ML20024G174

Amend 14 to License DPR-22,authorizing Operation of Facility W/Addl 8 X 8 Fuel Assemblies

ML20024G174
Person / Time
Site:	Monticello
Issue date:	10/30/1975
From:	Goller K Office of Nuclear Reactor Regulation
To:
Shared Package
ML20024G171	List: ML20024G170 ML20024G174
References
NUDOCS 9102070639
Download: ML20024G174 (42)

Text

{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ l Ia t Of W l' 1 .mTmgAm PM CNNY , ym,- DOCKIN HO. 50-263 FK)NTICEL1D NUCLEAR GENERATING STATION ~ AMENtMINT TO PROVIMONAL OMBtATING LICRMSF A % t No. 14 i .f dg;g. Liesnee IIs. DPR-22; 1. The Nuclever Regulatory Connaission (the Cosmaission) has' fouad-that A. The applications for amendment by t!.e Northern Statee Power Company (the. licensee) dated March 12, 1975 and August 4, 1975, along with supportive filings dated August 20, 1974,, July 9, 1PFS fJair it.and July 24, 1975, sad September 16, 1975, comply with the, standards and sequipements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulatione set forth in 10 CFR Chapter I; b. 'he facility will operate in confotinity with the application, the provisions of the Act, and the rules and regulations of %e Comis sit. i; C. There is reasonable assurance (i) that tha activities authorized by this amandment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in cospliance with the. Cessnission's~ regulations; and D. 'Ihe issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the

public, i

2. Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to titin license amendment nr.J Paragraph 3.B of Facility License No. DPR-22 is hereby araended to read as follows: l 9102070639 751030 PDR ADOCK 0500 3 p 'l

i V 5: 2 1 " 11 Technicol Specifient ions The Technical Specifications contained in Appendix l A, as revised, are hereby incorporated in the license. The licensee sh:.1] operats the facility in accordance q eit h the Technien! Specifications, as revised by issucJ char.gcs t hereto through Change 1;o. 2.1." 3. This license :mendment is effective as of the dat e of its issuance. Fol! Tile NUCLEAR REGULAT0nY CO:,NISSJ0:: l Er.r1 R. Goller, Assistant Director for Operating Henetors Division of Reactor Licensing

Attachment:

Chanpc Isa. 22 to the 'ic (Lnical Speci ficat jons Imte of 1ssunnee: 0GT 3 01975 l p ,7.,.. . n 8 w

( h ATTAClfithi 1U LICENSE AMENDMENT NO. 14 CHANGE NO. 221D THE TI!CHNICAL SPECIFICATIONS PROVISIONAL OPTRATING LICENSE NO. DPR-22 DOCKET No. 50-263 Replace the existing pages of the Technical Specifications listod below with the attached revised pages bearing the same numbers, except as otherwise noted. Changed areas on these pages are shown by marginal lines: V 2 6 7 4, g +x,, 10 13 through 22 inclusive 2; /,ddition 85 I D* 10a 103A 10Sn Del et e.' 10SC Deletc.i 113 113A Deleted 113B Deleted 114 189B 189C 189D 189E 189F through M inclusive Addition

• 189N, 0, P and Q
• These four pages are being reissued solely to change Section from 3.11 to 3.12.

They ve:e iireviously issued as pai'en 189D, C, D and T. i .. ges l

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D. _Irrmdinte - Ir cdiate r cans thnt the reJutred acti m c!T1 be initirted as seen er. prceticable co -iderirg the safe operatic, of the un't end tho ? m 'snee cf tb r Iuired cctien. th-Micction of a sir Ilated. E. 7. - t r me e punctien I Tc-t - An inst-t~. iunctit -: mear-r f pal into th ^ prin.,ry ne,sor te verify ac prcr-H ir ent channei rcrpense, alarm. end/cr iW t i r t irm acti n. F. Inntr" rent Coli' rntien - An in-trrnent enlibretic- ~r-' the cd'ust ent of an instrurert signal ac :uracy, and response ti--e to a kncwn ercut so that it correcponds, within nccopin';7 e n~c m^ 't Calibrrtien chall ence pass the entire (s) of the parcneter unich the Instr - v"7 i e rument incirdi: g actu,tica, c l a r ra er t ip. e' tine is nec.. art of the routine instru-ent cM ibration bst vil l be check"d ence per cycl.:. G. Liriti y Conditions for Onor-tien (LCO) - The li-i'irr cerditienn for enerntien specify the mini _r2 accertable icvc'e cf system pr_rfc ance r"ccs cry w m ure safe st artrp cnd operatien of the feel 11ty. Uhcr these conditier, are net, tie pic"* cc b' eperated sciely and ab omal situations car be scfely controlled. H. Li-itinr: ';afety Syste-Setting (TS9S) - The II--iti-s-fety systen rettings are settings en instru-mertation which initiate the sete,atic Prctective

tici st'a Icvc1 sech that the safety limits will net be exceeded.

The region between the safety lir'.t end these setting, represents cargin with nor 21 eperation lying below these settings. The ,rgin *-n been established so that with preper eperatien of the instrirnatation, the safety linits vill nev"r be exceeded. I. i.vir-na Crit tent P"ner Rat io ("CnPd - The mini:mm critical pcuer ratio is the value of critical p c.ac t ratio associated with the nest liniting asr-sly in the reactor core. Critical pcver ratio (CPR) is the ratio of that peuer in a fuel amenbi: which is calculated.by the GEXL i correlation to cause sont point in the necmbly to experience boiling transition to the actual aescrbly cperating pcuer. J. Mode - The reactor node is that which is established by the nede-scIceter switch. .K. Opernble - A sy-ten er cc ponent shall be censidered cperebic when it is capabic of perferning its intended function in its required nanner. L. Operating - Operating ecans that a syste er cc pc~nt is perfor-ing its required functions in its req" ired ranner.

1 One ~ tin Cvele - Interval between the c~1 ef ene r-reefin~ cutage end the end of the next subs e<;uent rcTreling mtap.

