ML19332C459
| ML19332C459 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 11/16/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19332C458 | List: |
| References | |
| GL-83-28, NUDOCS 8911280152 | |
| Download: ML19332C459 (3) | |
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StJETY EVALUATION E,Y TEE OFFICE Of kUCLEAR REACTOR REGULATION RELATEDTOGENERICLETTER8328,,,),TEtt,4.,5,3REACTORTRIP SYSTEM REllABILITY FOR DiTROIT ED1 SON COMPANY l
WOLVERINE POWER SUPPLY COOPERATIVE, INCOPPORATED FERM1-2 Dor. RET _p0,._,5,0-33,1 l
l 100 INTRODUCTION On February TE 1983, both of the scrato circuit breakers at Unit 1 of the Salem Nuclear Powcr Plant feiled to open upon an autor4 tic reactor trip signal from tFt reactot protectiot systet. (RPS). This incident was terrinated manually by the operator about 30 seterds after the initiation of the autotiatic trip signal.
The failure of the circuit breakers was detert.ined to be related to the sticling of the undenoltage trip attachment.
Fr ict to this on February 22, 1903, at Unit 1 of tit Salem Nuclear Power Plant, er autoratic itip signal was generated tesed on stear, generator ick-low level during plant startup.
In this case, tt.t rtactor was tripped manually by the opetator altcst coincidentall) with the autcr.it. tic trip.
Folic +ing these incidents, on February 20,1983, the NRC Executiu Director for Cperations(EPO),directedthestafftoinvestigateandreportonthegeneric implications of these occurrences at Unit 1 of the Salem Nuclear Power Plant.
The results of the staff's inquiry into the generic in.plications of the Saler Unit 1 incicents are rcported in NUREG-1000, " Generic Irap11 cations of the ATKS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Comission (KRC) requested (by Geocric Letter 03-23 dated July 8,1983) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to iespond to the generic issues raised by the analyses of these two ATKS events.
The licensees were required by Generic Letter 83-28, item 4.5.3 to confirra thaten-linefunctionaltestingofthereactortripsystem(RTS), including independent testing cf the diverse trip features, was being performed at all plants.
Existing intervals for on-line functional testing required by Technical Spttifications were to be revieweu to determine if the test intervals were adequate for echieving high RTS availability when accounting for considerations such es:
(1)uncertaintiesincomponentfailurerates;(2)uncertaintiesin comon mode failutt rates;(3 reduced redundercy during testing; (4) operator error during testing; and (5))cor:ponent " wear-out" caused by the testing.
8911280152 891116 PDR ADOCK 05000341 P
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y 2-t 200 DISCUS $10N The hRC's contractor, Idaho National Engineering Laboratory (INEL), reviewed the licensee Owners Group availability analyses and evaluated the adequacy of the existing test intervals, with a consideration of the above five items, for all l
plants. The results of this review are reported in detail in EGG-NTA-8341, "A Review of Reactor Trip System Availsbility Analyses for Generic Letter 83-28, Jtem 4.5.3 Resolution," cated March 1989 and saarized in this re) ort. The results of our evaluation of Item 4.5.3 and our review of EGG-NTA4341 are presented below.
The Babcock & Hilcox (B&W), Cembustion Engineering (CE),l reports either inGe and Westinghouse (W) Owners Groups have submitted topica 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 (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 by comon cause failure and human error.
The ebility to accurately estimate unavailability for very reliable systems was considered extensively in NUREG-0460, " Anticipated Transients Without Scram for Light Water Reactors," and the A7WS rulemaking. The uncertainties of such estimates are large, because the systems are highly reliable, very little experience exists to support the estimates, end 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 low failure probability, such as the RTS, common mode failures tend to predominate and for a number of reasons, accitional 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 testingcouldpossiblyincreasetheprobabilityofacommonmodeIailure through increased stress on the system.
Finally, not all potential failures 1
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.
30 CONCLUSION 0
All four vendors' topical reports have shown the currently configurated 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 testina than monthly will not appreciably lower the estimates of failure probability.
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s Based on our review of the 0wners Groups' topical' reports our contractor's i
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irgktendent an61ysis.: and the findings noted in WREG-0460, we conclude that
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thel existing.intervais', as recomended in the topic 61 reports, for en.line j
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Nat/onal A REY!tW OF REACTOR TRIP SYSTEM AVAILABILITY Eng/neer/ng ANALYSTS FOR r,tNERIC LETTER 83 28, ITEM 4.5.3, RtsotuT!on Laboratory i
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NOTICE The report was prepared as ar* account of work sponsored by an ageery of the United State Government. Neither the Umsted $ ate Governaset not any agency thereof, not any of thou employen, make any wartaaty, apreasd or imphed, or assumes any le841 hakhty or reponsahbry for any thed party's use, or the results of such use, of any informanen, apparsus, product er proc.
es disclosed in this report, or repressnu that its use try such third party would not infange pmately owned nghts.
