ML20012D202

Forwards E Beckjord 880928 Memo Re Background on Generic Ltr 89-10, Safety-Related Motor-Operator Valve Testing & Surveillance

ML20012D202
Person / Time
Site:	Hatch, Vogtle
Issue date:	02/26/1990
From:	Matthews D Office of Nuclear Reactor Regulation
To:	Hairston W GEORGIA POWER CO.
References
GL-89-10, NUDOCS 9003270047
Download: ML20012D202 (3)

Text

{{#Wiki_filter:m p s L_ o l February 2G,1990 Dockets Nos 50 424, 50 425, f l i 50-321 & 50-366 Mr. W.. G. Hairston, III t Senior Vice President. i Nuclear Operations Georgia Power Conpany i P.O. Box 1295 Birmingham, Alabama 35201

Dear Mr. Hairston:

i

SUBJECT:

CACKGROUND ON GENERIC LETTER 8910. SAFETY-RELATED MOTOR.0PERATED VALVE TESTING AND SURVEILLANCE - HATCH AND V0GTLE FACILITIES During the recent Georgia Power Conpany (GPC)/NRC Regulatory Interface meeting on February 8,1990, GPC asked what NRC considerations had been nede prior to issuance of Generic Letter 8910, " Safety Related Motor. Operated Valve Testing and Surveillance," dated June 28, 1989. Accordingly I am enclosing Mr. E. Beckjord's menorandum of Septenber 29, 1988, requesting CRGR review and containing nine related enclosures. Please note that only a copy of the cover of Enclosure 2 (NUREG/CR-5140) to Generic Letter 89-10 is included here. The docunent itself can be obtained from the Superintendent of Documents U.S. Governnent Printing Office, P.O. l Box 37082 Washington, D.C. 20013-7892 and from the National Technical Infornetion Service, Springfield, Virginia 22161. Sincerely. Original signed by David B. Matthews, Director Pmject Dimetorate 113 Division of Reactor Projects.1/11 Office of Nuclear Reactor Regulation

Enclosure:

As stated i {,; R. Ingram 14-H.25 NRC PDR T. Reed 14-H.25 Local PDR OGC (For inform. Only) 15-B-18 PDII.3 Reading E. Jordan MNB B-3302 S. Varga 14-E.4 ACRS (10) P-315 G. Lainas 14.H-3 HATCH PLANT File D. Matthews 14.H-25 V0GTLE PLANT File OFC' :LA:PDII-3

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Mr. W. G. Hairston, Ill Edwin 1. Hatch Nuclear Plant, Georgia Power Cornpany Vogtle Electric Generating Plant cC; Sh6w, Pittman, Fotts and Trewbridge Mr. R. P. Mcdonald 2300 N Street, H. W. 2xecutive Vice President - Washington, D.C. 20037 Nuclear Operations Georgia Fower Company Mr. J. T. Beckham P.O. Box 1295 Vice Presicent - Plant Fatch Birraingham, Alabama 35201 Georgia Power Company P.O. Box 1295 Mr. Alan R. Herdt, Chief Dirmingham, Alabernt 35201 Project Branch #3 U.S. Nuclear Regulatory Comission Mr. S. J. Bethey 101 Marietta Street, NW, Suite 2900 Manager Licensing - Hatch Atlonta, Georgia 30323 Cecrgia Power Company P.O. Box 1195 J. A. Bailey Birmingham, Alabama 35201 Manager - Licensing Georgia Power Company Mr. H. C. Nix P.O. Box 1295 General Manager, Nuclear Plant Birmingham, Alabama 35201 Cecrgia Power Company Route 1, Box 439 Bruce W. Churchill, Esq. Baxley, Georgia 31513 Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.W. Resident Inspector Washington, D.C. 20037 U.S. Nuclear Regulatory Commicsicn Rcute 1, Box 725 Mr. G. Bockhold, Jr. Baxley, Georgia 31513 General Manager, Vogtle Electric Generating Plant Regional Administrator, Pegion 11 P.O. Box 1600 U.S. Nuclear Regulatory Comission Waynesboro, Georgia 30830 101 Marietta Street, Suite 2900 Atlanta, Georgia 30323 Office of the County Commissioner Burke County Comission Mr. Charles H. Badger Waynesboro, Georgia 30830 Office of Planning and Budget Poom 610 Office of Planning and Budget 270 Washington Street, S.W. Pcom 615B Atlanta, Georgia 30334 270 Washington Street, S.W. Atlanta, Georgia 30334 Mr. J. Leonard Ledbetter, Director Environmental Protection Division Mr. C. K. McCoy Department of Natural Resources Vice President - Nuclear, Vogtle Project 205 Butler Street, S.E., Suite 1252 Georgia Power Company Atlanta, Georgia 30334 P.O. Box 1295 Birmingham, Alabama 352 01 Chairman Appling County Com.issioners County Courthouse naxley, Georgia 31513 I

3.. O ,o C ~. b-Mr. W. G. Hairston,111 .g. Georgia Power Company CC: Resident Inspector Nuclear Regulatory Commissicn P.O. Box 572 Waynesboro, Georgia 30630 'Jan.es E. Joiner, Esq. Troutmen, Sanders, Lockernen, & Ashmore 1400 Canaler Building 117 Peachtree Street, N.E. Atlanta, Georgia 30303 Attorney General Law Department 132 Jucicial Euilding Atlanta, Georgia 30334

.n l 0: f.Ip* *%g). UNITE 0 STATES y' j NUCLEAR REGULATORY COMMISSION j usecto.o e tosis %.v.... i SEPT 91Hg Y!wC:1 C

• FC : !:*ard L. Jordan Chair an Cemittee to Review Generic Requirements F:CV:

E-1: 5. Pety :rd. Director Office of 'iuclear Degulat:ry Desearch 3;;!;I' P: 0D05AL 70 EXPAND DEc!CDIC IN $17U TEST!NG ANO SUPVE!LLANCE

! U::E*EN?S FC: SAFETY : ELATED MOTOR-CFEDATE0 VALVES ("CVs)

'3ENE;:C !S$'.! ':.E.5.2. ":N SITU TEST!NG OF V4 VES") By th4s me crandum, the NDC staf' re:u,ests CDGR review and coments en the pre:: sal that all ::wer rea:t:r Owners i?clerent a pregram, as cuttined in a prc:: sed draf t ge eric letter (enclosure 1).. to establish, maintain, and theren'ter seriodically verify the design basis operability of all safety-related MCVs. The cre; ram is intended to extend and expand that pr gram Outlired in Bulletin 25-C3, "9otor-Operated Valve Comen Mode Failures Curing Diant Transients Due to Imer::er Swit:h Settings " dated November 15, 1985 and Su: le ent 1 dated Acril 27.19EE. The extended program will include all sa'ety-related MOVs and all *0Vs in safety-related systems that might become mis::sitiered and r*'ect system operability (referred to hereafter as ecsiticn-c*argeable POVs). In the expanded program. NRC will recomend that cweers spe:i'ically address a list Of potential degraded conditions for each

• 0V.

That list is found in A::ercix B of the draft generic letter, enclesure 1. TFe resolution e' this issue will complete the resolution of Generic Issue !!.E.5.1 " n Situ Testing of Valves." with respect to in situ testing of MOVs. This pre:: sal is censidered to be a Categcry 2 action item, as defined in the C:3: charter. In order to meet the schedule of the Comission's Five Year Plan, C:30 review and c:ments are required within five weeks of the date of this memerancum. contains a value-impact analysis to support this proposal. The aralysis irdicates that design basis failure rates for safety-related POVs are at leas *. an order of magnitude higher than previously estimated in several gereri: PFAs. The value-impact analysis also shows that the proposed program will be cost effective and will result in favorable radiation dose reductions. Our 1ewest estimates are that the industry would save 355 million dollars and that the expected value of public radiation dose would be reduced by 185,000 persen-rem as well. The ree:rvendations. outlined in the generic letter are in addition to the ASVE Ccde Section XI. Division 1 inservice tests and surveillance specified in 10CF.50.55af g). Therefore, pursuant to these recomendations, owners would

erfcm mere surveillance and tests than they currently perfom in order to assure POV operability. Hewever, all owners are currently obligated by Accendi 9 c# l0CF?50. to establish and raintain a test program that demonstrates tFe cesign basis c

e ability ef all safety-related compenents. The current ASME Sectf:n X: c:erability test for MOVs censists of c erating the MOV and i

j g }-B d d i

I i Edaard L..lordan 2 reasu * ; strcke time. Such testir; is not sufficient to previde meaningful i assurance of future POV coerability urter design basis co i conc!vsicn is base: cf kncalecgeable :erscrrel throughcut the incustry, examination of data over a number of years, and the value/ impact analysis, enclosurt ?. The sce:e of Bulletin 85-03 was criginally intended by the staff to cover all sa'ety-related meter c:erated valves, hcweve". the sco:e of the Bulletin was limitec te tr 3 e high Oressure safety syste*s. The evaluatien cf recessity f:r ex:arsion of sc :e of Sulletin S5-03 is part of the resolutten of Generic !ssue II.E.E.1. The analyses described in t'ais package clearly incicate the need to expanded the sc0:e of testing and ins:ection to all safety-related POVs and

siti n-changeable 90Vs.

