ML19325F168

SER Accepting Util Response to Generic Ltr 83-28,Item 4.5.3 Re Reactor Trip Sys Reliability for All Domestic Operating Reactors

ML19325F168
Person / Time
Site:	LaSalle
Issue date:	11/06/1989
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML19325F167	List: ML19325F166 ML19325F168
References
GL-83-28, NUDOCS 8911140343
Download: ML19325F168 (2)

Text

_

)

ENCLOSURE 1

)

SAFETY EVALUATION REPORT GEhERIC Lt.Tuf M. II ;M a.5.3 REACTOR TRIP SYSTEM RELIABILITI H F A.L E04ESTIC QFERATING REACTORS j

)

1.0 INTRODUCTION

On February 25, 1983, both of the scram circuit breakers at Unit 1 of the 4

Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal l

from the reactor protection system (RPS). This incident was terminated l

manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers was detemined to te releted to the sticking of the undervoltage trip attachment.

Prior to this incident, on Febraary 22, 1903, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was generated besed on steam generator low-low level during plart startur.

In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip.

following these incidents, on february 26, 1983, the NRC Executive Director for Operatier.s (EDO), directed the staff to investigate and report on the generic implications of these cceurrences at Unit 1 of the Salem Nuclear Power 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 investigatior, the Comission (NRC) reeuested (by Generic Letter 83-28 dateo July 8,1963' all licensees of operating reactors, applicants for an operating license, end holders cd construction permits te respond to generic issues l

raised by the ar.alyses of these two ATWS events.

Tbt licensees were reouired by Generic Letter 83-28. Item 4.5.3 to confim that I

on-line functional testing of the reactor trip system (RTS), including indepentient testing cf the diverse trip features, was being perfomed at all plants.

Existing intervals for on-lint functional tes.ing required by Technical Specificatiers were to be reviewed to determine if the test intervals were adequate for eehieving high RTS availability when accountint for considerations such as:

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

c i

l 2.0 DISCUS $10N The NRC's contractor, Idaho National Engineering Laboratory (INr.L), reviewed the licensee Owners Croup e.vailability analyses and evaluated the adequacy of the existing test ins.rvals, with a consideration of the above fiv9 iters, for i

all plants. The results of this review are reported in detail in EGG-NTA-83al, l

"A Review of Reactor Trip System Availability Analyses for Generic Letter 83-26. Item 4.5.3 Pesolutior," dated Parch 1989 and sumarized ir this report.

The results of our evaluatier, cf item 4.5.3 and our review of EGG NT/-8341 art presented below.

8911140343 891106 ADOCK0500g{3 l

DR l

I v

i

. < i The Babcock & Wilcos (B&h'), Combustion Engineering (CE), General Electric i

(GE), and Westinghouse (W) Owners Groups have submitted topical reports either 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 (STI).

The owners groups' analyses addressed the adequacy of the existing i

intervals for on-line functional testing of the RTS, with the considerations required by Item 4.5.3, by qusntit6tively 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, t

The ability to accurately estimate unavailability for very reliable systems ws considered extensively in NUREG 0460,

" Anticipated Transients Without Scram for Light Water Reactors", and the ATWS rulemaking. The uncertainties of such estimates are large, because the systems are highly reliable, very little experience exists to support the estimates and comon cause failurt l

l 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 systens with low failure probability, such as i

the RTS, coroon mode failures tenn to predominate and for a number of reasons, additional testing will nct appreciebly lower,RTS unavailability.

First, testir.g more freoutntly than weekly is generally impractical, and even so the increased testing could at best lower the failure probability by less than a factor of four compared to monthly testing.

Secondly, increased testing could possibly increase the prebability cf a comon mode f ailure through ine:,ased stress en the systet.

Finally, tct all potential failures are detectable by testing.

In sumary, NUREG 0460 provides additional justification to dencostrate that the current monthly test intervals are adequate to maintain high FTS availabili 9

3.0 CONCLUSION

i l

All four venders' topical reports hast shown the currently configured RTS to l

be highly reliable with the current rcnthly test intervais-Our contractor has j

reviewed these analyses and performed independent estimates of their own which i

conclude that the current test intervals provide high reliability.