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f TECHNICAL EVALUATION REPORT: A REVIEh' 0F REACTOR TRIP SYSTEM AVAILABILITY ANALYSES FOR GENERIC LETTER 83-28, ITEM 4.5.3, RESOLUTION 1
P David P. Ma:kowiak Johr. A. Schroecer i
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Idaho Falls, I:aho 83415 FIN 06001: Eval ation of Conformance to Generic Letter 83-28 for ors (Project 2)
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ADSTRACT The Idaho National Engineering Laboratory (INEL) conductec a
'i technical review of the commercial nuclear reactor licensees' responses to the recutrements 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 be covered by an adecuate analysis, will provice the NRC staff with a basis to close out this issue with no further review.
The licensees, as the four vencors' Owners' Groups, sucmittec analyses to the NRC either cirectly in response to GL 83-28, Item 4,5.3, or to provice a basis for requesting changes to the Technical Specifications (IS) that would exteed the Reactor Protection System (RPS) survet' lance test intervals (STIs). To conduct the review, the INEL cefined three crite*ia to dete*?.ine the adecuacy, plant applicability, i
and acceptacility of the results. The INEL examined the Owners Groups' recorts to cetermine if the analyses and results met the estaelished criteria.
Fort St. Vrain's responses to Item 4.5.3 were also reviewee.
ine INEL review results show that all licensees of currently coerating comme *cial nuclear reactors have acecuately demonstratec that their current cc-line RPS test intervals meet the recuirements of GL 83-28, Item 4.5.3.
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$UMARY The two anticipated trautent without scram (ATWS) events at the
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Salem Nuclear Power Plant in February of 1983, focused the attention of the Nuclear Regulatory Commission (NRC) on the generic implications of ATw's events. The NRC then published Generic Letter 83 28 (GL 83-28)
L which listed the actions the NRC required of all Itcensees holding operating licenses and others with respect to assuring the reliability of the Reactor Protection System (RPS). GL 83-28, Item 4.$ 3, required licensees to cemonstrate 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 S3-28, Item 4.5.3, requirements as Owners Groups with reports either in divert response to Item 4.$.3, or with a technical basis for requesting extensions to the surveillance test intervals ($ tis) that generally includeo 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
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analyses anc evaluate the overall adequacy of the existing test intervals.
INEL review results showing general compliance with Item l
4.5.3 will provice the NRC with a basis to,close out Item 4.$.3 without further review.
For the review, the INEL defined three acceptance criteria, reviewed the If:ensees topical reports, contractor review reports, and NRC safety evaluations, and determined the adequacy of the analyses and the RTS availability estimates with regard to the review criteria.
I The INEL review criteria te determine the licensees' Item 4.5.3 c:m;114nce were, (1) the five areas of concern of Item 4.5.3, (2) the analyses' plant applicability, and (3) the NRC's RTS electrical unavailability base case estimates from the ATW$ Rulemaking Paper, SECY+83-293.
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t Each owners Groups' reports were reviewed to ensure that all five areas of concern from item 4.5.3 were either included in the analyses or shown not to be significant with regard to RI$ availability. The INEL review also ensured that the' individual plants' differences from the i
analysis' models were taken into account and their effects we*e shown nct to significantly affect RTS unavailability. The Fort St. Vrain responses to Item 4.$.3 were also reviewed.
The Owners Groups' RTS unavailability estimates were comparee to-the NRC's ATWS Rulemaking generic RTS unavailability estimates to determine the acceptability of the Owners Grovos' conclusions that high RTS availability was demonstrated in the analyses.
The results of the JNEL review showed that all itcensees of currently ocerating commercial nuclear reactors have acetuately cetonstrated that their current on-line surveillance test intervals are consister.t with achieving high RTS availability.
