducted the review required by-GL 83-28,l Item.4.5.3 -and that the NRC f

considers the PSC and NRC reviews. adequate to meet the Item 4.5.3 ecuirements.

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5-REVIEW CONCLUSIONS e

All four LWR vendors have submitted topical reports either in response to GL 83-28, Item 4.5.3 or to provide a basis for RTS STI extensions, or-both.

For the most part, these reports have accressed all of the issues in Item 4.5.3.

Licensees 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 RTSs to be highly reliable with the current test intervals and prior to' implementing some of the requirements of'GL 83-28.

Implementation of these additional requirements will reduce the ATWS risk even further.

The'INEL has reviewed the relevant topical reports, TERs, SERs, acettional analyses, and the.indivicual licensee submittals with regard to GL 83-28, Item 4.5.3, reewirements and the review criteria. Based on that review, the INEL concludes that all licensees of currently opertting j

ccmmercial nuclear power plants have adequately demonstrated that their current RTS test intervals are consistent with achieving high RTS availability.

16

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6.-

REFERENCES 1.

U.S. Nuclear Regulatory C.ommission, Generic implications of ATWS Events at the Salem Nuclear Dower Plant,.NUREG-1000, April 1983, i

2.

U.S. Nuclear Regulatory Commission Letter, D. G. Eisenhut to All Licensees et al-., Reevired Actions Based on Generic Implications of i

Salem ATVS Events, Generic Letter 83-28, July.8, 1983.

i 3.

Combustion Engineering, Reactor Protection System Test Interva11 i

Evaluatten. Task 486, CE NPSD-277 Decemoer 1984 4

$. Visweswaran et al., BWR Owners' Group Response to NRC' Generic Letter 83-28. Item 4. 5. 3, NECD-30844, January 1985.

5.

R. S. Entinna et al, Justification for Increasino the Reactor Trio

\\

System On-line Test Interval, BAW-10167, May 1986.

i 6.

R, S. Enzinna et al.,JJustification for'Increasino the Reactor Trio System On-line Test Interval, Suoplement Numcer 1, 8AW-10167,

]

Supplement Numoer 1, Feeruary 1988.-

1 7.

. Combustion Engineering, RDS/ESFAS Extended Test Interval Evaluation, I

CEN-327. May 1986.

W. P. Sullivan et al.,_ Technical Soecification Imorovement Analyses for BWR Reactor D otection System, NECD-30851P, May 1985.

9.

R. L. Jansen et al., Evaluation of Surveillance Frecuencies and Out of t

Service Times for the Reactor Protection Instrumentation System,-

WCAD-10271,. January 1983, 10.

R. L. Jansen et al., Evaluation of Surveillance Frecuencies and Out of Se vice Times for the Rea: tor Prote: tion Instrumentation System,

[

Sueciement 1, aCAP-10271, Supplement 1, July 1983.

l 11.

R. L. Jarsen et al., Evaluation of Surveillance'Frecuencies and Out of Seevf:e Times for the Rea: tor Drotection Instrumentation System.

?

Suecieteat ;-;-A. WCAP-10271, Supplement 1-P-A, May 1986.

12.

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

But: hor. Accootance for Refe encine ~of General Electric Comoany (GE)

Topical Reoorts NECD-308aa, "BaR 0=ners' Grovo Resoonse to NRC Gene ic-Letter 83-28." anc NECD-3085;P. "Tecnnical Specification.:mo*ovement.

Analyses for BWR Reactor Drotection System."' April 28, 1986, n'

13.

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

I Sheocard, Acceotance for Referencino of Licensino Topical Resort WCAD-10271, " Evaluation of Surveillance F oouencies anc Out of'Seavice I' Pes for tne Reactor o*ote: Son Instrumentatier Systems." FeDruary i

21,.'985.

4 17

?'

o -- o 14.

U.S. Nuclear Regulatory Commission, Amendments to 10 CFR 50 Related to.

- Antici >ated Transients Vithout Scram ( ATWS) Events, SECY-83-293, July

-19, 19d3.-

~

15..J. P. Poloski and S. D. Matthews, Review of B&W Owner's Group Analyses for increasing The Reactor Trio System on-line Test Interval.

EGG-REQ-7718, Septemoer 1988.

16.

D. P. Mackowiak and B. L. Collins, A Review of the Combustion Engineerino Evaluatier. For Extendino the RPS anc 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-owr Reactor-Protection System, EGG-EA-7105, January 1986.

l 18.

R. E. Wright and B. L. Collins, A Review of the BWR Owners' Group Technical Specification Improvement Analyses for the BWR Reactor:

.4 Protection System, EGG-EA-7105, Rev 1. Maren 1987.

4 19.

D. A. Reny et al., Evaluation of Generic Issue 115. Enhancement of-the Reliability of-Westiacneuse Solic State Protection Systems, NUREG/CR-5197, January 1989, 20.

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

Eisennut-Response to Generic Letter 83-28, November 4,_1983.

f

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

Johnson, Response to Generic Letter 83-28. June 12,1985.

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• .s TECHNICAL EVALUATION REPORT: A REVIEW 0F REACTOR TRIP SYSTEM AVAILABILITY ANALYSES FOR GENERIC LETTER 83-28.

ITEM 4.5.3 RESOLUT!0N

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n..c u Davitt P. Mackowiak

^ March 1989 John A. Senroeder

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March 1989 r

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.....c Regulatory and Technical Assistance EG&G Idaho. Inc.

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P. O. Box'1625 Idaho Falls. ID 83415-06001'

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Instrumentation and Control Systems Branch l'

Division of Engineering and System Technology Technical Evaluation Report Office of Nuclear Reactor Regulation

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U.S. Nuclear Regulatory Commission Washington, DC 20555

....s........

.u a.e..e m The Idaho National Engineering Laboratory (INEL) conducted a technical review of the commercial nuclear reactor licensees' responses to the requirements of the Nuclear-Regulatory' Commission's (NRC's) GenericLLetter 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 orovide the NRC staff with a basis to close out'this issue with no further-review.

The licensees, as the four vendors' 0wners' Groups, submitted analyses.to the-NRC eith directly in resoonse to GL 83-28.. Item.4.5.3. or to provide.a bas'is for requesting changett to tne Tecnnical Soecifications (TSs)' 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 acceotability of the results.

The 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

?

nere also reviewed. The INEL review results show that'all-licensees of currently opera-ting comercial 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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