ML19262F381

Minutes of the Public Meetings on Generic Letter 89-04, Dated October 25, 1989

ML19262F381
Person / Time
Issue date:	10/25/1989
From:	Partlow J Office of Nuclear Reactor Regulation
To:
References
GL-89-04 NUDOCS 9001040128
Download: ML19262F381 (85)

Text

OCT I 5 1889 TO: ALL LICENSEES OF OPERATING NUCLEAR POWER PLANTS AND HOLDERS OF CONSTRUCTION PERMITS FOR NUCLEAR POWER PLANTS, AND INDIVIDUALS ON THE ATTACHED DISTRIBUTION LIST

SUBJECT:

MINUTES OF THE PUBLIC MEETINGS ON GENERIC LETTER 89-04 In June 1989, the NRC staff held four public n~etings to discuss Generic Letter 89-04, "Guid3nce on Developing Acceptable Inservice Testing Programs."

This generic letter, issued on April 3, 1989, provides guidance aimed at improving inservice testing (1ST) programs at nuclear power plants.

Attached for the use of licensees and construction permit holders in implementing the generic letter are the minutes of the public meetings.

These minutes contain a summary of the openiny remarks by NRC representatives at the meetings and responses to all of the questions raised at the four public meetings. Licensees and permit holders should review the entire package because specific staff guidance must be considered in the context of all questions and responses. These minutes are not intended to convey any new requirements and are 110t considered a backf it.

Please dirE=ct questions or comments regarding the meeting minutes to the dppropriate ~RC Project Manager.

James G. Pari~c~

As~ociate Director for Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated Distribution:

Central Fi le EMEB RF TScarl:Jrough TSullivan LMarsh JRichardson FMi rag lid JPartlow 9001040128 89102 5 t I *~; ;. , '**(,*t,. PDR MISC I &cE PNU Document Udrne: ALL LICENSEES

NRR Project Managers PMs for plants which are not listed on either Table of the GL should ensure that all plant specific TACs for review of IST programs issued prior to the issuance of the GL are closed out as noted in J. Hayes' May 17, 1989 memorandum. In addition, there were some inadvertent omissions in Tables 1 and 2 of the GL.

AN0-2 and Dresden 2 &3 should have been included as Table 1 plants and Catawba Units 1 &2 should have been included as a Table 2 plant. TACS for these plants, therefore, do not appear on the attached list except for Dresden 2 &3 which should be deleted.

Original Signed By:

James G. Partlow Associate Uirector for Projects Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/enclosures:

J. Taylor H. Thompson J. H. Sniezek Division Directors, NRR Associate Directors, NRR Project Uirectors, NRR Regional Administrators J. Conrad, CRGR C. Berlinger, DOEA S. Treby, OGC NRR Technical Assistants DISTRIBUTION Central Files Glainas JPartlow EAdensdrn GHoldhan JHayes SVarga POI I -1 R/F

• SEE PREVIOUS CONCURRENCE

_,,,

NAME :JHayes:sw :EAdensam :GLainas :SVarga :GHolahan

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:------------:------------:------------:------------:------------:------------:---------

DATE : 11 /03/89 :11/03/89  : 11 /03/89 : 11/06/89  : 11/09/89 OFFICIAL RECORD COPY Document Name: MEMO TO PMS

LIST OF TAC NUMBERS FOR PLANTS NOT IN TABLES l OR 2 OF GENERIC LETTER 89-04.

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DOCKET PROJECT RITS

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11 7~75'~

NO.

50-313

&g i i i 58 334 POWER PLANT Arkansas 1 Arka11sas 2 Seaver Valley 1 MANAGER Harbuck Pe! ltU"Y Taffl INIT NCH CPP PST 141 '157 50-412 Beaver Valley 2 Tam PST 1+1S:8 50-438 Bellefonte 1 Auluck RCA 6Q 439 Bellefonte 2 A~h,ek RGA 50-155 Big Rock Point Pulsifer RPV iQ 4ii a,ai~wee~ 1 Sa"d' SPS SQ 457 8rii~W99~ 2 SaAds SPS SQ 296 811e~tPIS Fel'l'Y 3 Gea,s GEG 58=324 Br~P1sw;e1< 2 Te~r;g,ty/Le EGT/NAL SQ 4i4 9:Y,eA 1 81 shaA LNO SQ 455 S;t,e,. 2 81 sha" LNO iQ 4i3 Galla\,a~ l Ale,tieA TWA 59 317 Galvert Cliffs l HeNeil SWH iQ 31i Cilvart Cl ifh 2 N,Nai 1 $WM 59 413 Gatawha l daeee~r KHd

-SQ 470 CE CESSAR=F l~ertyefi TBK

~Q~Q~61t17~5~-tC'C"E~Gre:E~SS~A1-HR-1B1:ttC:--~~---,K~e!flrt;~89fR~

1 T8~

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,etttA,...__ _ _ _ _. ,M1"*e~kifMMiHaA~*-----l:l:'I iQ 44i G&IRiR6~e Peak 1 Ma lle;i

• VHM-

---::::::=::==::::---~5Q~4~46~-C~o~m~i~R,~A~i~Pi~i~k-2ie------Na-~l~1~~.,-* VHM iQ 3li Geek 1* Giitter IJG SQ 316 Geek 2 Gitter -lJ6 50-298 Cooper O'Connor PWO 50-302 Crystal River 3 Silver HAS 58=346 ea,;9 ee,,e Wafflbaeh TW 50*275 B; ah le Ca,., et1 1 Ree et IIAR 59 323 e;at,1,e 6ar,ye" 2 Reeef IIAR 7t'Z6.i- se-01e Bre:sden l E11iekseR PS£ 7"'14!1 SQ 237 E)FesdeR 2 Siegel X8S

• 1.Y16!! liiQ 249 DresdeR J Siegel XBS 7t' Zle:f 50-331 Duane Arnold Ha 11 JRH

==:::::::::~-~Q~Q~6~6~g-~EP~R~l-~====~k~e~Rf'!-==--:WAL iQ ,4i ~irle;, 1 ReeYes EAR

-56=016 Fe,m; 1 EFiekseR Pit.

59 341 Fermi 2 Sta"' Sf~

50-333 Fitzpatrick LaBarge OWL iO 2&5 Vert CalRO~R 1 MilaAO PZM 58 267 Fert St, VraiD Heitner KLH--

56*605 SE*ABWR Scaletti DCS 50-244 Ginna Johnson AGJ 50-416 Grand Gulf 1 Kintner LLK POWER PLANT CHART 1 JUL 1 3 1989

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DOCKET PROJECT RITS NO. POWER PLANT MANAGER INIT 50-213 Haddam Neck Wang ADW iC 400 Mirri, l Se;ker RBI iQ a66 Mate~ 2 reeker LXC iC ii4 Mepe GPeek Shi'fPaki CSE 7 41- 77() 50-~47 Indian Point 2 Brinkman SBU 7'177/ 50-286 Indian Point 3 Neighbors JON 50 305 Kewatj"ee 6ody AOG 50 499 La Cre!!e c, icksou PBE 50-373 LaSalle 1 Shemanski PCS 50-374 LaSalle 2 Shemanski PCS 58=352 L;merick 1 Clark RJC SQ 3S3 Limeriek 2 Clark RJC Zf ZZ'/- 50-309 Maine Yankee Sears PMS iQ i69 McGw i Pe 1 llood OSH 60 i7Q MeSw ire 2 lleed 0511 z¥1zs 50-245 iQ 335 Millstone 1 Hilhte"e 2 Boyle 11 issi11g MLB 65V 50-423 Millstone 3 Jeff@ OHd 50-263 Monticello Long WAL 58-Ez!B Niue Mi le Pe;,it l Slt,uo" P*Ms 50-338 North Anna 1 Engle LBE 50-339 North Anna 2 Engle LBE 50-269 Oconee 1 Wiens LHW 7YZ i0 50-270 Oconee 2 Wiens LHW 50-287 Oconee 3 Wiens LHW

z. 50-219 Oyster Creek Dromerick AID 50-255 Palisades DeAgazio ABO 50=528 Pele ¥erde l Che" TC£

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-=====-~57661"11"*~53~8f---f:PL.aa4'1ee-JJV~ef11F(j!ffe!-;l3----tD~a.)l..v~i5..,____-fflffl 50 171 Peaeh Bot:t:om 1 c, ielcsou PBE

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ei~Q~2!+7+7-+i-Pee;aHe~h-1B~eHt~temmrt--r2---11~481erNt:-t;1't"_ _ _ -REM 50 2]8 Peaeh BetteM 3 Ha,tiA REM 50-440 Perry 1 Colburn TGC 50-293 Pilgrim 1 McDonald OGM 50-266 Point Beach 1 Swenson was 7'/:71 7 50-301 Point Beach 2 Swenson WOS 50-282 Pra;r;e l!la"e 1 0;1aP1Ai DCD 50-254 Quad Cities 1 Ross TER 50-265 Quad Cities 2 Ross TER 58=312 Reneho Seco Ka lma11 BCK 58=261 Rob;"!On 2 Lo RHL 7'1-7 90 50-272 Salem 1 Stone JTF POWER PLANT CHART 2 J

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G DOCKET NO. POWER PLANT PROJECT MANAGER RITS INIT 7~Z~ 50-311 Salem 2 Stone JTF

'Z'i:1 50-206 San Onofre 1 Tranmell CMT 59 351 5,a,. e.. ef ,.e 2 11;ekman -f}tti so a&2 SiR 9RefP'e 3 Hie kffla,. BtlZ iQ 44, i@al.H*eek l Ne,ses lJXN 59 444 Seal,JPeek 2 Nerse! 'JXN i;g 327 Se'fY&yaA t Be,.ehew JND 58-328 Setit:10, ah 2 Be,.ehew JND Zt Z74 50-322 Shoreham Brown SWB