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e ACRONYMS ATw$
Anticipated Transient Withewt $ cram j
B&W Sabcock & Wilcox BNL Brookhaven National Laboratory CE Comewstion Engineering GE General Electric I
HTGR High* Temperature Gas-Cooled Reactor ICSB Instrumentation and Control Systems Branch INEL Icaho National Engineering Lateratory l
LWR Light Water Reactor t
NFSC Nuclear facility Safety Committee NRC Nvelear Regulatory Commission NRR Office of Nuclear Peactor Regulation PORC Plant Operations Review Committee PSC Public Service Com:any cf Colorado PWR Pressurized Water Reactor R$$ MAP Reactor Safety Stu y Methocology Applications Program RPS Reactor Protection System RTS Reactor Trip System SER Safety Evaluation Re:c t STI Sweve111ance Test Interval TER Technical Evaluation Report t-W Westinghouse f
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CONTENTS AsstRitt..............................................................
14 sum *ARY...............................................................
iii ACRONYMS...............................................................
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INTRODUCTION.....................................................
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l.1 Hilt 0rical Background.....................,................
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1.2 Review PWrpole.............................................
3 2.
REVIEW CRITERIA..................................................
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MEVIEW MEIHOOULOGY...............................................
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4 REVIEW RESULTS..................................................
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4.2 CE E*8 ants..................................................
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I 4.3 GE Plants..................................................
9 A.4 WtitinghCWie Plant $........................................
20 4.5 Quantitative Review of Vendors' RTS Unavailab111 ties......
Il 4.6 Fort St. Vrain............................................,
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REVIEW CONCLUSIONS.....................................,........
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TABLES 1.
Comparison of Vendor and NRC RTS Unavailability Eltimates...................................,,...................
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TiteNICAL EVALVATION REPORT: A REVIEW OF REACTOR TRIP $YSTEM l
AVAILABILITY ANALYSES FOR GENERIC LETTER 83 28.
t ITEM 4.5.3 RESOLUTION 1.
INTRODUCTION t
1.1 Historical Backcreund l
In February of.983, two events occurred at the Salem Nuclear Gene ating statier that focusec Nuclear Regulatory Commission (NRC) attention on the generic implications of anticipated transient without scram (ATW5) events.
Firtt, on February 22, during startup of Unit 1 an automatic trip sigral generated as a result of a steam generator low-low level failed to l
cause a reacto* scram. The reactor was tripped manually by an operator almost ceincicentally witn the automatic trip signal, so the faat that the autcmatic trip had ' tiled to csuse a scram went unnoticed.
Three cays later on February 25, both of the scram breakers 6t Unit 1 failed to cpen on an automatic reactor protection system (RPS) scram sig'41. TFe coeraters took action to control this second ATW5 and succeecec in terminating the incicent in about 30 seconds. Subsequent investigation related the failure of the Unit 1 RPS to cause a scram to sticking of the uncervoltage trip attachment in the scram circuit breakers.
As a result of these events the NRC Executive Director for Operatices cirected the staff to undertake three related activities: (1) an evaluation of when and uncer what conditions the Salem plants would be al'c ec to restart; (2) a fact finding report of the events at Salem 1 anc the circumstances leading to them; and (3) a report on the generic ircHeations o' these events, To accress (3) abcve an interoffice, interdirciplinary group was fermec ia.cive'n; members fecm the Office of Nuclear Reactor Regulation's
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Human Factors $afety, Division of Engineering, Division of $4fety Technology, the Office of Inspection and Enforcement, the Office for l
Analysis and Evaluation of Operational Data, and NRC's Region ! Office.
This group published NUREG-1000I as a result of their efforts to resolve j
the following cuestions: (1) is there a need for prompt actical to address similar eovipment in other facilities; (2) are the NRC and its licensees l
1 earning the safety managenent lessons; ano (3) how should the priority and content of the ATW$ Rule be acjusted, j
i 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) reliability.
The actions covered fall into the following four areas: (1) Post-irip Review, (2) j Ecu1 pent Classification and Vencer Interface, (3) Post Maintenance Testing, anc (4) Reactor Trip System Reliability Improvements.
Item 4, above, is aimed at assuring that vender-recoamer.ded reactor i
trip breaker modifications arid associated reactor protection system changes l
are templeted 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, anc to ensure that on-line functional testing of the reactor trip system is performed on all light water reactors (LWRs).
l 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 cetermine 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 reduncancy curing testirg; (a) crerator errors during testing; and ($) component " wear-cut" caused by testing.
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The Babcock & Wilcox (B&W), Combustion Engineering (CE), General Electric (GE),andWestinghouse(W)OwnersGroupshavesubmittedtopical reports either in response to GL 83 28, Item 4.$.3'3 or to previde a basis ice requestin entensions.5,6,7,8.g.10,11RTS surveillance test interval ($T!)