2 1.0

2.0 SAFETY LIMITS LIIIITII!G SAFETY SYSTE!! SculliG5 2.1 FUEL r:ADDI:70 I!UCCRIT'( 2.2 TL CIADDI?;G I!Tr'MITY Ap lic.abilitv: Aa*'licability: Applies to the interrelated varichics Am lins to trip settings of the instruments sud associated with fuel therraal behavier. de--ices which nre provided t6 prevent the re:cter systea safety linits fren being execeded. Objective: OE actEve: To establich limits beJ c-: which the Te define the level of the process variables intes: i.ty of the fuel cladding is preserved. at which autenctic protective action is initi:ted to prevent the safety litatts fren be t.n: exceeded. Specifiention: M cification: A. Cere Titermal P<ner Limit (Reactor The limiting safety system settings shall be as Pressure '> 800 Psi n and Core Flcv is specified belcw: > 10/. of Rated) A.

cutron Flux Scram

. When the reactor pressure 1; > 800 Fsia 22 crd core fic e Ls > 107, of rated, the 1. APR'I -- The APRM flux scram trip setting existence of a dnin2:a critical pcwer shall be as shown in Figure 2.3.1 unless rn':io (!"CPR) less than 1.06 shall co".- the ec-bination of pcver and peak heat stitute lolation cf the fue? cladding flux is above the applicable curve in Fi:ure it*.cgrity safety limit. 2.3.2. Uiten the ceabination c? pc.cr m'i peck heat flu : is above the curve in Figure .'.3.2, 22 l the scr t setting (S) is given by: 6 2.1/2.3

t s 4 j. t I r 2.0 SATETY LIMITS LI:'ITI!:G S h a: SYSTE?! SETTINGS I t S = 486,000 P (7x7 fuel) [ l ^ 3 S = 1[3,000 P (8x3 fuel) E. Core Then:ral Pcwer Limit (Ecactor X Pre-cure !? 800 Psia or Core Where: Flev :5107. of Rated) P = percent of rated cover X = pcd heat flux - (LTU/I'R/m) i When the reactor pressure is :s 800 Pria shall be used. t-cr core flew is < 107. of rated, the l 2. IRM--Flux Scram setting shall be E 271 core thernal pcwer shall not exceed 237, of rated' thermal pc' cr. r tcd neutron flux. e ( 4 i r l 22 AFRM Rod Eleck - The APRM rod block setting I 3. chall be as shewn in Figure 2.3.1 unicss the conbination of power and peck heat flux is [ C. Fever Transients above the applicable curve in Figure 2.3.2. When the combination of pewer and peak flux is To insure that the safetv lim't established absv the curve in Figure 2.3.2, the rod block trip 2', in Specification 2.1.A is not exceeded, s tting (RE) is g ven by: each required scram shall be initiated by its primary source signal as indicated by EB = 437.400 F (7x7 fuel) L ~ the plant process corputer. y" i RB = 382,400 P (8x8 fuel) ~^ t 4 where: P = percent of rated power 2 X

• peak heat flux (BrU/IIR/FT )

I shc11 be used. t 2.1/2.3 C. Renctor Lov Ucter Level Scram setting shall be 21 10'6 ebeve the top of the active fuel. l 1 6 y - - ' - - - - - ~ ~ ~ - - - ' - - - - ~ -

LIIIITI?T; SAFETY SYS'E2f SETTING 2.0 SA" m M:2TS 1 Renctor Em Lew Water Level ECCS initiction D. Eccctor Water Irrel (Shutdeva Cen( tien) chall be > 6'6" < 6'10" nbove the tcp of Wher ver the recctor is in the shutdct.m th~ cetive fuel. conf ~ tic t with irradiated fuel in ti r rcccter t i 4 ves-31, the voter Iccci shall not be Ic s thn" that correspending to 12 inches ab ~7c the top ;f the active fuel witen it is roated in the core. This level shall be con-tin m ml, nonitored.*herever the recirc21ction ur" cr~ ict c;'eret'ng. E. Turbine Control Valve Fast Closurc wrm sh?ll initiate upon loss of pressure at the I recelerotion relay with turbine first stage pressure > 30'*. F. Turbine Ster Valve Scran shall be < 107. valve closure frem full open with turbine first stage prer.sure 2 30'.. G. I'ain Sten--line Isolation Valve Closure i 3crm shall be < 107, valve closure frc : ruil cren. I 1 s 2.1/2.3 l_

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W i l-i !c 1' Barcs Contitra<'d: of trrnsition boiling) and the Safety Li--it -is t'eri' ed frem a detailed statistical analysis consider-Iq all of the uncertainties in =mnitcrirg t're em e crerating stetc.incirding encertainty in the Seding trannition correlation as described in-P. ference 1. The uncertaintics ' enpleved in derivi-g _[ the Safety Limit are previded et the besitmin. of enfr. fuel cycle. [ t i recause the boi-ling transition correlatica is beced on a large gr. ntity of full scale data, there is a very high ccnfidence that eperatien of a fac1 arml:Ty at the MCFR Safety Lirit vculd not predrce i boiling transition. Thus, nitbeugh~it is net requir?d to ertablish the Garcty Linit, additional rargin i cxists between the Safety Limit and the actuni cecurrence of Icsr of cla0 ling integrity. e Ticnever, if boiling transition were to occur, c1c1 perforatien wculd not be expected. Cladding l terperatures vould increacc to opproximstely 1100N chich is belev the perforation te=perature of the cledding. material. This has been verified :by tests in the Gener:-1 Electric Test Eccctor (GETR) where f fuel similar in design to Monticello operatcd abere the belling transition for a significcnt period of l tire (30 minutes) without clad perforatien. ,,a If reactor pressure should ever execed 1400 psia during normal pcrer operating (the limit of applicability of the boiling transitica cerreintien) it would be assu-cd that the iuel cladding integrity Safety Limit has been violated. 1 In addition to the MCPR Safety Limit, cperation is constrained to a maxi:nsa LIIGR of 17.5 kv/ft fer E 7x7 fuel and 13.4 kv/ft for 8xS fuel. At 100". peect this limit is reached with a maxirr.ra total pecking factor of 3.08 for 7x7 fuel or 3.04 for 3x8 fuel. For tbc case of the maximu=t total pea 51ng factor execeding design, operation is pernitted e-.ly at less than 1007, of rated therr'al pcrer and q oniv with rc+ teed APRM scran and rod block settings rs required by specificatiens 2.3.A.1 and 2.3.B. [ i B. Core Thermal Power Limit (P.ecctor Preseure G S00 psi, er Core FIcw s 10% of Rated) At pressure belev S00 psia, the cor clevation pressere drop (0 pewer. O flee) is greater than 4.56 psi. At i tre pcuers and all core fice,. this pr~~rure'di % rc,tial is eninttined in the bypass regice of the I .m. 2.1 TASES 14 [ c e l'