-** 58-490 Sot:1th Texas 1 a;ek GFD 59.499 Set:1th l'e,cas 2 Biek 6F9 Z'IZf.t. 50-335 St. Lucie 1 Norrh JAN 50=389 St. LtJe;e 2 N-ot r h JAN SQ 395 SYRA@P 1 M.ye& cl GIi iQ 2iQ Swrry 1- Suckl&y 8'8 5g ,si iwrry a 81:1ekl e~ BEB 50-387 Susquehanna 1 Thadani MBT

~~~~ 50-388 iQ 48Q Susquehanna 2 T~Fee Mile IslaA~ l Thadani He,Aart MBT RIHf 59 329 n,ree Hile 15 lartd 2 Masrtik MTM z11z1.. z 50-344 Trojan Bevan RBS Zt. Z'J.l. 50-250 Turkey Point 3 Edison GEE Z~Z!1 50-251 Turkey Point 4 Edison GEE z~tol'J 50-271 Vermont Yankee Fairtile MBF iQ 424 '1egt le 1 llopk;ns dSli SQ 42e 'Jegtle 2 ltepk;ns ~

59 382 Waterfe,.d w;gginton 6~W Z6fl'Jl. 50-390" Watts Bar 1 Auluck RCA iQ 3g1 Wil tt, iiP 2 At:1lt1ck ,teA 5g &Ql RE5AR SP/98 ~eiijOrl TBK SQ (iQ WNP 1- Kertyon--~ TBK 58 397 WNP 2 Sa11Pwe,.th ~s se se&- WNPw3 www,-.-*-*--i<eny01't ;aK i;g 482 W91f GFeek PiGkett ~-01:P 74i42,50-029 Yankee Rowe Fairtile MBF iQ ~gi :liQR 1 Patel CBf>

iQ ~04 :1 jg~ ~ ----Pa te-l--- fill SQ QUi '11B~IA iS:iCkliQA F'BE-POWER PLANT CHART 3

TO: ALL LICENSEES OF OPERATING NUCLEAR POWER PLANT~ AND HOLDERS OF CONSTRUCTION PERMITS FOR NUCLEAR POWER PLANTS, AND INDIVIDUALS ON THE ATTACHED DISTRIBUTION LIST

SUBJECT:

MINUTES OF THE PUBLIC MEETINGS ON GENERIC LETTER 89-04 In June 1989, the NRC staff held four public meetings to discuss Generic Letter 89-04, "Guidance on Developing Acceptable Inservice Testing Programs."

This generic letter, issued on April 3, 1989, provides guidance aimed at improving inservice testing (IST) programs at nuclear power plants.

Attached for the use of licew*.:es and construction permit holder5 in implementing the generic letter* are the minutes of the public meetings.

These minutes contain a summary of the opening remarks by NRC representatives at the meeti11gs and responses to all of the questions raised at the four public meetings. Lir.ensees and permit holders should review the entire package because specific staff guidance must be considered in the context of all questions and responses. These mi11utes are not intended to convey any new requirements and ~re not considered a backfit.

Please direct questions or con111ents regarding the meeting minutes to the appropriate NRC Project Manager.

9. \ \~;&s:

t:

Ja es G. Partlow As oc1ate Director for Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated

'

1ST SERVICE LIST G. A. Arlottu NL 007 Region I:

R. Baer NS 217A J.P. Durr R. J. Bosna k NL 007 P. K. Eapen E. J. 8rown MNBB 2104 H. L. Brar.1mer 12 0 17 W. E. Campbell NL 2178 Region II:

F. C. Cher11y NS 217 A G. A. Bel isle K. C. Demps~y 9 H9 S. G. Tingen B. K. Grimes 9 A 2 C. I. Grimes 7 H 17 C. Ransom, EG&G, Idaho Region I 11 :

J. Huang 9 G 12 D. Danielsen R. C. Li 9 A 20 J. J. Harrison B. D. Liaw 7 E 23 G. C. Mi 11 man NS 217B J. D. Page NS 217B Region IV:

J. E. Richardson 8 H 1 I. Barnes C. E. Rossi 11 E 4

0. O. Rothberg NLS 302 H. K. Shaw 9 H8 Region V:

J. Sni~zek 12 G 18 C. A. Clark E. J. Sullivan 9 H2 R. J. Pate

Region I 6/5/89 NAME ORGANIZATION Paul Cervenka GPUN Jim Con no 1ly PSNH/NHY Bill Kittle PSE&G-Salem Cornelius Coddington PP&L Jeffrey Lomm NYPA Deborah K. Schultz PSE&G John Rigert LILCO Bob Knight GPUN Shafi Rokerya NYPA Noah Fetherston Yankee Atomic Saf io Toth NYPA Patrick Sheldon Yankee Ator11ic Francis Kamiski PSE & G Joann West Beaver Valley Clive Callaway NUMARC John T. Lindberg PP&L Douglas B. Ritter PP&L Eugene Perry Con Edison Jeff Neyhard Niagara Mohawk Joan F. Etzweiler Con Edi son J.R. Bashista HH-1 Albert A. Koehl NES V. C. Ruppert PECC R. Binz IV PSE&G D. Wallace NYPA Fitzpatrick K. Woodard NYPA Fitzpatrick R. Hal adyna BECO J. L. Sabina BECO W. G. Carro 11 BECO

Region I I 6/8/89

- - - - - - - - - - - - - - - - -ORGANIZATION NAME

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• W. E. Galbreath Duke Power J. A. Witherspoon Duke Power John Zeiler USNRC Paul Burnett USNRC/DRS/TPS J. J. Lenahan NRC M. Belford Southern Company Steve A. Saunders SERI Grand Gu 1f J. S. Jackson TVA C. L. Dunkerly Baltimore Gas & Electric J. M. Duke TSE Eben Burns BCP Techical Services, Inc.

Phi lip J. North Duke Power Company Robin Dyle Southern Company Services, Inc.

W. E. Campbel 1, Jr. USNRC/HES Gary Smith Sys tern Energy Wavel Justice System Energy Stephen E. Mohn Florida Power & Light A. L. Koon South Carolina Electric &Gas Gene G. Sow lt SCE&G V. C. Sununer Station Ken Kmetz Enercori Services John Zudans Florida Power & Light A. Rona1d Jacobstein Art Caud i 11 Georgia Power Company Herbert P. Walker Georgia Power Company Sid Burns /1 labama Power Company Bud Syx GeorgiJ Power Company Kris Miller Florida Power & Light Jim Holton Florida Power Corporation Mark Dryden Florida Power & Light Stan Pruitt Carolina Power & Light Al Schneider Enercon Services John Kin Virginia Power Peter Tay 1or NRC Arthur Szczepaniec NRC Karl Jacobs N.Y.P.A.

John B. Lee, Jr. Virginia Power S. L. Nader Duke Pm,;er Company John J. Hayes, Jr. NRC

Region III 6/13/89 NAME ORGANIZATION Bruce J. Sheffel Detroit Edison Company Larry L. Campbell Toledo Edison Pau 1 Shemanski NRC James R. Harkness Colilmonwealth Edison - llyron Station Laurence Attochman NUTECH Engineers Ua le L. Jones NUTECH Engineers Gctry E. Knapp Commonwealth Edison - Quad Cities Robert T. Kerestes Illinois Power Co. - Clinton Power Station Gary J. Roesner Union Electric - Callaway Donald W. Zebrauskas Commonwealth Edison Mort Khazrai Toledo Edison Stephen J. Coleman NDX Corporut ion Roger Dale Sogruoe LaSalle Station (CeCo)

David C. Uherek LaSalle Station (CeCO)

James F. Srni th NRC Uav id Maz 1iach fWX Corporation Kenneth Kelber Ceco LaSalle County Timothy P. Jaeger Combustion Engineering John Ozol Commonwea 1th Edi son Steve Sovich Duyuesne Liyht Company Dave Jones Duquesne Liyht Comµany Joe Edom Iowa Electric Light & Power Norm Peterson Iowa Electric Light & Puwer Mark Hdrris IMPELL Patrick M. Fin11emore Wisrcnsin Public Service Corp.

Gordon Svendsen Commonwealth Edison - Zion Station Jeff Grzeszczak Commonwealth Edison - Braidwood Station 3ary Bal Commonwedlth Edison - Rraidwocd Station Pat Tobin Northetn States Power Doug Kerr NUTECH Engineers Mark Horbaczewski Dresden Stdtion Vince Treagne Pu1nt Beach Nuclear Plant 8rent Metrow Illinois Dept. of Nuclear Safety Lawrence Sage Illinois Dept. of Nuclear Safety Steven M. Hutton Energy Testing Services 13 i 11 Carro 11 Pilgrim Station Joseph L. Sabina Pilorim Station John C. Rivers Cleveland Electric Illuminating Company Vi nee Co nee l Perry Power JJ 1ant Stephen Forsha Impell Corporation

,tephen P. Brown tWTECH Engineers Uennis Carlson Northern States Power Frank Dur,der Commonwtalth Edison - Dresden Station Rus.s Taniminga Conimonwea lth Edi son A. John Birk-it Consumer Por:~r Company ~ Gig Rock Point Jeff Cook Gmdha Public Power District George Schrader Consumers Power Comµany Uavid 1<.a11uc:h Impell Corpuration - ~r~sden Station Steve Bell I1linoi5 Puwer

Region IV & V 6/15/89 NAME ORGANIZATION Mary Miller NRC Robert McWilliams Arkansas Power &Light Ken Trippel Houston Lighting &Power - South Texas Project Steve Wideman Wolf Creek Nuclear Operating Corporation Alan Harris Waterford 3 Bruce Wadley TU Electric - Commanche Peak Don Ringle TU Electric Clifford Clark NRC John Arhar Pacific Gas & Electric - Diablo Canyon Terry Pe 11 i sero Pacific Gas & Electric Co. - Diablo Canyon Steve L. Scanvnon Supply System Ali Abbasi Southern California Edison John DeBonis Stone &Webster (c/o TU Electric)

Paul Croy So. Cal Ed.

Uon Hi ckrnan NRC Roe ky Schultz Cooper Nuclear Station Wayne Walling Gulf States Utility - River Bend Steve Asztalos Cygna Energy Services

PROJECT MANAGERS Project Mc1nager Ext. No. Backup Manager Plant - -- -Ext. -No.

Arkansas 1/2 Crd ig Harbuck 21341 Chet Pu~ lusny 21336 Beaver Valley 1/2 Peter Tam 21307 Frank Orr 21321 Bellefonte 1/2 TVA Raj Auluck 20759 Big Rock Point Robert Pulsifer 21330 Lynn Kelly 21305 Ora idwood 1/2 Stephen Sands 21396 Leona rd O1shan 23018 Browns ferry (TWA) Gerry Gears 20767 8runswich 1/2 Edmond Tourigny 21474 Tommy Le/Les Kintner21455 Byron Leonard Olshan 23018 Stephen Sands 21396 Ca 1laway Tom Alexion 21387 Randy Ha 11 21391 Calv~rt Cliffs 1/2 Scott McNei 1 21438 David LaBarge 21421 Catawba 1/2 Kahtan Jabbour 21496 Darl Hood 21442 CECessar Tor:i Kenyon 21120 Clinton John Hi c kma n 23101 Paul Shemanski 23017 Comanche Peak (SP) Melinda Malloy 20439 Cook 1/2 John Stang 21328 Tony Gody 21305 Cooper Paul O'Connor 23026 Doug Pickett 21336 Crystal River 3 Harley Silver 21470 George Wunder 21480 DavisBesse Tome Wambach 21323 Joseph Gi; ter 21379 Diablo Canyon 1/2 Harry Rood 21352 Roby Bevan 21361 Cresden 2/3 Byron Siegel 23019 Thierry Ross 23016 Duane Arnold Randy Ha 11 21391 Tom Wambach 21323 EPkl Paul Leech 21103 Dino Sea letti 21104 Farley 1/2 Edward Reeves 21457 Jack Hayes 21456 Fermi 2 John Stang 21328 Lynn Kelly 21305 Fitzpatrick David LaBarge 21421 Scott McNen 21438 fort Caltiourn Patrick Mnano 21347 Ft. St. ~rain Keneth Heitner 21333 Edward Tomlinson 23024 GEABWR Dino Scaletti 21104 Paul Leech 21103 Ginna Carl Stahle 21435 Patrick Sears 21429

&rand Gulf 1 Les Kintner 21458 Edmond Tourigny 21474 Haddam Neck Alan Wang 21313 Michael Boyle 21308 Harris 1 Dick Becker 21465 Edward Rec,ves 21457 Hatch 1/2 Lawrence Crocker 23049 Jon Hopkins 21494 Hop~ Creek Clyde Shiraki 21445 Stu Brown 21444 Indian Point 2 Don Brinkman 21420 Don Neighbors 21409 Indian Point 3 Don rleighbors 21409 Don Brinkman 21420 Kewaunee Joseph Gi i tter 21390 Tom Alexion 21389

2 Plant f'roject Manaoer Ext. No. Backup Manager

- -

Ext. No.

laSa lle Pau 1 Shemanski 23017 John Hickman 23101 Limerick 1/2 Hichard Clark 23041 Robert Martin 21426 Maine Yankee Patrick Sears 21437 Carl Stahle 21435 McGuire 1/2 Dearl Hood ~1442 Kahtan Jabbour 21496 Millstone 1 Michael Boyle 21308 Alan Wang 21313 fl11lstone 2 Guy Vissing 21314 David Jaffee 21312 Mi 1lstone 3 David Jaffe 21312 Guy Vissing 21314 Monti ce 11 o John Stefano 21309 Dom DiJanni 21324 Nine Mile Point 1/2 Marylee Slossom 21412 Robert Ben~dict 21402 North Anna 1/2 Leon Engle 21484 Bart Buckley 21452 Oconee 1/2/3 Leon Engle 21404 Kahtan Jabbour 21496 Oyster Creek Alexander Dromer1ck 21301 Ronald Herman 21320 Palisades Al OeAgazio 23063 Polo Verde 1/2 Terence Chdn 21366 Micheal Davis 21368 Palo Verde 3 Micheal Davis 21368 Terence Chan 21366 Peach Bottom 2/3 Robert Martin 21426 Richard Clark 23041 Perry Timothy Colburn 21389 David Lynch 23023 Pilgrim 1 Daniel McDonald 21436 Vernon Rooney 21440 Point Beach 1/2 Warren Swenson 21386 Timothy Colburn 21369 Prairie Island 1/2 Albert DeAgaqio 21323 Theodore Quay 21315 Quad Cites 1/2 Thierry Ross 23016 Byron Siege 1 23019 kdncho Seco George Halman 21367 Steve Reynolds 21366 River Bend 1/2 ~alter Paulson 23028 David Wigginton 23027 Robinsor, 2 Ronnie Lo 21463 Les Kintner 21458 Salem 1/2 Jim Stent! 21422 Mohan Thadani 21427 San Onofre 1 Charles Tramme 11 21363 Donald Hickman 21380 San Onofre 2/3 Oona 1d Hickman £1380 Harry Rood 23062 Seabrook 1/2 Victor Nerses 21441 Norton Fa i rt ile 21443 Sequoyah ( TVA) Thomas Rotella 20760 Shoreham Stewart Brown £1444 Clyde Shiraki 21445 South Texas George Dick 21326 Anthony Bourn ia 21345 St. Luc1e 1/2 Jan Norris 21483 Gordon Edison 21471 Summer 1 Jack Hctyes 21456 Ronnie Lo 21463 Surry 1/2 Bart Buckley 21452 Leon Engle 21484 Susequehanna 1/2 Mohan Thadani 21427 Jim StonP 21422 Three Mile Island 1 Ronald Herman 21320 Alexand~r Dromerick 21301 Three Mile Island 2 Michael Masnike 21373 Lee Thomas (717) 9481151 Trojan Roby Bevan £'1361 Charles Trammel 1 21363 Turkey Point 3/4 Gordon Edison 21471 Jan Nord s 21483

3-Plant Project Manayerl Ext. No. Backup Manager Ext. No.

- - - -

Vermont Jankee Vernon Rooney 21440 Daniel McDonald 21436 Vogtle 1/2 Jon Hopkins 21494 Lawrence Crocker 23049 Wat~rford 3 David Wigginton 23027 George Dick 21326 Watts Bar (TVA) Rejender Auluck 20759 Wapwr Thomas Kenyon 21120 Guy Vissing 21101 WNP2 Robert Sa11Morth 21304 John Bradfute 21381 Wolf Creek Doug P;ckett 21336 Paul O'Connor 23026 Yankee !<owe Morton Fairtile 21443 Victor Nerses 21441 Zion 1/2 Chandu Patel 21395 Lloyd Zerr 23100

MIIWTES Of THE PUBLIC MEETINGS TU DISCUSS GENERIC LETTER 89-04 11 GUIUANCE ON DEVELOPING ACCEPTABLE WSERVICE TEST ING PROGRAMS 11 On April 3, 1989, the NRC issued Generic Letttr BY-04 which provides guidance aimed .lt corr~cting several weakne~ses found by the NRC staff in inservice testing (1ST) programs at nuclear power pl~nts. The issudnce of the gcreric letter is part of an uvera 11 NRC staff effort to imp rove JST programs. The staff also has a long-term goal of making IST proyrd11,s essentially self-imple-menting such that estaLlishment of a proper !ST progran, \'IOUlJ be detf.rmined through audits and inspections at the µl~nt site rather than by stdff review before a prosra111 is implemented.

The NRC stdff held four public meetings to discuss Generic Letter 89-04 ~Jith holders of nuclear power plant operating licenses dnd construction pern1its.

These meetir!~S were 11oticed in the Federal Register (54 FR 23305) un ft,ay 31, 1989. In addition, the NRC Project Managers were requested to inform the individual lic;ensees of the meeting dates and locations. The meetings took place in King of Prussia, Pennsylvdnlu on June 5 for NRC Region l µlc1nts; Atlanta, Georgia, on June 8 for Reyion II µlants; Chicago, Illinois, on June 13 for Region III plants; and South San Fruncisco, Culifornia, on June lb foi*

Regions IV and V plants.

Transcripts of the Chicago and San Francisco me~tings were taken to pr0vide assistance in th~ pr~parrltion of meeting minutes. The minutes will be distributed to meeting attendees whu provided their address and holaers of nuclear power plant operating licenses and construction permit::;. In addition, the me~ting mir;ute,s together with the transcripts w11*1 be placed in the NRC Public Document Room.

At each meeting, a manageme11t repre~entative of th~ hRC Region \';here the meeting ~,as held provided opening remarks. Followiny those remarl<.s, Tad Marsh, Chief of the Mechanical En~ineering Br~nch (~MEB) 0f the Offict of Nucl~ar Reactor Regulation (NRR) discussed till; objective of i11s~rvice testing, it~

regulatory foundation, and prublems fuur,ci in IST prugrarns. He also provided a brief uverview of the rmc effort tu improve inservict:: testing at nuc1~.:1r power plbnts. Ted Sull1van, Section Chief ut the lnservice Testing section in the MEB, tlien pre~ented a detailed explanat.ion of Genc-ric Letter 89-04 and its applicability. Summaries of tht*se three prr*sentations ar~ pruv1<1ed below.

Copies of the s l 1d(;~ t.sed durin~: the presentations by Tad Mi.irsh ar1c.: h*d Sullivan are L1ttached tu thl'se meetiris rrinutes.

SUMMARY

OF OPENING REMARKS BY REGION MANAGEMENT (Bill Johnston, Region I; Al Gibson, Region II; Carl Paperiello, Region III; and Dennis Kirsch, Region V)

Iriservice testing of pumps and Vdlves is explicitly required by the NRC regu-lations. This testing however is not pcrformeo ri1erely to satisfy the Commission. Inservice testiny is highly important to the operational safety of a nuclear power plant.

It is well understood th<1t components important to the operational safety of the plant must function when needed. Two activities that provide assurance of the operability of these components are maintenance and inservice testing. In this regard, inservice testing is an equal partner with yood maintenance practices.

The NRC head4uarters and regional staffs are increasing their attention to iriservice testing. As evidence of this increased attention, Generic Letter 89-04 was issued to provide the first NRC generic guidance on inservice testing. This guidance was developed to address frequently encountered issues involving 1ST programs, relief requests. procedural implementation, and technical specification provisions for operability. This generic letter will be followed by additional guidance that the NRC staff is preparing on inservice testing.

As indicated in the generic letter, less emphasis will be placed on program review by the NRC staff for determining the acceptability of IST programs before this implementdtion. Rather, the focus will be on audits and inspections of the 1ST program and procedures at the plant site by NRC personnel. In light of their iniportance in ensuring the acceptability of the program and procedures, these 1ST audits and inspections will be more detailed than in the past.

SUMMARY

OF PRESENTATION SY TAO MARSH The objective of inservice testing is to dSSess the operational readiness of safety-related pumps and valves. The scope of Section XI of the ASME Code, however, is limited to Code Class 1, 2, and 3 components. Thus, there is a disparity between the objective of inservice testing and the scope of Section XI.

The Conmission regulations in 10 CFR 50.55a require compliance with the in-service testing provisions of Section XI. In order to account for improvements to the Code, the regulations were developed to require that IST programs be updated to the current Code edition and addenda every ten years. As has been seen, however, the IST provisions of Section XI have changed l1ttle in the last ten years and, in fact, hdve become quite stagnar,t. The regulations alsu allow licensees to submit for NRC review requests for relief from Code requirements where those requirements are impractical.

The establishment of effective JST programs is plagued by a variety of pr(JblE-ms. Many of these problems are the result of inadequate testrng requirements in the ASML Code. f0r example, the provisions of Section XI for performance testing of motor-operated Vdlves f:'Xtend only to stroke time.

Further, the Code re<.juirements for pump vibrc1tion testing are weak. Code Class 2 arid 3 safet)' valves are not explicitly required to be tested. The Code iricorrectly implitis that check vcilves have c1 safety function in only one direction. The trending requireme::r,ts in the Code are insufficient. As is appbrent, the inservice t~~ting provisions of the Code are lacking in mar~

respects.

In addition to the Code guidance, th~re are several other sources of 1ST problems. For example, there ha~, been an absence of NRC guidance on inservice testiny. Previously, a reguldtory guide was begun by the NRC staff, but was never completed. Further, the large number of revisions to IST programs and relief requests thdt require NRC review ha~ caused a bocklog in the approval process. Contrary to Standard Technical Specification 4.0.5, licensees have been implementing relief requests prier to NRC approval. Lastly, inspection efforts b,Y NRC personnel have bee11 made more difficult by the unavailubility in 5ome instances of a Sdfety Evaluatior, Report (SER).

The quality of inservict testing programs varies significc.ntly from one nuclear plant to another. While some licensees have guud 1ST programs, other licensees lack coordination among the groups (including corporate personnel) involved with inservice testing. At certain plants, inservice testing has been combined with inservice inspection despite the fact that these are distinct activities requiriny personnel with differtnt expertise. This combination of inservice testing and inservice inspection might be the result of their being addressed together in Section XI. Another problem is that inservice testing is often used as a training ground for junior personnel with those individuals moving to other areas as they progress. Further, many plant organizations do not have a single individual or organizational unit responsible for inservice testing.

Unfurtunately, in~ervice testing is viewed, on occasion, only c1s an activity to satisfy the NRC. The 1ST program, however, cun be a true benefit to a licensee by initiating corrective uctiori before a component must be declared inoperable.

In this manner, inservice testing can dlso provide important information to be used in the maintenance program.

The issuance of Generic Letter 89-04 is a first step toward resolving the large number of 1ST pro~lems. It provides generic guidance on eleven significant IST issues involving alternatives to Code requirements, and interpretation of the Code and technical specifications. Guidance is also provided to assist licensees in the development of acceptable IST programs. The generic letter clarifies the approval status of current 1ST proyram and relief request submittals that are under staff review. Finally, the generic letter presents a method for preparing revisions to IST programs in an acceptable manner.

In aod~tion to Genuric Letter 89-04, sevrrdl NRC activities arP i11tended t0 improve inservicc testing at nuclear power pld11ts. In particular, efforts are underway to rtvise Section 50.55c1 of the Commission regulations to endorse ASME

tandards OM-6 on pumrs on<.! uM-10 011 valves. Thest standards provide imµroved

~uidance for inservice testing. Consid0ration i~ Jl~u bei~~ given to revision of the regulations in other resµects. On~ proposed changt wo~l, sin~ly separate inserv*iu.:' testing from inservice inspectivr1 *i11 pdragrvph (g) of 10 CFR 50.55ii t<Jr administrat1ve purposes. Another cild11ge under considerdtion would illvolve the long range plan for inservice testing, such as emphasizir1g the need for ir1service t£:~ting to provide assurance that a co111pon~nt wil1 rerform all of its 5afety functions as recessary. Ir. e1ddition to Generic Letter 89-04, other

£E:neric guidc.nc.e may bt~ prepared. Finally, it was noted that ctn ASME/Nl<C syr.1µosium on 111service tfsting was scheduled for August 1-3 of this year.

SUMMARY

OF PRESEHTkTICN BY TED SULLIVAN In U1P past. the process to obtdin NRC staff review uf an IST program and approval of the relief requ~:)ts could consun1e d considc*rable amount of time.

First, the NRC sti.lff would revitw the IS1 program submitted by a11 applicant for or holder of a nuclear power plant operating license. From this review, a list of questions would be sent to the utility through the NRC Project Manager. An IST review meeting wuula then be held at thP plant site. At the conclusion of the meeting, the staff would request thcH the pro<Jra111 be revised to respond tu issues ruistd at th~ r.ic1eting. Following receipt of that response dlld its revi~w, the staff would issue dn SER. G~neric Letter 89-04 is intended to improve 1ST programs and also to simplify the process for obtai11irig NRC approval of JST prograr.1 relief requests.

To help specify the method of response by the i11dividual licerisees, the operatin9 nuclear power plants an: categoriZl*d in Generic Letter 89-04 according to the status of the SER for their IST program. In this regard, the generic letter provides two tablr~ listing particular nutledr power plants.

For those plarits in Table 1, the stJff is nearing co1i1pletio11 of an SER. Thctt is, thl' IST revie\'1 n1eeting has tdken place fairly recently with a subsequent resllbmittal by the 1icensee. With respect tu Table 2, these listed plants have received ctn SER on their curreritly subr11iLted 1ST program. If a plant received an SER several )'curs ago but Si!.Jnificc1ntly revised the progriHil i1 1 the meanti111t:,

that plant was not listed in Table l or 2. Similarly, if a plant hdd not reu:ived an SER on a prior l~T program but had submitted one or more significant program updates, that plant was also excluded from Tdbles 1 and 2.

About half of the operdtrng plants ure not listed ,n either Table 1 or 2. The staff is aware of ~1jnor errors in the tables but these have been resolved through the NRC Project Managers for those pldnts.

Plants listed in Table 1 or 2 do not need to submit a confirmation letter in rt:sponse to Gene1*ic Letter 89-04. Nevertheless, it is essential that these licenstes review the plant procedures tu Lnsure their consistenc.y with the provisions of the generic letter. For plants listed in Table 1 Ot' 2, the SER for the particular plant will constitute aµpruvdl of the IST progrdm relief requests, including any deviati0ns from the ASME Code.

In the case ot plants not listed in Table l or 2, the generic lHter provides the mean!-.> for approval of IST program relief requests from the ASME Code.

Certain steps should be complettd, however, for the appr~val to be valid.