In general, the owners groups' analyses were not cone on a plant specific tasis.
Instead, the analyses aedressed a particular class of reacter trip system and then discussed the applicability of the analysis to specific product lines. The NRC reviewed these reports for, among Cther things, their applicability to GL $3 26, Item 4.$.3 and summarized their findings in Safety Evaluation Reports 12,13 ($ERs).
1.2 Review D *Cose U
this report cocuments a review of the Owners Groups' tcpical reports, the NRC $ERs, anc ether analyses done at the Idaho National Engineering t.ateratcry (INEL) by persenrel in the NRC Risk Analysis Unit of EGLG Idaho, Inc.
The INE!. cencucted the review at the recuest of the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Irstrumentation and Control Systems Brarch (ICSB). The review was performee to determine if the Owners Groups' analyses demonstratec high RTS availattitty for the current test intervals, if the analyses includec the five areas of concern from GL 83 28, and if all of the plants were covered by tne analyses. The results of the review, if all plants are shewn to be covert: cy an adecuate analysis, would provide the NRC with a basis for i
c1:59g cut GL 23 28, Item a 5.3, fer all U.S. commercial nuclear reacters witnout further review, l
The body of this report presents the review and its findings with regare to the stated Cbjectives.
Section 2 describes the criteria usec in tne review to cetermine the aceQuacy of the analyses. The review Pet *c00 logy is discussed in Section 3.
Section 4 presents the review results.
The review conclusiers are given in Section 5.
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REVIEW CRITERIA To conduct a review, one must have criteria, or standards, on which a judgment or decisions may be' based.
In this section, the INEL availability analyses review criteria are presented.
GL 83-28 established the three criteria used in the INEL review.
GL 83-26 stated that:
(1) all licensees et al., (2) must cemenstrate high RT$ avai'acility for the current test intervals by documented review when
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(3) acccunting for such considerations as the five areas of concern listed in Section 1.1.
While GL 83-28 established all three criteria, it only l
defined two of them-who had to de a review and what the review had to take into account. The third and most subjective criterion, "high availability", was not defined.
i To establish a definition of high availability, the INEL used the electrical unavailability base case estimates presents.d in Table A-1 of Appendia A to $ECY-83 293. M Unav&ilability is defined as 1.0 minus avallatitity. A low unavailability is equivalent to a high availability.
Mest ana' lyses calculate a system unavailability rather than an
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availability, Therefore, our criteria for a "high availability" will be expressed in terms of low unavailability for compatibility. These RTS l
ur. availability estimates from Reference 14.were used for two reasons.
First, they were used because they were developed by the NRC's ATWS Task Force as a reevaluation of the bases for the RTS unavailabilities used in ATW$ rule value-impact evaluations. Second, as stated in Reference 14, this hRC analysis I
"... bases the RTS unavailabilities on worldwide experience to
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It is believed that this gives a reasonable estimate of RTS unavailability that incluces the common cause contributions that are believed to dominate. The experience based values are cistributed across the four vender 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 ATW3 analysis provide a framework with which to censider the topical report analyses estimates. The numerical l
estimates in the SECY-83 293 fcr the four venders combined with the five areas of concern from GL 83-28, Item 4.5.3, form the criteria usee for this review to determine if the vencers' analyses and estimates met the recuirements of Item 4.5.3.
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3.
REVIEW METHODOLOGY The INEL conducted this review by examining the vendors' topical reports (References 3, 4, 5, 6, 7, 4, 9, 10, and 11), the technical evaluation reports 15,16,17,18 (TERs) done as a part of the NRC topical report review process, the NRC's $ERs (References 12 and 13), and l
NUREG/CR+5197, Evaluation of Generic Issue 115, " Enhancement of Westinghouse Solid State Protection System."I' This was done for three l
reasons.
First, the reports were esamined to find out whether or not the vendors' analyses addressed the areas of concern from Item 4.$.3 and reflected a high RTS availability.
Second, they were examined to determine what plants were covered by the venders' analyses. Third, the Generic
!ssue 11$ report provided an independent, upcated estimate of the availability of the W solid state RTS for comparison to the review criteria.
Coe the plants covered by the vencers' analyses or the NUREG/CRa$197 analysis, the appropriate analysis and availability were compared to the review criteria established in Section 2.
If the analysis aceawately addressed the areas of concern and demonstrated a high RTS availability, the plant was accepted as having met the requirements of GL S3-28 Item 4.5.3.