pores Cantinued: 1 Since the pressure drep in the bypsss regic 1 ir essentially all elevation head, the core presr.ure i ~ drep at Icv r wers and cll fic wil' 8 ay, N acerter than 4.5' pni. Analyres s' ace titat with r 3 Nundic flew of 2 %10 lbc/' r. cundic m ~ c~re - cr : nearly inC<rrnient cf bundle power and has n value of 3.5 psi. Therefere, Jue to 9 6.M rm drivirg herd, t:n. bundle fim will be grecter th n 3 2h10 15c/he irrerpective cf total core f i c; em independent of bundle pcect for the range of burdle pererr of cencern. r tll sca!c UIM ts et dva taken at pressures fren 14.7 psia to 800 rnia indicate that the fuel ac m bly critic?1 c cr r "Sx103 Ibr/hr ir r,7rexi:stely 3.35 P:t. "ith the der hn peakira fnctarc thi c r r e-r" c cora t'c c r: o '. p rer of mere then 303. Thus. core ther: 1 p wer lirit e" 2R for - ,c t c - - es belcir FC9 pria or core fIcw loss than 107 n is conservative. C. Ther Transicut Plant safety analyces h ve n' ice that the screms initiated by exceeding safety system settien will assure that the Scfety L M t of 2. I. A or 2.1.B will not be exceeded. Centrol red scrna tiacs and screty system settines cre .ac:--d neriedically to assure that a rcre 'till prcceed as analyzed. As a further check, tic riant precces ct..w tcr will be ured as a fast data-acw inftion systen, when available durir~ a scran, to verify that the screm was initiated by the I prinary source signal. ne ec puter I nor ally avail 61c for this functien. However, it is recognized that the plant ray cperate witheit the computer in servico, in which event the coa-fir atcry data vill not bc.risilabic and the ve-tifiention specified by 2.1.C will not be required. The thermal pmeer transient renulting whet c scr.-- ir acec plished other than by the expected scrc-: ~ signal (e.g., scram frem neutron flux folicwing clor e cf the r.cin turbine stop valves) decs not recesserily cauce fuel damage. For this specifierti-a, ehen a scram is only acccmplished by r:eanc of a backup feature of the plant design, a epecific ulysis is required to determine whether or not a Safety Linit has been violatcd. Tac ccnct pt of net approachinr a Safety Lielt, providing scram r.ignal: are operable, is suppsrted by the exteraive plant safcty analysis. 7. penctor Mate

• Level (Shutdcen Conditic-) Durira periedr when the resctor is situt dcw 1, consideration

-ut also be g;iven to trater level req"ir< :en t s du ' t o the effect of dec.*y heat. If reactor water ievel shculd drop belev t h e t e-- of th" native fr ' d~rirq thi: ti-c, the ability to cool the core is reduced. Thfa reductie, i-cor-coo' carnbil :y :ould lead to clernted cladding temperatures v! cled perferntlen. The cert will N cecied neffi tently to prevent cl.-J elting sheuld the water Tet_1 Le reduced to tue-tb rd :be cer" ' e if t. r,telishment of the safety linit at 12 inches sove the tep of the fuel nrovides cdecente nnrgr. This level will be continuxsly ronitored when-I ever the recirculation purps are not crersting. I 2.1 FA"25

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ym.cc : l 2.3 7hc abnormal operational transients applicabic to operatien of the Menticello tinit have been analyzed threugitout the sp-ctrun of planced eperatin; conditione up to the themipewer Icvel of 1670 Nt. The raalyccs were bened upon plant eperatien in rccerdanc ~ith the cperating rap given in Figure 3-2-3 of the RSAR. n e licensed na.xtran pcrer 1c el 1670 W: rer e crts the r xir-m steady-state pc rer which. hall not hu wingly be excneded. r Censervatism is incerporated in the transient analyses in estimating the controlling factors, such as void reactivity ccefficient, control rod rcrem verth cr.~a delay time, peaking factors, and axial pcer chepen. These factors are selected censcrvativeic with r"rpcet to their effect on the applicabic transient results as determined by the current aralyris r-del. This tranrient ecdel, evolved over many 3 cars. nas been substantiated in cperation en a concervative tool for evaluating reactor dynamic per-fermance. Eccults obtained frc a General Electric boiliS sater reactor have been compared with predictions node by the model. W e ccmpariscus and renilt-are surmarized in Reference 1. 7 7he absolute value of the void reactivity coefficient used in the analysis is conservativcly esti isted to be about 257. greater than the nominal maximam value e gceted to occur during the core lifeti=c. The Doppler reactivity feedback coefficient has censervatively been derated to 907. of the expected value, n e scram worth used has been derated to be equivalent to approximately 807. of the tetal scram worth of the centrol rods. The scram delay time and rate of red insertion assrned by the analyses are conservatively set equal to the longest delay nnd slowest irsertien rate acceptable by Technical Specifications. Ec effect of scram worth, scram delay time and red insertion rate, all censervatively applied, are of greatest significance in the early portion of the negative reactivity inscr ien. The rapid insertion of e regative reactivity is assured by the tir:c requirc=crts for 57. and 207. insertion. The early portion of the scrcm stroke accc plishes the desired ef fect by inserting sufficient negative rcactivity to turn the t ransient around. The times for 507 and 907 irsettien are given to arsure proper cc:piction of the orpected performarce in the earlier portien cf the tr msient, and to establish the ultimate fully shutdesn s:ec0y-state conditien. 7eference (1) 17 2.3 LMES