In addition, l.

the analyset in NUREG-0460 beve shown that for a number of reasons, more l

frequent testing than conthly 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 theexisti.19 intervals,ndthefindingsnotedinNUREG-0460,weconcludethat as recomended in the topical reports, for on-line functieral testing are contfstent w'th achieving high RTS availability at all i

operating reactors.

l l

l u

ENCLO5URE 2 EGG-NTA 83Al j

March 1989 l

, 4

)

)

TECHNICAL EVALUATION REPORT

/deho i

Nations /

A REvrEw Or REACTOR TRIP SYSTEM AVAILAB:L:TY ANALYSES FOR GENERIC LETTER 83 28, ITEM 4.5.3, Eng/neer/ng RESOLUTION Laboratory i

1

'deresef

o. vie p. x eie 4ae i

e, rne v e.

i John A. Schroeder Deca ~r.ra of Earr;,

i l

t b

I i

l i

t 1

i 1

1 n

I n*

• gggy _

Prepared for the U. S. NUCLEAR REGULATonY COMMISSiO^

nove te~ow uw 00f Costs::

%: %f AC:7 NiD0tti l

l 1

EG3 NTA 5342 i

e r

TECHNICAL EVALUATION REPORT: A REVIEW OF REACTOR TRIP SYSTEM AVAILAE!LITY ANALYSES FOR GENERIC LET*ER 63-25, i

ITEM 4.5.3, RESOLUTION i.

i Davde P. Ma:Lewial John A. $ hececer i

EGiG I:ahe, Inc.

Icabe Falls, !:an: 63415 l

l l

FIN 06001.

Evaivation of Confern.ance te Generi: Le er 63-2B for ors (Freject 2)

L 4,

L j

ABSTRACT The Idaho National Enginc4 ing Laboratory (INEL) condu:ted a techni:41 review of the commercial nucle &r reactor licensees' responses to the recuirements of,the huclear Regulatory Commission's (NRC's)

Generic Letter 63-26 (GL S3-26), Item 4.5.3.

The results of this review, if all plants are shown te be covered by an adequate analysis. will provice the NR; staff with a basis to close out this issue with no fuether review. The licetaees, as the four vend:rs' Owners' Groups, submitted analysts to the NR; either cirectly in response to GL $3-25, Ite* 4.5.3, or to provice a basis for requesting changes to the Te:9nical Spe:ications (TS) that would extted the Reactor Protection System (RP$)

surveillance test intervals ($ tis:

To conduct the review, the INEL ce'daec three criteria to dete* sine the adecuacy, plant applicability, and attert&oility of the results. The INEL esamined the Owners Grovos' reperts to cetermine if the analyses and results met the establishec criter**

Ft t St. Vrain's respo.ses to Item 4.5.3 were also reviewed.

The lNEL review results show that all licensees of Currently operating comter:ial n.: lear rea:ters have ade:vately cemenstratec that their currer.t et-line R S test.ntervais meet the re:virements of GL $3-28, Itee 4.5.3.

m

- - a

SUMW.ARY s

i The two anticipated transient without scram (ATWS) everts at the

[

$alem Nv: lear Fomer Plant in February of 1983, focused the attention of the Nuclear Regulatory Commission (NRO) on the generic implications of ATWS events. The NRC then published Generic Letter 83-28 (GL B3-26) which listed the actions the NRC required of all licensees holding I

operating licenses and others with respect to assweing the reliability of f

the Reactor Pr tection System (RPS). GL 83-18. Itai 4.5.3, reRuired i

licensees to cemenstrate by review that the current on-line functional testing interva1> are consistent with achieving high reactor trip system (RTS) availability.

The licensees respended to the GL 83-28, Item 4.5.3, recaireeents as Owners Groups with eeperts either in cirect response to Item 4.5.3, or with a technical basis for requesting extensions to the i

svevelan:e test intervals ($T!s) that generally in:19ded the Item s.5.3 i

re:wirec c:aiews.

5 The NRO's Instrueentation and Control Systems Branch (10$B), Of fice of NL: lear Rea: tor Regulation (NRR), recuested the Idaho National E*gdret ing Lat: at:*y (INEL) te review the licensee availability tralyses ame eva' sate tne overall acecua:y of the existing test intervais.

INE. review reswits shewing general compliance with Ite-t d.6.3 will Or0 Vide tre NR; n'th a tasis to clost oVt Item 4.5.3 without

!;rthe-restew.

r r. ne cevie=, tre INEL cefined three a: eptan:e criteri;, re it.ed e

the l':ensees teo':41 re: rts, centta: tor review reperts, erd N4 safety evaluations, and determined the acequacy of the analyses Anc the RTS availatility estimates with rega d to the review criteria.

The INEL revie criteria te determine the licensees' Item 4.5.3 c: ;11an:e were, (1) the five areas of concern of Item 4.5.3, (2) the analyses' plant appit: ability, and (3) the NR;'s RTS electrical uravailabili y base case estimates free the ATWS Rulemaking Paper, l

t SE Y-E3-293.

l l

l l

1 l

I

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 RTS availability. The INEL review alt-o 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.$.3 were aire revie*:ed.

The 0.ners 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 availatility was demonstrated in the analyses.