The results of the comparisons for plants covered by a vencor analysis are given by vender in Section 4 For plants not directly coverec by a vencor's analysis, an accettable means was founc to extent the analyses to cover the plants. This was cone fer two plants: Clinton 1 (GE) anc Maine Yankee (CE). The means by which.
the analyses were extenced to cover these two plants are aise discusse: by vender in Section 4 One plant, Fort St. Vrain, a high temperature, gas-cooled reactor (HIGR), was not covered by any of the four vendors' analyses and required soecial consideration. The INEL examined the responses from Fort St. Vrain recuirec 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 a.6.
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REVIEW RESULTS This section summarizes the results of the INEL review of the vendors' analyses with regaret to the five areas of concern and plant applicability, The vendors' estimates of RTS availability are compared to the review i
availability criteria. Also, some insights concerning RTS availability, gained from an examination of RTS importance measures from selected PRAS, are examined.
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4.1 B&W plants l
The issues of GL $3*28, Item 4.5.3, were addressed by the B&W Owners j
l Group and the results were submitted to the Nk; by the individual utilities in their responses to GL 83-28.
Topical Report BAW-10167 (Reference 5) was submitted to the NRC to provice a technical basis for increasing the on-lire ST!s 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 aedress the GL 83-28, Item 4.5.3 issues. However, some information that was resolved in the generic letter analysis was not repeatec in the subsequent Topical Recort because it was not relevant to
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the proposec Technical Specification changes. To make BAW-10161 applicable i
to both OL E3-28, Item 4.5.3 and $T!/A0T issues, the Owners Group submitted EAW-10167, Supplement 1 (Reference 6), to the NRC. Supplement I completed the E&W analysis by acdressing all remaining Item 4.5.3 issues. The BAW -10167 and Supplement 1 analyses included the implementation of the
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autcmatic shunt trip on the reactor trip circuit breakers as requirec by GL 83-28. Item 4.3.
The INEL has previously reviewee the BAW-10167 and Supplement 1 analyses and documentec the review in a TER, EGG-REQ-7718 (Reference 15).
For the TER, sensitivity stucies which incluced all of the Item 4.5.3 areas of c:n:ern were conducted on the RTS mecels.
The sensitivity stucy results showed the mocels to be insensitive to variations in the failure rates asscciatec witn the Item 4.5.3 areas of concern.
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.c The INEL reviewee BAW-10167, BAW-10167, Supplement 1, and the TER and determined that the B&W aralyses adequately covered all five areas of concern and that all currently operating B&W reactors are included.
4.2 CE Plants j
Lii:ensees with CE reactors responded to the reevirements of GL 83-28, 1
Item 4.5.3, as the CE Dwners Group by submitting CE NPSD-277 (Reference 3) to the NRC. The NPSD-277 RTS availability analysis specifically included i
all five areas of concern and all current y operating CE reactors except l
Waterford 3, which was not in commercial operation until September 1985.
J The CE Owners Group also submitted CEN 327 (Reference 7) to provide licensees with a basis for requesting RTS STI extensions. This later analysis expanded on the simplified mocols of NPSD-277 to include all RTS in:Ut parameters. All currently operating CE plants except Maine Yankee were covered in the CEN-327 analysis. The CEN-327 $TI analysis specifically included the NPSD-277 analyses of the Item 4.$.3 areas of concern except component " wear-cut" during testing. The CEN-327 analysis showed that the major contributors to RTS unavailability for the four plant classes are common cau&e failures of the trip circuit breakers which are tested on a monthly basis.
In Doth NPSD-277 and CEN-327, the CE RPS designs are grouped into four classes by signal processing and trip device differences, otherwise the i
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 lEN-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 adecuate for Item 4.5.3 resolution for all classes of CE plants.
The INEL has previously reviewed CEN-327 with regard to $TI extension i
effects anc cocumented the review in a TER, EGG-REQ-7768 (Reference 16).
The results of seasitivity studies done for the TER show the models to be insensitive to an orcer of magnitude increase in the cerconent incepencent 8
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failure rates. The insensitivity to increased component failure rates along with the CE analysis results showing trip circuit breaker common cause failures to be the major contributor to RTS unavailability provides a i
a basis for this review to conclude that RTS test-induced component wear-out is not an issue at CE reactors.
1 The INEL reviewed CEN-327 anc the TER and determined that the CE analyses have acetuately covered all five areas of concern or they have been shown not to contribute to RTS unavailability and that all currently cperating CE reactors are included.
4.3 GE plants Licensees with GE reactors responded to the GL 83-28. Item 4.5.3 recuirements as the BWR Owners' Group by submitting NECD-30844 (Reference 4) to the NRC.
