ML19206A786

Order for Mod of License DPR-73 Requires re-evaluation of ECCS Performance,Power Reduction to 7568 Mwt, & Conformance to Procedures in Util 780505 & s

ML19206A786
Person / Time
Site:	Crane
Issue date:	05/26/1978
From:	Boyd R Office of Nuclear Reactor Regulation
To:
Shared Package
ML19206A785	List: ML19206A784 ML19206A786
References
TASK-TF, TASK-TMR TM-0167, TM-167, NUDOCS 7904210288
Download: ML19206A786 (9)

Text

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UNITED STATE 5 UF APEP.ICA NCLE/P REGULATCSY CO: fi!SSIOi; In the : tatter of

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EETROFCLITAM EDIS0it CCnPAin, CT AL

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Decket tic. 50-320

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Threc.'iile Island du; lear Station, Unit-2

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CRGER FOR t'0DIFICATICri 0F LICCDSC I.

The :letropolitan Ecison Ccepany, et al (the licensee or 'tet Ed), is the hcider of Facility Cperating License Mc. EPR-73 which authorizes the cr.eration of the nuclear power reacter kncwn as Three i-;ile Island Fuclear Station, Unit 2 (the tacility or TMI-2), at reactor core pcwer levels not in excess of 2772 cegawatts thernal (rated power). The facility, using a Babcuck & Wilcox Cenpany designed pressurized.:ater reactor (PLR), is located at the licensce's site in Cauphin County, Pennsylvania.

II.

In accordance with the recuirements of the Ccnnissicn's ECCS Acceptance Criteria,10 CFR Part 50.46, the licensee subritted on.: arch 31, 1975 an ECCS evaluation for the facility. The ECCS evaluation subnitted by the licensee was based upon an ECCS Evaluation tiodel developed by the "abccck 2 '.ilccx Ccocany (20'.4), the designer of the nuclear stean supply systen for tnis facility. The BOW ECCS Evalcaticn incel 9ad teen previously fcund to confor-7 9 0 4 e.10 2857 orrica p summaMs

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>u-tc uie requirements of the Cen;;ission's ECCS i.cceptance Criteric,10 CF3 T' art :P.l.co anc :.prendix.<. The evaluation indicated that with tne linits set furth in tne facility's Technical Specifica icns, the i.CCS coclins perfor.:ence for the facili y wculd conform with ti.e criteria conteinca in IbiJn Par LG.4d(::) unich govern cciculated peak clad temperature,

. ax i uc:a claauing oxication, naxir :. nydrc;ca cencraticn, ccclable jecnetry ani long-tern cccling.

Cn I::ril 12, 1972., W inforred the I;ftC that it had deterninec that in tha event of a snall creak less-of-ccolant accident (LOCA) on the aischarge sice of a reactor coolant pump, high pressure. injection (HPI) ficw to the ccre could be reduced sccowhat. Subsecuent calculations incicated nat in such a ca e the calculatcc peek clac tercerature micht exceed

.. C b.o Previcus small break analyses for Mi 177 fuel assertly (F/d icucred lcc?

plants hac ioentified the limiting snail break to be in the suctica line of the reactor coolant purc. P.ccent ana'yses have shed.n that the cischarge line ::reak is ccre li..;itine than the suction line break.

The Three :;ile Island i,uc! car Station, Unit 2, has cn ECCS ccaficurri.nn b.lich consists of tuc high pressure injection trains.

Lcn tr: ! 1 bas en HPI pump and the train injects into t'.o cf the fcur reactur c:olan