O s r i t.. 4 t. t i P.asca Continued: i, For a,alyses of the themal censequences of the tra sicnts, the Operating MCPR Limit (T.S.3.11.C) is 9 cencervatively assumed to exist prior to. initiation of the transients. 6 This choice of t' sing conservative values of controlling parameters end initiating transients at the design 4 pover Icyc1, prc<'uces more pessimistic ansvers tha:' ve. tid result by using ex-ected values of centrol j parreters and enalyzing at higher pcuer 1 vels. Deviations frca as-left-settings of setpoints are expected due to in'icrent instrument errer, cperator setting error, drift'of the setpoint, etc. A11ceable denations are assigned to t' c limiting safety system settings for this reason. 1he effect of settings being at their sliceabic deviation extrenc t is ninicial with respect to tnat of.the ccuscrvatis=n discursed obeve. Althcugh the operator will sat the setpoints within the trip settings specified, the actu:-1 values cf the varicus setpoints can vary 1 22 j frem the specified trip setting by the allevabic deviatien. d A violation of this specification is assumed to occur only when a device is knowingly set cuta.ide I of the lintting trip setting or when a sufficient number of devices have been affected by any means such that the autcc, tic function is incapabic of preventing a safety limit from being exceeded while in a reactor' mode in which the specified function rust be cperable. Sections 3.1 cod 3.2 list the reacter modes in which the functions listed abcee are required. r A. Neutron Flux Scra= The average pcwcr range mcnitoring (APRM) system, which is calibrated using beat L balance data taken during steady state conditiers, reads in percent of ratca thermal peaer (1670 fMt). 'hcause fission chc=tbers provide the ba :ic inpur sinnals, the AITJi syste : responds directly to average ncutron flux. During transients, the instantancers rate of heat transfer from the fuci (reacter th2rmal power) is less then the instantenec~s neutren flux due to the tir:e constant Of the fuel. lhorefore, d'tring abnormal cperatienal transients, tbc thermal pcwcr of the fuct will be less than l 2.3 TASES IS

w .o. + e i U .) W l. l }. 4 x q Bases Continued: 1 1 that indicated by the neutron flux.at the scrrt retting. A.nalyses demonstcate that, with a 1207. t j-4 scram trip setting, none of the abnormal operatienal. trantients analyced violate the fuel Safety -Limit and there.is'a substantia ~ margin frc feci de ge. Thereferc, the ute of flow referenced ceram trip provides even additienal margin. l i l An increase in the APRM scran trip setting would decr ease the margin present.before the fuel cladding Intcgrity Safety Linit is reached. The APPCI scrcm trip setting was determined by an analysis of margins required to provide.a reasenabic range for maneuvering during cperation. ' Reducing this cperating mar;;in veuld increase the frev ency of rpuriers scrams which have on cdverse effect en reacter safety because of the resultivg thernal s t r~m e s. Wus, the AFTOI scram trip setting was { selected because it provides adequate nargin for the fuel cladding integrit, Safety Limit yet i a11ews operating margin that reduces the possibility ef unnecessary scrams. Therefore, it is intended. to ultimately replace (with prior I;RC app-eval) the autc aatic flow referenced scrca with i a fixed 120 percent scram setting. 22 The scram trip setting must be adjusted to ensure that the LIIGR transient peak its not increased fer any combination of maximum total peaking factor and recctor core thermal power. The scram setting i l Is ' adjusted in accordance with the formula :in Specification 2.3. A.1,when the maximum total peaking j. factor is greater than design. If the AFRM scram set ting should require a change due to an cbnormal peaking condition, it will be done by increasing the APPdt gain and thus reducing the cIcpe and intercept point of the ficw referenced rcrm curve by the reciprocal of the AP3M gain '~ l change. Analyses of the limiting transients shew thet no scram adjustment is required to assure that the MCPR Safety Limit (T.S.2.1.A) is not execeded when the transient is initiated from the i Operating MCrn Limit (T.S.3.ll.C).. I For operation in the startup mode while the reactor in at Icw pressure, the IRM scram setting of i 207. of rated power provides adequate themal estgtn between the setpoint and the safety limit, 257. T cf rate 1. The rargin is adet;uate to acc<rr clate anticipated maneuvers associated with pcwcr plout ntartup. EIfacts of increaning pressur.- at cere es 'ce void content are miner, cold water frem scurces availabic during startup ir net much celder than that ai rcady in the system, I 2.3 rdSES 19

b i. a ~ ~ ~. t i i i i Baser Continued: i i -l tc perature coefficients are recil, end centrel red patterns are constrained to be uniform by I operating precedures backed up by the rod verth rininiter. Werth of ' individuci reds 'is verv lov f n a u tirena red pattern. Thus. of all possible sc rec of reactivity input. uniform centrol ' rod uitudrawal 17 the most prehble caune of sig 3ific at p acr rise. Because the flux distribution associated with uniform red withdretals does not invo'xc high local peaks. rnd-because several reds be moved to chanc,e pm er by a significant percentc~c of rated pwer, the rate of peacr rise is rmsu very sicv. uenerally, the cert flu

i. in near cruilibritet with the fission rate.

In en assumed untforn rod withdrawal approach to the scram icvel, *.he rate of power rise is no more than 57, of rated power per =inute, and the Ipli s, sten veuld be more than adequate to assure a scran before the pcver could exceed the safety limit. The IRM scrc,rc=ains active until the mede switch is placed in the run position. 'Ihis switch occurs when r~ c.cter pressure is greater than 850 psig. e The analysis to support operation at various po cr and flow relationships has censidered operatica ~ with either one or two recirculation pergs. Du--ing steady-state eperation with one recirculation ~, putry operating the equaliner line chall be open. Analysis of transients from this operating con-dition are Icss severe than the same trnnsients from the two pump operation. The operator will set the AP!GI neutron flux trip setting no greater than that shown in Figure 2.3.1. i!crever, the actual setpoint can be ar rmch as 37. greater than that shown on Figure 2.3.1 for recirculation driving flows ' less than 307. of design cod 27. greater than titat shown for recirculation driving ficas greater than 507, of design due to the deviations discussed on page 18. 4 B. ApR11 Centrol Rod Block Trips Reactor pcver Icvel may be varied'by noving control rods or by i varying the recirculation flew rate. The APaM systen provides a centrol rod block to prevent rod withdrcwal beyond a given point at ccnstant recirculation fica rate, and thus to protect against the condition of a MCFR less than the Safety Linit (T.S.2.1.A). This rod block trip' setting, which is autenatic,11y varic;I with rceirculation Icop ficu rate, prevents an increase in the reactor prver level tn creessive values due to control red withdrawal. Tine fice variable trip setting { provides subr antini rargin frc, frel dmar acming a steady-seate operatir at the trip settin~, ( .r 2.3 BASES k 20

3..