The results of the INEL review showed that all licensees of currently ocerating comme cial nuclear reactors have acequate'y ce crstrated that their current on-line surveillance test intervals are censistemt with achieving high RTS availability, i

iv l

e

'c

,e.,

.-[*

ACRONYMS ATW$

Anticipated Transient Without Scram B&W Babcock & Wilcox BNL Brookhaven National Laboratory l

CE Cemeustion Engineering l

GE General Electric HIGR High-Temperature Gas-Cooled Reactor IC5B Instrumentation and Control Systems Branch INEL Idaho National Engineering Laboratory f

LWR Light Water Rea: tor l

NF$C Nu: lear Facility Safety Com?.ittee NRC Nuclear Regulatory Commission

'NER Office of Nv: lear Reactor Regulation PCRC Plant Operations Revie. Committee i

PSC Public Service Comoany of Coloraco FWR Pressurizec Water Rea: tor

[

t R$5MA P Rea:ter Safety St;:j Met'.o:ciogy A;pli:ations Program l

RPS Rea: : Pr:tection System Ei!

Rea: tor Trip System i

SER Safety Evai'.at'en Re:: t

$71 Sweveillan:e Test interval f

TER Technical Evaluation Re: ort j

k W

Westinghouse l

t 6

i i

Y

o

.o b

CONTENTS

. s AE$ TRACT...............................................................

ii

$UM.W.ARY................................................,,,,,,,,,,,,,,

qqq 1

ACRONYM $.............................................................

L-V 1.

INTRODUCTION.....................................................

1 i

e 1.1 Historic 41 Background......................................

1

[

t I

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

3 i

i 2.

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

4 i

t 3.

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

6 4

R E V I E W R E S 'J L i $.................................................

7 i

4.1 B&W EI4r45 8

4.2 CE PI4r.t5 7

i 4.3 GE EI4P.t$

9 4.4 Westinghouse Elarts........................................

10 4.$

Qw3Mtit4*ive Review of Vender $' RT$ Un4vailabilitit$......

Il i

4.6 FC't $t Vrtin............................................

}4 6.

REVIEW CONCLUSIONS..............................................

16 6.

m.I...,4.LS i

M t%;

17

)

TABLE $

l.

Comperison O' Vender and NRC RTS Unavailability E5'imatt$

13 L

l

[

1 l

i A

l l

bi l-l 1'

i

L o

TECHNICAL EVALUATION REPORT: A RFVIEW OF REACTOR TRIP $YSTEM f

c AVAILABILITY ANALYSES FOR GENERIC LETTER B3-28 s

2

! TEM 4.5.3 RESOLUTION l

t 1.

INTRODUCTION l

l.} Historical Background 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.

I i

First, on February 22, during startup of Unit 1 an automatic trip sigrat generated as a result of a steam generater low-low level f ailed to casse a reacter scram.

The reactor was tripped manually by an operator j

alrest coincioentally with the automatic trip signal, so the fact that the

{

avicmetic trip had failed to cause a scram went unnoticed.

Three days late

• en Februa y 25, both of the scram breakers at Unit 1 failed to ope 9 on an automatic reactor protection system (RPS) scrae

(

s'; eat.

Tre coeraters teck action to centrol this second ATW5 and succeeced in terrinating the incidert in about 30 seconds.

Subsecuett investigation related the failure of the Unit 1 RPS to cause a scram to stickir; of the undervoltage trip attachment in the scram circuit breakers.

As a result of these ever's the NRC Executive Director for Operatier.s directed the staff to undertake three related activities:

(1) an evaluation of when and under what conditions the Salem plants would be i*

allowec 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 imcu catiens o' these events.

To accress (3) abcve an interoffice, interdisciplinary group was formed it.clud'n; remoers from the Office of Nuclear Reactor Regulation's 1

t J

o (NRR's) Division of Licensing, Division of Systems Integration, Division of Human Factors Safety, Division of Engineering, Division of Safety

{

Technology, the Office of Inspe: tion 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) in there a need for prompt actions to address similar equipment in other facilities; (2) are the NRC and its licensees learning the safety managen.ent lessons; and (3) how should the priority and content of the ATWS Rule be adjusted.

j i

As a result of the NUREG-1000 findings, the NRC issued Generic Leuer 83 282 (GL B3-28).,The actions described in GL 83-28 address

~

i issues related to rea:ter trip system (RTS) reliability. The actions coverec fall into the following four areas: (1) Post-Trip Review, (2) e Er.ip'rer.t Classifi:ation and Vender Interface, (3) Post-Maintenan:e Testing, anc (4) Rea: tor Trip System Reliability In.provements.

