ML19242B728
| ML19242B728 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 08/03/1979 |
| From: | Delgeorge L COMMONWEALTH EDISON CO. |
| To: | Parr O Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7908090344 | |
| Download: ML19242B728 (34) | |
Text
{{#Wiki_filter:~ Commonwealth Edison One ff!!#3ME31f13!3 Ch' cap.13nois [g Address Reply to: Post Office Box 767
% ,.- Chicago, Ilhnois 60690 August 3, 1979 Mr. O. D. Parr, Chief Light Water nonctors - nranch 3 Division of Project Manegemcat U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
LaSalle County Station Units 1 and 2 Initial Test Program Requirements NRC Docket Nos. 50-373 and 50-37_4
Dear Mr. Parr:
Enclosed are the Co:amonucalth Edison responses to NRC questions informally submitted to this applicant regarding the suoject Initial Test Program. These responses repre ent the culnination of an effort which involved numerous meetinc in your offices and telephone conferences between laembers of your staff and the LaSalle County operatio staff. We IP.v e been 3cd to believe by the NnC reviewer (Mr. B. Cl:iyten) thet the enclosed responses fulfill all existing require'aonts . applicable to LaSalle County Units 1 and 2 in this arer.. On the basis of this belief we are proceeding to complete the test program. As you must well appreciate, it is vital to the expcaitious completion of the Initial Test Program, and nore specifically the Prcoperational Test Program, that your con-currence in our planned approach be fornalized. We expect, therefore, that the open item (No. 63 in the latent summary provided by Mr. A. Bournia on July 26, 1979 and documented in tbc O. D. Parr letter to L. O. DelGeorge dated January 22, 1979) will be formally closed. Eight (8) copies of the responses to your informal questions are enclosed for your use. Appropriate revisions either have been made or will be included in Amendment 46 of the LaSalle County FSAR to document the resolution of your concerns. The material contained in the FSAR amendments will have been nub'nitted in 40 copies as required. q a Y~ 7908000 3 7'/ [
Commonwealth Edison NRC Docket I:os. 50-373/374 Mr. O. D. Parr: August 3, 1979 If you have any additional questions in this regard, please corcanunicate them as quickly as ponsible. It is now vitally important that your concurrence in this matter be appropriately communicatc' to the Regional Office of Inspection and Enforcement. Very truly yours,
/ /
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L. O. DelGeorge Nuclear Licensing Administrator Boiling Water Reactors enclosure 4 A
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- 2. 1he inferration providcd in Ecction 14 should be u;u':f.ed and/or coriccled for e,y<str s dn , n o n f. s i h.: t w i l l ' . r e l . d o n to
, l i t i g r i e .i n l i c i r s t r .1 : : n,i,nte . lit!. ut s r .i ('.i.!$) aft.:r ll.-'
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, Coaem ning /a .l; on !!c I/Ot 13 e d d.e t , Eil1: on ei r,iLted to the i nn i n 1 l a ' i<.n of the Al terna te R v tor Scron l'ys te'a for I.SCS by 1 < l t< r of ."ept.. 30 1 ' #2 from H.1,. Polip>r t o F 1 Ca se.
Thir fix include.' m f e t;, - d ed i . ' e d r a r . - range pre:.:;u re :en:-orn, a rodundant clnnr.eled b te':-up ' rr imbili ty .i th added p" er-en relayn, omn ra ted nnd enlarced < u r:- -d :.eln rco h .' .i d - and piping, the reci rcula t ion pur.p trip and a feed':a tor :.e t -back f . > turn. Tlnt Ali;S fix decreased the AT/lS prebabili ty by a facto: of 10-2 f rc.T. the precent valve and <:till retained the GLC nyctem ar. a ranuall:,-c;er; t< d niticator. ]n l' arch 1977 Edir,en requested !!RC's renconse on thi:, AG513 cor: aliment for- J Sct,, but Ins received no responae from the f!RC, nor ins any c. the other eleven utilities v ho eccaited in 1976 to install AliSS for AT'ilS preclusion. To da te, no cost-ef fective nitignor has '.>een defined ..'hich can Inndle AT/lS vci th failures as pos tulated dwing such trannients.
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- 3. To clarify the i n f o r. i t i r.n in' ction 14 ? relatim to "
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- a. lhose that will be used for sa fe shutd.. .n .n.d w olik.cn of the r ea<.tcr under norml plant conditions uid for
, rnaintaining the rc' actor in a safe ccndition for an extended shutJc.in period; or b, Those that will he used for safe shutdown and cool-do'.in oi the reactor under transient (infrtquent or raaderately f requrnt events) condi tions and po ,tulated accident conditions and for maintaining the reactor in a sa fe rondition for an extended shutde.-in period following such co,ditions; or
- c. Those that will be used for establishing conformnce with safety limits or limiting conditions for operation that will be included in the facility technical specifications; or s .