would te increased beyond those recomended in Bulletin 85 03. Owner This is based en ex;erience that shews that switch settings are only ore part of the problem of assurirg F0V c:erability. The Orcoesed action is intended to apply to all pcwer react:rs, both licensed or under c nstruction. This rackage has be.en reviewed by the Office of Nuclear Reacter Regulation WDh tre Of' ice for Analysis and Evaluation of 0:erational Cata (AE00), and the Office cf General Ccunsel (OGC). NRR and AE00 concurred. OGC has no legal ebjectices. A su-ary of the CDG package is contained in enclosure 3. A regulatory analysis, that accresses the items listed in NUREG/GR 0058. Revision 1. is contained in erciosure 4 The inforettien contained in enclosure 4 is su;plementary to that contaired in enclosure 2. This was done in order to comply as completely as

cssible with the outlint of recuired contents for CRGR packages, as listed in the CRGD Charter.

The regulatory analysis included in enclosure 4 specifically addresses the provisions of the proposed generic letter, enclosure 1. NUDEG/CP-5140 (enclosure 2) is concerned with extension of the provisions of Bulletin 85-03 only and was written without benefit of knowledge of the scope of the procesed generic letter. provides a probabilistic risk l' assessment (p0A) and value-impact analysis in support of this proposed action. Although :nly adjustment of switch settings is explicitly considered in enciesure

2. the PcA applies to assuring POV operability. Since POV operability cannot be assured by examination of switch setting alone, the PRA contained in enclosure 2 is considered applicable and appropriate to support the proposed action.

The justification for including position-changeable POVs was discussed with and 1 l a: r:ved by C#GR (CPGP Feeting No.133) for the Bulletin 85-03 MOVs. Positien-changeable POVs were always part of the Bulletin 85-03 program and Su::le ent ! to Eulletin 85-03 was specifically issued to clarify and amchasi:e the NDC's ecsition that these MOVs should be assured to be operable in those syste*s covered by the bulletin. The proposal to include the position-changeable POVs o' all safety-related systems is su: ported by the same arguments that l su;;;rt thefr inclusien in Eulletin 35-03 and its su:plerert.

~ h = t F.. Ecmarc L. s'cedan 3 N E I IIII l ~ Erc!csvre ! c:ntains a Orieritiratten and sc*eculing evaluation of the :recesed Orc; ram in act:rcarce with the CF3R Charter recuirement. Enc? sure 5 c:ntains an interim sumary of MOV f ailere rates based on replies 0 to fulletin !! 03.

• his suarary indicates that the failure rates estiested in the value w:act analysis, enclosure 2. are com:atible with these actually reportec in the licensee responses t: Bulletin 25-03. centains an inde:endent DES staff evaluation re: ort en the value-ir:ac analysis (encicsure 2).

Iht evaluatien fincing substartially su::cris the rethedology and cccclusicns cf the value-im:act analysis. Enclosure ! c:ntairs an annotated Sa'e4y !ssues Wana covering the M V ertien of Gereric Issue II.E.6.1. gerent System (5!*S) re:crt c ntairs a ec:y of Su:plement I to fulletin 65 03. A cocy of Eulletin 25 C3, as originally issued, is included in the value-impact analysis l enCICsure U. Eased en the evaluati s described in enclosure 2, 4 and 5 we conclude that: (a) there is a substantial increase in the overall protection of public health and safety to te derived from adottien of the propcsalt and (b) the direct and irdirect costs of imclementation, for the facilities affected, are justified in view of this increased protection. It should be noted again that the cost of ir lementation of this program will be minimal and probably will result in a net savings. W Eric 5. BeckjeP4. Directer Office of Nuclear Pegulatory Research

Enclosures:

See next sheet. Centact:

0. Fothberg, RES/EIS x23816, MS NLS-2:7A e

1

l- =,. o Edward L. Jer:an a SEP 2 9 tili Enciesures: . Oraf t Geae'ic Letter to LWR ?wners "Sa'ety elatec *CV Testing ard SLrveillarce." 2. NL':EG/CP !;a0, "Value Im: set analysis fer Extensien of NRC Bu'letin !! C3 to Ocytr All !afety Delatte MOVs". ( 3. C 30 :ackage Sum?ary - Pre;csal to' Escand r fedic In Situ Testing and e Surv ') lance :ecuirements fcr Safety-Related *0Vs. a. De;ulatory Aralysis - Precesal to Exce9d :triccic In Situ Testing and Surveillance :ecuirements for Safety-Eela ted *CYs. - 5. Pricriti stien and Scheduling Evaluatien - Proposal to Expand Feriodic In Situ Testing and Surveillance Pecuirements for Safety Pelated MOVs. 6. Firal Depert Status of IES~ 85-;3 Sumrary of Failure Eates as of July 29. 1988. 7 Mercrandum frem R. Housten to R. Sosnak, " Review of BNL Draft NL'3 E 3/O R-5140. " 2. 5:*S Decor RI?!5310-001 for Generic Issue II.E.6.1 annotated. 9. N* Sulletin 55-03. Supplement 1: "*0V Common v de failures During e ?Iant Transients Due To Improper Saitch Settings, s

l 1 i Enclosure ! CRAFT GENE:!C LETTER

-::: sal to Ex:ard Periodic In Situ Testing and Surve11:ar:e :ecuire erts for Safety elated Motor-Operated Valves (F0Vs)

TO: All "olde s of nuclear D0wer reactor crerating licenses (OLs) cr

rstru: tion :ermits (C?s).

!"!JE:7- $2'I'v :ELa7E I M070RC:ERs;EDVALV!7!!7IN3ANDSUPVE!LLANCE Ev!!etin !!-C2, dated Nove-ter 25, 1985, and Su: ele.?ent 1 of Bulletin E5-03, dated 1:ril 27, :355, recorrend that owners develop and implement a program to ensure that valve rotor o:erator switch settings (tercue, tercue bypass, 00siticn limit, ovealcad) fer motor-operated valves (MOVs) in several specified syste-s, are. selected, set and maintained so that the MOVs will coerate under desige basis' Conditions for the life of the plant. NRC assessments of the reliability cf all safety-related V0Vs, based on extrapolations of the currently available results ef valve surveillance performed in response to Bulletin EE-03, indicate t'at tre program to verify switch settings must be extended in creer to assure operability of all safety-related fluid systems. Our evalua-tien of the data indicates that, unless additional measures are taken, failure of sa'ety-related POV's to operate under design basis conditions will occur .?uch r:re often than had been previously estimated. The recommended program will crevide for MOVs to be tested, inspected, and maintained so as to provide the necessary assurance that they will function when subjected to the design basis c:nditiers that will be encountered during both normal operation and abn rmal events within the design basis of the plant. By this letter NRC extends the scope of the program outlined in Bulletin 85-03 and Su:Diement 1 of Sulletin 85-03 to include all safety-related MOVs. Those MOVs in safety-related systems, that could be unintentionally moved (that is, not locked in Desition or with power removed) to a positien that could interfere with the design basis operation of any safety-related system (to be referre; to bereafter as position-changeable MOVs) are also to be included in the program. 1/ The term " safety-related" refers to those systems and components that are relied c:en to remain functional during and following design basis -events te ensure (1) the integrity of the reactor coolant pressure boundary. (ii) the capabilitf to shut dcwn the reactor and maintain it in a safe shutdewn certitien, and (iii) the capability to prevent or mitigate the consecuences of a:ci:ents that c:uld result in Octential offsite exposures comparable to the 10 CF8 Part 100 guidelines. (See 10 CFR 50.49) 2/ Cesicr basis events are defined as cceditions of normal c:eration, in:ludith antici:ated c:erational oc:grren:es, design basis accidents, external events, and natural chercrena for which the clant must be designed to ensure functa:ns (i) through (iii) above. (See 10 CFD 50.49)

