ML19354E173

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

ML19354E173
Person / Time
Site:	Davis Besse
Issue date:	01/19/1990
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML19354E171	List: ML19354E170 ML19354E173
References
GL-83-28, NUDOCS 9001290164
Download: ML19354E173 (2)

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I SAFETY EVALUATION i

GENERIC LETTER B3-28. ITtn 4.5.3 REACTOR TRIP SYSTEM EELIABILITY FOR ALL DUME5 TIC OPERATING REACTORS TOLEDO E0150N CUMPANY l

AND THE CLEVELAND ELECTRTCILLUMINATING COMPANY DAV15-EE55E NUCLEAR POWER STATION UNIT NO. 1 DOCKET NO. 50-346 l

1.0 INTRODUCTION

On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system (RPS). This incident was terminated manually by the operator about 30 seconds after the initiation of the automatic l

trip signal.

The failure of the circuit breakers was determined to be related l

to the sticking of the undervoltage trip attachment. Prior to this incident, i

on February 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic 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 i

j 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 Nuclear Power l

Plant. The results of the staff's inquiry into the generic implications of the Salem Unit 1 incicerts are reported in NUREG-1000, " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant".. As a result of this investigation, the Comission (NRC) requested (by Generic Letter 83-28 dated July 8,1963) 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 that on-line functional testing of the reactor trip system (RTS), including independent testing of the diverse trip features, was being perfonned at all plants.

Existing intervals for on-line functional testing required by Technical Speci+1 cations wars to be reviewed to determine if the test intervals were adequate for achievfng high RTS availability when accounting for considerations such as:

(1) uncertainties in component failure rates; (2) uncertainties in comon mode failure rates; (3) reduced redundancy during testing; (4) operator error during testing; and (6) component ' wear-out" caused by the testing.

2.0 DISCUSSION TheNRC'scontractordroupavailabilityanalysesandevaluatedtheadequacyof Idaho National Engineering Laboratory (INEL), reviewed the licensee Owners 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, l

"A Review of Ratcter 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 cf Item 4.5.3 and our review of EGG-NTA 8341 are presented below.

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The Babcock & Wilcox (B&W), Owners Groups have submitted topical repo Com6ustionEngineering(CE),GeneralElectric (GE),andWestinghouse(W) in response to GL 83-28, Item 4.5.3 or to provide a basis for requesting Technical Specification changes to extend RTS surveillance test intervals

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(ST)). 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 re table and that the unavailability was dominated by comon 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 stch estimates are large, because the systems are highly reliable, very little experience exists to support the estimates, and comon uuse 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, cerimon mode f ailures tend to predominate, and, for a number of reasons, additional testing will not appreciably lower RTS unavailability.

First, testing more freouthtly than weakly is generally impractical, and even i

so the increased testing could at best lower tae failure probability by less than a factor of four compared to monthly testing. Secondly, increased testing could possibly increase the probability of a comon mode failure through increased stress on the system.

Finally, not all potential failures are detectable by testing.

In sumary, NUREG-0460 provides additional justification to dencnstrate that the current monthly test intervals are adequate to maintain high RTS availability.

3.0 CONCLUSION

All four venders' topical reports have shown the currently configured RTS to be highly reliable with the current trenthly 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 a theexistingintervals,ndthefindingsnotedinNUREG-0460,weconcludethat as reconsnended in the topical reports, for on-line functional testing are consistent with achieving high RTS availability at all l.

operating reactors.

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ENCLOSURE 3 EGS NTA.8341 March 1989 TECHNICAL EVALUATION REPORT

/daho Net /onal A REVIEW OF REACTOR TRIP SYSTEM AVAILAg!L:Ty Eng/neer/ng

$sSu!!ol R G NER!c LEntR 83 28, ITEM 4.s.3, Laboratory Monaged c'y the U S.