- d. Those that are classified as engineered safety fenures or will be used to support or ensure the npera tions of engincored safety features within design limits; or e.
Those that are assun.ed to function or for '..hich credit is taken in the accident analysis for the facility, as described in the FSAR; or f. Those t':at will be used to process , s tore, control, or limit the release of radioactive materials. _ Her.ponne:
/J 1 tes ts l i s te.d i n '"nbl e 14.2-100 are essential to the above criterin vel t h the f'ollce.inC LXerPtionn.
- 1. Control Fcd &quence Excha:q;e.
2 T,t o n Prcalu',' tl en .
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.See revised Table 1/,.2-91. 74 ) 4.,
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- 1. The reactor protection system (I.'J S) proopn: a t.i nna l *: e s t d emen st ra tes the capability of the RDS to initiate a scram signal in colif o t. . alice to System design. A dd i t i 3 n.111 y , a test is p2 rf or;- ed t o demonst r:it e RPG response to mo t.or- g on e ra t or (MG) set caastdcwn. A scram .esponse time test is performed f or each RPS sencor.
SYSTFM I!!ITI AL CCMDITICNS MiD F ' -:R EQUI SI TES i
- 1. All constructi.on tests are completed and approved.
- 2. Verify instruments within the RPS boundary have been calibrated.
- 3. Reactor manual control and CRD hyoraulic systems are operational bef ore scram t esting.
- 4. A-c and d-c electrical po ler is available.
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- 5. RP'S M3 sets are in service.
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v SAFETY PRECAUTIONS . ' O. . O ^
- 1. Vetity that all safety and constuction tags have been removed p f rom all equ ' ~nent to be operated or energized.
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~J v 2. Ensure that idequate electrical safety precautions are 1 observed whi e working on energized equipment. . 3 5o ? o- TFST PEccEDURE
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- 1. Verify proper operation of the RPS MG sets.
- 2. Verif y all RPS sencor logic systems and scram relay operation.
- 3. All scram re <.: e t time delay ot oratj ens ;1r e demons t z.a ted. ,
- 4. D< :r on s t ra t e t ':e abi lity of the cy c tizm to x: ram the re.:c t or W tl:in a ci":ci Ii C l'~'3.
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- 6. Verif y pro;ur Gyst: n o p e r?:s t i on fi n.a no ul an<I alternate pcuer supplies an<i <!cri:rg bu i.tchi ng transients.
ACCr.PT ANCE CS IT GI A
- 1. RPS MG sets operate in accordance with design.
- 2. All systera logic functions, interlocks, and time delay functions are within design tolerances.
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- 3. RPS responds correctly to simulat ed scra:a condition input ;
signals and annunciator and indicating systems f unc ti on - p rope rl y. o. e, 4. gj+5qRg6bueymc.h'=-9.a'. m sm w r-, .1 :,1,:+s7-tje[E ydhjOf c .2 = ; y A SM n, ,h p me:O-by M:J'Ch$s 1c'ef/'. > vTc sic' '!J. I'v2. .V+ngmt r La, .lw.;s .
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- 5. Power supply switching and interlocks function properly.
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S D-- E ll 101 TI- ST Oltl ECT IV ES
- 1. The objcetive of the iorhine electrohydrau1ie centrol test is {
to demonst ra te the reliable operation of the
- ys t em to the ,
extent ponni ble wi t hont reactor steam. l i SYSTEM I N 1'I I A L CON C I'l I ONS At1C l'R Elt EQtil SI TES
- 1. All construction t. c o t s are completed and approved. l l
- 2. Electrical p ov.e r is available to motors, control circuits, ,
and instrumentation. l l l
- 3. Turbine builling clcsed cooling water system rs opera tional. j i
li . Peactor protection system functional. l i p 5. Instrument calibration check sheeta complete. - SAFETY I'lt EC AU'l I ONS
- 1. All safety and construction tags are removed irom equipment to be operated.
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- 2. Ensure that adequate electrical precautions are fallowed when .
worki ng on energized equiptrent.
- 3. Ensure that main steamlines are vented before opening main steam stop valves and control valves.
'l E ST I'I;CC l DUil E
- 1. All cent r ols, ala2ms and i n t.o r l ock s are checked for proper opera t ion.