First, the 1 icensce is expected to review the IST prograr.t ,111d procedures agc:inst the positions in Gtneric Letter 89-04 and then revise as necessary to conform to those positions. A confirmation letter 1s to be submitted bj' the lic1::nsee within six nwnths of the issuanct of the gtneric letter to indicat~

conformance with its provisions. For any nc-cessary equipment mod1fication, the licensee should provide iri its confirmation letter a schedule for completing thuse modifications t~at is consistent with the time period specified in the generic letter. The NRC staff does not intend tu p~rform detailed reviews of the confirmation letter and any alternatives oiscussed in those letters. Thus, ctn SER will not be issued. Nevertheless, NRC personnel coula review this document~tion duri~g plant inspections.

Wht.:rt: a generic letter position is impractical for a particular licensee, a m~chanism for approval of rin alternative tu that pu~ition is pruvided in Paragraph B of the ge~eric letter. This mechanism requires evaluation of the mu"intenance and degradaticn histOI'.}' of the component. In this regard, all four criteria listed on page 3 of Generic Letter 89-04 must be addressed and documented in the 1ST program. If each criterion cannot be addressed, then Paragraph 8 is 1,ot the proper medns to ubtain approva, ot an a ltt*rnative to a yeneric letter position. Further, thE ~se of the Paragraph B mechanism tor cbtai11in9 approvill ut an alternative to a position in the generic letter is limited to areas within the scope uf those positions. Deviations to the ASME Code outside the scuµe 0f the generic letter positions will require submission of a relief request for review by the NRC staff.

It i~ recognized that the staff approach simplifies the review process for

µrev1ously submitted relief requests that are not covered by the generic letter

µositions. When the NRC stt1ff prepared tht: generic letter, it wa~ aeter111i11(*d that technic,.tl guidance would be provided on 1:leven issues. This deterniination

~,as based on the total numbt:r of relief request!> and their particular safety significance. Therefore, if d plant not li~ted in Table 1 or 2 had a program submitted and docketed before April 3, 1985, ,my relief requests outsid~ the scope of tl1e generic letter positions are Jpproved prcvided that they are not subsequently changed.

At present, some plants might have aspl'cts of their ;~i- program that have not be~r. ,. q;proved by the NRC staff. For those plants, lictr,sees should sµecify i11 their confirmation letter the relief rc4u1ring NRC staff review and approval.

and the ti111e frame in which thdt relief is needed. The statf will make a concerted effort tu complete those rt:views within the sprcified tim~ frame.

Overall, the goal is to have each 1icensee in11,ilc*menting a fully apµrove IST program.

A cur1Y of the currtnt 1ST prc,gram for each plant should he provideci tG the imc staff. In adaition, each lictnsee st.ould provide an updatt:d cc,py of its 1S1 prograr;1 tu the staff \'1hen substantive chct119cs are made to tht* proyrar:1. The

!.iubrnittal shou*ld cledr1y identify those oeviaticns from th1: ASME Coe.it Lhat are approved through the 1.1t:<.hctniSri1 of the generic 1ttter. Othc,r deviatior,s from the ASME Code that have re:ceived staff approval ur must unaer!.JO staff review should be su indicated.

Licensees should evaluate deviations from the ASME Code included in the current IST program to determine if plant conditions continue to require relief from the Cude. If the situation has changed, then approval of that relief through the generic letter may not be appropriate. Of course, where a licensee has received an SER on a particular rtilief request, that SER may be followed even if it appears to conflict with the generic letter. Where the staff believes that the relief is inappropriate, discussions may be held with the licensee to request a program revision. In significant cases, the staff may institute backfit procedures.

Generic Letter 89-04 is intended as a vehicle for the future as well as the present. For revisions to the 1ST program covered by the generic letter positions, the generic letter should be used as guidance for approval of the revisions. If a program revision is outside the scope of the generic letter positions and the licensee intends not to follow the ASME Code, a request for relief must be submitted for review by the NRC staff, which will then prepare an SER.

Upon implementation of the generic letter, some NRC staff resources will be shifted from 1ST program reviews to providi~g assistance in the inspection of 1ST programs. An inspection instruction will be prepared with a focus on the generic letter positions. The NRC staff has a goal of conducting an inspection of the 1ST program at each plant on a five-year schedule.

QUESTIONS Following the presentations at each meeting, the NRC stdff responded to the extent possible to questions subr.iitted before the meetirlg, as well as to written and verba 1 questions and comments from the audience. These quest ions have been grouped according to their subject and then answered by the staff.

In some instances, the staff responses at the meetings have been modified or expanded to answer the question in a more complete manner. The applicable regional meetiny (together with the question number for that meeting) and, where known, the name of the individual asking the question, are noted in parentheses after each qu~stion.

QUESTIONS ON ATTACHMENT 1 POTENTIAL GENERIC DEFICIENCIES RELATED TO IST PROGRAMS AND PROCEDURES TO GENERIC LETTER 89-04 Position 1: Full Flow Testing of Check Vdlves Question 1 Iten1 1 of Attachment 1 to the generic letter request that flow through a check valve be known for a valid full-stroke exercise test. Does this mean a direct flow indicatio11 ancJ a recorde:d flow rate~ 15 the only acceptable method for the test? For exc.mple, BWR minimum flow lines are not instrumentPd with flow indicators. (Region 1 meeting, Question 119 at the meeting, questio11er:

Dave Wallace of Fitzpatrick)

Is direct flow rate instrurne11tation required for vt:rificati0n of full-stroke capability for all check valvesr For example, the diesel cooling water check valves? (I #46)

Verifying full flow through srn.111 check valves in auxiliary systems or gas systems is typically impractical. As an alternate, will the NRC accept a qualitative evaluation of system response or performance in the place of flow medsurements? ( II #le, John Zudans, Florida Power & Light)

For check valves wherf: design accident flo\<1 is not specified, what guidance can you give for full-flow testing? (III h28, Don Zebrauskas, Commonwealth Edison Co.)

Response

Any quantitative measure that has acceptance criteria that demonstrate th~

required flow throuyh the check valve may be used to satisfy the full-stroke requirement. An indirect measure of flow may be acceptable. For example, a change in tank level over a specified period could be used. In another case, the accepti.lnc~ criterion could be based on a change in flow rate of an instru-mented line when flow is admitted from a non-instrumented line containing the check valve being tested. In any event, some form of quantitative criteria should be established to demonstrate full-stroke capability.

Question 2 Why isn't knowledge of total flow through multiple parallel lines acceptale, when the total flow through each path was known when it was established? (I

1. 13, J. W. Connolly, PSNH-Seabrook Station)

Regordi1g full flow testing of check valves, why is knowledge of total flow through parallel flow lines unacceptable? This seems to challenge conserv~tive Technical Specification requirements for flow balancing. (Ill #34, Gary J.

Roesner, Callaway Nuclear)

Response

The objective of ins1::rvice testing is to evaluate and investigate the possi-bility of degradation of components and to take C()rrective action before the components fail. VErification of tota 1 hec1der flow r<1tt might not identify a prublem, developing or occurring, with ar. i11ciiv1audl check valve in one of the parallel flow paths. With respect to the bulancing of flow, the Technical Specification requireme11t is based un the flow fron' one loop l>E.:ing lost through a break. Consequently, that flow pdth is restricted or throttled to minimize significant diversion of flow. The Technical Specif1c.:ution requirement was riot intended to verify individual check vulve operc1t.iility. The licensee is expected to Justify th~ use of a test method that does not verify full struk£' of individual check valves.

Question 3 Can check valves with external uperdturs ~r,d position indicators be tested only with these devices and never exercised Hith flow ur (iisassembled (I #47)

Is it the intent of the NRC to require full-stroke fluw testing of all check valves or is it acceptaLle to perform manual ~xtrcisiny and partial stroke testing of chec~ valves as permitted by IWV-3~22(b)? (II #la, John Zudd11s, Florida Power & Light)

Position 1 in,plies thu.t the only methoa acceµtable to the NRC for full stroke exercising is a full flow test. No mention is made of check valves with external features which can be used Tur full stroke exercising. Do the 6 criterir1 presented have to be addressed in the IST program to justify usin9 an external operator? (III #43, Pat Tobin, Northern States Power, Monticel1o)

Response

The ASME Code in IWV-3522(b) allows full stroke test1ng of check valves either with flm, or with a mechanical exerciser. Full flow testing is preferable where practical, but Position 1 of Generic Letter 09-04 was not iritended to imply thdt the ASME Code provisions for mechanic.dl exercising were not acceptable. Such mechanical exercising is clearly acceptable and is certainly preferable to valve disc1ssembly as a means of ensuring valve operability. If an external operator is used to exercise a check valve, the provisions of IWV-3522(b) must be met, but the six criteria in Position 1 of the generic lett~r need not be addressed.

Question_i

~hat is considered the maximum requir~d accident condition flow? (I #14, J. W.

Connolly, PSNH-Seabrook Station)

In reference to Item~ 1 and 2 of Attachment 1, please clarify the term "maximum required accident condition flow." (IV & V #22, John DeBonis, Stone &

Webster/Comanche Peak)

nesponse The phrase "maximum required accident condition flow" is inter,aed to mean at least the largest flow rate for which cr~dit is taken fur this component in a safety ana lys1s in any flow contiguration. The safet.} c1na lyses are those c.:ontained in the plant's Final Safety Analysis Report (FSAR), or equivalcr.t, but are not limited tu the accident and transi~nt analyses.

Question~

Is it the intent of the stated position of Attachment 1 that a satisfactory test of a valve in the open direction requires only meusurement of full accident flow through the valve and not the 111ec1surement ot differential pressure (with associc.1ted acceptar,ce criteria) as per IWV-3522(b)? (II #lf, Juhn Zudans, Florida Power &Light)

Response

The A~*~'.E Code does not r1::quire the measuru11ent of valve differential pressure when exercising check ~alves with flow. It should be recognized, however, that such d measurement might provide useful information for ev~luating the condition uf the valve.

~stion 6 For Lheck valves which drt: never required to open fully (i.e., thermal expansion or siphori breakers), verification of design (sJfety) function is tile testing required for forward flow. Is this correct? (III #42)

Response

In addition to verifying its safety function performance, quantifiable acceptance criteria should be developed for ttie testing of these components.

for example, a pressure decay test with specified acceptance criteria woula be considered d reasonable test.

~stion 7 In reference to Item 1 of Attachmt::nt 1, for non-parallel full flow test, does the , low obtained need to be documented quantitatively, or can 1t be qualitative (i.e., greater than~- gallons per minute)? (IV &V fr23, John DeBonis. Stone & i..'ebster/Cornanche Peak)

What is an dcceptable flow condition when, for example, the safety analysis requires 250 gallons per minute {gpm) flow but 600 gpm can be delivered? Would passing greater than, or equal to, 250 gpm be a valid full flow test, or would 600 gpm need to be delivered? (IV & V #24, D. G. Dobson, Texas Utilities/Comanche Peak)

Response

The full flow test is intended to demonstrate that the necessory flow rate can be tlchieved and to detect a,~ degradation of the ch~ck valve. Therefore, acceptance criteria for the test should involve more than the achievement of flow above a minimum rate. The acceptance criteria should also include the allowable variation of test results. To enable the test results to be compared, the ir1itial parameters for the test should be standardized to the maximum extent feusible. The acceptance criteria for the full flow test and the bases for those criteria should be documented and available for review by NRC inspectors.

Question 8 In reference to Item 1.3 of Attachment 1, please clarify what the rmc would expect a "qualification program" to include (i.e., how exte~sive). (IV & V

1. 25, John OeBonis, Stone &Weber/Comanche Peak)

Response

Position 1 of Generic Letter 89-04 indicates that, where full flow testing is impractical, it might be possible to qualify other techniques to confirm that the check valve is exercised to the position required to perfurrn its safety function. One of the stated conditions fur this approach is that the licensee should describe the test method and r~sults of the program to quolify the alternate technique for meeting th~ ASME Code. The language of Position 1 in this regurd was chosen to allow the licensees flexibility in qualifying alternatives to full flow testing. In general, the licensee should demor,strate that the alternate test is quantifiable ana repeatable. The alternate test should also meet the intent of the ASME Code. This qualification of the alternate test should be documented by the licensee and available for review by rmc inspectors. The Nucleilr Industry Check Valve Group (NIC) is said to be investigating the qualification of various testing techniques, such as ultrasonics and radiography for check valves. The resuits of those and other industry efforts might bf: of value to the individual licensee in providing for the use of alternatives to full flow t~sting.

Position 2: Alternative to Full Flow Testiny of Check Valves Question 9 Does the Generic Letter Attachment 1, iteri1 2c.: use of "orientation" refer to physical orientation (e.g., horizontal or vertical) or plant orientation? (I

1. 15, J. W. Connolly, PSNH-Seabrook Station)

Response

Orientation, as used in Generic letter 89-04, refers to the physicdl orientation (horizontal or vertical) 11s well as the physical relationship tc, major components. For example, a check valve at the discharge of a pump has a different ori(:ritation than one at the pui11p suet ion.

guestiun 10 When manually exerc1s1ng per position 2c, i~ this dune per Code or just a physical stroke checking for binding? (I #16, J. W. Corinully, PSNH-S~alJrouk Station)

When valves are disassembled and manually exercised in lieu of full-flow testing, is adhererice to the quantitative aspect~ &nd acceptance criteria cf IWV-3522(b) requir~d? (II #le, John Zudan~, Florida Power &light)

Response

The staff believes the requ"iremrnt 111 IWV-3522 (b) of the AS1-1E Cude to measure the force or torque while ma111Jt1lly exercising checl-; valves only applies to manual exercising from outside the valve where the observation of the valve internals cannot Le r.iade. This medsurenient permits a quantitative evaluation of the perforn~nce of the valve in thdt changes in the "easured force or torque may be indicative of degradation of the valve internals. While thE:1 valvt is ii, a partially disassembled condition the valve internals should lie inspected and the conditio11 uf the mc,ving parts evaluated. This inspection and evaludtion should include verification by hand that the valve disk is free to move, l>ut mt:!asurement of force or torque is not required. Following reassembly, d partial flow test is E:!Xpected tc., be performed.

yuestion 11 Uoes the utility have the optio11 of either inspection through disasstmbly or performing functional testing to sat1sfy IST requirements? Can either be used regardless of the previous test1ng mode? (I #31, John Wiedemann, PSE&G)

Respons~

Disassembly, together with inspection, to verify full stroke capc1bility of ched: valves is en option only where full struke extrcising cannut practicully be performed by flow c,r by the other positive means allowed by IWV-352t:.

Additionally, partial stroke t:xercise testing with tlow is expected tu be performed after the di sassE:'mb ly and i n~pt*ct ion is c.ornp leted but before returning th~ valve to service. If the previous test w~s perfot*"ed using flow, the licensee is expected to docurnent the JL~tificatiur, tor any change from that test r,1ethod. Alsr., fur the case where plant conditions prevent full stroke testing with flo\'1, the licensee:, should periodically evaluate whether pldnt conditions have been alterc:d in such a way that full stroke tE:st.ir,g using flew is possible. If so, the licenseE: should revise the test prucedure5 to provide for sue.Ii test ir1g.

Question 12 In light of the stated position of requ1r1ng check valve internal inspection at least one~ every six years, is it perndssible to schedule the inspections for the total group of valves on a six year frequency vs. each refuel outage? This is especially important where plant preparations for inspection of multiple valves are essentially equal to those for a single valve and they represent a considerable cost i11 terms of monetary outlay as well as schedule and availability impacts. (II #ld, John Zudans, Florida Power & Light)

Response

Position 2 of G~neric Letter 89-04 takes advantage of the benefits that can be ot,tained through sampling techniques. The NHC staff, however, recognizes that the position 111ay have a significant impact on outage time. For example, some plants have combined inJection header check valves that are physically located rn a position relative to the reactor coolant system (RCS) loops such that their disasserably would require draining the RCS to a level that would necessitate core offluad. In order to alter the inspection frequency as suggested by this question, lictr1sees should use the criteria in Position 2 to justify and tv documer,t the proposed disassembly schedule. The justification

~hould address the significance of thf loss of benefits of sampling in light of the condition, servict history, and application of the valves. For additional discussion of this issue, see the response to Question 19.

Questio11 13 Does disassembly/inspection require certified visual testing personnel, or can deta i 1ed inspect ion procedures be performed by maintenance personne 1 without certified in5pectors? (II #L5, Jim Holton, Floridd Power Corp.

Do personnel perforrning the visual inspections addressed on Position 2 have to be VT-3 certified, ANSI 45.~.6 (i.e., Mech Inspector) certified, or may engineering personnel competent in check valve technical requirements perform this visual in~pection (III #2, Larry Campbell, Toledo Edison)

Response

The personnel performiny the disassembly/inspection must be qualified to evaluate the condition of the valve and to assess its continued operability.

The licensee is responsible for the development and implementation of a program to ensure that 1ST personnel are appropriately trained and qualified for performing the valve disassembly/inspections. Generic Letter 89-04 alone does not impose any requirements for visual testing certificationc; (such as VT-3) beyond those currently in the ASME Code. Nevertheless, licensees must implement the provisions of ANSI/ASME N45.2.6, Qualifications of Inspection, 11 Examination, and Testing Personnel for Nuclear Power Plants," according to their conrnitments based on the implementation section of Regulatory Guide 1.58.

The NRC staff encourages those licensees that have not formally committed to following Regulatory Guide 1.58 to review the ANSI standard and regulatory guide for guidance in developing a program for the qualification of inservice testing personnel.

Ou~stiun 14 If a check valve within a sample group is disassembled/inspected ir1 a non-refueling outage, does the next valve need to be inspected at the next refueli119 outage, or can it still be scheduled for its original refueling outage? (II #26, Jim Holton, Florida Power Corp.)

Response

This question is difficult to an~~er without more detailed information. In gentral, in order to alter the disassemb1y/inspettion schedule as suggested by the question, the lic.ensee ~liould justify ancJ document the propos!:d change.

The justification should addres~ the effect of the proposed disassembly/in-spection schedulf-! on the samp"lrng program. The justification should rely on the maintenance:. history and known valve cor,dition from previous inspections rather than subjective qualitative jull9~111ent. Position 2 in Generic Letter 89-04 indicates the criteria that need to be addressed.

Quest i.Q.!!_15 Is it the intent of Position 2 of the GEmeric Letter 89-04 that during valve testing by disassembly, that the valve bt: completely disassembled and each internal valve part removed, if possible, and 100% of the part vis~ully inspected, or may only the valve bonnet be rcn~ved and the valve internals inspected in place without the removal 0t the internal valve parts unless evidence of discrepant conditions are found which then would require further insp2ction and probable removal of the part? Note: Inspectiun of the valve internal parts without re~ov<ll uf the part would be by direct visual inspection, use of mirrors, or by remote i~spection equipment such as horoscope fiberoptics. (III #1, Larry Campbell, Toledo Edison)

Response

When performing check valve disassembly and inspection to satisfy the requirements of the ASME Cude for inservice testing, disassembly is required only as far a!:> necessary to assess the condition of the valve and to allow manual exercising of the disk. (It must be recognized, however, that the Code requirements for inservice inspection are different from those associated with inservice testing.) If a partial stroke exercise with flow can be performed, this testing is expected to be perforrued after the disassemLly and inspection are completed but before returning the valve to service.

Disassembly and inspection uf a check valve is not considered a "test" as implied by the question. liisassembly is not a true substitute for an operability test using flow, but is allowed as dn alternative to a flow test where that test is not practical. Llisassembly c,nd inspection does, however, provide a valuable means of determining the internal condition of the valve.

A recent example of the value of disassembly and inspection involved the identification of broken bolting materictl in Anchor Darling check valves at two nuclear power plants. This occurrence is discussed in NRC Information Notice 88-85, dated October 14, 1988.

The Nf<C staff is encouraging tht: development and use of alternate techniques tu evaluate the position of check valve disks. The Electric Power Research Ir,stitute (EPRI) and the Institute of Nuclear Power Operatiorrs (INPO) are recommending an inspection periodically for check valv~[. that are subjected tCJ potentially harsh service conditions. The "4RC staff encourages tht:se activities as well. The industry group NIC is also investigating methods to

~e:w.onstrate the operability of check valves.

Question 16 Even though the check valve flow testing can be perforr,1e:d as required by ASME Section XI, may the valve test be: µerformed by disassembly as permitted by Position 2 in Generic Letter 89-04 when it is considered by tile utility that testiflg by disassEmbly will provide the same or greater assurance that the valve will fur,ction properly':' (Note: If possible, partial valve stroking qudrterly, or at co1d shutdown, or after re-assembly would be performed.) If the answer is yes, {a) c.an the test frequency, sample, etc., as descril.ied in Generic Letter 89-04 Postion 2 be us~d in lieu of t'.SME Section XI requirement-even if the St:etion XI tst could be perforrnetl, i.e., at cold shutdown; {b) must a rr1fof reque~t be processtd or may this "test by disi3ssembly" be noted ir. tht: valve 1ST program subrn-ittal to thP NRC; and (c, n,ust a relief request IJE:

processed or may the frequency si:rniple, etc.; be noted in the valve IST progran, suLmittal to the NRC? (Ill #3, 4, 5, 6, Larry Campbell, Toledo Edison)

Response

The various methods a imt::a at evaluating the operabi 1ity of check va lv£:~ are not cqudlly acceptable to the NRC stdff. At the outstl, the ASME Code requires a full stroke exercise using flow (or o mechanical exerciser) to be performed quarterly. Where full stroke exercising*cam,ot be performed quartE:rly, the Coc.le allm,s the performance of this test during cola shutdowns. Full stroke exercising during rtfu~ling outages may be an acceptable alternative if the test cannot be perforraed at cold shutdown, but this approach would require submission of a reiief request. For thosE:' C.dses where full stroke exercising cannot be performed quarterly, during ccld shutdown, or during refueling outag£:s, disi\ssembly and inspection in conformance with Position 2 of Generic letter 89-04 is allowed dS an alternative. If the provisions of Position 2 are followed, a relief request need not be submitted for NRC review but this deviation from the ASME Code should be docunEnted. (See also the response to Question 15)

Question 17 May the valve testing by disassembly/visual inspect.ion identified in Position 2 of Generic Letter 89-04 be applied to reverse flow testing of check valves?

(Ill #7, Larry Campbell, Toledo Edison)

Response

Position 2 of Generic ' "t ter 89-04 addresses the use of d hassernb ly and inspection as an alte1 native to forward flow testing of check valve~. The use of disassembly and inspection to verify closure capability (i.e., back fluw) may be found to be acceptable depending on whether verification by flow or pressure measurements is practical. As the generic letter does not addr~ss this use, however, the submission atid approval of a relief requ£:st before implementation is required. Oisa5sembly and inspection is not acceptable for demonstratio,, c,f leak-tight integrity.

Quest ion 18 We are only able to perform a partial flow test of the accunrulator discharge check valves due to limitations based on system configuration. Do we have to supplement this test with disassembly of the check valves? (III #20, Wisconsi~

Public Service Corp.)

Response

The safety injection accumulator discharge check vc1lves are typicaliy very difficult tu exercise with flow to tr.e position required to ~erform their safety function. If a µartial flow exercise i5 all that can be perforr.itC:, then some other techn1que, as aiscussed in Position 1 of Generic Letter 89-04, 1.iight be develuµed 1.o periudica 1ly vt:rify the capability of these valves to move to their saf£::ty fonction position. If this is not feasible, the licensee is expecterl to follow the provisions for the disassembly alternative contained in Position 2 of the generic letter.

Question 19 Regarding disassembly of check valves, please define "extreme t,ardship" when speaking with reyard to extensicr: ct disassembly interval. (Ill #36, Gary J.

Rotsner, Callaway Nucledr)

Response

The existence of "extrerne hardship" that would allc,w extensfor: of the disassembly schedule in Position 2 of Generic Letter 89-04 is depenaer,t uri the particular circumst,rnces dt the plant. To dE.:termine whether extrer11£: hdrdship exists, the licensee should conduct a detailed ev~luation oi the various competirry factors. First, the lice11see should determir1t the effect on plant safety that would rtsult fror,1 the proposed scheautt: extension. Tht: nictinterldnce hhtory of the cumpunent and other inforr.:ution relevant to its relidbility shjuld be rev1ewed to determine whether the decrr:ase in assur..nice of plant safety resulting fron1 the scheduh: extensior. is justified. A n~ed '.:.ci cffload the reactor core, such i:iS when testiny the combined injectior, hedder check valves at sc.,me plants, or to optrdtt: c:t mid-lev1::l of the n:uctor coolant luups 111ay be considcrt:cJ. The rad*iution exposure thcit woLJld result fror:, the disdssembly and inspt:ct1011 is a factor to bl. c.ons1der~d under the ALAkA (As Low As Reasonably Achievable) principle, but it should b~ judg~a i11 combination with al1 of the other factors.

Question 20 Position 2 goes into the scheduling of disassembly/inspection in a very detailed manner. Are other scheduling schemes acceptable as long as they have each valve disassen~led/inspected within 6 years? Would approval of an alternate schedule have to be i11 the form of an SER or acceptance of details provided in a confirmation letter (existing schedule for disassembly/in-spection agreed upon ir1 1ST program review with ~RC, but SER never issued)?

{Ill #44, Pat Tobin, Northern States Power, Monticello)

Response

As stated in Position 2 of Generic Letter 89-04, the burden is on the licensee to demonstrate the extreme hardship necessary to comply with the identified sample disassembly/inspectior1 schedule. The staff considers the sampling aspect of the position to provide assurance of the continued operabilit.>' of the valves that are 11ot inspected during any given outage. Therefore, the 1icensee should justify through the provisions listed in Position 2, any deviation from the stated schedule. That justification should be provided in the 1ST program submitted to the NRC staff, but need not be included iri the confirmation letter. Where the provisions of Position 2 for an alternate disassembly schedule are followed, it is acceptable to implement the alternative and an SER will not be issued. The NRC staff, however, may review the alternative and its justification during plant inspections.

Position 3: Back Flow Testing of Check Valves Question 21 With reference to generic letter item 3, if a leak test is performed to verify Category C check valve seat position, would any leak rate bt:: acceptable Sv long as the system r11eets its winimum requirements to perform its safety function?

(I #18, Al Koehl, NES)

Response

When performing a te~t to verify closure capability of a check valve th<lt dors not have a defined seat leakage lin1it, the achievement of the necessary system flow rate through the intended flow path might bt an adequate demonstration of the clo~ure capability of a check valve. For exan1ple, when verifying the closure capability of the check valves un the discharge of paral1el pumps, achievement of the required safety tlow rate from one runniny purnp with the idle pump's discharge check valve providing the barrier for recirculation flow would be considered an acceptable test configuration. In addition, the licensee should evaluate the consequences of the back flow through th~ check valve. This ~valuation should consider the loss of water fro11, that systE'm c:i11d conr.ectir,g systems, the effect that the leakage might have on components and piping downstredm ot the valve, and any iricrea~t in radiolc,gica*1 exposure r~sulting from the leakoy~.

Question 22 Are the items listed in Attachment 1, number 3a, d, e, f, specific to PWR's?

The nomenclature is not familiar to BWRs. (I #24, John Lindburg, PP&L)

Section 3 of Gerieric Letter 89-04 deals with back flow testing of check valves.

It has c list of several valves that NRC states provide a safety function.

Some of these valves do not appear to provide a safety function and we would lih: to hear the flRC's reason for classifying these: valves as safety related.

(III #19, Wisconsin Public Service Corp.)

Response

All of the listed systems do not necessarily apply to each plant. A licensee should evaluatE: at least the listed systems to determine if they apply to its facility and should wake any necessary modificatioris to its IST program. In regard to a particular question, items 3d, e, and fare specific to pressurized water reactors (PWRs) while 3a (feedwater header check valves) may be applicable to both boiling water reactors (BWRs) and PWRs. One example provided in Position 3 to the generic lette:r is the volume control tank outlet check valve in the chemical and volume control system. This check valve may serve dn ir.1portant safety function at some PWR plants to separate the non-safety grade water source from the safety grade source.

Question 23 In regard to Attachment, Position 3, how is individual seat le~kage determined for 10 CFR 50, Appendix J, Type C, tested valves? Tech Specs specify only penetration totals. (I #35, J. W. Connolly, Seabrook Station)

Response

IWV-3426 of Section XI of the ASME Code requires that a permissible leak rate be specified by the plant owner (licensee) for a specific valve. If leak rates are not specified by the licensee, permissible leak rates are provided in IWV-3426. It should be noted that Section XI provides no criteria or guidance for licensees on the method to establish or to specify the permissible leak rate of a particular valve. Apparently, the Code recognizes that leak behavior of a valve varies according to the type of valve, the vendor, the valve size, the service conditions, the safety-related functions, and other factors, and that there is no simple leak rate rule that may be applicable to all valves.

In general, the leak rate limits should be set within certain bounds. If the leak limits are too low, unnecessary repairs or adjustments to the valve can result. If too high, failure of the tests required by Appendix J to 10 CFR Part 50 could occur, leading to concerns for leak-tight integrity of the containment. Appropriate permissible leak rates can only be developed and refined by analyzing and trending the leak rate data of specific valves or leak rate data from similar valves at other plants. Therefore, the NRC staff is not in a position to specify leak rates. The licensee should document its methods for establishing the initial permissible leak rates and procedures for improving the leak rate limits.

• . 18 -

Question 24 In regard to Attachment 1, Position 3, does this backseat check require a ful 1-stroke exercise and is it performed at the Code specified frequency regardless of normal plant positions? (I #36, J. W. Connolly, Seabrook Station)

In reference tu Item 3 of Attachment 1, does a valid back-flow test on a check valve first require tht valve to be exercised to the open position then back tested, or is it valid to merely perform the bi.lck flow test? (IV & V #29, U. G.

Dobson, Texas Util1ties/Comanchc Peak)

Response

If a pa rt i cu lar valve performs u safety function only in the c. lose=d posit ion, demonstrution of a full-stroke uµen before verification of closure capability is net required by the ASME Code. This closure verification is required ti"' be performed at the frequency specified by the Code. If ll) the valve perforr.~:. a safety function ir the closec position, (2) the normal position for the valve is closed, and (3) this positior, can be verified during normal plant operation, then quarterly aocumentation of this verification satisfies the Code requ i re111ents. If a valve perforr,1s a safety fum:t ic,n in both the operi ar,d closed position~, however, the Code requires that the valve be exercised to the open positio11 and then be verified to c.lose.

Question 25 Previous to this, it was permissiblE to verify closure of stop-check valves

imply by operation of the stem (shaft). Is this acceptable instead of reverse flow testing? (II #1~, John Zudans, Floridd Power &Light)

Response

Verification of closure capability of stop check valves by using the handwheel meets the ASME Code re4u i rements. This, however, is not the pref erred method of test. The NRC staff considers reverse flow testing to be a more reliable indication of valve operability.

Question 26 Regardiny bac;k flow testing of check valves, what is the position of the generic letter in the phrase "verify by other means?"

(III #39, Mort Khazrai, Toledo Edison)

Response

The majority of the wording in the sentence in which this particular phrase appears was taken directly from IWV-3522 of Section XI of the ASME Code. The NRC staff included the phrase "by other positive means" to be consistent with the wording of the Code. When Generic Letter 89-04 was written, the staff did not have in mind any particular techniques that it would consider acceptable.

- 1SI -

Position 4, Pressure ls0l~tion Valves Question 27 Is it tht intent of Generic Letter 89-04 tlwt the only Reactor Coolant System Pressure lsc,lation Valves (PIVs) to be included in the 15T µrogram are those listed in the lechnical Specifications and those which are Event V PlVs? (JII tlB, Larry Campbell, Toledo Edison)

For µlants licensed prior tu 1979 which do not list all RCS Pressurl Isolation Valves in the1r Technical Specifications, is it the intent of Position 4 of Generic Letter 89-04 that only PIVs listed i~ the Technicdl Specifications and PIVs which are 11 Everit V" be included in the IST Program? (III #9, Larry Ca~pbell, Toledo Edison)

Uoes the NRC anticipate requirins (in the future) that all RCS PIVs be included in the 1ST program? (III #10, Larry Campbell, Toledo Edison)

Response

The position in Generic Letter 89-04 represents only a liniited area of the staff's conce;rns regarding PIVs. The gener.ic 'letter position only apJ)lies to those PIVs listed in individual plant Technicdl Specifications. However, the staff recognizes that the PIV~ in the Technical Specification!:> fur many plants, particularly older plants, arc a subset uf the PIVs ir, the plant. In view of th1s fact and other concerr,s regarding PIVs, the staff has recently undertaken a program to reevaluate various aspects of PIVs, including testi~~- Sample inspections arc underway as part of this NRC program.

Question 28 What, if anythiny, is being done with tht! lir.ensee resµunses to Generic Letter 87-06? The generic 1etter references PIVs in Sect1or1 4; however, it appears thdl there are no ch~nges required due to G~neric Lelt~r 87-06. Is this true?

(1Il h18, Wisconsin Public Service Corp.)

The responses to Generic Letter 87-06 are being used as input for the 1*1::solution of Generic Issue 105, 11 IntC:rfaciny Systems LOCAs at Light Water keactors," under inve!)t1gatiun by the NRC Office of Nuclear Regulatury Research. No further licensee action is required at thi!> time with respect to Generic Letter 87-06.

Position 5, Limiting Values of Full-Stroke Tin~s for Power-Operated Valves Question 29 , Position 5 in part states: 11 The O(:Viation should not be so restrictive that it results in a va,ve being declared inoperable due to reasonable stroke tin~ variations. However, the deviation used to establish the limit should be such that corrective action wuuld be taken for a valve that may not perform its intended function. 11 Given that MOVs operated by AC induction motors fail if slowed by mon* than dpprc,xirnately 10%, a valve normally stroking in 15 seconds will fail to operate by a change of 1.5 seconds. By comparison, a reasonable deviation from normal stroke time of 15

~econds caused by error in 111easurement might be 2 seconds. The fact that the reasonable deviation for this 15 second valve is larger than the possible actual deviation before failure makts the two qtJoted goals of Attachment 1, Position 5, mutually exclusive. Request res~lution.

(I #32, D. B. Ritter, PP&L)

~esponse The stdff agrees that strokf:: t irnes for f\C motor-operated valves probably wi 11 not chan9e appreciably before failure, especiall} for MOVs that have relatively short stroke times. If the ASME Code-identified testing does not provide useful inforr.1dtion for E.:valuating the c.:oritinue:d operability of thest- valves, then the licensee shoulc.i propose ar: alternative to the Code requirer:1t:11ts that does proviae such inforn1ation. The Cude requires the licensee to establish limiting values of full stroke tirn: for all power-operate<.! va*1ves and also requires medsurenient of stroke time to an occuracy of within 10 percent for this particular case. The Code does not prohibit tht-.: 111£:asurernent of stroke t i me more a c cu r a t e l y o r th e s e t t i ng u t the 1 i ri1 it i r. g va l ue a t l e s s t ha n 2S

µercent dbove the normul stroke time. The NRC anu industry recogniz~ thut the Code-specific criterid are riot sufficient for assuring operability of AC 1i1otor-operc.1ttd VcilVe'.>. In light of th*1s recoynition, the staff issued Bulletin 85-03 to require that licensees est6t>lish programs to ensure that op~rator switches for MOVs in certdir, important plant syst1:111s are se:lected, set, and 111aintarncd properly. P.s a result, in part, of thl! responses to that bulletin, the scope of thL~ eftort has been expanded in Generic Letter 89-10 to include many uther MOVs important tc plant saftty. rrnc staff t.1ctio11s suc.h as these will bt: need t..o contpE.!nsate for Htciknesse'., ir: the IST provisions of the ASME Code until ~n adequate IST standard is avaii~ble.

Question 30 In re~ard tu Attachment 1, Positiur1 5, what is crns1dtred ct r1:it~onable dt:viation fro11 the reft:rence strokl t1rne'? (1 H37, ,J. rl. ConrH1lly, Sedbru0k

~tilt ion)

In regard to Attachment 1, Position 5, can the deviation be different for valves with different functions and/or actuators? (I #38, J. W. Connolly, Seabrook Station)

What is mednt by 11 recsonably limiting value of full-stroke time? 11 (I #48)

What methods are considered acceptable for establishing the limiting value for full stroke times for power operated valves as given in Position 5 of Generic Letter 89-04? (III #50)

In reference to Item 5 of Attachment 1, is there any generic guidance on what is acceptable to the NRC 011 this item? (IV & V #11, T. F. Hoyle, Washington Nuclear 2)

What is 11 reasonable 11 value for deviating from the reference stroke time established for valve testing? (IV & V #16, Arkansas Nuclear 1 ana 2)

Response

The NRC staff has attempted to provide the general philosophy for establishing the limiting stroke time. The establishment of specific values for the limiting stroke time is aeµendent on a variety of parameters relevant to the particular valve and the conditions at the plant. The parameters include operating characteristics, operating envirorm1ents, actuator types, and valve stroke times. In that the test should confirm the operability of the compunent and not the system, the limiting value is not to be considered a function of the valves 1 s safety significance. As the limiting value is specific to the valve, the staff is not in a position to provide Vdlues for limiting stroke times. The licensee needs to use its best judgement in assigning these values.

The justification for the assigned values is expected to be documented and available to the plant site for review by NRC personnel. One i1Spect of the staff review will be a compansun of the limiting stroke tin1e to the technico.l specification value.

quest io!_l__l!_

In regard to Attachment 1, Position 5 (paragraphs 2, 3 and 5), why are Tech Specs or Safety Analysis lir.1iting criteria not acceptable for valve operability if maintenance is triggered by co1,1ponent evaluation? (I #41, Eugene Perry, Consoliddted Edi~on)

With respect to the application of stricter acceptance criteria for valve stroke times, apparently the NRC has some idea as to ihe philosophy dnd limits that would be acceptable. This informatiur, should be shared with licensees.

(II #17)

Define the liri1iting value uf full-stroke time." Is this number the 11 operability number for the valvE: even ii thl:' Tech Spec stroke is much higher?

(II #14, Mark Cardile, Geor~ia Power)

Response

The Tec...;,ll i cal Specifications provide assurance that important plant systems are capable of performing their safety functior,s in a timely manner during selected plant accidents. The provisions of Section XI of the ASME Code are intended to ensure the continued operability of particular plant components. The distinct bases for these two documents lead to criteria that may differ significantly.

Nevertheless, the Technical Specifications and ASME Code are both needed to provide confidence that the nuclear power plant can be operated safely.

Therefore, the more restrictive criteria of the two documents must be followed even though this might result in a component or system being declared inoperable. The response to questions on position 8 of Generic Letter 89-04 alco address the relationship of the ASME Code to the Technical Specifications.

Question 32 Is it requireo to measure stroke timt:s of valves that are not provided with remote position indication? (II #11, John Zudans, Florida Power and Light)

Response

The ASME Code requires the measure111ent of stroke time for all power-operated valves regardlf:'ss of whether they have remote position indication. The staff has endorsed this requirement. Without specifics, the staff is not in a position to corrmer,t on alterr,ate techniques that may be found acceptable.

Question 33 When considering comparison of power-operated (stroke time) valves according to valve type, valve actuators, valve size, etc., we find there is no consistency when using this comparison. However each valve consistently tests well. We are currently looking at a quantitative method of establishing maximum allowable stroke times. ls this an acceptable method?

(II #28, Jim Holton, Florida Power Corp.)

Response

If we understand the intent of the opening sentence of the questior1, we agree that c.riteria for setting the limiting value of full-stroke time may vary for each valve type, stroke time, size, etc. The use of a quantitative multiplier on a reference time may be an acceptable method for setting these values.

However, as discussed in some of the responses above, the licensee should dccument the justification for its quantitative methods of establishing maximum allowable stroke times. This justification should be available at the plant site for review by NRC personnel.

- ~3 -

Question 34

~Jhen the stroke timt CJf a power operlit~d Vdl11e exceeds its [limiting value for]

stroke time, as established in accordance with P0sition 5 of the Generic Letter 89-04, but is still within its plant Technicdl Specification or FSAR stroke time limit, can performing an evaluation which determinE:5 if the valve may remain operdble be used to satisfy PCJsition 5 in lieu of making it mandatory that the valve be declared inoperable? (Ill #12, M. J. Richter, Conu1101~wealth Edison)

Response

The limiting value of full stroke time i~ requir~d to be established for all power-operated valves. The limiting val11e should be that point at which the licensee seriously qutstions the continued operability of the valve. It is expected to be a value determined to be redsonable for the indivioual valve based un that valve's characteristics and past performance, but not to exceed any safety analysis requirer,1~nts. The value should not he based solely on the system requirements or values specified in safety analyses for system performance. When the identified lir,liting value is exceeded, the licensee shall declare the component inoperdble and shall Enter any dpplicable Technical Specificatic,n limiting condition for operation (LCO). Following the dcc'laratiun that the valve is inoperable, the licensee may perform an anulysis to identify the root cause of the problem with the valve. If this analysis cltarly demonstrates that the valve remc1ins capable of performing its safety function, the analysis might constitute the corrective action required by the Code. The dnalysis s*;oulci be docume:nted.

Question 35 If the 1imiting value of full stroke time is less than the 11 alert limit 11 identified in th~ Cude, does the trending still have to be done? (III #51)

Respunse If the 1ir.1iting value of full strokf: time is exceeded, then the licensee shall declare the valve inoperable and shall perform currective action. Where the limiting value is less than the 25 percent or 50 percent "c..lert limits 11 for trending as specified in the ASME Code, trending as envisioned by the Code becomes a moot point. The licensee could identify a r£:duced percentage alert limit for this valve to provide curly warning of problems with this valve, but this is not required either by the Code or by Generic Letter 89-04.