Item 4, above, is aimed at assuring that vencer-recommenced rea: tor trip bretter modificetions and 45sociated reactor protection system changes are C:ft:leted in pressurized water reactors (PWRs), that a comprehensive preg am of preventive'r.aintenan:e anc surveillante testing is implementec for tre reactor trip breake's in PWR5, that the shunt trip attachment 4:tivates automatically in all PWRs that use circuit breakers in their l

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

l The s;e:ific requirements of GL 83-25, Item 4.5.3, are that existing i

intervals for on-line functional testing required by Te:hnical l

See:ifications shall be reviemec 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) redy:ed redundancy during i

testirg; (4) operato" errors during testing; and ($) component " wear-cut" caused by testing.

l E

l

.-.--n

The Bab:o:L & Wilcox (B&W), Combustion Engineering (CE), General Ele:tric (GE), and Westin; house (W) Owners Groups have submitted topical re;crts either in response to GL S3-28, Item 4.5.3'3 or to provide a basis for requesting Ri$ surveillance test interval ($T!)

extensions.5,6,7,6,9,10,11 In general, the owners groups' analyses were not cone on a plant specific basis.

Instead, the analyses addressed a t

particular class of reactor triD system and then discussed the applicability of the analysis to specific product lines.

The NRC reviewed these reports for, among other things, their applicability to GL 63-28, Ittm 4.5.3 and summarized their findings in Safety Evaluation i

Reports 32.23 ($ERs).

1.2 Revie. Pu oese This rer:rt c::urents a review of the Owners Grours' topical reports, see NRC $ERs, and other analyses done at the Idaho National Engineering Laterateey (INE'.) by persentel in the NRC Risk Analysis Unit of E3&G Icaho, Inc.

The INEL con ucted the review at the re: vest of the U.S. Nu: lear Regulatory Commission, Office of Nuclear Reactor Regulation, Instrueentation and Control Systems Bran:h (IC$B). The review was L

pe f:rmed to cetermine if the Owners Groups' analyses demonstrate: high Ri$

l availacility for the current test intervals, if the analyses include: the fhe a eas of cen ern free GL S3-25, and if all of the plants were : vered by the analyses.

The results of the review, if all plants are shown to be

(

ceve ed by an adecuate analysis, would provide the NRC with a basis for c1csir; out GL S3-25 Iter 4.5.3, fer all U.S. comeerciel nu: lear rea: tors with;ut fu ther review, The body of this ree:rt presents the review and its findings with i.

regar: to the state: coje:tives.

Se: tion 2 describes the criteria used in the review to determine the adequacy of the analyses. The review methocology is discussed in Se: tion 3.

Section 4 presents the review results. The revie :onclusiens are given in Section $.

3

9 C

2.

REVIEW CRITERIA o.'

l To conduct a review, one must have criteria, or standards, on which a judgnent or decisions may be based.

In this section, the INEL availability j

analyses review criteria are presented.

t GL $3 26 establisbec the three criteria used in the INEL review.

GL B3-26 stated that:

(1) all licensees et al., (2) must dem:nstrate high 3

RTS avai'acility for the current test intervals by documented review when (3) acceunting for such consice*ations as the five areas of concern listed in Section 1.1.

While GL S3-28 established all three criteria, it only cefined two of them who had to do a review and what the review had to take into ac: cunt. The third and mest subjective criterion, "high availability", was not defined.

To es'.a:11sh a definition of high availability, the INEL used the elect *ical unavailabil'ty base case ettimates presented in Taole A-1 of A;cendix A to SECY-G3-293.I' Unavailability is defined as 1.0 minus availability. A low unavailability is equivalent to a high availability.

Mest analyses calculate a system unavailability rather than an availability. Therefore, our criteria for a "high availability" will be excressed in terms of low u* availability for compatibility. These ATS unavailability estimates free ft'ence 14 were used for two reasons.

o First, they were used because they were developed by the NRC's ATWS Task Fe ce as a reevaluatien of the bases for the RTS unavailabilities used in A~as rule value-imcact evaluatiers.

Second, as stated in Reference 14, this hRC analysis

"... bases the RTS unava11 abilities on worldwide experience to l

cate.

It is believec that this gives a reasonable estimate of RTS unavailability that includes the comraon cause contributions l

thst are believed to dominate.

The experience based values are l

distributed across the four vender designs basLd on a cemparative reliability analysis that evaluates the major cif'erences among the designs."

r - - -.

p Lt s..

i

[

The esttnates from the NRC ATWS analysis provice e framework with c'

which to censider the topical report analyses estimates.

The numerical estimates in the SECY 83-293 for the four vendors combined with the five i

i areas of corcern frem GL 83 28 Item 4.5.3, form the criteria used for this t

review to determine if the venders' analyses and estimates met the

[. -

requirements of item 4.5.3.

i v.

e i

l 1

I i

e

?

i i

i i

I i

p b

l I

W I

h

/t j'

t co n

I*

3.

REVIEW METHODOLOGY The INEL conducted this review by examining the vendors' topical reports (References 3, 4, 5, 6, 7, 8, 9, 10, and 12), the te:hnical evaluation reports 15,16,17,18 (TERs) done as a part of the NRC topical report review process, the NRC's SERs (References 12 and 13), and NUREG/CR-5197, Evaluation of Generic Issue 135, " Enhancement of Westing *:use Solid State Proten..'n System."II This was done for three reasons.