08Vid P. M4CkOwiak John A. $Chroeder Department of Energy hEmsg,,,

Prepared for the

U.S. NUCLEAR REGULATORY COMMISSION

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NOTICE Tlus report was prepared u an account of work sponsored by an assary of the Unsied States Government. Neither the Usated Sase Governness not any asemey thereof, not any of their employes, make any warranty. apresed or imphed, or assumes any legal habibry or reponsibibry for any taurd pany's use or the resulu of Mach use, of any informauen, apparatus, prtduet et pret.

est daclosed in thu report. or represenu that tu use Dy such thard pany would not intnnge pmately owned nahu.

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EGG-NTA-8341 TECHNICAL EVALUATION REPORT: A REVIEW 0F REACTOR TRIP SYSTE AVAILABILITY ANALYSES FOR GENERIC LETTER 83-28, ITEM 4.5.3, RESOLUTION David P. Mackowiak John A. Schroeder i

EG&G Idaho, Inc.

Idaho Falls, Icaho 83415 FIN 06001: Evaluation of Conformance to Generic Letter 83-28 for ors (Project 2) i e

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l ABSTRACT The Idaho National Engineering Laboratory (INEL) conducted a i

technical review of the commercial nuclear reactor licensees' responses to the recuirements 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 acequate analysis, will provide the NRC staff with a basis to close out this issue with no further review.

The licensees, as the four vendors' Dwners' Groups, submitted analyses to the NRC either cirectly in response to GL 83-28, Item 4.5.3, or to provide a basis for requesting changes to the Technical Specifications (TS) that would entend the Reactor Protection System (RPS) survei' lance test intervals ($T15). To conduct the review, the INEL cefined three criteria to dete mine the adecuacy, plant applicability, and acceptability of the results.

The INEL examined the Owners Groues' reports to determine if the analyses and results met the establishec criteria.

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

The INEL review results show that all licensees of currently crerating comme cial nuclear reactors have adecuately demonstrated that their current en line RPS test intervals meet the requirements of GL 83-28, Item 4.5.3.

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SUMI%RY The two anticipated transient without scram (ATWS) 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 ATW5 events. The NRC then published Generic Letter 83-28 (GL 83-28) which listed the actions the NRC required of all licensees holding operating licenses and others with respect to alluring 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, requirements as Owners Groups with reports either in direct response to Item 4.5.3, or with a technical basis for requesting extensions to the surveillance test intervals (ST!s) that generally included the Item 4.5.3 recuired reviews.

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

INEL review results showing general compliance with Item 4.5.3 will provide the NRC with a basis to close out Item 4.5.3 without further review.

For the review, the INEL defined three acceptance criteria, reviewed the licensees teoical reports, contractor review reports, and NRC safety evaluations, and determined the adequacy of the analyses and the RTS availability estimates with regard to the review criteria.

The INEL review criteria to determine the licensees' Item 4.5.3 com;11ance were, (1) the five areas of concern of Item 4.5.3, (2) the analyses' plant applicability, and (3) the NRC's RTS electrical unavailability base case estimates from the ATW5 Rulemaking Paper, SECY-83-293.

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Each Owners Groups' reports were reviewed to ensure that all five areas of concern from Item 4.$.3 were either included in the analyses or shown not to be significant with regard to ATS availability. The !NEL review also ensured that the' individual plants' differences from the analysis' models were taken into account and their effects were shown not to significantly affect RTS unavailability. The Fort St. Vrain responses to Item 4.5.3 were also reviewed.

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

6 The results of the INEL review showed that all licensees of currently operating commercial nuclear ~ reactors have adequately demonstrated that their current on-line surveillance test intervals are consistent with achieving high RTS availability.