- 2. The EllC f lui d nyut em is veriiied to opeiaLe proper 1y.
- 3. Eemot.e-ut c ra t ed va lv< u are cincked for proper o} o t a t.i o n .
ft . I':Itc con t ro l l i ng f unct ion is verified using ;i mu la t .>d input nignair. (% c1 Q> c)c ' 1 's . < - -: 0 :4
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ACCEL >TA Nf:E CH I'1 ER I A
- 1. The t ur bine 1-HC syst em, i n cl ud i ng a:- soci a t.ed i n t e r l o:: k <,,
i r .i ls and ins t ru: rent at. ion, ['erforms in accordance w i t. h d e t .i g n . I
- 2. All kPS inputs function accoiding to design.
- 3. All ala rms and annuncia to r c f unct. ion in accordance with design.
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1rovide ssuraru e of a reliable syst. 1 1L is ear ;.osi'i.n U,3t n Occep t ance c ri t s.:rion o f a t l eas t f i ve LCa secu t.i',e sutt. ,sful cold s tarts be es tablished for the tes t.. IlOSpOnGO: f1id-tht-liit0-se cep t -ou r . respunne-to- Q /,23. 28 7. . t'e do_ no t .i n t,;nd _.to en tabli ch-a n --a coep Lanec - c ri Ler i on of-bcoorme u t i ve- suen m,<;rul--cni d.
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l s'U c N il' To vorify tle p : . si,e r uiu r.:( ion of I h. ' Fea<:t<>i ('< i r e 1: <il a l i cin C o o ] i ntj (RCIC) ' ' 'i t e m over i t.a e sected op. r a t i nnj [in e , .u: e range. DESC R I PTI O!J 5 e.C XM S EfG' ON h.i([ 't # 1 'f lK . j P. ohn-Re] C- sys t em test-consists of-Lwo pa r t s : - - i n j <le t i o n - t o -- Lle e on2Onsate <; t o r a g e tank; and njection to the reactor vr> n 1
'he CST injections consist of controlled and < pii ch alarls al i
cactor pre <,sure rangtng from 150 psig t.o rat"d, with gorresponding punp discharge pressures throttled between ?50 g psig and 100 psi above rated pressure. During this I (>a r t-o f-- the- tes t i ng , proper- 6p[eac tor of the ; y s r em 'J i 1 1 - be - eration yerified and adjustments made as required to meet this criteria. cold quick start and 2 hour < continuous operation wil1 e demonstrated. A coldquickof start requires a minimum of days of no RCIC operation. The reactor ve.nel~ injection Eil_1 consist of a cold quick < tart of the synt.om with'all blow-ro u ted-Lo --t he-rea c tor-ve]s sel-a t--2 0 % --to-- 2 51- powe r . ~'u ACCEPTAtJCE CRITURIA Level 1
- a. The average pump discharge flow must be equal to or greater than the .1007, rated value after 30 seconds have elapsed from initiation on auto starts at any reactor pressure between 350 psig l and rated. I -
- b. h'ilh pump d i scha rge a t any pressure between 250 psig and 100 psig above rated reactor pressure, the required fIow is 600 gpm. (The 100 psig is a conservatively fiiqh value for 1ino l on';es ; the measured value may be used jf available )
- c. The RCIC turbine cha)1 not t. rip on overnpeed during auto or manual <; t a r t : .
1f any Level I criteria are not mot, the reactor ,c i 1 I be a l l o .v e d to operate only up to a re ;L r icted power levej, until the problem is colved. _Le v e l_ .2- - - - C. O I s/J Ju-2 f p a. Tho Ln:bine g l .a n d ;oal co: de n. . . r n) .; t . m N 11 he eq.ablo rif p:: eve n t i ng i t e. m 1. a k . i q .
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<si1% co2d RCIC Iwloare.(co9d" l,c.,v; dfi.,cd a, a m in i, g a cec day s ,,, n d n y un a cf Rca c op e ra l--;c n ) T!do c.RiI. anal ve .,s cA y : c f/c.e.
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to onane deacly g/tl'e con ks & e aclj u s %,i] s . 0 Yer % 9 4 con Iro&1- s e lhny s 't re- <le le n > < ,w < hao CST i nj e ch on s ct re done w f4 from cold' kc tc c n di k ,s. %e Cs r i>,jecl-;on is donc u; /1,
% ves.,- reac{or vecaeA done p cc m ee. <d~ i s o fu i ad % ofkr u; f7, m vess el pcc c sa ve et/- t cJi <l.
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-for (u:Fuee c eveif % ce fe afing .