.1-h l .a 't l i I t B ees contimted: i i over the. entire recirculatien flea re: gc. y;~,,,, 7, in o the Safe y.tmit increases as the ficv . o u i -

)ecreases or t e specified trip settim' versur fIcw relationship

therefore the worst casa l'rP a ch ecu d occur during stendy-state operation-is at 1037. of rated thermal pcwer because ok t5c'P [ [ oc trip setting. The cetual pear di-tribution in the core is established by uc fled control red seque-'ces and is monitered cen'in*:cusiv b'v the in-ccre ILDI s tem * ~ .e l tne mar.irnri total peaking f act~-- exc ee d-th". c s .., a tu e, th' e ren block settin~ is adjusted in N. ccordance w;,,-h the formula in Specification 2.3.;. If the APP 24 red block sett ng should re'tuire a c innge.due. to an abnormal peshng cendition, it will Se done by increasing the APFli Uain *nd snus reducing the slope and intercent noint o-. +;c-r r c. referenced rod bicek curve by the i t o j re tiprocal er the AFRU gain change. The operntor will set the APPZ red block trip settirns no greater than that shewn in Figure 2.3.1. Wueve r, the cetual setpoint can be as uch cs 27 grenter then that. :'icwn en Figure 2.3.1 for recirculation driving flows less than 50; of desic,n mnd 27. greater than that shewn for recirculatien I 22 driving fleva greater than 307. of desien due to the deviations dincussed on Page 18. C. P,eretor Im Veter Level Scrnn The reactor Icw water level scrcm is set at a point which will assure 4^ that the water Icvel used in the bases Ier the safety limit is maintained. i The' operator will set the icv vater Icvel trip settirg no lower than 10'6" above the tcp of the active fuel. However, the actual setpeint can be as much as 6 inches lezer due to the devictions discussed on page 18. D. Reactor Lew Lew Unter Level ECCS Initiation Trip Point The crergency core coeling subsyster s are [ l designed to provide' sufficient ccoling to the core to dissipate the energy associated with the Icss j 1 of cool nt accident and to limit fuel cind terperature to well belev the clad nelting tenperature to [ a :mtre t hat core gecmetry re airs i r tnc t and t e i f-'lt nny cled r-rt al-water reacticn to less than 17.. i Oc det p of the ECCS c rpexr.t s t o - e t th c -iterien was de;*endent en three previce :Ly ret [ "w p ar.ric t e r n : the nmirr'n bre. k size, the Icv vata Ic"cl screa se: point, cr,d the ECCS initiation set-l point. To lever the setpoint for initiation ~ th ' CCE ceuld pr* vent the CCCS co ponents fro:t l e l 2.3 FASES 21 l l 1 l

E cnr Continu"d: rectinr. thei-criterion. To raise the ECC5 initiatien setpoirt - ould be in a rafe direction, but it uculd reduce the rarpin ertablished te crevcnt cetn tien of tho FCCS duries no,'al cperatien er during normily enpceted trrr,ients. L e cperater vill set tile J eu lev vater Icre! 7CC? in.tiation trin setting 2 6'6" $ 6'10" above tha l t.op of the cr tive feel. II~.ce te r. tha n turi set ~ir can be c5 as 3 inches Ic,cr than the 6'5" s^tpoint and 3 inches creater t:r-t' c 0*10" netnoint due to ti:c deviatiers di.scussed e- " Tc 18. E. Tyhine Control valve Fast Closyre 9er-The turbine control valve fast cicrure scr:rn is previded to noticipate the r,pic. increc,e in r v eurc cH --utrc:t flux re ulting frea fast c1ccure of the ~3 sequent failure of the bpass. This transient tu'.'ine control valv~n due te a load rej"ctien r-is les severe than the tt&bine step cal.c closure "ith bypass failure an' therefere adequate r: r rin ":ists. 7' F. _ Turbine Stop Votre Scr.'n n e turbin^ stop valve clocure scran trip anticipates the pressure, ncrtron flu:c and heat flux increare thet ceuld racult fr-renid closure of the turbine step valves. With scran trip setting of 5107. of valve clesure frm full open, the recultant increase in surface heat l flux is limited such that ICPR remins 6cvc the Safety Limit (T.5.2.1. A) cven during the worst ene transient thet assumes the turbine bypann is clered. G. IMin Steen Line Isolation '.'cive Clerure Scrm_ En rnin steam live isolation valve closure scran raticipates the pressure and firx trarmientr !u :h c: cur during ren:c1 cr inadvertent isolation closure. With the scre.n set at 102 valvc cIesure there is no increr,e in neutron flux. .!I. "ccctor Coolent Low Precsu ra Initiatec "ain Ste"- Iso!ation Valve Closure The low pressure isolation of the main -tcen lines at 850 psig "ac n r <r I d e u' te -ive prctecti n cgairst rapid reactor depressuricat ion ~1 the resuItin3 rapi1 cc,lde. of t: 2 ve, " '. V "ta:e var t "en of the scr~:' feature which c curs hen t he noin ster. ! ? re isolati - v-i v-nre closed to provide for reactor shutd an so thet Lich prver eturation at Iw reveter r~csure d c: net occur, th w providicq prctection fer the fuct c!redir-int ority safety tirit. Op"ratien of t:m reactor at pre nures I wer than 850 psig requires 2 '" 2.3 FMES

l P.,r." C. on t ' r*> cd : l 1 thnt the reactor node switch be in the stcrtup penitien where prctection of the fuct claridin; fr:cgrity screty linit is pre.-irled by tic Irf t : ich na:: tron flur scrin. Thus, the cchinatien of ,Ir stean li c Jc r prerst rr inciation -r.d Oci '#en alce c?.cLure scr e arsures the avoiIn'.ility

': t! c neutro, cerm protcet'.ca over the ent.re ~ng' of applicability cf the fuel cieddir; integr '.ty r nfety tid.t.