I t 2 'Fe ruc' ear Ocaer PO!:"*C erated V4Ive maintenance and testing. industry Fas several c# saculd te use'ul to individual plant cwners in follcwing the recorrendatiensT c:ntaired 'e eir. Assurance e'.*0V c;erability is a cceplex t:Dic. as develo: cent of strong testing and maintenance Oregrams, manageme 5 and c:crcitation cf engineering raintenance and testing. be viewet by all c ncerred as a long.te m enc 0ing program. This effort sh:uld ( e :*0; rem 10 res:end to this letter should etetsin at least action iters e tnrcuge g. tele.c Eeview and cc:ument the design basis for the operation of each MOV. a. The review steuld include the determinaticn of the design basis conditiers to be er:ected during both crening anc closing the MOV for both nomal c:eettiens and abnomal events, to the extent that these MOV coeratiens ard events are included in the existine approved design basis. The desien basis is that cocumented in certinent licensee submittals such as FSAR analyses and c:reved c erating Orc:edures and eeergency procedures. The ebfective of the design basis review should be to detemine that the installec ecuiceent is capable of satisfying the design basis. The review should not be restricted to a detemination of *stimated maximur design basis pressure. The design basis review should :nclude a critical review of all of the :ertinent design and-installation criteria that were used in choosing the particular MOV. These criteria should then be c:r:ared witr actual c;erating conditions and expected design basis c ncitiens. For example, the effects on MOV cerfemance due to steam, wete. temperature, wiring si:e, centrol system design, and power sueely unde all :ertinent conditions should be estimated for expected design basis conditions. In addition, when determining the design basis differential pressure er flew for oosition-changeable MOVs, the fact that the MOV must be able to rec ver frem mispositioning shculd be considered in the evaluation cf the design t; asis parameters. b. Using the results from item a. above, establish the correct switch settings. This should include establishing a program te review and revise, as necessary, the methods for selecting and setting all switches (i.e., tercue, tor:ve bypass, position limit, overload) for each valve operation (c enirg and cl,0 sing). The inteet is to prcvide assurance that a ;rogram exists for selecting s and settir; valve ocerator switches to ensure high reliability of safety syste ,*0Vs. Actions should be taken in accordance with the apprc riate

crti:ns c' t'e :lart's technical s:ecifications if either (1) chargin; t'e swit:5 settings is not suic!ert to ensure POV c;eration under the design basil'c:nditices er (2) the assessrents o# the necessary charges t:

the swit:r settings indicate that the *0V, as presently adjusted, Eay P:t

i l ) i 3 1 I te : stat'e of :eratirr urter the cesign basis conditions. i Such find!rgs, arc correttive actions taken, shculd be retained as part of the a rceriate

er aaent re:Ords for ; e FCV and should be made available fer NDO audit on 1

'"e cwner ray alte elect to irclement accitional actions, such re: vest. as administe:tive er procedural centrols or eculpeent modificatiens in

der to minimi:e the likelthoed of MOV malfunction.

I dividual "0V switch settings sheuld be changed, as ae:recriate, t

:se established 'n item b. atcve.

'a'hether the switch settings are c?a r;ed er ret, tre MOV should be de Onstrated to be crerable by testieg t*e MOV at t*e design basis differential pressure and/or flew determired it ite

3. abcve.

Justificatien sh: bid be provided for any cases where testir; with

• e design basis differertial pressure or flow cannet practic6bly be pe rf o rced.

This-fustificatien should include a description cf the alternatives to design basis differential pressure testing or flow testing that are used to verify the correct settings. Occumentation of.iustification should be retainec as part of the acercDriate permanent records for the MOV and should be made availeble for NDC accit on recuest.

• his dis:Ussien is not intended to establish a recuirement for valve testing fer the c nditien simulating a break in the line centaining the V0V.

w: wever, to the extent that such MOV ceeration is retted upon in l-the design basis, a break in the line containing the MOV should be c:nsidered in the analyses described in items a. and b. above. The resulting switch settings for pipe break conditions should be verified, L te the extert practical, by the same methods that would be used to verify l cther settirgs (if any) for MOVs that are not tested at the design basis differential pressure or flew. i Each 90V shculd be stroke tested, to the extent practical, to verify that the settings defined in item b. above have been properly implemented even i' testing with differential pressure or flow cannot be performed. A descriptien of acceptable methods to demonstrate design basis ocerability for those MOVs that cannot be tested at design basis l rcssure or flew conditions, is contained in Appendix A. 1 i Prepare or revise procedures to ensure that correct switch settings are d. determined and maintained throughout the life of the plant. Ensure that at:licable industry recommendations are considered in the preparation cf these procedures. This is intended to be completely consistent with action item 3.2. " Post-Maintenance Testing (All Other Safety Related C renents)." cf Generic Letter 83 28. "Recuired Actions Based on Generic Im !':ations of Salem AT'a'S Events." These procedures should include rev?siens to monitor MOV performance to ensure the switch settings are c Prect. This is particularly im:ortant if the torcue or torque bypass switch,settir; has been significantly raised above that recuired.

t:.. 4 a 'be - Cedures for setting and *aintaining POV switches should at:Ourt ':r normal = ear, deterieration o' any of the critical MOV component earts, degre:td electrical C0nditions, dirt accumulation, corrosion e::3.u'att:r, aed tem erature effects as well as any misadjustreats of the switc*es themselves that rey have occurred ever time. The ialected assurance procedures shculd give plant persernel the capability to diagr:se the conditions that would be the source of the misadjustment cf t's swit:5 settings. It.is irsufficient to erely verify that the swit:h settings are un:hanted ft:- ;reviously established values. ibe swit:5 settirgs stuuld te verified, in ac:crdance with the pr: gram implerentati:n ard veri'icatien s:'edule cutlined below, to be ap:ropriate fcr the c:r:'ti:n e' ea:n MDV, The A!VE Co e Sectict XI streke-timing test re uired by ;D CF3 50 is ret sufficient to satisfy the intent of this action ite*, The swit:n settings reed not be verified each time the ASME Code strekt timin; test is terfor ed. A nu?ber :' de'iciencies, misad,iustments and degraded conditions were e. cis::vered by plant cwners, either as a result of their efforts to cerely wit' ! lletin E5-03 or from other experiences. A list of these :nditiens (in:luding impaccer swit:h settiegs) is included in Ac:endix B to this

letter, The program described in this letter should address, as a minimum, these c:nditions listed in Appendix B for all of the safety-related and resition-changeable MOVs.

The pregram should include s;eci'f: analyses, tests, and surveillance, as accropriate, to assure that ea:5 of the listed deficiencies, misadjust ents, and degradations are either prevented or identified and corrected. f. All actices t? ken, includina repairs, alterations, analyses, test, and surveillance, should be ful5y justified and documented for each POV. The de:umentation should include results and history of each as-found ceteriorated condition, malfunction, test inspection, analysis, repair, or alteration, g.

n addition te recording trends for the individual POV cerformance parameters, ownees shculd periodically (at least every 2 years or after each refueling cutage after program implementation) tally totals for occurrence of the c:nditions listed in Appendix B in order to estimate trends of MOV operability.

• 11 documentation should be retained as part of the appropriate permanent re:Ords and should be made available fer NRC audit upon recuest.

.The

• 0 gram to respond to this letter should be accomplished in accordance with the schedule out, lined in action items h, through k. belcw.

h. Addressees sheuld have the documents listed below available in the a:0-0;riate :ermanent records. Fer plants with an OL, this information should be available for NRC audit within 2 years er one refueling cutage cf t*e gate Of this letter, whichever is later. For plants with a CP, this in':rmatter should be available 'er NDC audit within 2 years of the datt

' this letter cr Orier to OL issuance, wnichever is later. The 10:vments sh0uld in:lude:

I I

I f

• Fe resu!ts O' the design basis review tl'escribed in action ite* a.

1. 0e 4II safety-related YOVs. anc csition-changeable MOVs as Cescribed, drd 2.