First the reports 4

e.amined to find out whether or not the venders' analyses addressed ti,.res> of concern from Item 4.5.3 and refle:ted a high RTS availability.

Second, they were examined to determine what plants were covered by the vendors' analyses. Third, the Generic Issue 115 report providec an indepencent, upcated estimate of the availatility of the W solic state RTS for comparison to the review criteria.

For the plants covered by the vencers' analyses or the N'JREG/CR-5197 analysis, the 40propriate analysis and availability were compared to the review criteria established in Se: tion 2.

If the analysis adequately addressee the areas of eencern and, cemonstrated a high RTS availability the plart was ac:epted as having met the requirements of GL E3-28, Item 4.5.3.

The results of the comparisons for plants covered by 4 vencer analysis are giv9n by vendor in Section 4 For plants not diee:tly coverec by a venoor's analysis, an acee: table means was founc to extend the analyses to cever the plants. This wat e:ne

'e-two piants:

Cidnton ; (GE) anc Maine Yankee (CE).

The means by which the analyses we'e extenced to cover these two plants are also discJssed by vender in Section 4 One plant, Fort St. Vrain, a high temperature, gas-cooled reactor (HTGR), was not covered by any of the four vendors' unalyses and requitec see:ial censideration. The INEL examined the respsnses from Fort 5:. Vrain re:virec by GL 83-28, Item 4.5.3 to determine if the responses demonstrated an a::e:tably high RTS availability. The review of the Fort St. Vrain responses is giver. in Section 4.6.

6

4 4

REVIEW RESULTS This section sum:..arizes the results of the INEL review of the vendors' analyses with regard to the five areat of concern and plant applicability.

f The vendors' estimates of RTS availability are compared to the review availability criteria. Also, some insights concerning RTS availability, gained from an examination of RTS impo:tance measures from selected PRAs, 3re examined.

4.1 B&W Plants The issues of GL 83-28, Item 4.5.3, were addressed by the B&W Owners Group anc the results were submitted to the NRC by the individual utilities in their responses to GL 83-28. Topical Report BAW-1016*/ (Reference 5) was submittee to the NRC to provide a technical basis for increasing the en-line STIs and allowed outage times (A0Ts) for B&W RTS instrument strings. The analysis pres 2nted in BAW-10167 was built upon the previous 5

analysis cone to address 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 Report because it was,not relevant to tne proposec' Technicai Specification changes. To make BAW-10167 applicable to both GL E3-28, Item 4.5.3 and STI/A0T issues, the Owners Group submitted BA's-;0'.67, Supplement 1 (Reference 6), to the NRC.

Su;;1ement I completed the B&W analysis by accressing all remaining Item 4.5.3 issues. The ban -;0167 and Supplement 3 analyses included the implementation of the automatic s*urt t-ip on the reactor trip circuit breakers as required by GL 83-28, Item 4.3.

Tne INEL has previously aviewed the BAW-10167 anc Supplete t I a

analyses and d:cumented the review in a TER, EGG-REQ-7718 (R n ence 15).

For tne TER, sensitivity stucies which included all of the Item 4.5.3 areas of concern were conducted or. the RTS mocels. The sensitivity stucy results s* owed the models to be insensitive to variations in the f ailure rates associatec with the Item 4.5.3 areas of concern.

7

c 4

i The INEL reviewee BAW-10167, BAW-10167 Supplement 1, and the TER and determined that the B&W analyses adequately covered all five areas of concern and that all currently operating B&W reactors are included.

s 4.2 CE Plants Licensees with CE reactors responded to the requirements of GL 83-28 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 all five areas of concern and all current 1y 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 reavesting RTS STI extensions. This later aralysis excanced on the simplified models of NPSD-277 to include all RTS input parameters. All currently operating CE plants except Maine Yankee were coverec in the CEN-327 analysis. The CEN-327 STI analysis specifically included the NPSD-277 analyses of the Item 4.5.5 areas of concern except comoonent " wear-cut" during testing. The CEN-327 analysis showed that the major contributors to RTS unavailability for the fc" plant i

classes are common cause failures of the trip circuit breakers which ere tested on a monthly basis, e

In both NPSD-277 anc CEN-327, the CE RPS designs are grouped into four classes by signal processing and trip device dif ferences, otherwise the l

logic and ohysical 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 l

CEN-327, Waterford 3 is includee in RPS Plant Class 3.

Between NPSD-277 l

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

The INEL has previously reviewed CEN-327 with regard to ST1 extension effects and cocumented the review in a TER, EGG-REQ-7768 (Reference 16).