The RTS availability analysis specifically included the five areas of concern and covered both generic relay and solic-state RTS designs which includes all currently operating BWRs. GE stated that the relay RPS configurations for BWR plants have the same primary cesign features.
Therefore, the ger.oric relay RTS models used in NECD-30844 do not ciffer significantly from the specific BWR plants. GE used the Clinten 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 E=R Cwners' Group also submitted NE:D-30851P (Reference 8) to the NRC.
Tne analysis in this second report used the base case results from NE:?-3084a to establish a basis for reouesting revisions to the current Technical Scecifications for the RTS.
The INEL had previously reviewed NECD-30844 and NECD-30851P with regard to both Item 4.5.3 and STI extension i
acceptaoility and documented the review in a TER, EGG-EA-7105 (Re'erence 17). Ove to insuf ficient information, the INEL review could net cce:1ete the solid-state RTS review and accepted only the relay RTS analysis results. The NRC reviewed the topical reports and the TER and g
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issued an SER (Reference 12). The NRC accepted the analysis results as a reference for T$ changes related to the RTS and as resolution to GL 83-28 Item 4.5.3, for GE relay pleMs 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.
i 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.
4.4 Westinchouse plants Licensees with Westinghouse reactors did not respond directly to the 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 Westinghouse methodology attempted to balance safety and operability and was applied to a typical Westinghouse four loop reactor plant with a solid state RTS in WCAP-10271.
The methodology was extended to cover RT$s 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 Brookhaven National Laboratory (BNL) and issued an SER (Reference 13) limiting their acceptance to changes to only the analog channel $ tis at l
Westingnouse plants.
Tne W methodology used fault trees to model the RTS.
The models incluced the following five ma,jor contributors to RTS trip unavailability:
L 1.
Unavailability of components due to random failures l
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Unavailability of components due to test a0 l__
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Unavailability of components due to human error i
5.
Unavailsetitty of components due to common cause failure.
While the y analysis did not directly include any sensitivity studies cence ning these five areas, the component unava11 abilities were increased as the test interval length increased. The STI analysis results showed a I
factor of 3 to $ increase in the RTS unavailability estimates for the longer test interval. Two conservatisms exist in the models that are
- retevant, first, no credit was taken for early failures that would be cetectac and, seconc, no credit was taken for the diversity inherent in the y Ris cesign. These two conservatisms, had they be'en inc19eed in the mecel, would cause the increase in the RTS unavailability estimates to be smaller than the ceserved factors.
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 investigatec the effects of this issue en RTS uaavailability. Despite the differences among the other vendors' RTS cesigns, they all found the effects of test induced component waar-out on RTS unavailability to be insignificant. Based on the other vendors' I
analyses, the INEL concluced that the effects of test-induced component wear-out on y RTS unavailatility would also be insignificant, Therefore, i
the INEL consicers all y plants to te coverec by adequate analyses, a.5 0.antitative Review ef Vencers' RTS Availabilities
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Sc far, only the adequacy of the vendors' antlyses has been disc.ssed.
No determination has been mace of the' acceptability of the nu*erical estimates from the various RTS availability analyses.
In this section, the INEL review considers the four Owners Groups' RTS availdbility estimates to cetermine if they are inceed indicativ of i high availaDility."
11
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?e' In Table 1, the four vendors' RTS unavailability estimates are compsred to the review estimates of low unavailability as defined in Section 2.
The MW and GE vendors' estimates are given as an overall RTS unavailability per demand by plant model and RTS type, respectively. The CE and W vendors' estimates art given on a similar' basis with an additional consideration that was not necessary for the MW and GE analyses.
In the CE and W analyses, RTS unavailability was estimated for all input For the CE and W unavailabiitty estimates in Table 1, the INEL parameters.
used the unavailability estimates for high pressurizer pressure, the parameter analyzed in Reference 19 as the limiting parameter for an ATWS in terms of the number of input channels and diversity of trip signal.
The differences in the relative values of the three PWR vendort' RTS unavailability estimates can be attributed to design differences among the RTSs.
M W and CE RTSs have four analog channel inputs for each monitored parameter with four trip logic channels while y 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 parameters in the W design. Also the 2 of 4 trip logic in tb1 MW and CE RTSs is more reliable than the W 1 of 2 trip logic. The combination of these two de:;ign differences make the W RTS unreliability somewhat higher than the other venders' RTS unavailabilities.