• e 0 *:7'a? des:rittiCn tD accCPDIish items b. through g. above, fer ali sa'ety re!sted MOVs and pcsitien-changeable MOVs as described, Each :! art with an OL sh0uld at:Or lish all analyses, verificaticts, tests, i.

ard its:ections that " ave teen instituted in Order to cce:ly with Actic-Ste's a. b., c., and e, above, wit *in 3 years or tro re'ueling cutages of the datt cf t*is letter, whic*ever is later. Each plant with a CP shog?c ace: Polish these acticns within 3. years of the date of this letter or price.t0 CL issuance, whichever is later. The ;r: gram fer verification of switch settings outlined in action item

d. ateve, as well as other tests or surveillance that will specifically identi'y eteatial s'0V degradatiens or misadjustments, as outlined in acticn itet abcve, should be ac::~;11shed af ter traintenance or acjustP**t (;ncluding :acking adjustment) of each POV. and reriodically i

t*erea'ter. at leest every three years or every seccnd refueling outage, whic'ever is lenger, k. 'n vee:gnition of the recuirements for cre-planning, refueling outages that start within six months of the date of this letter need not be c:Unted in establishing the schedule to meet the time limits recorrrenced in these action items. Pursuant t: 20 Cro 50,g:(f), addressees are recuested to provide infomation to NPC as outlined in acticn items 1.. m., and n. below. 1. CYners are recuested to advise NRC in writing, within 90 days of the date of this letter, whether the above schedule and recorrendations will be ret. For any date that cannot be met, the owner should advise the staff cf a creposed revised date,,iustificatien for any delay, and any planned ccm:er: sating safety actions to be taken during the interim. Owners are recuested to notify NRC in writing within 30 days after action m. itet h. has been accomplished, n. Owners are recuested to notify N C in writing within 30 days after action iters a.. b., anc c. have been accerclished. This gereric letter is intended to supersede the recorrendations centained in Sulletin !!-03 and its su: clement. fulletin ES-03 addressees need not make any further res::nses regarding the bulletin or its supplement. The information which wov1d have teen submitted to NDC in response to the bulletin or its supplement ,sh:uld te re,taired in ac:ordance with the recuests of this generic letter. l

= e = F "ecureated results of tests er etker surveillance that were used to satisfy 16e re::' tated a: tiers :f Sulletin PS.03 ay be used, to the extent a: 0licable, te sa tis'/ t*e a::i:n ite s stated herein, This reevest is ::vered by Of' ice ef wanagement and Budget Clearance t Nurter 3:50-*:: whi:h ex:fres Oe:e-ter 32, 1959. The esticated average burde*, 6.aurs is 200 tar.h0urs Ot* ewPer res00nse, including assessrtrt C# t6e rew re::.*erdati ns.. star:hin; d6ta s:urces, gatherin; ord araly:4r; the data, and Ort:aring the re:Vired letters. CC5Fents On the at:uracy c# this esti ate a*d su;;tstiens 10 duce the burden may be directed t: t"e Of#ite Of Va*4 ! e't ard Sudjet, 200* 22l3, New IKtcutive Office Buildir;. Vashir;t:P 0.0. 2 503, and the '.'.S. Nuclear Regulatory COErissiCn, Ee::PCs and De0:rts.*4aage ent f ranch, Of' ice e' admir.is tration and Pesourets Straietent, Washington, D.C. 2C555. I' you have any :vestiens regarding this P:tter, Dlease contact the ND: Or:Je:: cara;ea er the technical ::rtact listed belcw. Oennis Crutchfield leting Associate Direct:r for Prefects Office of Nuclear Reactor Regulatien Contacts:

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PL11etin !!."2) :: ce :estrate desich basis c:erability fer trose MOVs that cannet be teste: at cesi;n tasis Oressure or flow conditices:

!!!a ' 0m ins t'v eate: tests :' at Irast ' cur identi:41 "T.'s. cc ndu:ted urde" desiin tasis ::*:itt:rs 0" 10wer Ortssure or flew C0rciti:ts eV**a : I4ted t: Cesi;r basis c:Editi0ft, should be used 10 tstablish the r Oui'e: *er# ' an e criteria. 10enti:41 *CVs are 90Vs ecui :ed witn tte e sa?e maru'acture" an: Oc:el Cf c eratcr. These MOVs are also Of the sare va've tyre. same velve taru'a:!urer, same valve model, same valve stem ciareter, are sa e orifice ciaeeter, the POV tnat cannot be tested under t*e er:ected desige tasis ci'ferential pressure or ficw c:nditiens should be instrumented and shuuld de? nstrate e:erating enaracteristics at i IC* Ortssure. er ICw-flow. Cr no-pressure, er no-flow Conditions that Or vide eb,ie:tive evidence, tased on the established performance Criteria, t*at

• e MOV wCl functicn as re:Utred under its design basis conditions.

2. Pata from instrumented tests of at least 20 similar POVs. conducted under design basis conditions or Icwer pressure or flow conditions extrapolated to design basis conditions, shculd be used to establish. the required performance criteria. Similar WOVs are MOVs ecuipped with the same manuf acturer operator. These POVs are of the same valve type and are of reasonably similar si:e (:50 percent of nominal orifice diameter). The 90V that cannet be tested under the expected design basis differential pressure conditions should be instrumented and should demonstrate c:erating characteristics at low-pressure, or no flow, or no-pressure, or no-flow conditions that provide objective evidence, based on the established performance criteria, that the MOV will function as recuired under its design basis conditions. 2. Cata from an instrumented test of an identical F0V.. conducted under both design basis conditions and lower pressure or flow conditions, should be used to establish the recuired performance criteria. The MOV that cannet be tested under the expected design basis differential pressure or flow c:ncitions should be instrumented and should demonstrate operating Characteristics at low-presture, or Icw-flow, or no-pressure, or no-flew c0nditions that provide objective evidence, based on the es'tablished rerformance criteria, that the POV will function as recuired under its design basis ce,nditions. Wetheds cther than these outlined above may be proposed by NRC or plant owners to demonstrate 907 operability under design basis conditions for these MOVs that cannot be easily tested under such conditions. NRC is currently spensoring a series of ce'siin basis ;ressure ar.d flew tests of prototyre POVs (G 27} that may result in additiens or char:es to the eeth:ds cescribed above. Several industry organi s tiers ar'e planeir; t'est programs as well. e results of such tests may inci: ate trat ot*er rethods o' demeestrating des ;n basis operatility of safety-related PT.s are also a;;rceria te. l

6 - c, .O 4 8 2 :eedir ! c' Ora't Gertric Letter Eu-ory of Commen Motor.0:erated Valve Deficiencies, "isadjust.?ents, and Oegraded C0nditiens IP 0"re:t tor:te switch bypass settings 2. .nC rrect tor:te swit;h settings 2. L'nbalanced t:r:te switch a. 5:rica ta:k 930 or inc0rre:t $ ring pack : reload 5. InC0rrect ste? 04cking tightness 6. Ex;essive inertia 7. LCese or tight s te?. nut ICckngt 8. Incerrect li?ft switch settings 9. $te* wear

10. Eent or br0k n stem t

II. Worn er br0 en gears k 12. Grease Orobte s (hardening, migratien into spring pack, lack of grease, excessive grease, contamination, non scecified grease}

12. "ctor insulation er PCtor degradation la.

Incorrt:t or segraded wiring !!. Disk / seat bindirg

16. Water in internal parts or deterioratien therefrom 17 Meter undersized (for degraded voltage conditions or other conditions) 18.

Incerrect valve position indication

19. Misadjustment or failure of handwheel declutch mechanism 20.

Delay problems (incorrect relays, dirt in relays, deteriorated relays, miswired relays)' 21. !ncerrect thermal overload switch settings

22. Wern or breken bearings.

22. Ercken or cracked limit switch and torcue switch components 22 Missing or mocified torcue switch limiter plate 25.

Improperly sized actuators

26. Hydraulic Lockup 27 Incorrect retallic materials fer gears, keys, bolts, shaf ts, etc.

28. Cefective er deteriorated motor pcwer supply

29. Cefective motor control logic

20. Excessive seating or backseating force application -

31. ncorrect riassembly.or adjustment after maintenance and/or testing

22. Unauthori2td modifications or adjust.t.nts a

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RES/CR 5140 4 - - -

L NUREG.52146 l VALUE lMPACT ANALYSIS FOR EXTENSION OF NRC BULLETIN 85 03 ,s TO COVER ALL SAFETY RELATED MOVS a 1 J.C. Higgins, C.J. Ruger, E.A. Mac00ugall, and D. Huszagh i i Date Published - July 1988 ,tc DEPARTMENT OF NUCLEAR INERGY, BROOKHAVEN Natl 0NAL LABORATORY ii. UPTON, NEW YORK 11973 a.-

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"3 i E :'esu-e 3 i CR3; ;A;(ASE SLMMARY (ITEM IV.E. CRGR CHARi!R, R!vl510N 4, \\ AERIL, 1567)

• • sal to Ex:a c Seri::i: In Situ Test'rg anc 5.*.ei11ance Re:. ire ents f:r Safety Relate:

i Mot:r-Coerate: V a l v e s (MC'.'s ) (i) IMe *:: sec generic letter as it is to be sent out to lice 95ees, is c:staine: in enciesure 1 of the c:ver memorancum. lette* ill extene tre Ev11etin 85 03 program to all safety-relate:As creposec, MO.s

sition-ena*,geable MOVs in safety relatec systems.