The eperator eill set this nresrure trin at gre-tcr :10, or equal to 350 psig. Ucvever, the actuni ...a trip setting con be as cr.:ch ns 20 psi Icect dec to t!'e devictienn discusced on page 13. Re fe rr:'ce s 1. Linferd, R. E., " Analytical 1:cthods of P! ant ir:nsient Evalustiens fer the ':eneral Electric Boiling 1.' ster nenctor," NGO-10S02, reb.. 1973. i 27A 2.3 FASES

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3.0 LIMTIT:C-CO::DITICIC FOR OFERCICII h.0 CURVEILIJu]E RE~5IIREECS l chutdo.n ch211 be initiatcd im::x:diately ' ~ ~ ad the reactor preccure chall bc riidaced to 150 psig within 24 hour thereafter. F. I eactor Core Isolatien Cooling System (RCIC) F. Curve:llance of Reactor Core Inclation Coolln3 Cynten (ECIC) Surveillance of the RCIC System chall be perfccmed as follovc: 1. ':'c c t ing l. D: cept as spc cified in 3.5.r.2 belov, tia ECIC c.">cter H all be operable whencver j i Item Frequen~r the reactor preccare 10 ater than 150 Fur.:p operability Once/.ont.h pcig anc irre dicted fuel in in the reactor ""' " e1. Ittor operated Once/:nonth v tive operability To be considered opyrchle, the Eu,.C n. cyctem chcal be capable of deliverinC Flow rate test After major pir.p maintenance and LOO gpm tr.to the reactor vencel. every three months Simulated sutohttic Each refueling actuation test cutage (tecting valve ~ orarability) 2. From and after ti:e date that the RCIC cy:- 2. Een it in determined that the ECIC cy:- tem is made or fcand to be inoperable for tem in in' operable, the ITCI cycter chall be demonstrated to be operable ire.ediately any reacon, reactor operation is perniccible and daily therea'ter. ^nly during t':e cucceeding 15 dayc unlecc . :ch cy -;cm _; cccur made cperc ble, providca.

• hat during ach 15 day: c11 active compo-

,ents cf the IOCI cyctem are operable.

a0 ). h f'4. h ~11 .. -. - ~..... D

4 4 I 3.0 LIFITING CONDITION 5 TOR OPER!sTION 4.0 SL"WEILL\\NCE REQ"'REMENTS e i l t .I. Recirculation System I. Recirculation System 1. Except ch specified in 3.5.1.2 belee, whenever b Once per r:onth. when irrlated fuel is in the irradicted fuel is in the reactor, vitii reactor recctor with reactor coolant temperature gree f 0 ceoln it t~,perature greater than 212 F cnd both than 212"F and both reactor recircul.-tion j recctor recirculation purips operating, the pumps operating, the recirculatiott system cross j recircule* ion syste t cross tie vniva interlocks tie velve interlocks shall be demonstrated to rbn11 be aperabic. be operable by verifying that the cro7 tie j valves cannot be opened using the nor-al control j 2. The recirculation :yst m cross tie valve-Inter-switch. l locks may be inoperable if at least one cross tie valve is maintained fully clo:.e'l. 2. Iihen a recirculation system cross tic valve l ' interlock is inoperable, the position of at i least one fully closed cross tie valve shall j 3. Valves in the equalizer piping between the be rc_ corded daily. l recirculat bn loops shall be closed. Reactor v., j operntion with one loop shall be linited to 2* hours. 22 l i ~ l. i 3.5/4.5 - 3033 l, 4 i

~ Ihres Contir,ued L 5: G. Thrc:ncy Cooling Availability The purpose of Specification G in to accure th1t sufficier.t core cooling equipment io available at all It is during refuelin; cutaces that rajor.w ne.nnte is perfcreed and during cuch time that tiers. cll core and cont"inment cooltr3 cutsyctems :~ty b. er or rervice. Crecificatico 3 5.G.3 allows all corv :nd contal:rtcnt coolirc eu' cjcttan to be 1:gr ;1e ::rovided no work is being done which has the potential for dra:ning the reactor veccel. Tru : eve. ' r2 quiring cort cooling are precluded. Specification 3.S.G.4 recognizec that concurrent with contml md drive =ainte:rince during the arrueling outrce, it ney be nececenry to drain the supprecrion c"-ter for taintenance or for the inspection requimd bv 2pecifiention h.~i. A.l. In this citurtier, a rufflcient it.ventory of water in c:ainteined to accure adequate core cooling in the u-J.ikely event of locr of control md drive houcing or inctrument titirble sen1 inter,rity. II. Deleted I. Recirculation System The capacity of the Emergency Core Coolant System is Nsed on the potmtial consequences of a double ended recirculation line break. Such a break involvec 3.9 sq. ft. when the cross tie valves are closed and 5.3 sq. ft. when the crocs tic valves are open, rpecification 3.ll.A is based on an ECCS cvaluatica assuming a break area of 3.9 sq. ft. ;. the limitations o f 3.ll. A do not apply to the larger break area. ~ Therefore, at 1 cast one cross tic valve must remain closed with two pump operation to reduce the potential break area. l l l 113 3.5 BASES