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hurhv3 w asch .njec'N on s zu rco cker- s[co.n.,, adm; s s ica valu cs q ja .,n a in c l. {cebalu h;r a li >1e s co e cko5cd <Me m e<:r il m re a cI< r N, c r f h aLa % ,. - ; & c e c.avva t!,,~ th . co.c- %0
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[ ut: stance rit:rino .or l.uiR is unrniat"f to
- c.e 3c:yal tast a The ac<.ep t inco cri' Orion should oe codified such t. h a t U;e ; JR deter rained f ren actual test results will be congared to a predicted MCPR for the actual tes t case.
Response
~he iuss of fee?" ate-
- eating test has a not f'. u:< < r i t e r o n that is adjus'ed to 'ho feec.ater tempera'ure. i .i tial'~'.is reac + 0r n~rior and 'he ac tual cinrse in critorian is a directl.. m aaurablo vnna
. es are the pararotera used to adjust it to actu,1 reactor etnditicns.
The simulation input parameters along with the necessary parametric information will be included in the Transient Sa fe ty Ana lysis Des iijn y Report (MPL-A4-5010). T h . c-+w t i s i ci. c a s .. . o . ...u ' . '. C . hy-
~~ ge t-i g. , m.,a e o ,ny ...fI , s irne a rTEDTEfUrrscracna t- ~rr*"+We I' >
.s Jua-<st ,.,de pea k hea t f l ux ' ' a. . ... as a function of ini tial pov.er, initial dcme pressure, and the feedwater temperature drop. The predicted heat flux f rise is then determined by applying the initial power, initial ccme pressure, and actual temperature drop correction factors to the base case calculation. There would be no ..cro.orent in the w. orstandin" of the +est rosul's b:. ad: ling a criterion t hat is based on a correla ion of roactor pararo ers. Such an 2pproach o'lld .juct cloud the basis of the criterian and onnfuc2 ' its interpretaticn b;- inserting a correlation in nlace of 'est icte. (It snould also ce noted that each core location sceu] d hwe a di f foren' bogin" ins and end coint). It scarcely needs to be sta ted tha t an o 11 : a *. i on of the tharml limi t s 'refore and af ter the test s ould be adequa te for
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.s to be justifi>d and related to a cident analysis a .- at ;ns . ine acceptanca criteria for the test shculd also be proviced, as requested, along with the roquired deurce of onvergence be t..een pn:dic ted results and actual results and the bases given for the degree of convergence.
Response
The proposed noeter level has been chanced to > 9%. See the modified Table 14.2-122. 1',ltg., c/ & & xw q f.ls MSIV cWub ww . ^ Ao e~La L{% TL Lu 6" n%M- h L ' iA A Ao AL h Eu.uat1 Ag mL Av & A .A L q~ EfL c693 w. t e 7= T3o + b U f) 3 . T= mss u &no 76 . Y,o = r, ~z L oo % dnl . n,a(3 e- t 1, G u J
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events, 3 ) the r m.m,t area cri f.erion referencinc the flow cono u un curve Tecific i in Fi re '.4.P '7 does not fl. ast. d m - w ours.- - s 'i n i rr ,t n=n inortie %m: to ho n: rec ' tibl .ti th nnal; sis . .'. ) turbine e'zersp yd for the goneratnr ) unea for LOA of the plant's electries'. cjstem for both turbine t. riptrip and and 5 generatur
) respanne load re.iect even+.s a?sc.
the hard::are v:here vors' ca:e 'he abst ract should be rnodi fied to identif; design or technient soecifica tion values viill provided. be used in predicting the outcome of t.he test and jus tification Respons . The renerator output ci rcuit breakers. ] cad reject till be initiatod by cr,ening tha genorator
"'his v:ill result in the instantaneous discennection possible overspeed of the Enerator from its load and veill cause the v/orst cenci'.icn.
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- rip j,e:;t, p redi y { iOr.r a "e a r:cw t.eri: eJ i n in i t.l al joc.ec, initial do .. p r. , m re , scram d e l e,; o , and ce; m r. ie tivi t., i n t," r t i o n ra t.e . ei n- 3 iform.d i."n La-
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' :'l A R 't ! ! P 'l L ST 111H POS E To denonstrate t)m response of 1he reactor an<1 i t '; control f systems to protect-ive trips in' the turbine a nd ge ne ra tor . I i
_D_l SCR I PTIOli- 4 The turbine stop valves are tripped at select ed reactor power I levels. The main generator load rejection will be initiated in such a way that a load in.ba lance tri p occ urs. Several re actor and turbine operating purameters will be monitored to evaluate the restonne of the bypass valves, relief valves, ii P S , and the reactor recirculation system during the subjec t transients. Add it iona lly , reac tor pressure, simulated heat flux, and reactor power will be monitored to determine the peak values of each of these par ameters during both the turbine and ge ne ra tor trips. The ability to ride through a load re jection within the bypass capacity without a scram will be demonstrated. i l ACCEPTA!;CE CRITERIA l t Level 1 i i m e . 4 k'\ qb I i ~/ b
\ , - 5 \ .