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ATW$

Anticipated Transient Without Scram l

B&W Babcock & Wilcox BNL trookhaven National Laboratory CE Comeustion Engineering GE General Electric HTGR High-Temperature Gas-Cooled Reactor ICSB Instrumentation and Control Systems Branch INEL Idaho National Engineering Laboratory LWR Light Water Reactor NFSC Nuclear Facility Safety Committee NRC Nuclear Regulatory Commission NRR Offics of Nuclear Reactor Regulation PORC Plant Operations Review Comrittee PSC Public Service Com;any of Colorado PWR Pressurized Water Reactor RSSMAP Reactor Safety Stucy Methocciogy Applications Program Rp5 Reactor Protection System RTS Reactor Trip System SER Safety Evaluation Re:c-t STI Surveillance Test Interval TER Technical Evaluatier Report W

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CONTENTS 1

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

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

111 ACRONYMS..............................................................

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INTRcouCT:0N.....................................................

I 1.1 Historical Background......................................

I 1.2 Review Purpose.............................................

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2.

REVIEW CRITERIA..................................................

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REVIEW METHODOLOGY...............................................

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4 REVIEW RESULTS...................................................

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i 4.1 B&W Plants.................................................

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

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

9 4.4 Westinghouse Plants........................................

10 4.5 Quantitative Review of Vendors' RTS Unava11 abilities......

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4.6 Fort St. Vrain

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REVIEW CONCLU$10N$..............................................

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REFERENCES.......................................................

17 TABLES 1.

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

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TECHNICAL EVALVATION REPORT: A REVIEW OF REACTOR TR!p $Y$ TEM -

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AVAILABILITY ANALYSES FOR GENERIC LETTER 83 28.

ITEM 4.5.3 RESOLUTION i

1.

INTRODUCTION 1.1 Historical Backaroun_d In February of 1983, two events occurred at the Salem Nuclear Generating Station that focused Nuclear Regulatory Commission (NRC) attention on the generic implications of anticipated transient without scram (ATWS) events.

First, on February 22, during startup of Unit 1 an automatic trip signal generated as a result of a steam generator low-low level failed to cause a reactor scram. The reactor was tripped manually by an operator almost coincicentally with the automatic trip signal, so the fact that the automatic trip had failed to cause a scram went unnoticed.

Three days later en February 25, both of the scram breakers at Unit 1 failed to open on an automatic reactor protection system (RPS) scram sigral.

The operators took action to control this second ATW$ and succeeded in terminating the incident in about 30 seconds. Subsequent investigation related the failure of the Unit 1 RPS to cause a scram to sticking of the undervoltage trip attachment in the scram circuit breakers.

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

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

allowed to restart; (2) a fact finding report of the events at Salem 1 and the circumstances leading to them; and (3) a report on the generic imclications c' these events.

To address (3) above an interoffice, interdisciplinary group was formed includ'ng members from sne Office of Nuclear Reactor Regulation's 1

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t (NRR's) Division of Licensing, Division of systems Integration, Division of i

Human Factors safety, Division of Engineering, Division of liafety Technology, the Office of Inspection and Enforcement, the Office for Analysis and Evaluation of Operational Data, and NRC's Region ! Office.

This group published NUREG-10001 as a result of their efforts to resolve the following Questions: (1) is there a need for prompt actions to address

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similar ecuipment in other facilities; (2) are the NRC and itu licensees i

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 28 (GL 83 28). The actions described in GL 43-28 adc'ress issues related to reactor trip system (RTS) reliability. The actions covered fall into the'following four areas: (1) Post-Trip Review, (2)

Ecuipment Classification and Vender Interface, (3) Post-Maintenance Testing, and (4) Reactor Trip System Reliability Improvements.

Item 4, above, is aimed at assuring that vender-recommended recctor trip breaker modifications and associated reactor protection systsm changes are ecmpleted 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 that use circuit breakers in their reactor trip systems, and to ensure that on-line functional testing of the reactor trip system is performed on all light water reactors (LWRs).

The specific requirements of GL 83 28, Item 4.5.3, are that existing intervals for on-line functional testing required by Technical Specifications shall be reviewed to determine if the intervals are consistent with 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 testirg; (4) ocerator errors during testing; and (5) component " wear-cut

caused by testing.