The results of seasitivity stucies done for the TER she, the models to be insensitive to ar orcee cf magr.itude increase in the coeoorent incependent 8

+e

+

e--

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

The INEL reviewed CEN-327 and the TER and determined that the C:

ar.alyses 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 reactors are included.

L 4.3 GE Diants Licensees with GE reactors responced to the GL B3-28 Item 4.5.3 recairements as the SWR Owners' Group by submitting NECD-30844 (Reference 4) to the NRC.

The RTS availability analysis specifica'.1/

incluced the five areas of concern and covered both generic relay and sclic-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. Tnerefore, the generic relay RTS models used in AECO-30544 ce net cif'er significantly

.am the specific BWR plants. GE usec t*.e Clinton 1 era ings for the solic-state RTS moceis. Since C11nten 1 is currently the only GE plant with a solid state RTS, no plant unique analysis is necessary.

I The E.R C aers' Group also suceitted NECD-30851F (Reference 8) : the l

NRC.

Tne analysis in this second report used the base case results from NECD-30544 to establish a casis for recuesting revisions to the current Technical Specifications for the RTS. The INEL had previously reviewed NECD-30544 and NECD-30551P with regard to both Item 4.5.3 and STI extension I

acce;tatility and documented the review in & TER, EGG-EA-7105 l

(Re'erence 17).

Due to insufficient information, the INEL review could not ecm:'ete the solid-state RTS review and accepted only the relay RTS l

analysis results. The NRC reviewec the topical reports and the TER and l

l l

9 l

l 1

e a

r

issued an SER (Reference 12). The NRC accepted the analysis results as a reference for TS changes related to the RTS and as resolution to GL 83-28, item 4.5.3, for GE relay plants caly. The INEL later completed the solid state RTS analysis review and issued Rev 1 to the TER (Reference 15), thus accepting the analyses for all classes of GE plants.

l 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.a Westinghouse Plants Licensees with Westinghouse reactors did not respond directly to the u

requirements of GL $3-28, Item 4.5.3.

Prior to the Salem ATWS, they had submitted WCAP-10271 (Reference 9) to the NRC to provide a basis for re: vesting 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 RTSs for tw'c, 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 STis at Westinghouse plants.

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

L 1.

Unavailability of components due to random failures 2.

Unavailability of components due to test 10

J a

3.

Unavailability of components due to unscheduled maintenance j

4..

Unavailability of components due to human error 5.

Unavailability of components due to common cause failure, j

While the W analysis dic not directly include any sensitivity studies conce-ning these five areas, the component unavailabilities were in reased 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 crecit was taken for early failures that would be c2tectec and.seconc, no credit was taken for the diversity inherent in the W RTS cesign.

These two conservatisms, had they been incluced in the mecel, woulc cause the increase in the RTS unavailability estimates to be smaller than the ocserved facters.

i I

Test-induced component wear-cut was not tcdressed in any menner in the h' RTS analysis. However, the RTS analyses done by the other vendors, Referent *s 3, 4 and 6, specifically investicated the effects of this issue er RTS unavailatility.

Despite the cifferences among the other vencers' RTS cesiges, they ali founc the effects of test induce component wear-ou; en RTS unavailability to be insignificant. Based or the other vendors' analyses, the INEL concluced that the effects of test-incucec'comperent wear-cut en W RTS unavailatility would also be insignificant. Therefere, tre INEL consicers all W plants 10 be coveret by aceQuate analyses.

a.5 0.antitative Review of Vencoes' RTS Asailabilities Sc far, only the adequacy of the vendors' analyses has been cisc.ssed.

No determination has been made of the' acceptability of the nu e- :al estimates from the various RTS availability analyses.

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

e

4.

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

The B&W and GE vendors' estimates are given as an overall RTS vnavailability 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 necessary for the B&W anc GE analyses.

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

For the CE and y unavailability estimates in Table 1, the ]NEL used the unavailability estimates for high pressurizer pressure, the parameter analyTed 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 venders' RTS unavailability estimates can be attributed to design differences among the RTSs. B&W anc 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 B&W 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 the B&W and CE RTSs is more reliable than the W 2 of 2 trip logic.

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

The comparison shows the B&W, CE, and GE RTS unavailability estimates are lower than the NRC's estimates while the y estimates are the same as the NRC's.

The ]NEL review recogni:es the Vendors' estimates and the NRC's estimates are influenced by a number of factors.

These facters include, (1) tne data uncertainties for both the NRC and Venders analyses, (2) the sea-city of actual RTS failures world wide, (3) the modeling assumptions and simplifications used by both the NRC and the Vendors, and (4) the ciffering levels of mocel development between the NRC analysis and the Venders' analyses and between dif ferent Venders' analyses.

These factors 12

t e

J TABLE I.