N ., ' ( L tC S--F M R jgg ..g y y;, ) .g 9 ln %, 3 ' 5 7 Y c3> . \ ; e A
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T. 41 1-: 14,2 1/4
,,j
/y .,/y (C< mi. ' d)
;3 , n eiha: o r.
pt r;a ;a . t 1: 3 3 ej s ;in - t ; u t. .nt i _l o d i : cj of the ; L ea n l ii,e following 1. h o _ e transients,
- b. The t .s. o 1 oc i reula t ion pu:ap dr tre f]ou coad down t r a n t, i e n t: during the first 3 ;oconds :1o nt he equal i to or faster thiui tha t specified in Figure 11.2-7. I fMy '
.;rt15 E I g, g. g. Tic positive change in vessel dome pressure occurring l within 30 seconds after either generater or turbine ;
trip nust not exceed the Level 2 criteria 1y more t . . than 25 psi. T he positive change in simulated heat flux shall not exceed the Level 2 criteria by more than 2A of rated I value. l g , g ,g, For turbine and generator trips there should Le a l delay of less than 0.1 second following t he beginning of control or stop valve closure before the beginning l of bypass va lve opening. The hypass valves should be opened to a point corresponding t o ap[irox i mat ely 80 % of their capacity within an additional 0. 2 se con d (or an elasped total of 0.3 second iron the 1oginning of turbine control valve or turbine stop valve closure motion). Level 2 N' 'Ldwo ,(. opud c/ce.) neb mab A/u pA a& w Yu My 4 , .j4 - om ,
- a. The MSIV's sha ll no'_ he tripped closed at any time during the test transients. _- .
- b. The positive change in vessel dome pressure and in s i mu la ted heat flux which occurs within the f ir st '30 seconds a f ter the initiation of either gener a tor or turbine trip must not exceed the predicted values.
Predicted values will t. e r e f er enced to actual tost condit i cno of initi al power level and dome pressure and will uno BOL (beginning of lite) nuclear data. Worst caso d-;i gn o r t echn i ca 1 sj iec i f i ca ti on values o f all hardware performance shall !)e unod in the predjetion, with the o>.ceptjon of control rod insertion tim and the del ay f r om hoginning of turbine control valve or stop val v e m o t.i o n to the gene ra ti on of the scr nu signal. The tiredjeted pr es Sure alld hea t flux kill 1)e co r r ec t, r t ] for ihe act na 1 r.ca urod va 1 ne r. of t bese t wo pa j .nno1 o r s. C af c- t%t.lf .L ,1 fi[o c uv w d <; A. if M x, Q,' O
; in. .
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. Ai 2 i)fEa, t e,a., <4j 3,M N = ;
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" !L' V. i t h ' < l'11 i on Of I ": trCUl'!JnD J,l ':: 3, .I e t >
,! G Id e DD I.rol V l'cr /,) EUct.pl'. Isle criteria for t r: . :fe: f ro:' CDP, to 15 ;i:-; toc, >r suanl i o ,
"hou!d be catahli:Jel in ter:r, cf allce.'ible r2nge of put.> >nead c: hen the transfor Occurs.
Ilesjonse:
- 1. Plant instru;nentation (,ic t pu'no and recire. pump d?) ill 'a clorely nonitored during the pac;er reduction described in revised Table 1/.2-127.
4 n@med-i n-/-n-n Flo c. :ont rol valve and jet pump cavit, tion veill sho'c. up es a significant increase in the noise ievel cf the jet pnno dP instru'nentation and recire pucp cavltation ciill result in similar observations on recire pu p dP instrumena tion. br u su~u'1s c {r Tablu 14 2 ~ 1**l 7-
- 2. W+es&:mgeMitg-dn. e .; p-inerth c . - ottl-ltingmtudicd.