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The tabcock & Wilcox (B&W), Combustion Engineering (CE), Gezeral.

Electric (GE), and Westinghouse (W) Owners Groups have submitted topical reports either in response to GL 83 28, Item 4.5.3'3 or to provide a basis for requestin extensions.IO'gIRTS surveillance test interval ($TI)

In general, the owners groups' analyses were

'not done on a plant specific basis.

Instead, the analyses addressed 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 83 28, Item 4.5.3 and summarized their findings in Safety Evaluation 12 Reports,,13 ($gg,),

r 1.2 Review purcese i

This report documents a review of the Owners Groups' topical reports,

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the NRC SERs, and other analyses done at the Idaho National Engineering Laboratory (INEL) by persennel in the NRC Risk Analysis Unit of EG4,G Idaho, Inc. The INEL conducted the review at the request of the U.$. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Instrumentation and Control Systems Branch (IC58). The review was performed to determine if the Owners Groups' analyses demonstrated high RTS availability 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 covered by the analyses. The results of the review, if all plants are shown to be covered by an adequate analysis, would provide the NRC with a basis for closirg out GL B3-28, Item 4.5.3, for all U.S. commercial nuclear reactors witnout further review.

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The body of this report presents the review and its findings with regard to the stated objectives. Section 2 describes the criteria used in the review to determine the adequacy of the analyses. The review methodology is discussed in Section 3.

Section 4 presents the review l

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 judgment or decisions may be' based.

In this section, the INEL availability analyses review criteria are presented.

GL 83-28 established the three criteria used in the INEL review.

GL 83-28 stated that: (1) all licensees et al., (2) must demonstrate high RTS avai' ability for the current test intervals by documented review when (3) accounting for such considerations as the five areas of corcern 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 third and most subjective criterion, "high availability", was not defined.

To establish a definition of high availability, the INEL used the electrical unavailability base case estimates presented in Table A-1 of Appendix A to SECY-83-293.I" Unavailability is defined as 1.0 minus l

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 j

expressed in terms of low unavailability for compatibility. These RTS i

unavailability estimates from Reference 14 were used for two reasons.

First, they were used because they were developed by the NRC's ATWS Task Force as a reevaluation of the bases for the RTS unava11 abilities used in ATWS rule value-impact evaluations.

Second, as stated in Reference 14 this NRC analysis

"... bases the RTS unavailabilities on worldwide experience to 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 vendor designs based on a comparative reliability analysis that evaluates the major cif'erences among the designs."

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The. estimates from the NRC ATW5 analysis provide a framework with-which to consider the topical report analyses estimates. The numerical estimates in the $ECY-83 293 for the four vendors combined with the five areas of concern from GL 33-28, Item 4.5.3, form the criteria used for this review to determine if the venders' analyses and estimates met the requirements of item 4.$.3.

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REVIEW METHODOLOGY The INEL conducted this review by examining the venders' topical reports (References 3, 4, 5, 6, 7, 8, 9, 10, and 12), the technical evaluation report: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 Westinghouse $olid State Protection System."I' This was done for three f

First, the reports were examined to find out whether or not the reasons.

venders' analyses addressed the areas of concern from Item 4.5.3 and reflected a high RT$ availability.

Second, they were examined to determine what plants were tavered by the venders' analyses. Third, the Generic Issue 115 report provided an independent, updated estimate of the availability of the W solic state RTS for comparison to the review criteria.

i for the plants covered by the venders' analyses or the NUREG/CR-5197 analysis, the sopropriate analysis and availability were compared to tha review criteria established in Section 2.

If the analysis acequately i

addressed the areas of concern and esmonstrated a high RTS availability, the plant was accepted as having met the requirements of GL 83-28, Item 4.5.3.