Qu£:stio11 36 In reference to 1tPm 5 cf Attachment I , is Item 5 in fact a rewn te of the stroke time criteria that are tu be applied in accordance with OM-10?

(IV &V #31, D. G. Dobson, TE:xas Utilities/Comanche Peak)

Response

The informdtion in Position 5 of Generic Letter 89-04 was not intended simply to b~ a rewrite of the information in ASME Standard Uf*l-10. This i::051tio11 has evolved over the years and is considered rcdsonable by the staff for establishing limiting values of full stroke time for power-operated valves. As such, the position represents a clarification of existing ASME Code requiremer,ts. For its part, ASME Standard OM-10 does not provide guidance for the establishment of the limiting value of full stroke ti1,1e. lni5 5tandard, however, does require that a valve be declared 111operable imnit:,diatel.,. upon discovering that it f~ils to exhibit the requirtd chanye af obturator position or exceeds the limiting value of full stroke time.

Quest ion 37 Since establishing maximum stroke ti111e limits may in some cases at first prove too restrictive, is it acceptable for corrective ~ction to be an engineering evaluation which increases the time limit (based on more detailed analysis)?

(IV & V #33, Alan Harris, Waterford 3)

Response

The Commission rt:gulations in lu CFR 50.59 allow licensees to perform enyineering evaluations of plant structures, systems, and componelits. If the stroke tine limit is exceeded, the valve must be declared inoperable and dny applicable Tec.hnical Specification lir*;iting coridition for o~eration entered.

At thdt point, an engineering analys*.s may be performed to verify that the valve is capable of performing its *,afety tur,ction. This analysis should include more than a determination 1.hat the new valut: is less than the FSAR or Technic,11 Specification limit. For example, a root cau!:ie *investigation should be performed to deterr11ine the reasons for the stroke time increase.

~stion 38 We have been informed that we could omit the valve stroke time limits from our IST Submittal. Where can we find guidJnce on what is really required in a submittal (minimum scope)? (IV & V 113i, Pdul Croy, Southern California Edison, San Onofre)

Oo specific valve stroke time requirements (or limits) need to be specified in the 1ST plan, lJr is specification in implementing procedures sufficient? If procedures are sufficient, can existing limits referenced in the plan be r~moved in a future revision? If plan specification is required, is this limitl!d to Te:hnical Specification and safety analysis stroke time limits, or must owner specified stroke time limits that are required also be in the plan?

(IV & V #36, Terry Pellisero, Pacific Gas & Electric, Diablo Canyon)

Response

The specific. limiting values of fui1 stroke time for each power operated valve as determined according to Position 5 of Ge11eric Letter 89-04 are not required to be identified in the 1ST program. These liwiting values, however, should be providea in a document such as the individual test procedure or a general procedure that identifies the criteria tor establishing these valu~s. The c~ncern for the specification of lih11ting values is the rEsult uf weaknesses that the NHC staff has fourit.i while reviewing IST procedures. As a general rule, IST programs should contain sufficient information to indicate what parameters are being measured, how tests are being performed, and the bases for the acceptat,i 1ity of any departurt:-s from the ASME Code. For example, the program should indicate forward flow testing or back flow testing, or both, for check valves.

Pastian 6, Stroke Time Measurements for Rapid Acting Valves Qi..Estion 39 With reference to the Generic Letter item 6, paragraph 4, where does tht:

two-seconds come from and whdt is the bases for the two-second only criteria, could this be a minimum of ') or 4 seconds? (I h19, Al Koehl, NES)

Response

The two-second criterion is bdsed on the staff's considerdtion of the response time of personnel and equipment and the difficulties irivolved in aµp'iying the ASME Code requir1::ments i11 this situation. Any alternative to Position 6 of Generic Letter 89-04 or the ASME Cude requirements may b(; submitted, along with a sound basis, for staff review through a relief requE:st. As relief requests containing alternatives to the Cude requirements are expected to address the fundamental puq.>0se uf ir1service testing, see the summaries of the openir1g presentations for a discussion ot this subject.

Question 40 Generic Letter 89-04 states that previous analysis (IWV-3417{ct)) can be replaced with a consen,;tive "refE.'rence value" comparison. Generic.. Letter 89-04 stdtt:s this should !Je docu111entf.'d in the I~T proyrdrl!. Should this chctnge be made by re 1ief requts t or by d text change to the proyri.1111 body? ( I #23, Jeff llyhard, Nine M1 le Station)

Generic lE::tter pos1ticin on power 0µ1::rdted valve strokt:- t iL:t:s of grt:atl*r than ten seconds is t0 pla~e the valvE* i11 increased frequency if stroke t inie is greater than 25~ of the bcJSl' line strukt tin,.:.

(Ill #38, f-lort Khazrai, Toledo ldison)

Response

When the staff prepdred the discussion in Position 6 of Generic Letter 89-04, the objective of the first paragraph WdS to set the stage for the discussion on "rapid acting" valves, and it was not intended to address a 11 aspects of stroke time for power-operated valves. Nevertheless, the staff believes that the use of a reference value stroke time as a base line for comparison of routine test values is a better method of evaluating change in valve performance than that specified ASME Code IWV-3400. Therefore, if a licensee wishes to use reference values rather than previous test values for comparing stroke times for valves with normal stroke times equal t~ or less than ten seconds, the generic letter provides the vehicle for this deviation from the Code and a relief request need not be submitted. As the generic letter does not address valves with norma 1 stroke times g,*c.:Jter than ten second$, a licensee must submit a relief request for staff review and approval before using reference vctlu~s as a base line for stroke times for these valves.

Question 41 Can an MOV or power-01,1era ted valve have a duu l classification under rapid II acting" and "less than 10 seconds? 11 ror example, we have valves that stroke closed in less than 2 seconds and oper, in less than ten seconds. Therefore, is the classification and the previous test (or reference test) percentage based on opening time or closing tir.ie? (I #34, Jeff Neyhar<l, Nine Mile Station)

Response

If the vaive performs d safety function in both positions, and the stroke time in one direction is less than two seconds, then for that stroke direction, the licensee may use either the acceptanct criteria of ASME Code ur the staff's position for rapid acting valves. Where the c::troke tir.1e for the valve in th' other direction is greater than two seconds, the acceptance criteria for that stroke time range, as identified in the Code, should be followed when testing the valve in the yrfater-than-two-second direction. Similarly, the alternative concerning measurements of changes in stroke ti 1e allowed by Generic Letter 1

69-04 may be used for th< stroke direction that has a strokE..* time of less than ten seconds. (NOTE: Although both MOVs and power-operated valves are mentioned, the question is more applicable to air-operated valves. Normally, MOVs do not have widt: ly differer,t stroke tirries for the open and close directions.)

Position 72 Testing Individual Control Rod Scram Valves in BWRs No questions.

Position 8, Starting Point for Tin~ Period in TS ACTION Statements Question 42 lU CFR 50.55a(g) states that 1ST programs comply with Section XI.Section XI states for valves that If the condition is not, or cannot be, corrected within 11 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the valve shall be declared inoperative." This is in direct disagree111ent with the Generic Letter which states that the LCO must be declared immediately. How do you justify this disagreement with the Code?

(I #5, Dave Wallace, Fitzpatrick)

Generic Letter 89-04 implies that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period for declaring valves opera~le versus inoperable does not apply. Can the utility continue to use the 24 liours before declaring a valve inoperable? (I #27, Jeff Neyhard, Nine Mile Station)

Position 8 specifically states that licensees cannot use the 24-hour grace period for declaring a valve inoperable (IWV-3417(b)) and must make such declaration immediately upon recognition of exceeding a stroke time limit.

Position 5 states that the intent of developing more restrictive stroke time limits is to identify a valve problem "before the valve reaches tht: point where there is a high probability of failure to perform if its safety function is called upon. Per Position 5, exceeding the more restrictive limit does not imply that the valve is inoperable but that the probability of failure is increased. With this philosophy, the 24-hour grace period is even more reasonable. (II hB, John Zudans, Florida Power and Light)

This question is in reference to Item 8 of Attachment 1: "Starting point for time period in Technical Specifieutions ACTION statement." This item eliminates the 24-hour clock for valves which exceed Section XI limits. In most cases, the Technical Specifications limits are higher than the Section XI limit. This item needs discussion. (II #15, John Kin, Virginia Power)

Response

The Standard Technical Specifications in Section 4.0.5 specifically state that the more restrictive requirements of the Technical Specifications take precedence over the ASME Code. For example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable. That definition takes precedence over the ASME Code, which allows up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before declaring inoperable a valve that (1) is incapable of exhibiting the required change of disk position or (2) has exceeded its limiting value of full stroke time. Therefore, if a valve is tested and the data indicate that it is inoperable as defined by the required action range, then that valve must be declared inoperable at that time and not 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later. This elimination of the 24-hour grace period before declaring a valve inoperable is consistent with the requirements of ASME Stdndard OM-10.

Question 43 When d piece of equipment enters the required action range, why must the Tech Specs action statement be entered without some time to reflect on \'1hy it has entered the required actiu11 range? A reasonable approach would be to establish a limited reflection time, for example thE: existing shift, to review how the test was conducted and review previous tests to see wh~t the problem is. In declaring equipment inoperable wher, it really may not be upon review of how the test wa~ conducted, gerierdtes r:eedless paperwork and impacts INPO avai 1-abi l ity statistics (i.~., HPCI, RCIC, RHR). (l f/28, Bob Binz, PSE&G Hope Creek)

Response

For some time, th~ NRC stJff has bee11 concerned with the unrestricted grace period for declaring a compor;ent inoperable allowed by the ASME Code. One example of this yruce period is the 24-hour delay allowed by IWV-3417 of Section "1 followir,g c1 failure of a valve tl., E:>r,ibit the required change of disk positio~. The staff's concern i11 this area has been expressed to individual licensees or, many occasions. In orc;;er to provide guidance that is consistent \'11th the Star,ciard Technical Specificdtions and that can be applied generically, the staff developed Position 8 of Generic Letter 89-04 which states th~t the unrestricted grace period in the ASME Code is unacceptable.

Once a component is dee.lured inoperable, the action statement in the Technical Specifications would provide time for evaluation of the situation, including performiny the test, before chanye is required in plant operating mode. A licensee may propose alternatives to the NRC staff's position. For example, a valve stroke time that is less than the limiting str0ke time could be establishea as an alert time. If the alert time is exceeded and the limiting time is not, the licensee would initiate a 24-nour period for evaluating the condition of the valve before decldring it inoperable.

Question 44 Address the conflicts between the background of the generic letter which states "The intent of testing is tc, detect degradation affecting operation and assess whether adequate margins are maintained" irnd Position 8 regarding thl' starting point for Technical Specification ACTION statPnit!nts. This will require declaring components inoperab'le which c1re car<1ule of fulfilling their safety function (i.e., uperable). (II #33, Philip J. North, Duke Power)

Response

The stc1ff does nut see a conflict between the stdtement in the background and Position 8 of Gen~ric Letter 89-04. Testing is intended tu detect degradation of d co111punent ,.111d to provide assurance that adequate ri1argrns are ma intaineu.

rJhere tl'sting indicates that a component has undergont:: such degradation that its operability is in question (e.g., thlc limiting value of fu1l stroke time for a valve has been exceeded), Position 8 of the generic letter requires that the: component be declared irlCJperdble.

Question 45 Referring to paragraph 8, after testing a pump and declarin9 it inoperable, is it acceptable to replace the process instruments with test instruments which are more accurate then retest, rather than recalibrating process in~truments?

(IV & V #14, Arkansas Nuclear 1 and 2) kesponse Accuracy of the instrumentation is an important consideration in the per-furma11ce of a test. In :iddition, the t£:st must be µerformed in a nwnner that allow~ the test results o be compared fur trends. This consistent per-formdnce of d test is son times referred to as 11 repedtability. 11 Where instru-ments with different characteristics (such dS Hith respect to range and accuracy) iire used for each test, the ab1 lity to monitor the results for trends m1y be lust. Therefore, the staff prefers that the same set of instrumPnts be used in perforriiing tests 011 a particular component. This can be accomplished most readily by use of µroperly calibrated process instruments installea in the system. The installation of test instrumentation that are more accurate than the pn.1cess instruments is al lowed by the ASME Code. For the example cited by the question, after declariny the pump inoperable because of the test results from ~rocess instruments, the operability of the pump may be verified by more accurate test equipment. Because the same instruments should bE used for tests to 1,1onitor the results for trends, the licensee should recalibrate the process i115truments for their continued use ur should establish a procedure to use the more accurate test rnstruments from that point forward.

Question 46 In reference to Item~ : , ttachment 1. it states that the prov1s1ons to recalibrate in IWP-' / H.,\1 1 ' can only be done after the component is declared inoperable. What i1, * .. l'ing a pump test, before test data is taken, it is clearly observed ti ' a gauge i~ malfunctioning. Do I need to declare the pump inoperable, or can I stop testing and recalibrate?

(IV & V #36, Ken Trippel, Housto11 Lighting & Power/South Texas Project)

If it is obviou~ that d test has been run incorrectly (i.e., a recorded pdramet~r is out of the range of the device being tested), do we still enter the action statement before re-running the test? (I #26, Bill Kittle, PSE&G

- Salem)

Response

If a test is under way (regardless of whether test data have been taken) and it is obvious thdt a yauye is malfunctioning, the test may be halted and the instruments should be pror,1µtly recalibrated. One example r;1ight be a wildly fluctuating gauge. It should be nuted. however, that, in many situatfons where anomalous data are indicated, it may not be clear that the problem lies with the gauge. In these cases, the licensee shculd attribute the proLlem to pump performance. The licensee wuula then declare the pump inoperable and eval~ate the condition of the pump during tli<! time allotted by tile applicable Technical Specificdtion.

Position 9, Pump Testiny lJsi119 r,:ir;imum-Flow Return Line or Without Flow Measuring Devices Question 47 With reference to the Generic Letter item 9, in cases where only the minirauni flow return lir,e is the available path, wc,uld the generic letter be revised to consider rtducing the 5 minute tir.1t required for stabilizing the punip i.is required by IWP-3500(a) tCJ a lesser time such as 2 or 3 minutes in order to minimize the possibility of pump damage occurring duriny the pump's operational test? {I #20, Al Koehl, NES)

Response

The staff du:s not inte11d to revise Generic Letter 89-04 to change any currtnt positio11s or to address additionul issues. If there is a problem concerning compliance with the A~ME Cude, requests for relief from the Colle may be submitted.

Questio11 48 If mini-flow recirculation line!:. are instrumer,ted for flo\\, are quarterly tests alone, which m~asure flow, differential ~ressure, and vibratiori, acceptable?

(IV & V #18, Waterford 3)

Response

Mini-flow rec1rculc1tior1 line tests are riot prohibited by Section XI (Jf the P.SME Code. The staff, however, believes that a 111ini-flow test car, be detr11i1t11tal to c1 pump and is not a desir~Lle test cGnfiguratior,. These tests produce d~ta of r.,arginal value and provide little C1..;nfidence in the contir,t..H:d operability uf the pump. The stctff would prefer a r.iure compn~hensive test µerformell ut some reduced frequency rather thori relJing u11ly on the mir.i-flow tesL that is perforr.1ed quarterly. This JJdr~1cular issue may he a t(Jplc ot anuthe:r generic letter addressing inservice test ins i11 th~ future.

Question 49 Many mini-recirculation lines have no means to adjust flow to a reference value prior to taking data. Thus, this recirculation flow is relatively fixed.

Since Table IWP 3100-2 limits are placed in differential pressure, what criteria should be used to place limits on flow? Even with a fixed-flow system, measured flow wi 11 def!.10nstrate so111e variation test-to-test due to instrument repeatability, operator interpo1i1tion of needle position on meter face, etc. Table IWP 3100-2 lin1its do not seem appropriate for flow in this case. To allow both flow and differential pressure to vary within 13'.t ranges doe~ not appear to meet the intent of Section IWP. (IV &V #19, Waterford 3)

Response

In most cases, mini-flow recirculation lines do not have flow adjustment capab*ility. The ASME Code recognizes this in rnP-3110, which permits the use of one or more fixed sets of reference values for pump testing. The Code identifies dCceptance criteria for both differential pressure and flow rate in Table IWP-3100-2. It is not permissible for both parameters to vary during a test. With one parameter set at a reference va *,ue, the other parameter is compared tu the acceptance criteria.

QuLstion 50 It is more desirable to test pumps at ~ubstantial flow conditions than on mini-recirculation lines. Should entire trdins of safety systems be declared inoperable and 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> dction statements entered solely to realign these systems for inservice testing? Does the oLtaining of "better" pump data Justify the increased risk to the public during the time the system cannot perform its safety function? (IV & V #20, Waterford 3)

_Response As stated in the question, it is more desirable to test rumps with sul.l!)tantial flow than in mini-flow recirculation configurations. The NHC staff, however, does not agree with the questioner that the performe1nce of i11service testing results in increased risk to the public. Iriservice testing is intended to provide assurance of the cunt inued operdbi lity of pumps and valve!). To provide this assurance, it is considered acceptaLle for a Technical Specification action statement to be entered on infrequent occasions in order to test a component. Where a system must be taken out of service to perform a test, it is likely that, in the event of a plant emergency, the system could be realigned fur operation in short order. Where one train of a safety system will be disabled for an extended period or both trains 0f th~ s}stem must be made inoperable to perform a test, the licenset: should propo$e d testing schedule that provides for verification of co111punent operability vdth testing performed during period (e.g., refueling outages) when availability of th~

system is not essential to plunt sdfety.

Position 10, Containment Isolation Valve Testing Question 51 In regard to Attachment 1, Position 10, why can't valves other than containment isolation valves (CIVs) that are 6 inches or larger be exempt from the needless requirement of IWV-3427(b)? (I #40, J. W. Connolly, Seabrook Station)

Dues the exemption from IWV-3427(b) pertain to pressure isolation valves (PIVs) dS well as Appendix J valves? (II #4, John Zudans, Florida Power c111d Light)

Do PIVs have relief from IWV-3427(b)? Item 10 on Attachment 1 only discusses CI Vs ( II 1 #46)

Response

The reliet from IWV-3427(b) of the ASME Code granted through Generic Letter 39-04 only applies to CIVs under containment leak rate testing. This position was written in response to numerous relief requests concerning CIVs from licensees that cited difficulties in trending leak rate data. We were not aware of similar difficulties with PIVs during reactor coolant system leak testing. The relief from th~ explicit requirements of IWV-3427(b) should not be taken as an indication that the NRG staff is disregarding the value of trending CIV leak testing data. Until more information is available on appropriate leak rate limits and on reasonable scatter of data, however, Position 10 will remain in effect for CIVs. The NRC staff anticipates developing a more comprehensive position of the subject in a future generic communication to licensees.

Positiur, 11, 1ST Program Scope question 52 IWV-1200 specifically exempts control valves from testing. Why are these valves included in the list of examples in IST program scope as part of ? ( I #6, Dave Wallace, Fitzpatrick)

Response

IWV-1200 of the ASME Code does not exempt valves that have a required safety function from the provisions of Section XI. Code interpretation XI-1-83-59 states that it is a requirement of Section XI that flow control valves that have one or more defined safety-related functional requirements be classifieed Category A or B, dS applicable, and tested in accordance with the requirements of Subsection !WV. This philosophy applies tu all control valves that have one or more def 'ined safety-related fun ct ion a 1 requirements.

Question 53 PleaSl! clarify the last three lines of Generic Letter item 11 of Attachment 1.

( I #10, Shafi Rokerya, r;-!w York Power Authority)

The scope statement of Position 11 is much too vayue. The position with respect to program scoµe must be clarified and explained to provide further guidance and should also dddress the backfit issue. In addition, in th~ past, it hds been the practice of adding additional components t0 the scope of 1ST Programs vit1 the authority of 10 cm SLJ.55a(y)(ii). How will this be addressed in th~ future? (II ,s, John Zudans, Flori~~ Power & Light)

Do satety-related component~, cutside cf Class 1, ,, and 3 need to lie tested in accorddnce with the Cude c1nd be inc. luded in tlw IST program, or is it the iritent to have ~u111e form of testing to demonstrate operability. (III #29, Vince Treague, Point Beach)

In reference tu !ten, l 1 of Attachn,~nt 1, plca~e clarify tht intent of the last seriter*ce of thi!:. item: 11 Therefore, whi~e iO CFR 50.550 delineate5 the testing requirements for A5ME Cude Class l, 2, and 3 µumps and valves, the testing of pumps and valves is not to be limitc-d to only those covered by 10 CFR 50.55a.

(JV & V #10, T. f. Hoyle, Washington Nuclear 2)

How will the NRC review pump anct valve testing not included in the scope of the IST µrogri.un't Will the ASME Code n*quirernents be applied to these components?