O o i Danes h.9: The testing interval for the core and centai,n :ent cooling systems is based en a quantitative reliability analysis, judr-ent, and practicali ty. The core aelin rystems have not been designed to be fully t? stable during cperation. Ter example. the cer* rpr y f;nal dnirsion valves do not open until r< actor precevre han fal3en to h50 prig; thun, r'urinc cer;tien eren if hidi drivell pressure were sinulatel, the fin tl yalves muld not open. In the cas< of the W autc=ntic initiatien during powcr eperation muld result in pumin,~ cold water Into the re9-tor ;cr 1, which is not desirable. The systens can,be automatically actuated during a refueling outage and this will be done. To incr7ase L' , the availnbi'lity of the indiv dual ec pon<.nts of .V core and containment cooling systens, the co:penents I Alich cake up the system, i.e.. inntrrentation, pmpc. valve operators, etc., are tested more frequently. The. instrumentation will initially te functien.11y tested once per acnth until a trend is established arl thercafter according to Figure h.1 (cee 3ection 3.1/h.1) with an interval not greater than threc conths. The pumps and motor-operated. ealves are %sted each tenth to assure their operability. I'm ce<-hination of a simulated autentic acturtien tent each refueling cycle and nonthly tests of the pumps snd valve operators is deemed to be adeq:nte testJ.ng or these cyctena. With components or subsystens out-of-service, overall core and containment cooling reliability is main-tained by derenstrating the operability of the re aining cooling equipment. The degree of operability to be demonstrated depends on the nature of the ressen for the out-of-service equipment. For routine out-of-rcrvice periods caused by preventative rai t n.nce, etc., the pu=p and valve cperability chrcks will be perforced to deconstrate operability of the remaining components. However, if a failure, design deficiency, etc., caused the out-of-service perio.1, then the deconstration of operability should be thorough enounh to assure that a similar problen does not exist en the recaining components. For exacple, ,t if an out-of-service period were caused by failure of a pu=p to deliver rated capacity due to a design deficiency, the other pumps of this type nicht be subjected to a flow rate test in addition to the operability checks. I C l 4.5 r.u ss

'2

s 3.0 LIM ~ JC t W iTIOSS FCR OF " 7IT "' _.d E N EILI/5C EL'UIEr"E!TS 3.11 u1LCr0R l't!EL ASSDSLIES I4 RFlCTC? FUEL ASSciBLIES I .11 y licS Hit-Ann li c:) S i. li ty % e Limiting Conditions fo r Ope ra tion ne Surveillance Requirements apply to associated with the feel rods apply to ? the patracters which t-onitor the fuel there paramet rc n'iich monitor the fuel f rod operating conditiens. red opcratir' conditions. Objective Objecti_ve We objective of the Limiting Conditione i ne objt 2tive of the Surveil 7.cnce Requir., eats for Operation is to ascure the perfor-ir to rpecify the type and freqtiency of rurvcil-ruince o f the fuel rods. lance to be applied to the fuci roh. 22 y ~~ Specifications Specifications A. Ave rane Plana r Llnea r '! cat Genern-A. Avn.ege Plcnar Linese fica? Genera-tion Rata ( A PtIlf"U tion idte (IJLl!GR) During steady state power operation,

1. De APLI!GR for each type of fuel as a the AFL11rR for each type of fuel as functica of average planar expecurc ' vill a functicn of average planar exposure be determined daily during reactor creration shall not exceed the liniting value at1257. rated themal power.

shown in Figures 3.11-1. If at any t1xe it in determined that the limit-

2. T(acnever the plant technical staff determiner ing value for APLIICR is being exceed-that more frequent surveillance of AFL'IGR cd, action shall be taken innediately is necessary, it shall specify an aupented to restore operation to within the surveillance program commensurate with e

prescribed limits. His occurrence l reactor conditions. shall be reported to the Office of Nuclear reactor Regulation within 30 days of the date of occurrence. 3.11/4.11 a

s ~ 3.0 LT:'IIWG CD:'IIIG3 FOR 01'EPJsIICU 4.0 .".~ 'XEILLECE RI'Qi'1RDENTS D. Lacal LFGn ! D. Local Litcn Durint. cteady rLate petect operatien, the 'in-ar

1. Esc local 73GP. an a function of core hetcht j '

fuct accen51y at any axial Ircation shall net cparatien at > 237. ef rated thermal pevar. hert generatien rate (UIOR) of an-red in r :f 6:11 be enecked daily during reactor cre ed the manin= allowble I' rn ac c:.leC atcd g by t!e following eq"atica:

2. Genever the plant technical staff t'ete-rines

~ that more frequent c.urveillance of lecci uiGR $ LUCR 1-fg. ] fL 't U:R is necessary, it shall specify an P LMGR L _ ( P,/r2x ( LT / max d avrmented surveillance progra i ec:-'ensurate l l' i vi':h reacter conditions. d = Denign UIGR I i = 17.5 ku/ft for 7x7 fuel g - 13.4 kw/ft for 8n8 fuel [A[ = Ma--inun po ecr spiking penalty {P-max = 0.026 for 7x7 fuci = 0.021 for 8x8 fuci LT= Total core length = 12 ft L= Axial position above battem core If at sny time it is deternined that the liniting value cE UICR is being exceeded, cc: 1 m sbc': be taken irrediately to rectore creration to. ith in i , precer Oed limits. 3.11/4.11 Iti C !I

m J.C LI: '.*U'; CC'DT'T-:~;5 F9.", C W 2. \\T " " i

6. r* c 7,*ji:n g j + ; 7 pyng;g:TT C.

MI" t- *m Crit icni Pmeer IMiid'CPP)- C. Mir i~ rr Cri tica l Pover P.tio Q< iCP"_'1 l Du rine; sten:'y stat' ,~ecr ep r,tien. I 3

1. ?CPR c'n11 be c' ecked daily be dur!"g

ictor pever cperatien at ti. 0: era t i r n '5PR

,.1.~1'- 2 1.47 Car F-9 f rel and 21.13 Cor I ( r? 2: ~'- r.7 :cd thot -a l pcvar. 7 7 fuel at nite 1 pr w r o m] O e. I :: ; l Fer c< e fican ether th7e r: h - the O. Gene f r r the pl"nt technient ctaff dete nines 22 I C'e rc t ing !".~i? TJ it shalf be *he f j that mere frecuent su,feillince of PCF' in e.2 value ~ ult iplied by 1:f. f } pccc o-- y, it cbn11 rnecify en au;m:nt~1 r is alv,a ::y Fign o 3.11.2. j sur>cil ic ice prer,rc.m cor er:rurate vi.th r eactor eberc E