The comparison shows the M W, CE, and GE RTS unavailability estimates are icwer than the NRC's estimates while the y 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). tne data uncertainties for both the NRC and Venders analyses, (2) the scarcity of actual RTS failures world wide, (3) the modeling assumptions and simplifications used by both the NRC and the Vendors, and (4) the differing levels of model development between the NRC analysis and the 1
Vendors' analyses and between different Vendors' analyses. These factors 1
1 12
< ;,_s TABLE 1.
COMPARISON OF VENDOR AND'NRC RTS UNAVAILABILITY ESTIMATES a
Unavailability Estimates Unavailability Estimates Vender (Failures / Demand)
(Failure s / Demand) -
B&W Davis'Bessie Model-1E-10" 3E-5d Oconee Class Model 1E-6*
3E-5d CE Plant Class 1 2E-7' 2E-5 Plant Class 2 3E-6' 2E-5 Plant Class 3 3E-6' 2E-5 Plant Class 4 2E-6' 2E-5 GE f
Relay Plants 3E-6 2E-5 Solid-state Plants 3E-6 2E-5 W
Relay Plants 5E-58 d
SE-5 Solid-state Plants 5E-59 d
5E-5 f
All estimates are rounded off to one significant digit.
a.
b.
From Reference 14, Table A-1, base case RTS electrical unavailability i
est mates.
I 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; l
base case test interval, high pressurizer pressure unavailability estimate.
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f.
From Reference 4
(
g, From Reference 19, solid state RTS base case. Applied to relay-plants basec on similarity of design (see Refe ence 11, Section 3.2.2 anc 3.2.3).
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help explain the differences between the Vendors' and the NRC's point L,
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estimates of RTS availability.
l 4.6 Fort St. Vrain Fort St. Vrain responded to GL 83-2C, Item 4.5.3 in a letter to 20 Eisenhut dated November 4, 1983
,,g,ggng
" Existing intervals for on-line functional testing j
required by the Technical Specifications are currently under review by Public Service Company of Colorado (PSC) and the Nuclear Regulatory Commission Region IV staff. The current testino frecuency at Fort St. Vrain has been dictated by the i
Nuclear Regulatory Commission staff." (Uncerline accec)
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 j
following reply to the NRC in a letter to Johnson, dated June 12, 198521;
" Existing 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 5.4.1) are currently being reviewed by the Plant Operations Review Committee (PORC).
This-Technical Specification change is expected 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 proposee changes to the Technical Specifications, on-line functional testing requirements were reviewed based on past experience.
Possible changes to the testing intervals in certain cases where available-test data may support such changes has (sic) been i
discussed at length with the Nuclear Regulatory Commission staff. The Nuclear Regulatory Commission staff has informed 1
Public Service Company of Colorado that no such changes would be ac:eptable at this time."
The INEL review interpreted these responses from Fort St. Vrain to mean the N_RC nas establishec Fort St..Vrain's RTS current test intervals, R
the current test intervals have been evaluated by PSC, and the NRC will not allow changes to the test intervals at this time.
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From these responses, the INEL concluded that Fort St. Vrain has-p ' *;.
' conducted'the review required by GL 83-28, Item'4.5.3, and that the NRC l,;.
considers the PSC and NRC reviews adequate,to meet the Item 4.5.3-requirements.
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5' REVIEW CONCLUSIONS
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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 individual responses to Item 4.5.3.
The analyses in the topical report have shown the. currently configured RT$s 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, accitional analyses, and the individual licensee submittals with regard to GL B3-28, Item 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 adeawately demonstrated that their current RTS test intervals are consistent with achieving high RTS
. availability.
l' 16 n.,-
.e 6.
REFERENCES 1.
U.S. Nuclear Regulatory Commission, Generic Implications of ATWS Events I
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 Actions Based on Generic Implications of Salem ATWS Events, Generic Letter 83-28, July 8, 1983.
3.
Combustion Engineering, Reactor Protection System Test Interval Evaluatien, Task 486, CE NPSD-277, Decemoer 1984 p
4.
S. Visweswaran et al., BwR Owners' Group Response to NRC Generic Letter-83-28, Item 4.5.3, NECD-30844, January 1985.
5.
R. S. Enzinna et al., Justification for increasino the Reactor Trio 1
System On-line Test Interval, BAW-10167, May 1986.
6.
R. S. Enzinna et al., Justification for increasing the Reactor Trio System On-line Test Interval, $upplement NumDer 1, 8AW-10167, Supplement Numoer 1, FeDruary 1988.
7.
Combustion Engineering, RPS/ESFAS Extended Test Interval Evaluation, CEN-327, May 1986.