(IV & V i/15, Arkansa~ Nuclear 1 and 2)

B~~p_Qnse Criterion 1 in Appendix A tu 10 CFR Part 50 requires, among other thiri9s, that components important to safety be tested to quality standards co~nensurate with the imµortdllCC:! c,f the safety fur,ct ions to be performed. Appendix 8 to Part 50 describes the quality <1ssurance program, which includes testirig, for safety-related compo11e11ts. Paragraph (g) of 10 CFR 50.550 requires the use of s~ction XI of the ASME Code for inservice testing of componl::r1ts covered by the Code. For other components important to safety, the licensee also has the burden of demonstrating their cuntrnued opE:rability. The list prc.,vided in Position 11 co11tains examµles of components that have bee11 shown by our experience to be frequently omitted from a routine testing progra111. The licer,see should review the safety siynificance of these identified comµonents to ensure that the inservice testing is adequate to den1onstrate their continued operability. NRC inspectors wil I evalua.te the adequacy uf such testing. The Code-required IST program 1s u reasonable vehicle tu provide a periodic demon-stration of the oµerability of pumps and valves not covered by the Code.

If non-Code tumponents are included i11 the ASME Code IST program (or some other licensee-developed inservice t~sting program) and certain Code µrovisions cannot be met, the Conmtission regulations (10 CFR 50.55a) do nut require a "request for relief" to be submitted to tne staff. Neverthl:'less, documentation that provides dSsurance of the UJlltinued operability of the non-Code components through ,he performed test~ ~hould be available at the plant site.

Question 54 Tht: Diesel Generator air !itdrt systelil direction that was iri the initial dnft of Generic Letter 89-04 has now been dropped. Can we ren,ove the test i 119 from our program? (Not that we would, I feel it is a good pr~ctice} (I #22, Jeff Neyhard, Nine Mile Station)

In Position 11, why were the emergency diesel generator support system components deleted froffi the list in th~ final version of the letter? (II #3, John Zudans, Florida Power & Light}

ResponsE:

Typically, the Emergency Oit'sel Generator air start system is r.ot Code Class 1, 2, or 3 and, therefore 10 CfR 50.55a does not require the testing of these components to be performed under the prov h Hms of the ASME Code. Emergency Diesel Gen~rator air start, cooling water, and fue:l oil transfer systems, however, an.: considered safety related. As such, Appendic.ts A and B tCJ Part 50 require that they undergo component testing.

Question 55 Are the items listed in Attachment 1 number llc, d, and e specific to PWRs? (I

1. 24, John Lindburg, PP&l) kesponse The listed itl:r.is were not intended to dpply to every plant. Each licensee 5hould review the list and determine thu~e items applicable to its facility.

ln response to the siiecific question, items 11c, d, and e do not apply to BWRs.

0THER QUESTIONS DURING GENERIC LETTER 89-04 MEETINGS Schedule for Implementing the Generic Letter Question 56 The scope of ttie Generic Letter is broad and requires more than tht allotted 6 months fur response. What guidance can be given for extension of the response dater (I #8, Dave Wallace, Fitzpatrick)

How much is expected to be done at the end of 6 months? ( I#50)

What is the schedule requirement for implementing additional or revised testing arising from the activities related to the generic letter? Keep in mind that the results of reviews and evaluations must be available ~rior tu revising and implementing the related procedures. (II #9, John Zudar,s, Florida Powtr & Light)

Do the requirtn~nts to conform to the stated positions of the generic letter within 6 11onths of the date of the letter mean that all procedure~

have to be revisea and dpproved within this 6 month period, or is it acceptable to have procedures in the process of being revised within the 6 naonth period? (III #15, M. H. Richter, Commonwealth Edison)

Due to outage schedules and co11straints, are there any provisions for not completing all equipment raodifications within 18 months of the date of confirmatory letter, or the first sche:lluled refueling outage fol lowi1;y the confirmation letter? (III #16, M. H. Richter, Cofi~onwealth Edison)

How are extensions of the October 3, 1989 deadlim: viewed; what factors arl:

considered on such requests? (1II#21. Point Beach Nuclear Plant)

Do ~tilities have to contact their Project Managers to schedule ifim~diately a meeting to resolve any requested relief requests outside the generic letter (prior to requirea test frequency) to obtain ..1pproval ar.d avoid violation after submittal, or will there be a grace period? (III#41)

Response

With regard to plants not listed in Table 1 or 2 of Generic Letter 89-04, the ir,tent has i..,een that, by the end of six months, (1) the l~T prugram would be revised t0 incorporate il11 the requirements uf the generic letter, (2) the procedures woulc be written and implemtnted, (3) the c0nfirmation letter and any necessary additional relief requests would be submittcu to the NRC, and (4) a schedule wuuld be provice:d for any plant modifications necessary to comply with thE: requirement~. It has been additionally intended that any r,ecessary equipment modificat*ions be completed within 18 months of the date of the c0nfirmation letter or the first scheduled refu~ling outage fo 11 owing the confirmation, whichever occun, l ii ter.

We have ret:eived several comments stating that this schedule may not be achievable. For example, one licensee noted that acceptance criteria need to be developed before procedures can be prepared and implemented. Fol lowing preparation of the procedures, several weeks were said to be needed to provide the necessary training to plant per~onnel on various shifts.

Another licensee indicated that the resources necessary to implement the yeneric letter had to be determined to justify to management the need for contractor assistance. Even where licensee management accepts the justifi-cation for contractor assistance, it was said that few highly qualified contractors in the area of inservice testing are available. With respect to equipment modifications, one licensee hypothesized a situation where a refueling outage began soon after the confirmation letter and the next refueling outage would be a month or two beyond the 18-month limit.

Several reasons that the NRC staff does not consider sufficient to justify not meeting the schedule in the generic letter were also given by meeting attendees. These insufficient reasons include (1) the lack of activity relative to Generic letter 89-04 until the NRC meetings took place and (2) the lack of a designated individual responsible for 1ST at the plant when the generic letter WdS issued. If any particular plant anticipates a problem in n~eting the schedule, this should have been discussed with the NRC Project Manager. ln determining the necessary schedu 1e ex tens ions, licensees should have limited the request for schedule relief to the smallest set of revisions to the 1ST program and procedures, and n~difications to equipment. The information submitted to the NRC by the licensee to justify a delay in meeting the schedule establlshed in Generic letter 89-04 should have contained at least (1) a dEscription of the actions to be completed by October 3, 1989, incluoing an interim schedule of accomplishments t,y system and component, (2) a description of the action for which an extension in the schedule is being requested with the specific proposed schedules for the program, procedures, and any necessary equipment modifications, and (3) a discussion of the specific reasons for the need to extend the schedule, including the hierarchy of the proposed schedule extensions as established by their importdnce and dependence on the completion of other ac.tions.

Question 57 Does the NRC exµect the licensee to take any specific action prior to receipt of the SER? (IV & V #1, T. F. Hoyle, Washington Nuclear 2)

Is it the intent to have all implementing procedures of changes required by Attiichrnent 1 be completed within 6 months? Does this apply to Table 1 and Table 2 plants'? (lV & V h6, T. F. Hoyle, We1~hingtor1 Nuclear 2)

Response

The positions in Generic Letter 89-04 address both proyram and procedural issues. Pusition5 4, 5, and 8 are related to procedures and would not be covered by a review of the I ST ~rogram. The remainder ut these µos it ions are related to both the IST prograr1 and the procedures. For Table 1 plar;ts, we believe that it would be reasonable for the generic letter provisions to be implemented with111 5ix months of issuance of the SER. The precise

schedule, however, will be specified in the SER. The schedule for Table 1 plants is keyed to the SER because the licensee needs an opportunity to review the SER before having to corrunit to an implementation schedule.

Nevertheless, the staff encourages Table 1 plants to begin verifying that plant procedures are consistent with the generic letter before receipt of their SER. Table 2 licensees should verify that plant procedures are consistent with the generic letter positions within six months of issuance of Generic Letter 89-04.

Confiro@tion Letter Qut:stion 58 With our confinnation letter will be a couple of relief requests. How will they be handled? Can we assume relief is granted? Oo we have to wait for your SER? (I #30, Joann West, Beaver Valley)

What is the level of iriformation expected in the response to the generic letter? How detailed must it be? (II #22, Garry Galbreath, Duke Power) ls "relief" required for items per Generic Lt=tter 89-04 which differ from the ASME Code? (Ill #22, Point Beach Nuclear Plant)

Response

A confirmation letter from a particular licensee may contain several forms of information, depending on the IST program. The confirmation letter should address the extent to which the licensee's program and procedures meet the positions attached to Generic Letter 89-04. It is anticipated that most licensees will have to modify their 1ST programs as a result of the generic letter. The revised prograrn should accompany the confirmation letter. In cases where a generic letter position that approves an alterna-tive to the ASME Code is being followed, a relief request is not required, but the deviation from the Code should be documented in the IST program along with its method of approval (i.e., through the relevant generic letter position). As a suggestion, licensees may reserve the use of the term "relief request" for those cases where specific staff review and approval are needed before imµlementation.

If a licensee cannot meet one of the generic letter positions, an alternate test method may be performed, providing the provisions of Paragraph B of the generic letter are met. This Paragraph B approach for generic letter positions does not require a relief request but the justification should be retained in the IST program. In that the generic letter does not supersede the regulations in any way, the option still exists to submit requests for relief from the Code for program-relJted positions in the generic letter.

For plants not listed on Table 1 or 2 {i.e., plants that will be submitting a confirmation letter), any requests outside the scope of the generic letter

that were submitted before April 3, 1989 are approved by the issuance of the Generic letter. If a relief request is subr.1itted after April 3 or a relief request submitted before April 3 is modified, the requested relief may 11ot be implemented until receipt of staff approval. The date by \'1hich these r~lief request approvals are needed shoula be specified in the ccnfirmaticm letter so t~at their revfrw may be prioritized.

Verific~tion of Generic Letter lmplen~ntation Question 59 When and how is guidance going to be provided to the Regional offices on inspectio11 and enfcrcement of the issues stated in the Generic Letter?

(I #3, Dave Wallace, Fitzpatrick) kegarding the dpproval of the 1ST Progrun1 scope irnd related relief requests, it appears that NRC is not planniny to perform detailed r~view and *is merely stating that their responsibility re. 10 CFR 50.55a is satisfied by the generic letter supplemented by plant site inspections. This eliminates the pre-approval discussions done previously; however little guidance is provided to give licensees* confidence that the subJective opinior1s of the various inspectors can be anticipated before the fact. It would help if there were some mechanism whereby a utility could rec.cive an official opinion/determination with respect to proyram scope and relief request queries in a timely manner. (II #6, John Zudans, Florida Power & Light)

With respect to inspections, will there be an inspection module developed, or is this to be an "ad hcc" type of inspection? (II #27, Ron Jacobstein, Florida Power &Light)

To what extent is the rrnc planning to muke their guidance uniforin policy for all inspections? It is very important that uniform policy be applied at all facilities, regardless of the composition of inspecting teams. (II #38, John Zudans, Florida Power &Light)

Many alternatives that are given seem vague and subject to interpretation.

Who dee ides adequacy and what are the r~11if i cation of differences between licensees and the NRC? (111#23, Point Beach Nuclear Plant)

What guidance will Region/NRR auditors use in accessing IST Programs for Table 1 or 2 plants? Will they use the SER or the generic letter? (IV & V #3, T.F. Hoyle, Washington Nuclear 2)

Response

The NRC staff has been performing activities to provide assurance that app 1i cat ion of the generic.: letter by the inspectors will be consistent.

For example, a meeting to discuss the generic letter was held in Rockville, Maryland, in April 1989, and each NRC Region office was represented. A temporary instruction (TI) will be written by NRC/NRR, providing guidance to the regional inspectors on prioritized inspection activities for 1ST and the Generic Letter 89-04. It is intended that the TI will be completed in six

to eight months. Periodic NRR/Regior, counterpart meetings wi 11 be held to ensure consistency on the 1ST subject matter. Additionally, the inspection teams are expected to be made up of NRC/NRP, NRC Regior1, and contractor persvnnel, thereby providing for consistent communication. These inspections will cssist the stuff in verifying the adequacy of the !ST program rather than Vt>rifying adequacy by the traditional staff revi~w. It is intended thi.t the inspectors will rely on the generic letter, the temporury instruction, and the p<<rticular SER for Table 1 and 2 plants. These inspections will not be ptrformed ,,n an ad hoc bctsi5. Although only relief requests will receive NRC revit:w before theirlmple111t:ntation, licensees mo.,* direct questions concerning intE:rpretdt ior: cf requirer,1ents on the 1ST prograr,1 dnd procedures to the NRC staff through their Project Manager.

Ques!ior. 60 If the Slk does not constitute NRC concurrence that the ~eneric letter requirements (at least thost> that are routinely GCtn.:sseu in the program submitta'f) are met, then how will issuance of SERs to Tilbh.? 1 or Tat,lE* 2 plants constitute NRC approval oi th£! IST prograr.1*c (II 1119, Sid Burns,

/\ labama Powt::r Company)

Will all SERs issued in the near future, or recently issued, incorporate all the issues in the generic letter? (II #41)

Besponse It is recognized thdt the positions in Generic Letter 89-04 go beyc,11d the ureds covered by past SERs on inservice testing. Positions 4, 5, and 8 deal Hith procedurol matters that are not rt-fleeted in the IST progrdrns and SERs.

Therefore, it cannot be expected that an SER would cor:~titute cor1currence that all of the generic letter positions have been met. The SERs for Table 1 and 2 plants explic;itly contain approval only for relief requests. These SERs can be considered dS µroviding 1ST prograr,1 upproval only* in that the

µrc1ctice has bet:11 to perform a thorough review and identify problem areas that need resolution.

Qedates ,md Revisions of the IST Programs Question 61 If relief requests exist that do what one, or any, of the positions state, should these requests be retracted with the corifirmation/resubmittal?

(II #29, Jim Holton, Flvrida Power Corp.)

Do "changes to the program" include administrative changes such i:lS rt:ferencing different procedures, or just intent of program? (II #32, Jim Holton, Florida Power Corp.)

In instances when a lict:nsee modifies their IST program beyond that currently submitted to the NRC, [as discussed in] Paragraph U of the generic letter, and reviews th~ modification against the positions found in Attachment 1, is it required that the 1ST program modifications be submitted to the NRC?

(III #14, M. H. Richter, Commonwealth Edison)

Our plant is on Tati1e I. We have revised the program to identify Ge11eric Letter 89-04 as a reference and made some minor changes consistent with the 1et ter. Do we need to resubmit the prog rani? ( l II #26, Steve Be 11 , I 11 i no is Power)

Are all future revisions to the IST progrdm required to be submitted to the Commission? Section D of the generic letter is silent on this subject.

(IV & V #5, T. F. Hoyle, Washington Nuclear 2)

Does the generic letter rnedn that program submittals are r,o longer required?

Under what circumstances are submittals still required? (IV & V #12, Arkansas Nuclear One)

Should we provide changes to the NRC as suon as made even if numerous "trivial" or "typo" changes are being issued? What about the "complete and accurate" requir~ment in 10 CFR 50.9? (IV & V #30, Paul Croy, Southern Caliturr.ic Edison/San Onofre)

Shau ld updated plans document spec if i c re 1ief requests tl1dt were approved on a prior datE? (IV & V #34, Alan Harris, Waterturd 3)

Since progrums are revised frequently and in a piece-meal fashion, does the NRC expect each change to be submitted as soon as it's made, or is once per year, once per two years, etc. adequate? (IV & V #35, San Or,ofre 1)

Response

The NRC st~ft should have the current IST program being implemented at each plant eve11 if this means that a licensee sends multiple submittals to the NRC edch year. The most up-to-date version of an IST program will not be used for the purpose of the staff perfur111ing complete program reviews as has been done in the past. Rather, it is needed to prepare for 1ST inspections and to assist in the review of relief requests. The staff wculd prefer to have a complete progri1m rather than individual changed µages. The idt!nti-fication in the proyrdm of the mechanism for approval of specific relief requests would be particularly helpful. That is, the program should indicate whether the approval is (1) through a position in Generic Letter 89-04, (2) by virtue of the relief request !Jeing outside the scope of the positions in the Generic Letter and submitted before April 3, 1989, (3) through the mechdnism described in Paragruph Bin the generic lett~r. or (4) obtained usir19 a relief request that \<1ill need staff approval by a specific date. Currently-approved relie1 requests that follow a generic letter position should not be retrc;cted but thE: source of upproval (i.E.,

the generic l~tter) should be identified in the IST progra~. Non-technical and minor tyµoyraphical changes may bt:: hE:!ld until the lic~11set has cullt:cted several such changes. Thi~ is considered to meet tht'.! intent uf JC CFR 50.9 fur complete and accurate ir,forniatio11. For plants not listed in Table 1 or 2, revisions to the IST µroyrar.1 should be sent Hlit::11 the confirmation letter is submitted.

Question 62 If valves are added to or removed from the system, does the change to the program require resubmittal? (II #32, Jim Holton, Florida Power Corp.)

Can components be deleted without prior NRC approval? (III #45)

Response

Neither the Commission regulations in 10 CFR 50.55a(g), in general, nor Generic Letter 89-04, in particular, require the licensee to obtain NRC approva 1 on each tt:s t on every component in the I ST program. As 1ong as the program is consistent with the regulations, the ASME Code, and the Generic Letter, relief is not required. To amplify, deletions from or additions to the IST program do not necessarily require NRC approval. The burden is on the licensee to verify that their !ST program is complete and all compone~ts that require 1ST are included and tested to the extent practical. If a particular component is deleted from the IST program, documentation of the reason 111 an appropriate place is recommended.

Question 63 Please clarify the intent uf the last sentence of [~ection DJ: "The modified program should comply with the disposition of relief requests in any applicable SER based on a previously submitted IST program." The sentence quoted abov~ seems to apply to Table 1 or Table 2 plants only.

Also, the sentence stems to allow the use of an extension of a previously granted rel"ief request. (IV &V #4, T. F. Hoyle, Washington Nuclear 2)

Response

Section O of the Generic Letter 89-04 applies to all plants. Previously approved relief reruests remain Vdlid. However, if a relief request has beer, aenied iii an SER, the SER usually provides information cm the reason the relief request was denied and recommendations on appropriate actions for the licensee. The last sentence of Section D is indicating that these recol'IUTlended actions should be followed.

Question 64 It is clear that if an NRC position is covered by Attachment l, then the licensee must either comply with or follc\i tht alttrnate provisions contained in Section B of the generic letter. But for proyram changes not covered by , [Section D] states that the provisions of IO CFR 50.55a(g) should be followed. This infers that a relief must be submitted. Further, in accordance with the plant Technical Specifications, relief must be granted prior to 1mplementation. (IV &V #4, T. F. Hoyle, Washington Nuclear 2)

Respon~e It is correct that, where an 1ST program change is proposed that is outside the scope of the positions in the Generic Letter and does not meet the Section XI requirements, the licensee must submit a relief request to the NRC for review. The program change may not be implemented prior to staff approval.

Question 65 For plants with SERs, can r.hanges to NRC reviewed and approved programs be made without additional submittals to the NRC? What if changes are in accordance with the generic letter? (IV&V#l3, Arkansas Nuclear 1 and 2)

Response

As described in the response to Question 61, licensees need to send any char1ges to their 1ST progrdm to the NRC. If these chaiiges are in conformance with Generic Letter 89-04, NRC review and approval are not necessary. The 1ST programs submitted to the NRC as a result of program changes should indicate the reasons for the changes and the relief requests, if any, that require staff review.

Relief Requests Question 66 If a relief request issued for one unit has been approved, can, or will the turnar0und time fur apprrval of the same relief request on a second unit (for a two unit plant) be reduced? (II #18, Herbert P. Walker, Georgia Power/Vogtle Project)

For future relief requests outside the scope of Attachment 1, what is the perceived ability of the NRC regarding turnaround time? (I1#23, Garry Galbreath, Uuke Power)

Response

New relief requests Hill be evaluated on a priority basis. Therefore, the licens1.:e should specify the date by which the relief is needed, and where possible, should provide additional information to assist in this review, such as "this relief request is identical to relief request number X in the Unit 1 1ST program." The staff recognizes that, on occasion, there will Le a need for rapid NRC response. The staff will mdke every available effort to be responsive to such needs.

Question 67 If revised relief request subm1ttals are not considered approved, then do we continue working to the pre~ently approved request? (II #30, Jim Holton, Florida Power Corp.)

Response

The approved relief request is controlling until the licensee receives approval of a reviseci relief request. As we have indicated above, if p:ant 01,>erations anci ASME Codt requirements dictate relief request approva1 by a certain date, the licensee should indicate that cJate in the submittal containing the relief request.