The letter mill aa: alsc c:atain rec 0mmen:sti0ns for owners to accress a list of specific M;V ceg*a:e: concitions anc perform surveillance on a regular schecule. (ii) C::d es of e'l reference materials cuotee or referrec to in any of the er:1:svres to the CRGR pac < age may te obtainee from 0, Roth: erg of the RES/EIS staff, x23816. fiii) The staff's position that the preposed program cees not constitute imecsition of new recuiremer.ts on owners beyonc their existin bases is ciscussed in the cover memorandum to this enclosure.g cesign Owners .ill rave to co more anc tetter testing and surveillance in or:er to assu e 'tnat M0ks meet existing NRC recuirements and their own licensing bases. Tre ore:csec program actions can be justifiec on a cost-tenefit basis, even if they are consicered to te cackfits, (iv) The preposec method of implementation of this program is by a generic letter, enclosure 1 of the cover memorandum, OGC, NRR and AE00 concurrence .ere cetainee on the submittal of this package prior to suemi.ttal to

CR3R, (v)

A Regulatory Analysis covering tne proposee program is contained in enclosure a of the cover memorancum, and enclosure 2 are

mplementary Enclosure 2 of the cover memorancum contains a value-1 i cact assessment for extencing Bulletin 85-03 and contains a PRA anc partial regulatory analysis succorting that position.

contains a ceterministic assessment of the complete program and also refers to enciesure'2 for suppert, (vi) Tre Or e:sec program will ce a:pliec to all operating power reacters anc t'.ese un0er construction. (di) D*iiri ti:ati:n anc sche:uling evaluation, for all plant categories, is Or0vice:'in Enc'esure 5 of the cover memorancum.

g..,.; c.97r N S ~ ' ' ' ' ~ - _ - - _ _ _ _ _. _ - - - - - - - - - - _. _ - _ _ _ - _. _ _ - - - - - - ~ _. _ _ -.. - - _ - - - - - - - - - - - - - - - - - - - - - = - - - - L- . ;g 4 -.% ) ,j, '4 f i '- j f E-h s s I b s, + y%}.i-Y' ^ 4 ut .{.m 2: e 'e w

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!3ULATORY ANALYS'!

-::: sal to tr:and Periodic In Situ Testing and veia*:e !rste tion :e uirements fcr Sa'ety :(lated Motor-0:erated Yalves IMOVs) 2.

State est c' t6e Dr:b'em t The N:: staff is :r::: sin; to issue a generic letter fen:lesure 1 to tr re cese to :**' t*at will ex:and testing and ins:ection of YOVs. e c:ve-E5 03, N 0 re:o : e ded t' at ewrers verify switch settings for safety relatedIn ful'et9 MCVs in t*e hi;' Oressure :colant infedtion/ core spray and emergen:y fee:* ate-(:C!O f:r EW:,' s;.:te s. ?>e exteasien described in the proposed generic letter will include all safety-related MOVs and those MOVs in safety-related systers that -4;ht be:o e PisDesitioned (position-changeable). A number of root cause I degraded c nciti;rs will be addressed in the Droposed expanded program, in additi:e to swit:" settings. to the A The letter will also cutline methods ac:eptable be tested under design basis pressure or flew.0 staf' fer cemenstrating M These acti:ns ar[necessary to reduce the failure rates of safety related M and thus prevent safety system failures. Best estimates indicate that the failure rate for safety-related MOVs is about 16 to 29 times the rate assu~ed in the NU EG !!!O study. These estimates are based on an evaluation of the data cotained from licensees in response to Bulletin 43-03 and from studies of the database Of the most utilized signature analysis service company, res:ec-tively. urther, a number of root cause degraded conditions for MOVs, in addition to swit:5 settings, have been identified over the past several years. Any of t*ese conditions can result in failure of a MOV to move, or to move c:mpletely, to the desired position under design basis conditions. Until recently, the reliable perfomance of F0Vs was assumed, by all concerne-, to te assured by the testing and surveillance standards currently used by the nuclear industry. It is now apparent that this assumption was incorrect and that additional reasures must be taken to assure design basis performance of safety-related MOVs. POVs are Dresent in virtually every fluid processing system in every nuclear power plant. (Limitor:ve of Lynchburg, Virginia).Over 90 percent of the valve operators we Identical operators are used in parallel trains arc in series, valves in individual trains of virtually all safety-related and n:n-sa'ety-rela ted fluid sys tems. l The operators function in the same 1 0:eratir;. maintenance and physical environment, in most cases. These comen moce ascects caer:: be reliably quantified, however, as a matter of engineering i,,udgment, treir im:act is large. Ea:5 *0V is.a ::.:ler electro-rechanical asse?bly whose operation affects and is af fected by tie .4 N' '2 M ctential it act on radiological exposure cf facility employees;~ IFe value/ fir.'ect analysis indicates the following overall

veational exposure estimates:

) a"::!'rE tsr:y4 t Hi;h fest ' L ew 3 2 - Ac:1 der'. Oc:::ational Exposure

5. 57x10 5.23x 0 3.9;x10 3

3' 2 Operational Oc:upational Exposure ,6.40x10 1.92x10 -2.89x10 3 3 Total _ Exposure 7.22x10 7.'5x10 -2.50x10 / A positive number indicates dose avoidance. A negative number indicates a dose increase. All figures are person-rem. 'e) Installation and continuing costs associated with the action', ) including the cost of facility dcwntime or the cost of j 7 censtructlon delay; J TFere will be an overall net c0st benefit associated with implementation of this program. IFe value/ impact analysis indi:ates the ~ following dollar savines associated with: 1. Industry implementation 2. !ndustry operation 3. NR0 Oevelopment 4 NRC Implementation 5. NRC' Operations Estimates of total savings for 127 plants over 25 years are: Low - $355 million Best - $418 million High - $453 million 1 'his is due primarily to estimated decreased industry operational c0sts. If one adds additional factors, such as plant availability improvement, the savings are even greater, f') The potential safety-impact of changes in plant or ocerational coeplexity, including the relationship to proposed and existing regulatory requirements and staff positions; eum e

3.;. .s 3 No :Fysical changes in the plant are anticicated to be recuin : hy the ;reccsed program. Normal plant operations associated O th starting, running or shuttirg dcwn the ree:ter are expected +o be enhanced by the increased operational reliability of safety-re M r-ar.d Ocsition cnangeable MOVs. The owner's program for in situ testing and maintenance will recuire. increased training of personnel as well as increase coordination of test and maintenance efforts. (g) The estimated resource burden on the ' C associated with the preocsed action ano the availability e,f such r sources; The. burden on NRC headcuarters is expected to be minimal because the ewners will not be recuired to submit programs for review. NRR manpower resource constraints dictate this approach. NRC pegional Offices may, as part of their audit program, review the individual. plant programs, implementation procedures and results. A Temporary Instructien, produced to assure unifonn review procedures in the Regional Offices, would be useful and would require headouarters efforts. SRP SectfEn 3.0.6, " Inservice Testing of_ Pumps and Valves". must also be revised to incorporate the recuirements of the Generic Letter. (h): The' potential im act of differences in facility type, design or age on the relevancy and practicality of the proposed action; lt is ebviously expected that older plants, which have older MOVs may discover more ino:erable MOVs or deteriorated MOVs than newer plants. Several clants that have large numbers of Rotork operators ' may recuire special attention'due to variations in the-maintenance and ' examination procedures for these operators. There is not expected to be much variation in overall impact on SWRs versus PWRs or on individual types of each,'and there is no reason to break down the-impact by such classifications as part of the_ resolution of this issue. -(i) Whether the proposed action is interim or final and, if interim, the justification for imposing the proposed backfit on an interim. basis; The actions proposed are considered final allowing for minor modifications based on experience and improvements in monitoring technicues, a 9 l

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I $6 a es, ;f, g g ;,,, 1965 -3t? 916

53 49$

gg7 a uT 1 n A tete'ently ist the nuttet af switch t'anges et ist the aggg,t g Othef feletts does not necessarily zean that none g: gutted it

tetably means that the ! ce.ste staalf did not fe;ctt any gata en the to;st To be sure, see the

:sta e n t s R e p o r t for the fatality

s, s-pagt-3 Of.3 - -_9: 5 4 Tot =the 43 plants r e p, g g i n g .l e date, ' 8 15 the seerage faalute fate in peggeng, i 44 as,the standard gov 4L8th of the f4414te rate in pottent = Ter the 34 plants repstting inopetable valves to date. 15 il is the asetage failure tale an pottent.. i 11 17): as the standard deviatten of the f ilure rate.1n pottent 4 i h Se