COMPARISON OF VENDOR Afl0 NRC RTS UNAVAILABILITY ESTIMATES 8 Vender RTS NRC RTS b

J Unavailability Estimates Unavailability Estimates Vender (Failures / Demand)

(Failures / Demand)

B&W l

Davis Bessie Model IE-10" 3E-5d C

d Oconee Class Model IE-6 3E-5 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 I

Relay Plants 3E-6 2E-5 I

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

Relay Plants SE-59 d

SE-5 d

Solid-state Plants SE-59 SE-5 a.

All' estimates are rounded off to one siqnificant digit.

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

From Reference 5, base case.

l-I d.

Includes automatic shunt trip on the reactor trip circuit breakers.

l l

e.

From Refe*ence 7, Tables t. 1-1, 4,2-2, 4.1-3, and 4.1-4, respectively; base case test interval, high pressurizer pressure unavailaDility estimate.

f.

From Reference 4 1?

L g.

F.om Referer..:e 19, solid stata RTS base case. Applied to relay-plant i,

based on similarity of cesigr. (see Reference 11, Section 3.2.2 anc 3.2.3).

l

('

P 13 l [.

~. -

v e' o,

{.1 help explain the differences between the Vendors' and the NRC's point estiraates of RTS availability, i

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, stating:

" Existing intervals for on-line functional testirg 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 testing frequency at Fort St. Vrain has been dictated ! '

Nuclear Regulatory Commission staff." (Uncerline accec), the In response to a request for information from the NRC conce.ning the Fort St, Vrain responses to GL 83-28 previously sent, PSC sent the following repl> to the NRC in a letter to Johnson, dated June 12, 196521:

" Existing intervals for the on-line testing recu: red by the Technical Specifications were reviewed by Public Service Company of Colc cdo.

A Technical Specification change te Limiting Conditions for Operation 4.4.1 (Plant Protective Syster.) and its associated surveillance requirements (SR 5.4.1) are currently being reviewec by tne Plant Operations Peview Committee (PORC).

j This Technical Specificat;on.nange is expected to be approved ay the PORC and the Nuclear Facility Safety Committee (NSFC) by June 30, 1985.. As part of the cevelopment process for these propesec h

charges to the Technical Specifications, on-line functional j

testing requiremerts were reviewed based on past experience.

l Possible changes to the testing intervals in certain cases where available test data may support such changes has (sic) been discussed at length with the Nuclear Regulatory Commission staff.

The Nuclear Regulatory Commission staff has informed Puolic Service Company of Colorado that no such changes woulc be acceptable at this time."

The INEL review interpreted these responses from Fort St. Vrain to mean the E has establishec Fort St. Vrain's RTS current test intervals, L

the current test intervals have been evaluated by PSC, and the NRC will not al'.ow enanges to the test intervals at this time.

i 14

r y

i 1

'4..

From these responses, the INEL concluded that Fort St, Vrain has condur.ted the review required by GL 83-23, Ite;n 4.5.3, and that the NI;;

considers the PSC and NRC reviews adequate to r.eet the item 4.5 3 requiremer.ts.

l I

b

{

9 rh.

e t

h 6'

P 15 s

4

e o

1 5 REVIEW CONCLUSIONS

~

\\

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, er 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 RTSs to be highly reliable with the current test intervals and prior to ir.plementing some of the requirements of GL 83-28, Implementation of these additional requirements will reduce the ATWS risk even further.

The INEL has reviewed the relevant topical reports, TERs, SERs, acettional analyses, and the irdividual licensee submittals with regard to GL 83-29, 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 adequately demonstrated that leir current RTS test intervals are consistent with achieving high RTS availability.

I 16

~

C" i

4 i

i 1

• s.

1 6.

REFERENCES 1

1.

U.S. Nuclear Regulatory Commission, 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 so All Licensees et al., Reeuired Actions Based on Generic Implications of Salem ATWS Events, Generic Letter 83-28, July 8,1983.

3.

Combustion Engineering, Reactor Protection System Test Interval

~

Evaluation. Task 486, CE NFSD-277, December 1984.

4 S. Visweswaran et al., BWR Owners' Group Response to NRC Generic Letter 83-28, Item 4.5.3, NECD-30544, January 1985.

5.

R. 5. Enzinna et al., Justification for Increasine the Reacter Trip Sy tem On-line Test Inte* val, BAw-1016". May 1966.

6.

R. S. En:inna et al., Justificatien for increr. sinc th. Reacter Trio System On-line Test Inteeval,$ueolement Numeer _1, BAw-10167, Supplement Numeer 1 Feervary 1988.

7.

Cemeustion Engineering, RDS/ESFAS Extended Test Interval Evaluat'en, CEN-327, May 1956.

8.

W. P. Sullivan et al., Technical See:ification Imorevement Analyses for BWR Reacter Drote: tion System, NECD-30851P, May 1985.

)

9.

R. L. Jansen et al., Evaluat on of Surveillance Frecuencies and Out of i

Service Times for the Rea::: crete: tion Instrumentation System, wCAD-10271, Janua*y 1963.