The results of the comparisons for plants covered by a vendor analysis are given by vender in Section 4, For plants not directly covered by a vencor's analysis, an acce: table means was founc to extene the analyses to cover the plants. This was cone for two plants: Clinton 1 (GE) and Maine Yankee (CE). The means by which the analysirs were extended to cover these two plants are also discussed by vender in Section 4 l

One plant, Fort $t. Vrain, a high temperature, gas-cooled reactor (HIGR), was not covered by any of the four vendors' analyses and recuired soecial consideration.

The INEL examined the responses from Fort $t. Vrain recuired by GL 83-28, Item 4.5.3 to cetermine if the responses demonstrated an accettably high RTS availability.

The review of the Fort St. Vrain responses is given in Section 4.6.

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REVIEW RESULTS This section summarizes the results of the INEL review of the vendors' 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 avat14bility, gained from an examination of RTS importance measures from selected PRAS, are examined.

4.1 B&W plants The issues of GL 83-28. Item 4.5.3, were addressed by the B&W Owners Group and the results were submitted to the NRC by the individual utilities in their responses to GL 83-28. Topical Report BAW-10167 (Reference 5) was submitted to '.he NRC to provide a technical basis for increasing the on-itne 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 address the GL 83-28. Item 4.5.3 issues. However, some information that was resolved in the generic letter analysis was not repeated in the subsequent Topical Report because it was not relevant to the proposed Technical Specification changes. To make BAW-10167 applicable to both GL 83 28, Item 4.5.3 and ST!/A0T issues, the Owners Group submitted BAW-10167, Supplement 1 (Reference 6), to the NRC.

Supplement 1 completed the B&W analysis by addressing all remaining Item 4.5.3 issues. The

. BAW -10167 and Supplement 1 analyses included the implementation of the automatic shunt trip on the reactor trip circuit breakers as required by GL 83-28. Item 4.3.

The INEL has previously reviewed the BAW-10167 and Supplement 1 analyses and documented 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 con:ern were conducted on the RT$ mocels. The sensitivity study results showed the models to be insensitive to variations in the failure rates associatec with the item 4.5.3 areas of concern.

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The INEL reviewed BAW-10167, BAW-10167, Supplement 1, and the TER and determined that the B&W analyse: adecuately covered all five areas of concern and that all current h operating B&W reactors are included.

4.2 CE Plants Licensees with CE reactors responded to the requirements 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 included all five areas of concern and all currently operating CE reactors except Waterford 3, which was not in commercial operation until September 1985.

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

analysis expanded on th~e simplified models of NPSD-277 to include all RTS input parameters. All currently operating CE plants except Maine Yankee J

were covered in the CEN-327 analysis. The CEN-327 STI analysis specifically included the NP$D-277 analyses of the Item 4.5.3 areas of concern except component " wear-cut" during testing. The CEN-327 analysis i

showed that the major contributors to RT3 unavailability for the four plant j

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 included in RPS Plant Class'3. Between NPSD-277 arc *EN-327, all of the CE plants are included in plant classes analyzed in CEN-327.

This review considers the analysis and results in CEN-327 l

acecuate for Item 4.5.3 resolution for all classes of CE plants.

The INEL has previously reviewed CEN-327 with regard to $TI extension effects anc cocumented the review in a TER, EG3-REQ-7768 (Reference 16).

The results of semitivity studies done for the TER show the models to be insensitive to an orcer of magnitude increase in the component incepencent 8

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failure rates. The insensitivity to increased component failure ratd:

along with the CE analysis results showing trip circuit breaker common cause failures to be the majqr contributor to RTS unavailability provides a i

a basis for this review to conclude that RTS test-induced component wear-cut is not an issue at CE reactors.

i 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 been shown not to contribute to RTS unavailability and that all currently operating CE reacters are included.