_ _ ___ _ _ _._. ~ mn 4,. h;g,,,o % g g h ] ^ [ gs'usq ., i A UNITE D sT ATEs f 3.c f.g NUCLE AR REGULATORY COMMisstON

iWf y'.w/ j '

wasm,.ctoy o c ::sss '/ jus n 2 m3 MIMOG C M FCE: R: tert Bosnak, Ce:uty Direct:r Division of En;1neering Office of Nuclear Regulatory Researcn FROM: R. Wayne Houst:n, Acting Director Division of React:r Ac:iden: Analysis O'fice of Nuclear Regulatory Resear:n

SUBJECT:

REYlEW OF BNL ORAFT NUREG/CR-5140

Reference:

Memorandum from R. B:snak to R.W. Houston, dated April B-IgBC, " Request for Sup; ort :0 Review BNL Craft Nureg." We are ~res;:nding to your reques; in the referenced memorandum to review a BNL craft re;;r: entitled "Value-Impact Analysis for Extension of hRC Bulletin 85-03 to Cover All Safety-Related MOVs." We initially read the January 1988 first craf t, and then tne April 1958 verson..Our review concentrated mainly on their use of the SARA ccde and the a;;r:priate data volumes fr:m NUREG/CR-4550, es;e:ially Surry (Vol 3) and Grand Gulf (5). ' e performed s:me calculations using SARA in an-attem:t to verify some of :ne values given in Tables 5-2 and.5-3. But some of the MOVs consicered in the FRA explicitly identified as such, and are classified as (part of) pipe are n:: se;ments. It woulc recuire cunsicerable effort'to locate these (hidden) MOVs and su:n. ime was not available; but from what we did simulate, the results -wera close to those in the tables, in lieu of uncertaking such a task, we had s:mt. extensive pnene conversetions witn Mr. J. C. Higgins of. BNL, incuiring as

neir s:::e and precedure for enalyzing the~ problem. ~1t was evident from

rese ecnversations tnat the'BNL team has cone a.very thorough analysis anc has.

c:rre::ly used the PRA data and SARA coce to perform tneir calculations. We clso.ste: :ne SARA contractor, EGSG/INEL, to contact Mr. Higgins regarding ce:eils of thn application of SARA, ano they reported that the BNL team had. fer via:ec 6. Solid approach nd seemed to have carried it out pr:;erly. A basic assumption in the BNL analysis _is that MOVs would only see a high

ressure cifferential in high pressure sequences (tnat is, transients and small LOCAs). M rcover, the assumption was mace thct all the MOVs in all the

ystems wculd see a high pressure differential in the nigh pressure scQuences. The

'first assur.ption may be nonconservative, since it is possible that in some 10w

res

ure se:vences, valves would see a hign differential pressure, where hign cifferential pressure is defined as a pressure differential comparable to sne-cesign pressere cifferentiil for the volve. The seconc assumption is c:nservative; valves in certain systems, such as the service water system or tne c mponer.: cooling water system, may not see o high differential pressure, even in high pressure secuences, in addition, the BNL study did not give

recit for manual local recovery of the f ailed MOVs. The amount of

- t~ je e a 4 j e, a '.. 3.=, ! .= - a rm , ~, y< ^ .= .,d,; 1 i Q ',,) -.+ := s e- $s..l .C. .= Q-w s ~ k 4 s N a % 3 w. m _o s. o. y, m. . Ns .u I"* M.-,. % c,~./ L. ~* 4 .g.-

• O u

---

'g

} i; 4 t, i, - T*- ba= ; ~ g 6 ' =. 1 =. =. p s sy 'P. t1 -..n D, -.a j s.=. .a .O v.e 1 -a t O, %g =- O. e - y .s. a w, e = o *\\,-~~::~ t-= )=-. w 1 I ~ h. 1.1 l h .- a ~~~ 4

~s ?-

sW .\\ g 6 [m O - D 'i"'- U Q l t = e s U I .'w'... ".=wsa m,' e e, s

e. w t:,,,

=. .se.- e 24 =z h ee.ner me ..ggm. e w.. g .= c w g e==so eg

==+w 3.a e_. -e mg= *:. 1.... iw-or

==. 0EI..;...w

• w*w -g6:=

w, 1 w .= #O y -.:-.** ti w 3:66g o-g g 3#.!:@ts re.I.== te-0 l~"I EI*:i .w ws . I*;*st :: e . =. =-. m., ^Cw**d I=*w s ..e =6 5-I w'de- .= w ee-e .... ggt.e e,__=-- =x l =. .E=.=.-,mw 9 -,.e =>- w .= -.S ..e = -.I... y h. w = -. =

m. z.-

=,.=... _... e - = O ., = * = O MM - = e.4

• 0 s.I.=.w wm e

an..we=.$> .c J =. l e. c,. w I e es g =s ew c c== m s 5, dme 1.8' = = E "E e. e 0,=#.a-aa.3as w=23-45...D* .m e 4, = a, a, t === w e4 -=ge .=ggi., e g o b w - e a = = = = < te.. = f =v.S = e s.. 3 3I.= s. EmabC=3

= = * =2 = 8 4

oI. -me a mye g m e.a. e _a. E. e O we 8e.e.:p.,oc, Ig - .a.'

1. Iw.6m.e.=. - weIce e

-a . Oa. m

==, s eg .e w. g

• hwe=

33 8 <e.ema5sXI=Cr m=.=amsm e e' m a k. a y ee3= .g a6 E st 2 4 = 86=w myg=re. E a a e w>ew t.e ws me = %a 3 6

=

w 55e. T.ai-se--w= a =

• C a a - O.s E. 9 ' w E m w w s I.6. a -

sw. e.3=en-w e to sw I ce w y =M=RE=G.==mw===.==ce e g e w @ g e = =.s,, G B IIS S S e.m e..W w D E de D

1. SSE o.'.e 6 4

,=, g e. wa.gg.g...gm.sg.,, g -..ama..O .== = =r.e--letew1. .W = = m e"

• 6 SwE=3 W

ew d eg g pg.qg.-g = .r_ D e e s = d e.8. =E g " m

==see ew-we e a a 3 i e m e =g = =e S u

• S=

g8 G===WS= hSJ emw. GE.8923.b aw g game..g-a e as m ms..=8Wh.Ie..=.w=e-4.3 ,* e6 32 32 3> = .' w u = =. 3 Swa w ..e.I.gggg, co b = . 3=3.m-q =.. ses aa 2 W = .3 g t=4e>=33MeC=83 S=w w3 3 gg e E O a i e..=g pg- .'= .e =we 3.=lmv e=e =3 I.= a =pa.w E

OI=e og8.

e 3

• = _p 3== = 3.= =a.m a..= = z.

.e-. ..s ne--dw m = - et. e = z .= .. - D e e... e. =. =. = = #.O Sam E ass m..=...g ...-== =4DGEOSSm==htesws-e .w==e=.,, >.=.. ,6.ee u .O = m 4 53 eda e egggeme' c z.. = y = 4 i = 3 N 4 e (), S-e o e t 4 g 4 - as = emo 94 G W w is. Q g t e 2 3 e a w a e L 3 O Q. t w w a i .o O O g Q (- w g Q l w e 3 m = 6 ew g m w e a j e m m a = - 4 3 O h 2

1. 3 9 km e

e e = > a, W a .D e e =w33y w. = a= w*e m e. 4 8 wew w a D e w s D C.l e w..J .wem e.313 SE =g Gew 4 g.M. n N O EME W w W Im e. = m' 8 = b

= y e==a ms me g= m 3 = 3 w soms wwghe gescas-333 SI. E =a sw. = w 3 y 2 e = -e >>ws a = e. =3.a s. . n em e=w a. m e = eem *. wee

==e D m.s e e. e = E e. 3 e.e m Sw..me a3 g - m O O t ee

O 3===30e e

= ee= e e M l emb =24=W .ew.4I S

==ywwe-{- J 193 Db.34wewt W esa 4..

5=S=

hea w w S.E e s a $ = 1 em. e e e DK =33 w=3 m d $Q=O-4 Gem 4= 3 we OeJWee@S O e i.a es eks e=33= Qeamgee.e E..a s w e W e ed 2 .w'we e g =e S.E g e y 2 m = 9 2

== -See l p w w e RW wtew=mS e S

==-34 e2.=- =g=wwSh5 meg = .O g a m3 =a e = a 5e a.