I's C

L. Jansen et al., Evaluation of Surveillance Frecuencies ane Out of

revice Times for the Rea

ter Prete: tion Instrumentation System.

Su:eieren 1, wCAS-10271, Supplement 1 July 1983.

J1.

R. L. Jarser et al., Evaluation of Surveillance Frecuencies and Out of Se vice Times fer the Rea:::r P-etection Instrumentation System, Sveeleven 2-G-A, wCAF-10271, Supplement 1-P-A, May 1966.

12.

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

Butcher, Acceotence for Refe-encin; ef General Ele:tric Cemeany (GE)

Teoical Reports NECD-30644, "En; Owners' Greue Response to NRC Gene ic Lettne 83-28," anc NELO-5085;P, " Technical Specification lme ovement Analyses for BAR Rea:te-Pretection System,"' April 28, 1986.

13.

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

Sheopard, A::eetance for Referentinc of Licensine Te;ical ReDert WCAP-10271. " Evaluation of surveillan:e Frecuencies anc Out of Se"vice Times for the Reactor Prete:tica Instrumentation Systems," E eruary t

21, 19E5.

~~

a; 0

1

'e 1

-3

~

I

\\

14.

U.S. Ni, ; ear Regulatory Commission, Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram ( ATWS) Events, SECV-53-293, July j

19, 1983, 15.

J. P. Poloski and S. D. Matthews, Review of B&L Owner's Group Analyses for increasinn The Reactor Tri; System On-line Test interval, EGG-REQ-7718, September 1988.

16.

D. P. Mackowiak and B. L. Collins, A Review of the Combustion Encinterino Evaluation For Extending the RPS and ESFAS Test J r.t e r v a l s,

EGG-REQ-7768, Saptember 1988.

17.

R. E. Wright and B. L. Collins, A Review of the BWR Owr.ers' Group Technical Specification Improvement Analyses for the BwR Reactor Protection System, EGG-EA-7105, January 1986, 18.

R. E. Wright and B. L. Collins, A Review of the BWR Owners' Grove Technical Specification Improvement Analyses for the Bm'R Reacto-Pectect*on System, EGG-EA'-7105', Rev 1, Maren 1937.

19.

D. A. Reny et al., Evaluation of Generic Issue 115. Enhancement of the Rel! ability of Westincrouse Selic State Protection Systems, NUREG/CR-5197; January 1959.

20.

Public Stevice Company of Coloraco Letter, O. R. Lee to D. G.

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

21.

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

Johnson, Re:oonse to Gene ic Letter 83-28, June 12, 1985.

l l

l l'

e 18 i

.-a-,

e;. +

)

T - r,,..

a. m s.eeee u.:

...c..

=en..

u 9

Y : e.

%g

,*/,%;'l sinuoGRAPHIC DATA SHEET EGG-NTA-8341 us.=c..e.e i o...e.ose 4.i.......

TECHNICAL EVALUATION REPORT:

A REVIEW OF REACTOR TRIP SYSTEM AVAILABILITY ANALYSES FOR GENERIC LETTER 83-28, I

ITEM 4.5.3, RESOLUTION

,.. ' " ""I' ' * ' ' '. '...

Nanch 1989

!. avid P. Mackowick

+1 o":. u.It

~~

John A. Schroeder m

l Ma rc'..

1999

..o. g.. w... ;,. 4.,.............. w..... n.. < u.

. =c.e c...- a c+..

...o Regulatory 1.nd Technical Assistance EG&G Idaho, Inc.

P. O. Bor 1625 Idaho Falis, ID 8:415 D6001

,.... c.. w..............,... +,,. c o. o

<. c.

",.. o..o= i Instrumentation and Control Systems Branch Technical Evaluation Report Divis on of Engineering and System Technology Office of huclear Reactor Regulation U.S. Nuclear Regulatory Comission Washincton. DC 20555

~..........c.u

+

.......e m. -

The Idaho National Engineering Laboratory (INEL) conducted a technical review of the' commercial nuclear reactor licensees' responses to the requirements Of the Nuclear Regulatory Commission's (NRC's) 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 vende'rs' Owners' Groups, submitted analyses to the NRC either directly ir. response to GL 83-28, item 4.5.3, or to provide a basis for reovesting changet to the Technical Specifications (TSs) that would extend the Reactor Protection System (RPS) surveillance test in+.ervals (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 current'v opera-ting commercial nuclear reactors have adequa:ely demonstrated that their current on-line FPS test intervals meet the requirements of GL 83-28, item 4.5.3.

i p,.,;,;;, -

. ::...................c..uc... :..

t e

.. n :..... :.. u. :. : s

, ei.. ni e.i. :..o Unciassified

! Unclassified

.........w

... e s t.

.