4.3 GE Plants Licensees with GE reactors responded to the GL 83-18, Item 4.5.3 recuirements as the BWR Owners' Group by submitting NECD-30844 (Reference 4) to the NRC. The RTS availability analysis specifically incluced the five areas of concern and covered both generic relay and solid-state RTS designs which includes all currently operating BWRs. GE stated that the relay RPS configurations for SWR plants have the same primary design features. Therefore, the generic relay RTS models used in NECD-30'4t4 do not dif fer significantly from the specific BWR plants. GE usec the Clinton I crawings for tne 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 EhR Owners' Group also submitted NECD-30851P (Reference 8) to the NRC. The analysis in this second report used the base case results from NECD-30844 to establish a basis for recuesting revisions to the current Technical Specifications for the RTS. The INEL had previously reviewed NECD-30844 and NECD-30851P with regard to both Item 4.5.3 and ST! extension acceptacility and documented the review in a TER, EGG-EA 7105

-(Reference 17). Due to insufficient information, the INEL review could net co elete the solid-state RTS review and accepted only the relay RTS analysis results. The NRC reviewed the topical reports and the TER and l

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o issued an $ER (Reference 12). The NRC accepted the analysis results as a reference for TS changes related to the RTS and as resolution to GL 43 28, Item 4.5.3 for GE relay plants only. The INEL later completed the solid state RTS analysis review and issued Rev 1 to the TER (Reference 18), thus accepting the analyses for all classes of GE plants.

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

i 4.4 Westinchouse Plants i

Licensees with Westinghouse reactors did not respond directly to the requirements of GL 83-28, Item 4.5.3.

Prior to the Salem ATWS, they had submitted WCAP-10271 (Reference 9) to the NRC to provide a basis for recuesting changes to the Technical Specifications regarding the RT3. The Westinghouse methodology attempted to balance safety and operability and I

was applied to a typical Westinghouse four loop reactor plant with a solid 1

state RTS in WCAP-10271.

The methodology was entended to cover RT$s for two, three, and four loop plants with either relay or solid state logic in t

WCAP-10271, Supplement 1 (Reference 10).

The NRC reviewed the Westinghouse topical reperts with the assistance of Brookhaven National Laboratory (BNL) and issued an $ER (Reference 13) limitir.g their acceptance to changes to only the analog channel $ tis at Westingneuse plants.

i The W methodology used fault trees to model the RT3. 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 10

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3.

Unavailability of components due to unscheduled maintenance.

4.

Unavailability of components due to human error 5.

Unavailability of components due to common cause failure.

While the y analysis did not directly include any sensitivity studies cencerning these five areas, the component unav411 abilities 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 longer test interval. Two conservatisms exist in the models that are relevant: first, no credit was taken for early failures that would be cetected and, second, no credit was taken for the diversity inherent in the y RTS design. These two conservatisms, had they be'en included in the l

mecel, would cause the increase in the RTS unav&ilability estimates to be smaller than the observed factors.

l Test-induced component wear-cut was not addressed in any manner in the y RTS analysis. However, the RTS analyses done by the other vencers, References 3, 4 and 6, specifically investigated the effects of this issue on RTS unavailability. Despite the differences among the other vendors' RTS designs, they all found the effects of test induced component wear-cut on RTS unavailability to be insignificant. Based on the other vendors' analyses, the INEL concluded that the effects of test-induced' component wear-cut on y RTS unavailability would also be insignificant. Therefore, the INEL consicers all y plants to be coverec by adequate analyses.

l 4.5 Cuantitative Review of Vendors' RTS Availabi11 ties So far, only the adequacy of the vendors' analyses has been ci s c;.s sed. No determination has been made of the' acceptability of the neerical estimates from the various RTS availability analyses.

In this j

section, the INEL review considers the four Owners Groups' RTS availability estimates to determine if they are inceed indicative of "high availability."

f G

11 i

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

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compared to the review estimates of low unavailability as defined in

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section 2.

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

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unavailability per demand by plant model and RTS type, respectively.