=6 ee

e2

== e= e w=e = a eswa wee = 9 esame e

s

== =so j ', o w = w E t o 3

>=ewSE w

w3w w. wwaw ews 1

= g i Sr 8 Dwew. SEMww W

caer w OwD wh=Ew.D=D 99$>mW64 Q=e f e-m See=S=3= G= w. w e W twE4 WEseme

• w

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Dws =

==SWeSO 90=w 4 g = 5, mw e.e w OOw.m .e .=.ESS.ewGw _. q www am... O _G = O.O m = m m.e w==E- . O w z ww e. w = W =W OewO OO.me3 ww b w eB GD $34O a.s 'f m. =a n= d h e =.E w' h k $ b u.4 WE = g a g 3O=O 1em w

= m e = I S = E. 3 5wI.a

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~ t

.5-=690

W====== O e t w ee== E.=4SO eg aa2a= e.a g w s>=wIe2 * = * = E D m.4 4 9 e t e W45== JS 4 e ma = w =

==ewemedh

== EVE =hWh*C 3 e=E=4 =2 >== ,a . g e e.33=e= + WQ = .iv.. -#=.O sme ZeO e E = 3 3 4 e e E.3 O 6.m a s. t em. e..,O.nr.e..e.a 4 w e C e =m ...e .c w w w -w += r .=== 0 e 5 .=

• e' f e8 "E

=~" "*wo-Ozo = e -a =

• AG=.5=

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• 3 w -

g w t = q 4 O O 6==h42e 23E 3 3 e* 4. ,2,.a{43 L' lL ~~

4 I s-tj. y I ' :. l + 'p1 s ,J g . ~..- .y -5 au s. 9 p D. w a 3 o, w .U {

~

O 3, 4 4 '2 .g f J= e-4 & s. 1 y 2

• y b i

s t G-3,- N 1 .=4.u o d- + 6 n i o.,p. J " '. ' {W J T1*:j t 6 1 0 4.' 8' 5,,

s. *. y 4 a W

a. w a. 4., E 2 s n' ,e, + $ 8 4.3 1"3 0# 3 * *# 4 d 4 .5a 's 4 d 3$g 5D5 tg ' b Y [. e s an c: se - v s. W r# d

n)

~ o e,

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a. T ya><;

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d. l

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r -

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3. e.

.a s! w E 4. I. . i. a 2 .3.. : I. e .E.t = -._r i e.- .= .. k 1.> E m e. s. 5 E t.,.= r a-. .t .... E .s o . 8 E w e-e t e.,, a A .'..E,.=.=. w ...L 8 W. .r,, ke 7 = s s e i . 31 w - - U W g; W g g h4 m w w me en y e. g m.l e ~ g W W W e e -e . m a s.. E3 3 2 2 h h E. G w e e,a Y m .e e O. e i r e z E: s ~ c r e

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== == 0

== W .D. .e= me. e e a w 31 e w.a gi w e en a ' 9 g me in ind w

• =

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== e e. O +

• . =

n as 2 a= >+ 2 e em

== 2' Os g e e a

• 3 ed 24 ime a

e Os ed: Cl E eng w and s a 21 ime O ' W =J g e e

== m e-

• a m

w .a, 2

== 4 s e as g 2' D Cd .O. e 3 3, J O ap

• • 9 26 w

e Os .3 -w e ed en. Si 4.. 2"" ene e aus W 4 w em 4 es one 2 = 1 28

• 2' C

O S w mie 4 .=. Di

== s= 6 ed w a 3 2 a= a em ed 2,

== 2 =* en 2 W 2 Ci .4 as e O >= .ee eE Di w en o us e S g At w me, _J .J en se =A me O se = ime ime

== me C' a 4 Os = ** E4 v4 m e d = > D > m .e =g as (* e a D e 2 O e 4 w es w 2 4 8 2'

== == ed a me W O .e e SE D W am 2 OS 31 O m- & a 2 2 ae = 0 a.s me g = al 2 e 2 e > 2 e 2 4 e Il

==e e e. g ene O =A me O **==== > 0 D Gm ee

== de .e 6 4: - & E' 3 eg D me O== 0 w A K. em W e b,,e en en ed 4. es me 3 S tF9 8 4 en O-

• = = -

og 4 e a e. 3 e es an e en 4 e g w 2 2: em O m,e 0 3 3, g en a.

==3 r g 5 en ** 4m L O ** ee

==m g e e e .e e E O

== art wi

== O Da e O J g O e 2 a O imJ' ed a 2 + e. ws a 3

==e one

== a me-C a a,, a O D tm 5 .g g as ad &= .=n = e E3 4 es e Y ..i .2 i. i

1. y o

.4 t.

' k-g, 'r e h 'y ...i l l3 .s l = t 9 1 .5 w E i

==. .c. 9 o e l-g l = + N e E v E e. 6 _I 2 f.W o,. s e 3 e9 a em 1 e

E l-l E

j a al - a . o.i o. ~ a w. l. 2 "l E 2 : E*3 t-a ~ a E *: = gu ei I~.E n'- "J v; o Ir = =- 6. i> e - e I = -. e w .e, = -- 3-6 .s u m e 6, mW o a a a.

u..

2 .i =. o.W e. I .J. tud On un i =- 22... A. g <= 2, = 88 l d un 9 .o O .and =.=e. h.A = 45 s s e a=*

• 9a
• w m

0 vs e O O Q wt en S "4 me' 2 ..Ad q, Of 3. .1 3 .m.l .e 3: w 4 e g g gg 4 ) W e si

== w e.. 2, -t** t .e w . a. d -M D'l W s u 3 mv 2 3

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• = W.

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== Qi CI

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es

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e a

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== s

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==F Es a., eA

== ==* en m a== O C4 2

• 0 89 EJ D

31 O ar me 3+ w9e w S4 89 D* Gn e nae o U O~ mn K .St .O' e a ,e O e $6 a=m es 8* E1 ( e

• e e

4 \\ = m

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1. 4 ewJ t

• 4 4h 0

me > end ed r E. m 5Y t me ' I Wm0 P . M. f W "M ed e. . ese F mi W Od - a $l ' hp3 W ~g eQ **j O N - em g k U w I @ um Of 2 O 04= a = he2 2 3.m e = g e e

=

~@ E E &, to es 4.- q. a-G w m o e W \\, R 2 3 g

== b. 4 m Q m-m. w b od E

• 1 4

3 O es 2 . u g en

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== g M M ' g 9

== s and 2'

== bl %# me em 4 as 's G E me D a = = >e 3 en en ame hid N me me ME E es - gg a em E 4 e a= b e9 ee. I a me bd O' Os e e 4

== W O O - es a m. E E e we e> e O e .a O z sa W td D

• e e

as F_ - - - - - - - - - - - - - - - - ~ - - - ^ ^ ^ - - - ^ ~ ~ ~ ~ ~ ~ ~ ^^ ~~ f

g

.f.<^ ..s. s f ENCL 0sURE 9 OvB No.: 2150-0C:1 NRCB PS-03, Su;ple e t l. UNITED STATES NUCL E AR R! AVL AICRY C0wul$$1CN OFFICE OF NUCLEAR PCACTOR REGULATION WASWINGTON, D.C. 20555 April 27, ISEB hRC B LETI!i it. 85 03, SUPPL!v!NT 1: POTOR-OFERATED YALVE CCua0N FCCE FAILURES DURING PL AN'T TRANS!!NTS DUE TO IMPRCFER SWITCH SETTINGS Addrestees: All holders of c;erating licenses or construction permits for beiling water reacters (EWES). Putrese: The eu ::se of this su;:le et to NRC Bulletin P5-03 (B 85 03), " Meter-Coerated Yalve Ce r:n Fede4ailures During Plant Transients Due to Irpre:er Switch Settings," is to clarify (1) which' valves are to be included and (?) the meaning of the phrase ".... inadvertent e:uipment operations-(such as in-advertent valve clesures or o;enings)..." as used in the bulletin. Backe :ua.d: B E5-C3, which was issued on November 15, 1955, was prom;ted by the Jure 9, 1955 event at the Cavis-! esse plant in whi:h the inability to ree:en two valves that had inacsertantly been closed led to a loss of bcth trains of the auxiliary feed-water sys*em. Discussien: Review cf the responses to B 85-03.from BWR facilities, including these from the ow*ers gr up, has incicated that there is a misunderstandin in regard t: the related issues of (1) which valves are to be included and (g) the meaning 2 of the phrase "... inadvertent ecui:nent operations (such as inadvertent valve closures or openings)..." as used in the bulletin. The first misunderstanding pertains to which valves are addressed by the bulle-tin. As written, the action pertion of B 85-03 applies to rotor-operated valves in selected systers that "... are recuired to be tested f: operational rea::1-ness in accordarce with 10 CFR 50.55a(g)..." At the time the bulletin was issued, the staff believed that the inservice testing programs recuired by 10 CFR 50.55a(g) were applicable to rest, if not all, of the safety-related valves in the selected systems. However, recent conversations with the cwrers-grou: and several licersees have in:1cated that a number of valves in taese systems are r.:rmally kept in'their sa'ety positiens and are not covered by t e EEC4210C15