Question 68 Uoes a relief request that is grandfathered but no longer required still need approval? (II #44)

Response

By grandfathered relief request, we as5ume that the question is referrir.~ to a relief request not covered by the positions in Generic Letter 89-04 but submitted before April 3, 1989. Withdrawal of relief requests, regardless of the prior approval status, is permitted without NRC review, presuming the 1ST program remains consistent with the regulations, the ASME Code, or Generic Letter 89-04.

Question 69 Is a continuous feedback system required to provide a mechanism to reverify that relief requl:!sts are still valid Lased on ongoing maintenance and plant modification activities? (III #52)

Response

The licensee is expected to havt a feedback system that will maintain the IST program as D living document that will be updated tu be consistent with cha11ges in plant configuration. If i:I particular relief request is no longer required because of changes in hardware, system design, or new technology, the licensee is expected to revise the progrdm to \'lithdraw the relief request. Conversely, if a system modification results in the addition of a component to the 1ST program, the feedback system should

• nsure that the Code requirements or Generic Letter 89-04 provisions are

.. 1et, or that a relief request is submitted, as appropriate.

Quest io~.1_9 Relief request requirements ort changed ir the Generic letter. Previously approved relief requt~ts are now beilly challen~ed because th~ NRC uses a different reviewer. Th*is appears to be a backfit issut. (I #4, Dave Wallac..e, Fitzpatrick)

If relief was granted by the NRC for an iter.; ciuring U;e first interval, is the ~ame rel~ef yra~t~d during the second inter*va~ even though the relief is nut rn compliance w1th GL 89-04? (Il-33, Joe Bash1sta, TMI-1)

In the 1st 10 Year submittal, an SER dpproved a relief request which is not consister,t with the alternative positions in Generic letter 89-04. Does the generic letter vGid previously approved alternatives/relief requests (via an SER) or r:;ay these i:1.lternutivE:s/relief requests not cc,rsistent with Generic Letter 89-04 still be considered valid ar.d so documented in the IST program?

(III #31, Toledo Edison)

When will it be known what the stcdt s position is on SER dpproved relief 1

requests that contradict Generic letter 89-C4 dictated testing? (III #33, Gciry J. Roesnl'r, Calla\o1ay twclear)

Response

We assume that the <.juestions are not referring tu interim reliefs but rather relief re4uests c;r, which the NRC staff prepares an SER. Assuminy that the reviewed informdtiori was complete, acturate, and remains up-to-adte, an approved relief request may be currently followed even if it conflicts with the Gerjeric letter. These types of situations will te reviewed in prepand. ion for inspections. Safety sigr,~ficant differences between the apprvved relief request and the Generic Letter will be aiscussed in an effort to obtain licen~ee ayreement to edopt the Generic letter position. Where agr1:ernent cannot be reached, the staff may consider initiation of backfit procedur~s. Relief requests are subject to review by the NRC staff at the ten-year update for consistency with current NRC rE:gulatory positions, including those contained in Generic Letter 89-04. Reliefs that are inconsistent with the generic letter would likely not be approved for a succeeding ten-year interval.

Question 71 What is the long term status of the 11 relief 11 system? (Ill #22, Point Beach Nuclear Plant)

Response

The section of the Commission's regulatior1s pertaining to the relief request system is 10 CFR 50.55a. This regu1atiur, is not, and cannot be, superseded by Generic Letter 89-04. A revisiun to thi'.:. re9ulation is under c.onsideration. With respect to the 11 relief 11 system ds described in the regulation, the stdff may, at some time i~ the future, issue additional guiddnce to provide a pre-approval mechanism much as the generic letter does in certain of its positions.

Question 72 To conform to generic letter positions, what dues document in the program" mean? Shou*1c, relief requests be generated with the understanding that the generic letter grants them? Or does a statement included in the program describiny how the deviation conforms to the generic letter suffice? (IV & V #21, Waterford 3)

Response

The 1ST program should include the deviation from the ASME Code that the licensee intends to take, and the basis for the change just as a program would normally contain. There should be sufficie11t information in the program to demonstrate that Generic Letter 89-04 is applicable to the situation in question anri that the testing being perforr.1ed conforms to the generic letter.

Question 73 Is the following statement correct? A relief reque~t submitted prior to April 3, 1989 but not discussed in e1ny SER and is not a subject of generic letter attachment 1 is dppruved fur use without any further utility reviews. (III #49)

Response

Relief requests that were on the docket before Aµril 3, 1989, for plants that are not in Table 1 or 2 in Generic Letter 89-04 and are topics trat were not discussed in Attachment 1 are approved by this generic lett~r. Any relief requests outside of the Generic Letter positions that are submitted after April 3, 1989, will require stdft review and approval before imple-mentation. The response to Question 74 expldins the basis for this approach.

Other statements regarding utility's required dctions fur the review of imple-menting procedures aaditionally apply .

.Q_uestion 74 What is the NRC's basis ior stating that approval is by virtue of the generic letter for previously submitted relief rEquests when such rel1~ts could be outside the scope of the positions in the generic letter and have not undergone NRC review? (lII t,37, Brent Metrow, Illinois Dept. of Nuclear Safety)

Response

From the general knowledge of the relitf requests, the NRC staff selected the technical issues considered the most significa1,t to be addressed by Generic Letter 89-04. The NRC st<.!h checked o sampling of the curr~nt IST programs to pr~vide confidence that those issues not ddtiressed in the Generic Letter were: not highly safety ~igr,ificant. Additional issues that would require the NRC sti.iff to perforr.1 a detailed regulatory ar,alys1s may be aadressed in future genen c guidance.

Question 75 Regarding a multi-unit site, if one unit has an approved SER which grants relief on items which do not meet all the criteria of the generic letter, can the approved SER provide a basis for the other unit to go ahead and implement the relief request prior to NRC re-review (assuming design differences do not exist between the two units)? (III #48)

Response

When relief is granted in an SER for one particular unit on a multiple unit site, that relief applies only to that one unit even if the other unit is essentially identical. If an SER is written for two (or more) units, the relief would apply to all units specified in the SER. The SER for one unit may not be used as a basis for implementing the r~quest before staff approval. See also the response to Questior1 66.

Question 76 If an SER that is received by a plant on Table 1 after the generic letter was issued denies a relief, and another plant that is not getting an SER has the same relief request grandfathered (approved), is this fair?

(II #42)

Re~ponse Such situations \'Jill be considered by the NRC staff when preparing for plant inspect ions. Safety s i gni f i cant differences between the approved relief request and Generic Letter 89-04 will be discussed at that time to try to obtain licensee agreement to follow the generic letter. If dgreemtnt cannot be reached, the staff will consider the need to initiate backfit procedures.

Question 77 Does the first sentence of [the 1ST PROGRAM APPROVAL] ~ection apply to Table 1 and Table 2 plants? The last sentence infers it does not. (IV&V#9, T. F. Hoyle, Washington Nucleor 2)

Response

The first se11tence of the "!ST PROGRAM APPROVAL" sect10n of Generic Letter 89-04 stc1tes that "[t]his generic letter approves currently submitted IST program relief requests fur licensN s who have not received 1

an SER provided that they (1) review their 1,1ust recently submitteci 1ST programs and imp lernentat iun procedures against the pos 1t icms de 1ineated rn and (2) within 6 months of the date of this letter confirm in writing their conformance with the stated p0sitions." This sentence applies only to plant~ not listed in Table 1 or 2.

Quest ion 78 In the approval process, when an SER conditionally gives relief and requires further plan changes, is an SER supplement provided, or is relief approved by letter, or is the relief granted based on conformance to the SER stipulation? (IV & V #32, Alan Harris, Waterford 3)

Oiablo Canyon's SER grants several relief requests with conditions. We are revising reliefs t~ meet these conditions. Will we need NRC approval of revi~ed reliefs prior to implementdtion? (IV & V #39, John Arhar, Pacific Gas &

Electric/Diablo Canyon)

Response

If the conditional approval specifically identifies what must be done to obtain relief, then conformance with the condition is complying with the relief. A revised program should be sent to the NRC stating that the conditions have been met. In that case, a follow-up SER would not be issued. Where the relief request is denied and the staff asks for more information (e.g., additional analysis or basis), then a specific request must be made to the staff for its review and approval before in1plementation by the licensee.

Recent and Upcoming SERs Question 79 For a Tdble I plant, can changes be made to the IST program in accordance with the generic letter, even though the SER has not been receivEd? (II #35, Al Koon, South Carolina Electric & Gas/Sumn1er Nuclear Station)

Response

Any licensee may revise its 1ST program to conform to Generic Letter 89-04.

The licensee should provide changes to the IST program to the NRC as discussed in the responses to Questions 61 and 65.

Question 80 Will the implementation schedule for procedure changes and hardware changt:s be specified in the SER? Will this schedule b~ similar to the generic letter; e.g., will the licensee have six months to effect procedure changes and 18 months/next refueling outage to make hardware changes? (IV & V #2, T. F.

Hoyle, Washington Nuclear 2)

.Response The implementation schedule for procedure and hardware changes will be contained in the SER. The NRC staff expects the schedules to be similar to thost.; in the Gerieric Letter 89-04. See also the response to Question 57.

Question 81 Before the SER is issued or for the first six months thereafter, is it per~issible for the licensee to use its current 1ST program as sub~itted to the NRC? (IV&V#3, T. F. Hoyle, Washingtc,ri Nuclear 2)

Response

Licer,sees should use the current version of their IST progrdm. The generic letter, ir: tffect, provi6es interin1 approval of the existing program for Table 1 licensees until the SER is issued.

Question 82 If a plant with an SER on its !~T program has a 10 year review up coming, how should that be handled'/ Resubmittal? (III #3b, Gary J. Roesner, Callaway Nuclear)

Response

A plant with an SER that is preparing a rev1s1on for the 10-year update should revise the program to be in conformance with the provisions of Generic Letter 89-04. The licensee does need to submit the program update tu the NRC. The progra,n should indicate which relief requests require NRC review and approval ar,d which relief requests are already approved through the gen~ric letter. Staff revi~w and apprcval of the unapproved relief requests are required before the licensee implen~nts the new program.

Alternatives to Positions in the Generic Letter Q_uestion 83 Are the new criteria always to be used even if it is not applicable?

Can it be partiolly implemented if the licensee feels the relief request is sufficiently justified by specific in house experience? (I #4, Dave Wallace, Fitzpatrick)

Response

Certain positions in the Generic Letter 89-04 are not fully applicable to all plants. For example, the components listea in positions 3 and 11 are not applicablt to all pla~ts. Further, Position 7 is applicable only to BWRs. Alternatives to the positions of the generic letter, or partial implementation as this question suggests, should be justified in accordance with Paragraph B of the letter. Specific i11-house experience is only one of the sources of informatio~ that should be utilized when evaluating alternative testing, and is not a substitute for the criteria in Paragraph B of the generic letter.

yuE:stjon 84 Will any deviations frurn the requirements ir, the Gerieric Letter be reviewed and an SEf.. *issued for those relief requests? (I #42)

Is a relief request required when onl) 2 or 3 o*, thr 4 items identified in Generic; Letter Item 6, page 3, cdn bE: 111et? (I #45)

Generic Letter 89-04 states in Paragrdph 8, that when licensets are unable to comply with the µositions of J\ttachment 1, evaluation of alternate testing shc.rnld address [four criteriaj. Is it mandatory fo1* each instance to address al I 4 of the above items? In some instances or situations, the above items may nut apply, or only a portior. may apply.

i!hen evaluc1ting an alternate test tu one of tht Positions of f1ttach111~11t l of Generic Letter 89-04, n,ay the alternate test t,e implemented without

µrior NRC approval providiny an evaluatio11 is performed and documentE:d and retained in the IST Prugram? lJoes the documer1ted dlternative test evaluation in the 1ST program have to be formdlly submitted to the NRC as an 1ST program revision, arid if so, in what time frame? (III irl:. M. H.

kithter, Commonweulth Edison)

On Page 2 of Ted Sullivdn's review, he indicated thdt the NRC will net issue SER~ in Attdch111ent 1 items ar,d Justified c.:lternatives. Are the justified alternative$ tht: 4 points on past cornponer,t hi5tor_y? Can I use these 4 alternativE:s to justify a deviatiuu tram tht Attachment 1 i,iusitions'!

If so, are these then approved by the generic letter? After issuing a confirmatory letter, ci:Sn I go through the ubove process to get II automatic" or µre-approval of Attachment 1 exceptior,s in the future? Can the 4 puints be used for non Attachment 1 iteu,s following a sinnlar µrocess? (II1#3L)

For relief requests not c.CJvered by thi~ 9eneric letter, is (in accorddr1c.c with Technical Specification 4.0.5) specific written approval required prior to imp1eme11tution? (IV&Vh8, T.F. Hoyle, Washir,gton Nuclear 2)

Re_s_p_!Jn s_e Assuming that St:c.tion XI will not be followed, P1ffa9raph B of the Ge11eric Letter 89-04 provides guidance for the situation in which u licensee is unable to comply with one of the positions of the generic letter because of design considerations or personnel hazurd (as opposed to inconvenience).

In such d situation, a licensee may devE::lop an alternative testins method provided an evaluatio11 is performed thdt dddresses four specific criteria.

The alternate test would not be acceptable unless the data associateu with those criteria are sufficient to justify its adequacy fur detecting degrada-tion and ensuring continued operability. Where the four criterid are satisfied, the alterriate test is considered approved by the generic

letter and may be implemented. The specific justification is expected to be documented in the 1ST progrdm submitted to the NRC, but need not be documented in the form of a relief request. This documentation will be subject to review for completeness, accuracy, and applicability during NRC inspections.

If at some time, the circumstances chanye such that the justification obtained through Paragraph Bis no longer valid, then the licensee must submit a relief request for ~taff review Lefore continuing the alternate test. Parayraph B may also be used when future revisions to the IST program relating to the generic letter positions are prepared. If all four criteria cannot be met, then a relief request must be submitted tu the NRC and the alternate test method cannot be implemented until staff approval is received.

For technical issues c,utside the scope of the positions in the generic letter, the alternativE: ~rovisio11s of Paragraph G may not be c.pplied onc.1, in these cases, a relief request must be subn1itteo fur NRC approval before implementation.

Question 85 Since 10 CFR 50.55d(g) i~ a top tier d0cument, is it st1ll permissible to use its provisions of the relief request prucess when thr: requiren,ents of the Code/generic lett~r cannot be met? Must thcs~ relief requests be dpproved prior to ir.1plementat1on in accordance with plant Technic<Jl Speciti-caticn 4.0.5? If a t*equired te:st cannot bE: done, should the utility use the exigency provision? (IV&V#7, T. F. Hoyle, Washington Nuclear 2)

Res pons_~

The provisions of 10 CFR 5U.55d(g) remain available for tht: licensee's use for submitt.ir19 relief rt.quests and ol.itctining upprovals. In dccordance with the Technical S~tc1fications, approval of relief requests is required before implementation. Relief requests should indicate the date by which approval is needed. Generic Letter 89-04 is µroviding anothi:r method vf receiving dp~roval of deviations fror.1 the ASME Code requir1:1,1t-nts. The licensee 1i1ay prepare a case to ju' _;ty postponement of a particular te:st on thE: basis of 1:xigency. At this r. it,t~ we are unaware of any aspE:ct of Generic Lettt-r 89-04 that would quality for the exigency pruvisior,.

Question 86

~as the gener1c letter issued as opp0s~~ to chJn~ing the reyulation?

Prior to regu*lation changes, will cc,n,rnents be solicited fruru the licensee~,':'

(IV & V #12, Arkansds Nuclear 1 J11ll d Gf:11eric Letter 89-04 is r;ut considcrc*d an alternative to the regulation but is a vehicle to obtain pn:i1µprov~o 1*elief from cu*tctin ASME Cude requirer.11~nts.

If the regulut10n is changed, the noru,dl rulernaking proce~'.> w111 be followed und comments ~Jill be sul1cited.

Requests for Additional Information (RAI)

Question 87 How do plants which have receii.ted requests for additional information (RAI) from the NRC but are not on the list of plants to receive an SER get RAI items resolved that are not addressed in the Generic Letter? (1#1, Dave Wallace, Fitzpatrick)

Does the Generic Letter or the RAI take precedence and which one must be comp l i ed with? ( I# 43)

We received 86 questions (RAI from NRC) of which some were general in terms. A couple dealt with justificatioo wording in which the questioner reconvnended a more detailed justification, although the alternate method would remain the same. Would we have to make these recommended changes and resubmit, or can we leave them alone? If revision is more of an administra-tive wording issue, then are they considered to require an SER?

( II #31, Jim Holton, Florida Power Corp.)

What do I do about an RAI that I received prior to the generic letter and issues in the RAI are outside Attachment 1? (II#43)

Response

There are a small number of pldnts that have received RAls and that have not had an IST review meeting to discuss the RAI. Utilities in this category are plants not on either Table 1 or 2 and that are expected to respond to Generic Letter 89-04 with a confirmation letter. Utilities that have received RAis do not need to respond explicitly to the RAls, but should use tht::m to assist in responding to the generic letter. The RAls provide an indication of possibly weak or questionable aspects of an 1ST program.

For those cases where the intent of dll NRC question is unclear. licensees may obtain clarification through the rrnc Project Manager.

Question 88 Some questions in a recent RAI are in conflict with previously approved relief re4uests. Which one must be compliE~ with? (I #44)

Response

Previously approved relief reque~ts remain valid despite what might appear to be a conflictir,9 position in an RAI. This statement assumes that the previously approved relief w~s granted on the basis of accurate and complet~

information available to the NRC staff at that time.

Modification of the Generic Letter Question 89 Is a NUREG to be issued on this Generic Letter to clarify underlying issues?

(I #7, Dave Wallace, Fitzpatrick)

Response

There is no current plan to preparf! a NUREG document to clarify any under-lying issues with Generic Letter 89-04. These minutes will be sent to all licensees and attendees who provided their address.

(Juestion 90 Will Generic Letter 89-04 be updated from tirn~ to time to provide additional positions on 1ST programs in areas such as the following? The ASME Section XI Code does not require leak t~sting for valves where leakage is continuously monitored, however, for PWR plants the NRC often requires leak testing for Category A valves suc.h as the RCS accumulator/core flood discharge check valves which are monitored continuously for seat leakage. (III #11, Larry Campbell, Toledo Edison)

The staff has no plan to issue a supplement to Generic Letter 89-04.

Another generic. letter on 1ST may be issued in the future, but would cover new topics or expand on the current scope of components covered by the 1ST program requi r~d by the ASME Code. The Code does require that valves whose leak tight integrity is important fc performance of their safety function be individually leak rate tested. From the staff's experience, most continuously monitored leakage detection systems do not verify the leaktight intf!grity of each valve in the flow path and the staff does not consider these systems to meet the Code requirements.

Backfit Concerns Question 91 The Generic Letter states that "In cases where conformance with the stated positions would result in equipment modifications, the licensee should provide in his conformation letter a schedule for *ompleting the required modifications." The Generic Letter goes on to r,tr,re acceptable schedules for completion of these mods. Are these modifications subject to the provisions of 10 CFR 50.109 backfitting? {1#2, Dave Wallace, Fitzpatrick)

Please confirm that the NRC's opinion and present position is that the generic letter is not considered a backfit for all utilities. {11#17, K. Jacobs, New York Power Authority)

-b3-noes the statf intend to do a hackfit analysis reyardiny this position? We currently have approved relief re~uests for the first Ten ~ear Interval in which the staff has found our lack of instrumentation acceptable. This appli~s to other positions as well. (II #34, Philip J. Horth, Duke Power)

Do the modificatior,s that are needed to contorm with the stated positions require a t,ac;kfit. If modif1catiuns are necessary to comply with the statc.:d positions, are relief requests necessary if it is deemed impracticil1 to make the modificati~ns? It not through relief, how do we deal with these issues?

What if no maintenance history is avai *1able to substantiate relief? (IV & V

1. 17, Arkansas Nuc lec1r 1 and 2)

Defend ur explain your basi~ for saying the generic letter does not require "Lackfit. (IV & V #26, Paul Croyt Southun California Edison/San Onofre)

Besponse Generic Lett~r 89-04 was presented ~o the NRC's Co~mittee to Review Generic Requirements (CRGR) as a backf1t issue, and certain positions were identified as changes to past staff positions. As discussed with the CRGR, the staff determined thdt those positions i 11 the yeneri c 1etter that represented changes frum previous staff positions were necessary in order to bring licen~ees into comi:iliance with the Commission's regulations. Therefore, according to 10 CFR 50.109 (a)(4)(i), a backfit analysis was not required to justify issudnce of the generic letter. If the positions in the generic letter cannot be met, the uption discussed in Paragraph l3 r.1ay be available.

Further, if the licensee will nut be following the generic letter positions, Paragraph B of the letter, and the ASME Code, the licensee must submit to the NRC stuff a request for relief from the ASME Code. Where a licensee is following a pr0vision of its operating license or a particular exemption from the ASME Code that was gra11ted by the NRC staff, a backfit analysis would need to be performed by the NRC stdff before requiring a.~ change to thdt 1 icer,see practice. With res11ect to the staff review of previously approved relief requests at the ten-year update of the IST program, however, a backfit analysis would not be necessary. ~ee the response to Question 70.