7. i a t c ny t ir a it is dete-i: ed thet i

cend it ions, } the li:-it-irr :: alt!c of NCF" is boirg cxeceded, action shall b< t.:kca t-edictely to restore opetiti<m to j . thin precc: ileI lirlts. 4 i i l l l 189 D 3.11/4.1.1

s' Baser 3.11 A, Av ra re "lenar Linear Heat Generatien prte (A DLPM W s specification assures that the pen claG inr temperature fc11cwing the pontulated design f~ric los.r-of-coolant accide t will not exce. ' the limit specified in the 10GFR50, Apperdix K. The pech cic6!ing temperature follcuing a pc'-tulated loss-of-ccolant accident is primarily a function e C the avercre heat generation rate af all the roda cf a fuel asscobly at any axial loca tion :'"d i s onl:- deerdent recerAnrlly cr the rod to red power distribution within an en a bly. Since expected 1ccal variatione in pcrer distribution within a fuel assembly affect the calculatcd peak c1Mding tempercture by Ic~; than + 20 F relative to the peak temperature for a typical fuci desirm, the lhit on the average linear heat generation rate is sufficient to assure that calculated trperatuces are withic the 10CFR50 Appendix K limit. The limitirg tatue for ArumR is given by this specification. It is recetnited that APUICR is a calculated parameter that is not continually monitored and alamed directly during core pcwer distribution changus. If at the time of the calculation it is found that the limits are beirg exceeded,. there la alwayr an action which will t turn the average planar U:GR to within prescribed limits, na cly power reduction. Under most circums' ances, this will not be ti e caly alte native. %enever L5e linit is exceedes' Oc t onitored value will be doct=:ented cnd availabic 22 for revici, audit and inspection of plant operations. B. Local UIR This specification ass tres that the linear heat generation rate in any rod is less than the design lit. ear heat generation if fuel pellet densificatica is postulated. The power spike penalty specified is based on the analysis presented in Section 3.2.1 of Refere:.ce 1 and in References 2 and 3, and assumes a linearly increasing variation and a <Ia1 gaps betueen core bo.; ton and top and assures.with a 957. confidence, that no more than ene fuel r'd exceeds the design linear heat generation rate due to power spiking. It is recorniced that UIGR is a calculated parameter that is not continually monitored and alar ed directly during core power-distributien chanres.. If at the tire of the calibration it is fcund that the Jimits are bei,g accedci, there i.c Gways an action which will return the UIGR to vitiin presc ri. bed limits, ncmelv utser reduc tion. L' Jar r: ort circtr.s t anc es, this will not be the enly a lt< rna tiv^. Whenever

• he limit ir enceeded tc - onitored value will ba decrmented and available for revics, audit and innpection of pDnt ope-ations.

11te only way to violcte the Litaiting Condition for Operatian is to knowiny,1y allow eperation beyond the prescribrd limits without taking the recessary oction to restore the L"GR to within prescribed limits. 3.11 7 m s IM E

s I Mses 3.11 (continued) C.

n l e Critical Power Patin (MCPM 1he ECCS c'.aluation presented in Reference 4 a,str ed the stea3y state MCPR prior to the postulated loss of coolent accident to be 1. I'i for all fuel types.

The Operating MCPR Linit of 1.41 for 8x8 fuel an ' 1.33 for 7x7 feel is deterrained fren the analysis of transients discussed in nases Sections 2.1 a:-d 2. 3. I': m N :aining an operatinn MCpR aberc these limits, the Sa fet, Linit of 1.06 (T.S.2.1.A) applicab'< - all iuel types is r aintained in the event o f t h e no : t limiting abnonna t operatienal t ror,ir t. For operation with less than rated cere flew the operating MCr'1 Limit is adjusted by multiplying tbc above limit by K. F e fe renc e 3 discusses how the transient tralysis f done at rnted conditionn enempasres t':e rede&I *~1c" situation tchen the preper Kr factor is applied. It is reccanized that MCPR is a calculated parameter that is not continually monitored and alanned directly during core power distributien n,d thermal-hydraulic changes. If at the time of the evaluation it is found that the li-itn are being exceeded, there is always an action which will remrn the MCPR to within prescrib^ 1 hits, nanely pover reduction. Under most 22 c iretas ta nc es, this will not be the only alter ive. Whenever the linit is exceeded the ronitered . :udit and inspection of plant operations. value vill be documented and available for re. The enly vay ta violate the Iimiting Cenditic, for Operation is to knowingly allew operation beyond the prescribed limits without teking the necessarv action te restore the MCPR to within prescribed limits. _Re fe renc es l. " Fuel Densification Effects in General Electric Eoiling Water Reactor Fuci," Supplements 6, 7, and 8, NEDM-10735, August, 1973. 2 Supple ent 1 to Technical Report en Densificatien of General Electric Reactor Fuels, December 14, 1974 (USAEC Regulatory Stafi) 3. C~=unication: V A Moore to I S Mitchell. ' tdified GE Model for Fuel Densification," itecket 50-321,:* arch 27, 1974. 4. " Ment i e llo Nucl ^a r Ge nera tin g, Plant Lesr-?f "colant Accident Ana'.ysis Ceaformance with 10 Cir 50 Appendix K, August 197'4 L 0 Pye-C:SP) to J F O' Leary, August 20, 1974. 3. " General Electric BWR Cencric Reload ArH ier tion for 8 x 8 Fuci," NEDO-20360, Revision 1 1, Novemb>r, 1974. I m" - 3.11 MSES

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, 1, 1.0 SUITIEILLANCE FEUDID'.EITIS

3. 0 LD1ITI?!G COI.3ITIOC FOR OPEMTION 4

1

2. Each cealed cource with renovable a. Ihch cealed cource, except ctortup l

contaminntion in execcc of the limit cources cubject to core-flux, con-in 312. A.1 chall te i=ediately with-tuinire radioactive caterial, other ~~ drawn from uce cnd: titn !!ydrogen 3, with a hc.lf-lifa Creater thra 30 days and in arce form

a. Either decontaminated and repaired, other than cas chall be tectea fur leakage and/or conta-dnation nt or intervalc not to exceed cix r:onths,

b. Disposed of in accordnnce with the regulationc of the Cor6 scion b. Tne periodic leak test repirerl does not apply to cealed cources th"t are stored and not being uced.

": cources c>:enpted from this tec+ *n be tested for leakage prior to any tire or transfer to another ucer unless they have beer leak tested within cix months prior to the date. of use or tranufer. In the absence of a certificate from a transferor in-dicating that a test hac been en.de ~' within six conths prior to the transfer, cealed cources shnH not be put into use until tested far leaknge. c. Startup cources chall be leak tested brior to and following any repair or maintenance and before being subjected to core flux. 1890 22 79 ,a.12/4.12 i had

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