3 See revised Table 1/. 2-127. fre'. Krd u. f, n . u.
- 4. The purpose of the 60!in to 15!!n power supply trar nfer la 10 reduce patter and f1cc. before 'J'e reactor enters the cavitatien region of the power-fluct nap. The cuccessful drmast ration tha t eavi tation doe., not occure prior to t his transfe: ensures the ndequac. of the transfer se tpoin t. Fu r t.h e r , the parer supply transfer set .ointa are verifed te be in accordance cith derign during the feod" :ter cont rol crearere'ie:nl test. l'o additional acceptance criteria are necessary.
J
\.
t c I 4
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LFCS-FSAR Anl:mm.UT 4 4 ti'd W M 1979
'\
i TA n i,F 14,2-1/7 n FC19 UI T r J un r> ys i' t, S t /m m P e q PU; dose The purloces of this te;t are ( 1) to obtain recirculation system retformance data under di f forent operai iona 1 condi t tom >, such ar pun.p trip, f1cw coastdown, pu:rp resta rt and flow induced vi b ra tion ; (2) to verif y that no recirculation system cav ita tion will occur in the ope ra ble regicn of the power-f309 .na p ; (3) to ve ri f y th a t , during the trip of one recirculation pump, the feedwater control systerr can satisf ac torily control water level without a result.ing t u r l?i ne t rip /sc ra m; and (4 ) to record and verif y acceptable perf ormance of recirculation two pump circuit trip system. DESCIUPTION The reactor coolant recirculation system consists of the reactor vessel and two piping loops. Each loop contains a constant speed centrifugal recirculation pump, a flow control valve, and two i sola tion val ves located in the drywell and ten jet pumps in parallel, situated in the reactcr dcwnecme r region. Each r ecirc ula ti on pump takes suction from the reactor downcomer and f discharges through a tra ni f old system tc the nozzles of the ten jet pumps. Here the flow is agumented 1y suction flow from the downcomer and delivered to the reactor inlet plenum. Rec irc ula ti on pump trips have several different effects upon the reactor. In case of higher Ecwer turbine or generator t ri p s , . there is power supply. an automatic opening of circuit breakers in the tump The result is a f a st core f low coa stdown that helps reduce peak neutron and heat flux in such events. This two t r i p, pump-trip test, initiated by a simula ted turbine generator verifies that this flow coastdown is satisf actor y prior to the high power turbine / generator trip tests and subsequen t operation. - A poten t i al threat to plant availability is the high wa ter level turbine trip scram caused by the level upswell that results after an une qected reeirco jat ion one t ump trip. The change in core flow and the res ul t.a n t power decrease causes void f or ma t ion which the level sensing systerr. senses a s a rise in water level. The cne pump tests, initiated by opening the drive motor brea k from the control rocm, are to prove that the wa ter level wi l l not rise , enough to threaten a high level trip of the main turbine or the feedwater pumps. { I Hoth the jet rumps and the recircul at ion pumpo till cavitate at i q ccnditicas o f hjob i 1 < .: m .d Itw po ,w r
- bare NP!m de. :nds are high
\ a nd li tt l e f ee'h 4 t er u.'< u i22; 3 r a:u rs , { in <<er, the iecircu20 tion f loc will h M ical]y run l a,c k n Jn .. .i ng a
,( j r-i {
Ds l i n. c .s w~ch l i a >
. - - . - . , . . . . ~ .
. . :ur ! ., n a.
!!A R Cl! 1979
% i..!.M l el 2 I?/ (('< n L ' d )
s
- c. ,:c. a n o in r- . ;t.:r 11 '
., t.o 1 ( c. zi r i h. - i<a.tr>r i , .t ih > ro b ek is ac., pl i:;h d lay a t r.o. ;I e r rit i . .. i i s:u l e t i on pu ..p o f. 3r miso l i . : . f t o:a 60 11 a t.o 15 IIz . T1. 0 i~a :c i ;uu m recirculation f l e c.1 is l i; ail ed by appropr ia t e s t o; s whir h zi11 run back the reeirealation flow a.ay from Ihe p m'.ib)e caviLation region. It will.be verified that t h.mo 1imito are sufficient to prevent opera tion where recircolation pump or jet pump cavitation occurs by reducing core po*.;e r v ia con t ro l rod insertion on the rod control 1ine expected to givu 100?, core flow at approximately 501 power until recirculation system runback occurs or cavitation is observed.
ACCEPTA11CC CRITERI A
~-
I,ev e l 1
%The t.wo pump drive flow coastdown transient during the first 3 seconds 1.iust be equal to or f ast.or than that specified l '
in, Figure 14.2-7. _ , , , . _Ife v 61 2' - .