Use of OM-6 Jnd 10 Questior; 92 When addressing cold shutdowns, OM-10 uses statements like sufficient 11 duration and shall continue." I/hen tryiflg to implement these statements, 11 11 operations personnel frequently ask what is the NRC s definition of a cold 1

shutdown of sufficient duration. Is cold shutdown testing expected to be back to back tests or can 1 or 2 day breaks be acceptable (i.e. shall continue is not edsily defined)? (I #39, Jeff l~eyhard, Nine ~lile Stdtion)

ln 1987 and early 1988, the NRC rejec.ted a gener,tl relief rt!quest to us£:

OM-6 criterio tor flow and delta pressure for µunips. Can we now revise our proyram to use the crittria of OM-6 and OM-10? It the answer is yes, do we need a relief requE:st'? (l f21. Jeff NEyhard. Nine Mile Statior,)

~Jhat is the tirnf-! frame for the 10 CFR 50.55d(y) change? Is the NRC willing to accept the currently approvl:d OM-6/0M-10? ( l 1#24, Garry Galbreilth, Duke Power)

Will any of the guidunre provide~ 1~ the generic letter chdnge with the implementation of Pdrt 6 and Part 10 of O&M? (II #40, J. Zudans)

Once OM-6 and OM-10 an: approved, wi 11 it be required to implement tht:!m immediately (within 6 months) or will they be imµlemented at the next

µrograr.1 update'/ {Ill #~i, Larry Hochman, Nutect.)

Re~ponse Rulemaking tc reference A~ME standards OM-6 and lU in the regulations is underway at this time. lt can be said, however, that, in some recent relief n:quest evaluations, the use of the pump allowuble range limits identified ir, OM-6 for flow rate and different id 1 p.ressure has not been found acceptable to the staff. The staff has not compltted its dSsessmPnt of the inter-relationship of Generic Letter b9-04 and OM-6 und 10. When appropriate:

references to OM-6 and 10 are incorpordted in the regulations. these standards may be used by the lic.tnsee as the rt gulatior,s permit the use of more recent 1

referenced st~ndards. We anticipate that rulema~ing to reference these standards will be issued for public conrnent in the near future.

Solenoid-Operdtc~cJ Valves (SOVs)

Question 93 To perfurm position indication testing on solenoid optrdted valves, is a light check dCceptable or must the positior, verification be performed by rur,ning the system or injecting air, etc. to prove vc1lve position? (I #29, Jeff Neyhdrd, Nine Mile Station) ls a remote position verification required for SOVs with no positive means availabl~? (III #47)

HespE~se Verification of remote position indication by IWV-3300 is required to ensure that the indic.at.ion accurately reflects actual valve µusition. This could take the form of a differential pressure test, flowrate measure~ent, or other change in some µarameter that posit1Vely shows that the valve is in the indicated position. An indirect verification, using techniques such as r~diography, mJy also be acceptable.

-.;er)'*

benera1~~ionJ yuestion 94 Please clarify what 1s mednt by "one part of a DrvdO effort" in the Background sectio11 ut the Generic Letter. ti ffll, Shc1fi Rokeryo, New rnrk Power* AUthori ty)

Response

Generic Letter ~Y-04 is part of a larger progrdm to improve IST throughout the industry and to provide additional information and clarification on the subject to all affected parties. The joint ASME/NRC Symposium 011 1ST held in Washington, O. C., in August 1989 is dlso part of this effort. Additional generic regulatory guidance may be prepared on other IST aspects. For a discussion of the "broad effort" that NRC is pursuing, refer to the summary of the presentation by Tad Marsh provided in these meeting minutes.

guestion 95 How do the Generic Letter 89-04 requirements differ from the ASHE requirement~?

(I #12, John Wiedemann, PSE&G)

Response

Generic Letter 89-04 is intended to provide fundamental information on the NRC's interpretation of certain Technical Specifications and ASME Code requirements, and to ider,tify certain alternative testing that the NRC staff finds acceptable. The generic letter also goes beyond the ASME Code in that it covers procedural issues ir, addition to programmatic i~sues.

ltie generic letter may contain Code interpretations that differ from those of certain licensees. Th~ one area that we are aware of in the generic letter that is different from the Code is contained in Position 8 on the starting point for the time period in Technical Specification action statements. This position is consistent with other Technical Specification starting points. This position is also articulated in the bases for certdin of the Standard Technical Specifications.

~estion 96 In a refueling outage that is greater than 3 months, how is the cold shutdown frequency handled? Can we perform the: cold shutdown procedure once during the outage or do we perform the cold shutdown procedure every 3 months during the outage? (1#17, Jeff Neyhard, Nine Mile Station)

Response

When a component is required to be in service during the outage, the testing is expected to be performed quarterly during the outage. When a component is not required to be operable during an outage, the testing need not be performed quarterly. In accordance with IWV-3416 of the ASME Code, however, those valves must be tested within 30 days before return of the system to operable status. Further, as required by IWP-3400(a), pumps must be tested within one week after the plant is returned to normal operation.

Question 97 Is radiography on check valves an acceptable method for determining valve position? (I #25, Bill Kittle, PSE&G - Salem)

Response

Radiography may be utilized if it clearly indicates the octual position of the valve disk.

Question 98 Most plants have been given relief from measuring pump bearing temperatures

µer IWP-4310. Is it the policy of the NRC that this will continue to be an item of 11 generic 11 relief? (II elO, John Zudans, Florida Power & Light)

Response

It is true that some plants have been given relief from measuring pump bearing temperatures on the basis of the impracticality of measuring temperature for specific pump desiyns. This issue has not been treated as an item of "generic relief" because each relief request has been individually evaluated. F0r the foreseeable future, NRC will continue to lValuate these relief requests on a case-by-case basis.

Where pump parameter mea suri ny instruments do not meet the specific requirements of the Code but do satisfy the fundamental technical requirements for testing, would it be acceptable to allow relief? (II #12, John Zudans, Florida Power & Light)

Bespons~

It would be difficult to answer this question without more specific information. There have been cases where r<!lief re4u(:sts in this arec.1 have been approved. In tho~e cases, however, the basis for relief has been that

the instrur11entation has been adequate to meet the fundamerital objective of detecting degradation. In relief requests of this tyr", the licensees should address the reason thdt the ASME Code requiremints are not currently being met and the basis for concluding that the fundam~ntal objectives of 1ST are being accomplished.

Question 100 The schedule for exerc1s1ng manual valves should be extended to something less than once each quarter. ls this feasible? (II #13, John Zudans, Florida Power and Light)

We are not aware of a basis for exercising mar,ual valves at a frequency di1*ferent from other valves. Because this subJect is not specifically related to Generic Letter 89-04, it was not addressed at any length r' *ing the mceti ng. If the license,

• are awart:: of reasons why the frequency shou *d be changed, we recommend tha . . this subject be explored with the ASME O&M Working Group on Valves.

~estion 101 It has been said that some plants have excellent IST organizations. Who are they? (II #16, Charlie Dunkerly, Calvert Cliffs)

Respon:.~

Dresder1 is one examµle of a facility with a good IST organization.

Q~estio11 102 How do we handle cold shutdown justifications in the future? (ll #20, Art Caudill, Georgia Power/Vogtle Project)

~esponse Cold shutdown justificatiuns were previously reviewed by NRR for adequacy.

In the future, th~y will be reviewed during 1ST inspections. The cold shutdown justifications are expected to be described in the IST µrogram the licensee provides to the NRC staff.

Question _!03 After this meeting, what is the process for getting further questions answered regarding the generic letter? (II #21, Garry Galbreath, Duke Power)

Response

These meeting minutes will be distributed, which should answer most of the industry's questions. If after reading the meeting minutes you still have questions, you may contact the cognizant personnel through the NRC Project Manager.

Quest ion 104 Uoes "needed to n1itigate the consequences of an accident" mean an accident as described in Chapter 14 of the Final Safety Analysis Report (FSAR)?

(II #36, Charlie Ounkerly, Calvert Cliffs)

Response

We assume that the question is directed to the chapter of the FSAR describing accident analyses performed by the licensee. Those analyses are intended to provide confidence that the publit health and safety will be protected in the event of certain accidents and anticipated transients at a nuclear power plant. The term "accident" is also used in different sections of the Commission's regulations. Fur example, Appendix B to 10 CFf< Part 50 establishes quality assurance requirements for the design, construction, and operation of "structures, systems, and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public." Part 100 describes structures, systems, and compone11ts that must be designed to remain functional during a "safe shutdown earthquake" as those necessary to ensure: (1) the integrity of the 11 reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of this part." M, can be seen, the term "accident" is used by the Commission to describe a broad range of possible adverse events at a nuclear power plant.

Therefore, although most of the accidents of concern to IST are addressed in the accident analyses chapter, licensees should be aware that there may be other accident analyses in th~ FSAR that need to be con5idered.

Question 105 This question is in reference to 10 CFR 50.55a(g)(4): " ... to the extent practical within the limitations of design, geometry, and materials of construction of the components." In reviewing this wording, along with the statements of consideration, do you think this rule was intended to impose plant modifications as a result of meeting subsequent editions and addenda? That is, once the staff evaluates a licensee's determination of impracticality, will the NRC impose plant modifications as alternate requirements? (II #37, Mark Dryden, Florida Power & Light)

f<esponse The NRC staff in the Mechanical Engineering Branch of NRR ha~ .a<.l lengthy discussions with the NRC Office of the General Counsel 011 this matter. The current interpretation of the rule is that it is not intended to require a blanket imposition of all plunt modifications that would be nece~sary to comply with subsequent editions and addenda. The rule <loes require an evaluation of the impact on the licensee, that is the impracticdlity of making the modifications, as part of an assessment of the requests for relief fror,1 the ASME Code requirements. The legal staff has stated that there is nothing in the regulations that relieves licensees from m~king all hardware modifications to the plant to comply with changes to 1ST requirements throughout a plant's life in later editions uf s~ction XI.

Some hard\vure modifications can be required. The diffic.ult issue to resolve is how much may be required. For example, major equipment or piping modifications may be beyond the limitations of practicality in meeting subsequent editions of the Code. We, however, regard modifications such as the installation of instrumentation to be practical as used in 10 CFR 50.55a(g)(4).

Question 106 For plants that do not have operating licenses, 10 CFR 50.55 requires that you apply t~c codes that are in effect 12 months prior to plant startup. Where does the 6 month conformance letter stand for construction plants in this situation? (Il t,39, Jackie Jackson, Tennessee Valley Authority)

Response

There are only two plants rxpected to receive operating licenses for which the staff's review of the IST program has not been completed. These plants are Comanche Peak and Watts Bar. These two plants will be treated essentially dS Table 1 plants in that a review will be completed and an SER issued. The reviews of the Comanche Peak and Watts Bar IST programs, however, may not be completed in the same time frame as the reviews for plants listed in Table 1. To obtain the scheduled completion dates for the JST program reviews, the Comanche Peak and Watts B* organizations should contact th~ir respective NRC Project Managers.

Question 107 Currently, we only test the JCS pump suction check valves JCS 3A(B) to verify they opeu as part of the JCS pump test. Originally, the only safety function recoynized was for the valves to open to provide.- water source, the RWST, to the ICS pumps. During au independent revie..., Jf thP 1ST program, it was determined that these valves niay also ha.2 a safety function to close when the pumps are taking suction from the RHR system. These valves, if they failed open, could provide another flowpath (to the RvJST) besides the normal flowpath to containment. This flowpath would also

allow potentially contaminated water from the containment sump into the RWST (NOT DESIRABLE). As part of our company's in-house safety syster.;

functional inspection, it was determined that if these check valves failed open, adequate flow to the containment would still be achieved. We are also converting the manual valves upstream of lCS 3A(B) into motor operated valves in order to prevent sump water form getting into the RWST. Do these check valves need to be leak tested? (111#17, Wisconsin Public Service Corp.)

Should Category A be applied to valves other than containment isolation valves (e.g., valves which isulate HVAC damper air accumulators: checks/~OVs)?

(IV & V #27, Wayne Wolling, Gulf States Utility/River Bend)

Respons1:

The NRC staff has a generic concern with the current practice of categoriza-tion of check valves. The ASME Code assigns all check valves as Category C.

If seat leakage of a check vulve is limited to a specified amount, the Code also requires that valve to be assigned to Category A. Whereas Category C check valves are required by the Code only to be exerc1sed 011 a p~riodic basis, Category A/C check valves must be ledk tested in addition to being exercised. The NRC staff has fuuna thdt, in many instc1nces, check valves are not being assigned tu Category A/C despite the fact that cred1t is taken by the licensee for the check valve providing an essentially leak tight function. The categorization of a check Vdlve is not dependent sulely on the function performed by the valve, such as whether it is a containment 1solation valve. When deterrnining the proper categorization of a check valve, a licensee should take all applicable aspects i1,to account. Fur example, the licensee should detern1ine (1) whether the flow requirements for connected syste111s can be achieved with the maximum possible leakage through the check valve, (2) the effect of any reduced system flows resulting from the leakage 011 the performance ot other systems and co111ponents, (3) the consequences of the loss of water from the system, (4) the effect that backflow through the valve may have un piping and components, such as the effect of high temperature and therma"I stresses, and (5) the radiological expc.,sure tu plant person11e*1 and the µublic caused by the lec1k. If any of the above considerations indicdte that Cdtegory C testing ma} not be adequate, licensees should a!>sign the check valve tu Category A/C and should comply with the associated leak testing r ~uirements.

Quest ion 108 What is the IWC's opinior., per Generic Letter 89-04, of n'in-4uantit1c.1!Jle demonstrations of perforrna11ce*t Fur example, a solenoid valve has no position indication that can be observed or tir.ied, but bec1rin9 temperutures show no overheating. (Ill 1124, Point 81:ach Nucle,ir Plant)

Response

The NRC staff is discouraging the use of qualitative criteria as an alterna-tive to the Code required component testing. Licensees should strive to develop a quantitative method of determining the ability of a component to perform its required functions. This recommendation is based on the goal of 1ST to detect degradation prior to failure of the component. For specific examples, see the response to Question 1. With respect to the specific question, more details would be necessary before arriving at the acceptability of the suggested method.

Question 109 Should LaSalle County Station be on Table 2 of Generic Letter 89-04?

If not, why? Zion Station underwent the same review 2 months after LaSalle and they appear on Table 2. (III #25, Roger Sagmoe, Commonwealth Edison Co.)

Response

Although the LaSalle nuclear power plant received an SER about a year ago, a significant revision to its IST program was subsequently submitted for NRG review. The NRG staff determined that a review of the IST program could not be completed in the necessary time frame. In the context of Generic Letter 89-04, LaSalle, therefore, has been classified as a plant that does not possess a current SER and will not be receiving an SER. As a result, LaSalle is expected to respond to the generic letter in accordance with the implementation provisions for plants not listed in Table 1 or 2.

Quest ion 110 What additional NRG guidance can be provided on testing skid-mounted pumps and valves (i.e., diesel generator systems: lube oil pumps/valves, internal engine cooling; RCIC systems - condensate/vacuum pumps with only one source of power, etc.)? Most of these pumps and valves do not have the necessary test instrumentation to support ASME Section XI testing and do not fall within the scope statements of IWP and IWV. Will modifications need to be performed? (III #30, RJger Sagmoe, Commonwealth Edison Co.)

1

Response

The purpose of inservice testing is to provide assurance of tht operability of components and to detect degradation in their performance. Where a particular component is integrated with other components in a system, it may be difficult to perform an individual test of that compor,ent. In specific cases for which individual testiny is not feasible, an alternate test should be proposed by the licensee. In developing an alternatP. test, the licensee should attempt to develop quantitative criteria to evaluate the operability and cond it i 011 of the colilponent.

Question 111 Is temporary flow instrumentation (i.e., portable flow meter) permitted in lieu of a modification to install permanent flow instrumentation? If so, is relief required? {III #40)

Response

The staff does not interpret the ASME Code as excluding the use of portable flow rate instrumentation, such as ultrasonic. We have seen difficulty, however, in meeting the Code-specified accuracy requirements with these instruments.

Quest ion 112 Is trending a requirement for pump~ Is it a requirement for valves?

The Code and the regulations do not address this, nor does the 9eneric letter. (IV&V#28, Wayne Walling, Gulf States Utility/River Bend)

Response

We define "trending" as the analysis of test data to detect degradation of the tested component and to enable preventive maintenance to be performed before significant challenges to component operability occur. The ASME Code contains few requirements for trending of t~st data. For example, the ASME Code in IWV-3417(a) provides for more freq~ent stroke-time testing of power-operated valves where an increase in stroke time is seen from a previous test. The NRC staff allows a reference value to be used for this comparison in Position 6 of Generic Letter 89-04. In IWV-3427(b), the Code provides for more frequent testing, and possibly maintenance, where the leak rate of a large valve increases beyond a specified amount from one test to another. In Position 10 of the generic letter, the NRC staff explains 1ts view that this provision of the Code may not be worthwhile and may be suspended. Although the ASME Code is weak in the area of trending, the NRC staff remains of the view that trending is a valuable tool in the IST program.

The Corrmission's regulations can be interpreted to require efforts in this area. More explicit guidance for trending may be developed in the future.

In the meantime, we recommend that licensees analyze IST data to take advantage of the benefits oi trending.

INSERVICE TESTING GENERIC LETTER 89-04 REGIONAL MEETINGS LGGISTlCS 0

ATTENDANCE SHEETS IN DACK V

l~AME TAGS

..,

CARDS FOR QUESTIONS - NAME, COMPANY, QUESTION 0

MEETING MINUTES WILL BE PUBLISHED 0

QUESTIONS - WE'LL ANSWER THEM ALL SCHEDULE:

0:00-10:15 OPENING REMARKS - REGION MANAGEMENT 10:15-10:30 BACKGROUND ON GENERIC LETTER 89-04 T. MARSH 10:30-11:00 APPROACH OF GENERIC LETTER 89 T. SULLIVAN ll:OO-l::3G QUESTIG~/DISCUSSION SESSION I

.12:)u- 2:00 LUNCH/NRC STAFF CAUCUS 2:0U - 4:00 QUESTIONS/DISCUSSION SESSION II 4:0G - 4:30 BREAK/NRC STAFF CAUCUS 4:30 - 5:00 CLOSING REMARKS - NRC

OBJECTIVE TO ASSESS OPERATIONAL READINESS OF SAFETY RELATED PUMPS AND VALVES JO CFR 50.55A 0

REQUIRES PUMPS AND VALVE IST PROGRAM IN ACCORCANCE WI TH ASME CODE, SECT I or~ XI 0

UPDATE IST PROGRA~,S TU TtiE CURRENT CODE ED IT I 0~

AND ADDENDA EVERY 10 YEARS 0

/\LLOWS THE GkANTING OF RELIEF REOU[STS FOR COCE RE~UIREMENTS THAT ARE l{1PRACTICAL STATUS

~ FEW PLANTS HAVE RECEIVED SERs 0

SOME OF THE ISSUED SERs ARE OUT OF DATE (SUPERSEDED HY LATTER SUBMITTAL)

PROBLEMS 0

I MADE~UATE TEST I NG REQU I HEhENTS I I~ CODE 0

NO WRITTEN NRC GUIUANCE ON 1ST 0

fiUGE VOLUME OF PROGRAMS/REVISIONS/RELIEF RECJUESTS HUGE EACKLOG 0

RELIEF RE~UESTS I~PLEMENTED WITHOUT PRIOR NRC APPkUVAL 0

INSPECTION EFFECTIVENESS HA~PERED 0

I ST PROGRAl"1 Ir1PLEMENTAT 1or. VAk IES - sur~ET IMES POOH

PURPOSE OF GENERIC LETTER <GL>

0 PROVIDES GENERIC GUIDANCE ON ELEVEN SIGNIFICANT 1ST PROBLEM AREAS 0

PROVIDES GUIDANCE ON DEVELOPING ACCEPTABLE IST PROGRAMS

~ CLARIFIES APPROVAL STATUS OF IST PROGRAMS (I.E., RESOLVES TS 4.0.5 ISSUE)

FUTURE NEW ASME STANDARDS O&M G PUl'1PS O&M 10 VhLVES MODIFY 10 CFR 50.55A(G)

FURTHER GENERIC LETTERS IS1 SYMPOSIUM - AUGUST l - 3, 1S89

APPROACH USED IN GENERIC LETTER (GL) 89-04 THREE GROUPINGS OF PLANTS TABLE 1 PLANTS 0

SER NEARING COMPLETION 0

SER CONSTITUTES APPROVAL TABL~ 2 PLANTS 0

SER ISSUED ON CURRENTLY SUBMITTED PROGRAM 0

SER CONSTITUTES APPROVAL TABLE 1 AND 2 PLANTS 0

DO NOT NEED TO RESPOND TO GL 0

NEED TO ASSUkE PROCEDURES CONSISTENT WITH GL

PLANTS NOT ON EITHER TABLE 0

GL CONSTITUTES APPROVAL PROVIDED LICENSEES:

- REVIEW PROGRAMS AGAINST hTTACHED POSITIONS, AND

- CONFIRM CONFORMANCE OR JUSTIFY DEVIATIONS FROM ATTACHED POSITIONS IN SIX MONTHS, AND

- MAKE ANY MODIFICATIONS WITHIN SPECIFIED TIME 0

ALTERNATIVES TO ATTACHED POSITIONS MAY DE IMPLEMENTED PROVIDED:

- MAINTENANCE AND DEGRADATION HISTORY EVALUATED

- DEVlATION JUSTIFIED AND DOCUMENTED 0

RESULTING IST PRGGRAM TO BE PROVIDED TO NRC 0

NRC WILL NOT ISSUE SERs ON

- CONFORMANCE WITh ATTACHED POSITIONS

- JUSTIFIED ALTERNATIVES TO ATTACHED POSITIONS 0

NkC WILL ISSUE SERs ON

- "E" Ri:LIEF REQUESTS ON AHEAS NOT COVERED BY ATTACHED

. POSITIONS

rROGRAM UPDATES/REVISIONS

., FOR PROGRAt': CHANGES COVERED BY ATT f1CHED POSITIONS

- SAME GUIDANCE AS ABOVE

" FOR PROGRAM CHANGl:~ NOT COVERED BY ATTACHED POSITIO~S

- STAFF ~ILL EVALUATE PER 10 CFR 50.55A(G) 0 RELIEF REQUE~TS PREVIOUSLY APPROV[D

- HILL NOT BE REEVALUATED

- APPROVAL REMAINS IN EFFECT INSPECTION AND ENFORCEMENT 0

INSPECTIONS TO LE CONDUCTED Fuk CONFORi*,ANCE WI TH 10 CF~ 50.55A, AS EXPLAINED I~ GL

- FCJCl.JS Cf\ ATT J\CI l[L PCS 11 I UNS

- OTHER AREAS MAY DE INSPECTED

UNITED STATES FlfllT CLAU MAIL NUCLEAR REGULATORY COMMISSION f'OITAOE

• REI PAID UINRC WASHINGTON, D.C. 20655 PERMIT No. 0-t7 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, *300