8.
W. P. Sullivan et al., Technical Soecification Imorovement Analyses for BWR Reacter Drotection 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 1983.
10.
R. L. Jansen et al., Evaluation of Surveillance Freauencies and Out of Service Times for the Reactor Protecti6n Instrumentatier. System.
Supolement 1, aCAP-10271, Supplement 1, July 1983.
11.
R. L. Jarsen et al., Evaluation of Surveillance Frecuencies and Out of Service Times for the Reactor D otection Instrumentation System.
L Sueolement 1-e-A, wCAP-10271, Supplement 1-P-A, May 1986.
l 12.
U.S. Nuclear Regulatory Commission Memorandum, G. C. Lainas to E. J.
I Butcher ~, Accootance for Refe-encinc of General Electric Company (GE)
L Tepical R.eports NECD-30844, "SaR D ners' Group Response to NRC Generic Letter 83-28," anc NECD-3085;P, "Tecnnical Specification Improvement Analyses for SWR Reactor Protection System."' April 28, 1986.
L 13.
U.S. Nuclear Regulatory Commission Letter, C. O Thomas to J. J.
Sheppard, Acceotance for Referencing of Licensing Topical Report WCAP-lC271, " Evaluation of Surveillance Frecuencies anc Out of Service T ees for tne Reactor ;-otection Inst *umentatier Systems." Feeruary 21, 1985.
l 17
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14.
U.S. Nuclear Regulato y Commission, Amandments to 10 CFR 50 Related to
-Antietoated Transients Without Scram (ATW5) Events, SECY-83-293, July 19, 1983.
15.
J. P. Poloski and S. D. Matthews, Review of B&W Owner's Greue Analyses for Increasino The Reactor Trio System On-line Test interval, EGG-REQ-7718, September 1988.
16.
D. P. Mackowiak and-B. L. Collins, A Review of the Combustion Engineerino Evaluation For Extending the RP5 and E5FA5 Test Intervals, EGG-REQ-7768, Septemeer 1988.
17.
R. E.- Wrignt and B. L. Collins, A Review of the BWR owners' Group Technical Specification Improvement Analyses for the swr Reactor Protection System. EGG-EA-7105, January 1986.
18.
R. E. Wright and B. ~ L. Collins, A Review of the BWR Owners' Grevo Technical Soecification Improvement Analyses for the BwR Reactor Protection System. EGG-EA-7105, Rev 1, Maren 1987.
19.
D. A. Reny et al., Evaluation of Generic Issue 115. Enhancement of the Reliability of Westireneuse solic state Protection Systems, NUREG/CR-5197, January 1989.
- 20.. Public Service Compary of Colorado Letter. O. R. Lee to D. G.
Eisenhut, Respense to Generic Letter 83-28, November 4, 1983.
- 21. Public Service Comoany of Colorado Letter, J. W. Gham to E. H.
Johnson, Response to Generic Letter 83-28. June 12,1985.
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TECHNICAL EVALUATION REPORT: A REVIEW 0F REACTOR TRIP SYSTEM AVAILABILITY' ANALYSES FOR GENERIC LETTER 83-28, 1 TEM 4.5.3. RESOLUTION
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March 1989 Davitt P. Mackowiak John A. Schroeder g
March 1989
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Regulatory and Technical Assistence EGAG Idaho, Inc.
P. O. Box 1625 Idaho Falls, ID 83415 06001-
...............e................u..<.c.
Instrumentation and Control Systems Branch Technical Evaluation Report Division of Engineering and System Technology Office of Nuclear Reactor Regulation
- **. a u. p u U.S. Nuclear Regulatory Comission Washington DC 20555
...s.........n 4......e m The Idaho National Engineering Laboratory (INEL) conducted a technical review of the comercial nuclear reactor licensees' responses to the requirements of the Nuclear Regulatory Comission's (NRC's) Generic Letter 83-28 (GL 83-28). Item 4.5.3.
The results of this review. if all plants are shown to be covered by an adequate analysis, will provide the NRC staff with a basis to close out this issue with no further review.
The licensees. as the four vendors' Owners' Groups, submitted analyses to the NRC either directly in response to GL 83-28, Item 4.5.3, or to provide a basis for requesting changes to the Technical Specifications-(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 acceptability 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 were also reviewed.
The INEL review results show that all licensees of currently opera-ting commercial nuclear reactors have adeaustely demonstrated that their current on-line RPS test intervals meet the requirements of GL 83-28, Item 4.5.3.
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