-r-
- a. The water level, IsPRM and transients of simulated heat flux, pressure drive and core flow chall not. l cym2, Xbe, si g n i iH ca n t l y- dev ia te-fhom-th a t p r ed i c ted in the Control System De<;ign Report for the one pum Lr i p .~( TSr G Jw( li
/!e ~c~)y/L p..f,R p&h %;> ~b /wp.~.-,n/
%.wl ~ mv vf ~l;D ., ) .
- b. The reactor water level margin to avoid a high -
level trip shall be >_3.0 inches during the one l pump trip.
- c. The simulated heat flux margin t.o avoid a scram will be >l 5.0's during the one pump trip.
- d. The recirculation system setback runback feature shall be adjusted such that. a flow runbach
$ (transfor of recirculaLion pump ['o te r supi>l j os i from 60 llz to 15 Hz) occurs prior to any observable bgc cavit.alion in the. recirculation system. ,,
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- 1. Conntruction t m; t s and pr r: ope ra t i ona l tests are co-'.:p l e l. e and appruved. 5)
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- 3. The reactor protectica :y s t e:n is operat.ional.
- 4. All ;yste:ac are operational to t.h e extent required to conduct tests.
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- 1. Dn ri n.c, the Icss -f offsite percer test the plant
. ...11 rem in isolnted fror the grid until control of vessel level and : eznre have been demonstrated r fer 30 ti:.;tes, v;hiche fer is crec er (Control of vr:. sel - .
i level and pressee is defined as vessel level cair sined by operat.cr j action above -he automati: initiation setpoint of -% ECCS systems end pressure cairtsined by cperator action belo 7 the ir est relief valce cetpoln- . Durin:a us t;.me, non-class ,u- ._ r e :s nr. . e s uo. n. ,. i t a. > i by the onsite excrgene;.' pc cer systen as requi red s ubject to the aveii'.Dili .y > of diesel generator capacity. : I Data taking fcr this tes', 7till begin vihen the uni- Is isolated frca the grid and vrill contin :e until the plant Iceds sre returned to the gria.
- 2. Se performa m cf the ~ site emergency po;;er syc: =- ireludinc bus transfers, diesel ster s 2nd lead sheddin,o and se-.= 'ing is checked i" +.he Emerger rc Pnv;er Eedrdency Preoperationsl ~ =r- (See Ta' le 1/..f - )
during sinulat+d less of effsite pov;er and LOfA res .s. The denorstraticr of ecepliance :I? the second Level l critericn cf : evised Table 2 4.2-1:'?. included in A . 44, provides additional assurance rf ratisfactor; ever711 _ perforcance of -he onsite energency poveer systen r.d diesel generators. ' 3 3e cost probsle cause cf the reactor scram is ;';e less of AC porter
,o the 3cak de:ectien ryster. This vii]l cause e 1 sir steen line isola'i~
- hich, in tur .. ..ill c.suse s scram on ?'SIV's < 9" ::en. These events are expected i: recur ;i 'nin the first several serrnis foller;ine the initiation of -he tnst Since this test ir basicl;. s re-ri fication of cverall electriesl dis'rit ; tion rystem perferrer c sni reactor tr-ncio
- t erfe r nne e .. i-" ' -i'-i nu of supporting :.,. te.r. -"" "ite in quertic' tes .o sirnifiere in sc:: pl.ishing; the . . r c Je'::~es. .nerefore.
.o acceptance criterio Ms tcen est ablishmi fe" " _ s t s r t.nei er.
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l HIPronE i t The purpose of tbis Fest is to mt.'sure too stabilit y of t bo core - power-void dynami c rer.ponne and to dononutra te that its bc havi or l is within speci f ied limits. I I i i DESCRTPTICU
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The c ore powc.r--voi d mod e that results f ro:n a combinaijon of the i neutron kinetics and core thermal-hydraulics is least sta ble nr e:r I the natural circula tion end of the 100% power rod line. Stability at this point will be investigated by ma r,i ng ra pid i reactivit y changes using control rod notching a nd (:r e s s u r e regulator manipulatien (si mu l a t e d regulator f a ilure to b ick up) . Notchinq even a ntrong control rod is expected t o produce only small effects while, the pressure regulator should p r od n e r- more signi fica nt disturbances. The pressure regulator ma r. 2 p u l a t i on will be done as pa r t cf the preneure regulator start u p test, but recording and analysis requirements o f bot h thi s tost and the f pressure regulator startup test must he accot.iplished for satis factor y completion of thin tcst. ACCEPTAUCE CPITFRIA ( Level 1 ' The decay rat io must be less than 1.0 for ea ch proc ?ns va riable that exhibits oscillatory response, ~~>M cc ' _, s Level 2
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'l hST OPJ ECT IV ES
- 1. Demonst rat e the reliabilit.y of the standby diesels and liW's diesel and a-c power source.