The CE and y vendors' estimates are given on a similar basis with an additional consideration that was not ateessary for the MW and GE analyses.

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

[

For the CE and W unavailability estimates in Table 1, the INEL parameters.

used the unavailability estimates for high pressurizer pressure, the parameter analyzed in Reference lg as the limiting parameter for an ATWS in terms of the number of input channels and diversity of trip signal.

The differences in the relativ6 values of the three PWR vendors' RTS unavailability estimates can be attributed to design differences among the RT$s. RW and CE RT3s have four analog channel inputs for each monitored parameter with four trip logic channels while W RTSs have three or four analog channel inputs for each parameter with only two trip logic channels.

The 2 of 4 analog channels for the MW and CE RTS designs are inherently more reliable than the 2 of 3 analog channels for some 1

parameters in the W design. Also the 2 of 4 trip logic in the M W and CE RT5s is more reliable than the W 1 of 2 trip logic. The combination of 1

these two design differences make the y RT5 unreliability somewhat higher than the other vendors' RTS unavailabilities.

The comparison shows the RW, 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 i

and simplifications used by both the NRC and the Vendors, and (4) the differing levels of model development between the NRC analysis and the Vendors' analyses and between different Venders' analyses. These factors l

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TABLE 1.

COMPARISON OF VENDOR AND NRC RTS UNAVAILABILITY ESTIMATES * '

Vendor RTS NRC RTS b

Unavailability Estimates Unavailability Estimates Vender (Failures / Demand)

( Failures / Demand)

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

3E-5 l

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

Plant Class 3 3E-6' 2E-5 Plant Class 4 2E-6' 2E-5 GE Relay Plants 3E-6 2E-5 f

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

Relay Plants 5E-58 d

SE-5 Solid-state Plants SE-59 d

SE-5 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 Reference 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 based on similarity of design (see Reference 11, Section 3.2.2 and 3.2.3).

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' gp help explain the differences between the Vendors' and the NRC's point estimates of RTS availability.

4.6 Fort St. Vrain Fort St. Vrain responded to GL 83-28, Item 4.5.3 in a letter to 20 Eisenhut dated November 4, 1983,,g,ggng

" Existing intervals for on line functional testing required by the Technical Specifications are currently under review by Public Service Company of Colorado (PSC) and the Nuclear Regulatory Comission Region IV staff. The current testino frecuency at Fort $t. Vrain has been dictated by the Nuclear Reculatory Commission staf f." (Uncerline acced)

In response to a request for information from the NRC concerning the Fort St. Vrain responses to GL 83-28 previously sent, PSC sent the 21 following reply to the NRC in a letter to Johnson, dated June 12, 1985 "Emissing intervals for the on-line testing required by the 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 b.4.1) are currently being reviewed by the Plant Operations Review Committee (PORC).

This Technical Specification change is espected to be approvec 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 esperience.

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 Regulatory Comission staff. The Nuclear Regulatory Commission staff has informed public Service Company of Coloraco that no such changes would be acceptable at this time."

The INEL review interpreted these responses from Fort St. Vrain to mean the N,RC has establishec Fort St. Vrain's RTS current test intervals, the current test intervals have been evaluated by PSC, and the NAC will not allow changes to the test intervals at this time, 14 l

1

i 1

u,

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0, From.these responses, the !NEL concluded that Fort $t. Vrain has >

]

conducted the review required by GL 33-28, Item 4.1.3, and that the NRC I

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

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5 REVIEW CONCLU$10NS i

r 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 addressed all of the issues in Item 4.5.3.

Licensees not covered by the topical reports have submitted i

incividual responses to Item 4.5.3.

+

t The analyses in the topical report have shown the currently configured RT5s 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.

i The INEL has reviewed the relevant topical reports, TERs, $ERs.

accisional analyses, and the individual licensee submittels with regard to

(

GL 83-28, 1:em 4.5.3, requirements and the review criteria. Based on that review, the INEL concludes that all licensees of currently operating commercial nuclear power plants have adottately demonstrated that their current RTS test intervals are consistent with achieving high RTS availability.