c...y-

• f 4

NDCB 85-03, Supplement I-April 27,1938 Page 2 of 4-inservice. testing ~ program. However, if the pre er precautions are not taken, i these valves which are normally properly positioned could be mispositioned, either be'ere or during the initial phases of an event. This would render b the safety system ineperable Urless the valves could be repositioned to the . proper' position. Therefore, the heading of the action section of R 85 03 has been revised to include all safety-related valves in the selected systems. 1 The meaning of the phrase "... inadvertent eQuipeant operation'(such as in-advertent valve closures or openings)..." used in action item a of the E l bulletin can also be misunderstood. This phrase stems fecm the desire to address the salient feature of the Davis-Besse event -- nacely, the inability to reposition either of two redundant valves that had been mispositiered R earlier in the event. Although it was not the intent of the bulletin to expand the design-basis events for plants, it was intended to ensure the - high reliability of individual safety systems. To this end, and given the I-chain of events associated with the Davis-Resse event, the staff felt that the enly way to ensure this high reliability was to verify the ability of all vahes to recover from mispositioning. Therefore, action item a of P B5-03 has been revised to clearly indicate that each factor-operated valve j: must be able to recover from an inadvertent mispositioning. This revision to B E5-03 may expand the number of valves addressed by soee licensees. In addition, s e of these licensees may have already completed their scheduled activities to comply with the bulletin. Therefore, the time limits for cerpleting all the activities (action item e) have been modified to allow additional time for these licensees who have already completed their planned activities. Acticas fer all BWD Holders of Oceratine Licenses or Cerstruction Permits: For safety-related motor-operated valhes in the high pressure coolant infectionf p ccre spray and reactor core isolation cooling systems not included in the acticns planned or completed.in response to the original bulletin, develop and implement a program to ensure that valve operator switches are selected, set, and train-~ tained pr::erly. This should include the following: -a. Review and document the design basis for the operation of each valve. This documentation should include the maximum differential pressure ex:ected during both opening and closing of the valve for both noreal and abnormal events to the extent that the events are included in the exieting, approved design basis (i.e., the cesign basis documented in pertirent licensee submittals such as FSAR analyses and fully approved 0;erating and emergency procedures, etc.). In addition, when determining misocsitioned,{ferentialpressureforvalvesthatcanbeinadvertently the maximum di the fact that the valve must be able to recover from i such mispositioning should be included. Any motor-operated valve that is net blocked frem inadvertent operation frem either the centrol reem. the meter centrol center, or the valve itself shculd l-be considered ca:able of being misresitionec. L l - + + - w- * -- - - -

.) {l** 4 l' NRCB 85 03, Sup:le v nt ; April 27, 1983' Page 3 of 4 b. Perferm actien item b of the original bulletin-for any additional valves identified above. The intent is to provide assurance that a program exists for selecting and settint valve cperator switches to ensure a high reliability of safety. system valves. If changing the switch settings is not sufficient to ensure the capability for repositioning 4 particular mispositioned valve, a jus;ification for. continued operation should be provided in the bulletin response if the licensee does not elect to implerent additional actievs, such as ad.inistrative' or procedural controls or equipment modificaticns, to minimize the likelihood of va1ve malfunction'. c. Perfom action item e of the original bulletin for any additional valves identified above, d.- Perfom action item d of the original hulletin for any additional valves identified above, Within 30 days of receipt of this supplement, suteit a written re;:rt to e. - + NRC that,'for any additional valves: (1) provides the revised results or item a, above and (2) contains a schedule for completion of items b through'd, above. 2. No changes from the schedule for complying with the original bulletin are anticipated for plants with-an OL that, as of the date of this su:;1eMnt, had not yet begun the refueling outage during which the activities in the original bulletin were scheduled to be accerplished. 2. Plants with an OL that, as of the date of this supplement, have co. - pleted their planned activities in response to the original bulletin have until the cornpletion of their next refueling outage to coeplete any additional activities resulting from this supplement. The final report covering the activities already completed in response to the original bulletin shall; be submitted in accordance with the original schedule. 3. No changes from the schedule for comolying with the orieinal bulletin are anticipated for plants with a CP. f.- Revise the report requested by the original bulletin to include any addi-tional valves. This revised report shall be submitted to the NRC within 60 days of completion of the program for the additional valves. Additicnal' c lated Gen +ric Coemunications: e B 85-03 identified a number of related generic communications. Since its issuance en Nevember 15, 1985, the following accitional related inforeatien notices have been issued: a. Infer ation Notice No. 85-29 " Effects of Changing Valve Motor-Operator Switch Settings,"'was issued on April ?S, ige 6.

f:: .. i+ '* t 9 hPCA 85-03, Su:;1erent 1 April 27, 19E0 Page 4 of 4 ~ b. Inferration Netice No. 86-93. "!!B 85 03 Evaluation of Motor-0perators teen. tifies Improper Torque Switch Settines," was issued on November 3,1955. Inferrat10n Netice No. 87-01, "RWR Valve Misalignment Causes Degradation ' l c. of ECCS in PWRS." was issued on.1snuary 6,1957 ? The written reports requested above shall be addressed to the U. S. Nuclear i Rege'.atory Com. mission, ATThi Docurent Control Desk, Washington, D.C.

2C555, under oath or affirmatien under the provisions of Section 182a Atomic Ener;f l

Act cf 1954, as amended, in addition, a copy shall be subr.itted to the a:oro-priate Re;!onal Administrator. This. recuest for inferration was approved by the Office of Panage ent ard Bud;et under clearance nurber 3150-0011. Corrents on burden and duplication should be i directed to the Office of Management and Budget, Re:erts Manage ent Room 37:2, New Executive.0f fice Building, Washington, D.C. 20503. Although no spedific recuest or recuirement'is intended, the. tine required to-I ccrpiete. etch action item above would be helpful to the NRC in evaluating the 'I cost of this bulle_ tin. I If you have any cuestions about this matter, please contact the technical l centact-listed belew or the a;;repriate NRR project manager. l i f Yd', [i, Director L Charles-E. Ross Divisien of Operational Events Assess ent .{ Office of Nuclear Reactor Regulation l. i Technical

Contact:

Richard J. Kiessel, NRR (301) 492-1154'

Attachment:

List of Recently Issuec NEC Bulletins l l I s I e m 4

. J,* . 'n j f e Attachment NRCB B5-03. Svegie,ng.1; April 27, 19P.a . LIST OF DECENTLY ]$$Ug0 NRC BULLETINS Eviletin g,g, 7 No. Subiect issuance issued to 87-0?, Faster;er Testing to 4/22/88 Su;plerent 1-Detemine Confomance All holdS C# Ol3 {ecificationsth Applicable Material. pe,, r re et -r 28-03 InadecuateLatchEngageentI 3/10/88 All holders of Ols in HFA Type Latching Relays or cps for nucient Manufactured by General power reactors. Electric (GE) Cer:any 85-02 Rapidly Pro agating Fatigue 2/5/88 All holders of OLs' -{rJksinSteamGererator or cps'for W-designed-nuclear power reactors with steam generators having carbon' steel support plates. 83-01 Defects in Westinghouse 2/5/E8 All holders of CLs Circuit Breakers or cps for nuclear power reacters. 87-CP Fastener Testing to 11/6/87 All holders of Ols. Cetermine Confon ance or cps for nuclear with Applicable Material power reactors. Specifications 87-01 Thinning of Pipe Walls in 7/9/87. All licensees for Nuclear Power Plants-nuclear power plerts holding an OL or CF. 25 04 Cefective Teletheracy Timer 1C/??/86 All NPC licensees That Fay Not Teminate Dese authorized to use ecbalt-60 teletheracy units. 86-03 Pctential Failure cf Pultiple 10/8/86 All facilities ECCS Pures Due to Single holding an OL or Failure of Air-Operated Valve CP. in Pinimum Flow Recirculation Lire-OL = 0:erating License CP =-Constructien Per-!t .}}