- 2. Provide auuurance that the diesels are capable of providing emergency electrical power during normal and simulated accident condi bions.
3 Diesel testing will demonstrate the proper startup operation by simulating a_ll automatic start signals , and demonstrate that the diesel-generator unit can i start automatically and attain the design voltage and speed within the acceptable time and limitations. 4 The diesel generators will be tested to demonstrate the full load carrying capability for an interval of 24 hours of which 22 hours will be at full load (2600 kW) and 2 hours at the 2 hour rating (2860 kW). (3 5 The diesel shall be load-shed tested at full load as well I as the largest single load to verify the diesel voltage and speed requirements are within acceptable limitations ! and overspeed limits are not exceeded. t
- o. The diesel generators will be tested to demonstrate their - i functional capability at full load temperature conditions 'l' by rerunning the tests in item g3above immediately follow-ing item 7 above.
'I 7 The diesel generator testing will demonstrate the ability to (a) synchronize the diesel-generator u.'"t with offsite power while the diesel generator is connected to the emergency load, (b) transfer the load to offsite power, (c) isolate the diesel- genera tor unit , and (d) restore the diesel generator to standby status.
S. The test will demonstrate the capability of the diesel-generator unit to supply emergency power within the required time and tha t this wi]l not be impaired during periodic testing. I yg e oM a
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. m ,_ - 1 U A R A11 E N t i:11.N T 44
!!ARCli 1979 1
i 1 T. % M 14.2-34 (Cr,n t. ' d ) ! i
- 9. The t. ,t i11 d. on :tre le ihat. til 1 die 01 . ni r :l o r:, wi11 f u:o r t ion i nd. pen.!on t-ly .c he n itaried nimu11m..,,m: 1y. l This lost will be performed once during preopm a tional testing to help identi.fy certain con mon f a i l u re . nodes undetected in a single diesel generator unit test.
10 The start, stop, and protective circuit logic, the air receiver changing times, and system redundancy will be verified for the diesel-generator air starting s ys t.em . The test procedure will also verify that the diesel-generator a i. r starting system provides the proper number
. of starts on each system.
11 Load rejection tests shall be conducted t.o assure continuous diesel operation during accident or loss of normal power conditions.
- 12. Testing of droop settings shall be conducted en standby diesel generators. Droop settings are to be made during this test. Verification shall show the diesel generators operability to carry full load at the drcop settings.
- 13. The required reliabili ty of each diesel generator unit
- h will be demonstrated ,iith a minimum of 23 consecutive valid tests with no failures (valid tests and failures as defined by Regulatory Guide 1.108 Section C.2.e).
*0ther than the Unit 1, HPCS die-el generator whose reliability was Test.
established et LSCS via the Vendor's Qaalification Prototype SYSTEM ItJITIAL cot;DITIOt:S At3D _ PREREOUISITES
- 1. All construction tests are completed and approved.
- 2. All instrument calibration sheets are completed and approved.
- 3. The f ollowing rystems and/or components are available:
- a. diesel fuel oil day tanks,
- b. diesel fuel transfer system,
- c. Iire protection system in DG room,
- d. diesel ventilation system,
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sA FET/ IP i CA trl'l'O NS 1. Verify ihat al l mechani ca.l. sys tems a re ope ra t iona l. 2. Verif y t ha t alltosafety the equipment and construction tags a re removed f ront be operated.
- 3. Prior to precautions the operation of the diesel vecify that all vendor have been satisfied. '
4. Established electrical saf ety procedures should be adhered to when working with energized equipn.ent. I TEST PROCEDURE
- 1. The test procedure will assure that the system will op^ rate according to design.
- 2. All h diesel starting and trips will be tested to assure proper opera tion.
3. All diesel generator auxiliary systems will be tested to demonstrate that they operate within design limits. !
- 4. Demonstrate the manual and automatic opccation of the diesel generator.
_ i 5. All interlocks, with design. controls, and alarms operate in accordance -i 1
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ACCEPTANCE CRITERIA
- 1. Acceptance criteria will ensure that system coistruction and operation wi]l meel design.
2. All auxiliary systemt f un c t ic>n in accordance with design. No t.e : diesel as This preoperationr1 t.e s t . . l so a pp li ot, to the lil'CS an c:nergency pr wer genera tor f o r- the H I'C S pump
. except where specificall: excluded.
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