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REFERENCES 1.

U S. Nuclear Regulatory Commissich, Generic Implications of ATWS Events at the Salem Nuclear Power Plant, NUREG-1000, April 1983.

2.

U.S. Nuclear Regulatory Commission Letter, D. G. Eisenhut to All Licensees et al., Recuired Actit,ns Based on Generic implications of Salem ATVS Events, Generic Letter 83'-28. July 8,1983.

3.

Combustion Engineering, Resetor Protection System Test Interval Evaluation. Task 486, CE NP50-277 Decemoer 1984.

- 4 S. Visweswaran et al;, BWR Owners' Group Response to NRC Generic Letter 83-28. Item a.5.3, NECD-30844, January 1985.

~

5.

R. S. Enzinna et al., Justification for Increasino the Reactor Trio System On-line Test Interval, BAW-10167, May 1956..

G.

F.. S. Enzinna et al., Justification for Increasino the Reactor Trio System On-line Test Interval. Supplement Nummer 1, BAW-10167, Supplement Numoer 1. Feorwary 1988.

7.

Combustion Engineering, RPS/ESFAS Extended Test Interval Evaluation, CEN-327, May 1986.

l 8.

W. P. Sullivan et al., Technical Specification Imorovement Analyses for

'BWR Reactor protection System, NECD-30851P, May 1985.

l.'

9.

R. L. Jansen et al., Evaluation of Surveillance Frecuencies and Out of service Times for the Reactor Protection Instrumentation System, WCAP-10271, January 1963, 10.

R. L. Jansen et al., Evaluation of Surveillance Frecuencies and Out of Service Times for the Reactor Protection Instrumentation System.

Supplement 1, wCAP-10271, Supplement 1, July 1983.

11.

R. L. Jansen et al., Evaluation of Surveillance Frecuencies and Out of Service Times for the Reactor Protection Instrumentation System.

Sucolement 1-P-A, WCAP-10271, Supplement 1-P-A, May 1986.

l-t 12.

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

l Butcher, Acceotance for Referencino of General Electric Company (GE)

Topical Reports NECD-30Baa, "BWR Owners' Group Response to NRC Generic

. Letter 83-28," and NECD-3085;P, " Technical Specification Emprovement Analyses for BWR Reactor Protect' ion System."' April 28, 1986.

13.

U.S. Nuclear R'egulatory Commission Letter, C. O Thomas to J. J.

Sheppard, Accectance for Referencino of Licensing Topical Report E

L WCAP-10271, " Evaluation of Surveillance Frecuercies anc Out of Service Times for the Reactor Peetection Instrumentatier Systems " Feoruary p

21, 1985.

17

?

l:

i

- ~

..-v 3 O,

14.

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

Anticipated Transients Without Scram (ATW5) Events. 5ECY-83-293, July 19, 1983.

]

15.

J. P. Poloski and S. D. Matthews, Review of B&W Owner's Grove Analyses for Increasino The Reactor Trio System On-line Test Interval, EGG-REQ-7718, Septemeer 1988.

D. P. Mackowiak and 8. L. Collins, A Review of the Combustion 16.

i Encineerino Evaluation For Extendino the RP5 anc E5FA5 Test Intervals, EGG-RED-7768, September 1988.

17.

R. E. Wright and B. L. Collins A Review of the BWR Owners' Group Technical Specification Improvement Analyses for the 8wR Reactor Protection System, EGG-EA-7105, January 1986.

18.

R. E. Wright and B. L. Collins, A Review of the BWR Owners' Greue Technical Specification Improvement Analyses for the BWR Reactor Protection System, EGG-EA-7105, Rev 1. March 1987.

19.

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

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

Eisenhut, Response to Generic Letter 83-28, November 4,1983.

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

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

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