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" 2* 1 - 72 3 - it vi G= e,,; L I,; e m lgl2.0 i 1== 1 ' l.25 1A I ii g.6 -= l 4 6" b El%4%s"&p ^ g 4t 4)"% zw>,yw///>, 4+.2.::s f,;,,t n a y a. v o ex v. <. igt' cp 3_ system (F.CS) ccid less on,he discharge sida of c.te F.CS pe;. (T:iere is also a third LoI cuup installec.) The two parallel di trains are cccr.ectac but are kept isolated by rar.ual valves (known as the cross-connect valve:s) that are normally cicsed. Upon receiving a safety injectica signal the liPI pumps are started and valves in the fcur infecticn line: are cpened. Assuming loss of cffsite power and the uorct single f ailure (failure of diesel to start) enly one HPI pump sculd ce availaale and two cf the fcur injection valves would f ail to open. If a small break is postulated to cccur in the RCS piping t'etwe.:n the RCS pump discharge and the reactor vessel, the high prcssure injection flow injected into this line (about half of the cutput of ena high cres-sure injection pump) could flow out the break. Therefcre, fcr tna worst ccr.cination of break ic:ation and single failure, cr.ly one-half cf the flow rate of a single high pressure injection pucp wculd centricute to caintaining the coolant inventcry in tne reactor vessel. This situs:icn had not been previously analy:ed and BC,. had indicaten that the linit: specified in 10 CFR Part 5C.46 may be e.tceeced. CTiW has stated that they have analyzed a spectrum of srcll bred.s in the pung discharge line and have deternined that to raet the limits,f 10 CFri Part 50.46, operator action is required to cpen the two manually-operatac crcss-connect valves and tc canually cpen the tuc retor-criven inclaticn i4a u,..

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[M.. h... NRCFORM 318 (946) NRCM 0240 D us s. eovannusny omntine omesa so7s-eas.eaJ Tois wouls allow the flow frcn the cre hPI purc to faed 611 fcur rccc:cr cociant legs. G W nas assurec that 30 percent of the ficu would 'ce icst thrcugh the break and 70 percent s ccic ccntribute to recovering the core. h has preparec a sucrary entitled " Analysis of Er.all Brecks in the ?aacter Ccolant Pur.p Discharge Pising for the M?: Lewerec Lcop 177 F1 Pl;nts," Mey 1,1973 (the EP.> Scrrary), which describes tr.a etnces used anc tha results cctained in the above analysis. The analysis redels ocerator action by assuninc a steo increcsa in ficw to the reactor vassel (witn balancec flow in the three intact loops) ten minutes after the LCCA reactor prctection systen trip signal cccurs. Cy letter catec i-ey 5,1972, Met Ed subai ec a copy of the EEn Suxcry for our review. In their submittal i*et Ed statec that they had reviewed the C&W Surncry ana detemired that the results were applicable to T"I-2 aM that cperaticn cf Ti11-2 up to 25EG regawatts thernal woule Se in full confor ance with 10 CFa Part 53.45. Ti ey also statec that acditional analyses will be av::'able to the Ccmiss an fcr gewer levels up te 100 percent gewer (2772 mega.' e ra ter.eal) by June 1,1971. In their surrittal of 90y D, 1973,.ie*, Ed also stated that they "ad d ified ceridin plant prccedures to previce the recessary crerater JCthCcs c0 3 ~,i d SC310 CCnsi5 tent iitn thau assured in tw 3calysis, canca

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_5_ and that they nac conducted Grills to verify tha: the assured operator resecase tire..cs achiavable. The Ccrrissicn's Uffice cf Inspection 'nd Enforcerent has confir'ed that appropriate prececures are in place and that crilis were perferred thich verified cperator response tine. Net Ed also cennitted to suboit as scon as pcssible a request for arcncren; of the TMI-2 Technical dpecificatiens as apercpriate to reflect ado;ticti of thase prcccoures, and cuanittec to subni: a prepcsal fer a rerranent solutien to this problen cy Aur.uct 5,1070. In their le::er of Pay 11, I'd8, Met Ed providea adaitional infornation clarifying aspects cf the ;roposec nanual actions. In the even cf a snall areak and a limiting single failure, manual action will be taken to begin caening the crosscennect valves and the isolation valvos within five ninutes and have them cpened and an adecuate flow split ob:cined within 10 minutes. To facilitate this operation tne ticensee nas ccrnitted to maintain cne of the series-ctnnectec, nenLally-cre atec crcss-connec valves normally open. The analyses pcrforced by 21.. assurec tnat thu ficw split was established at 650 secones by operator articn, ne concluce that t.he analyses are a reasonaale approximation of the everator acticn tuat actually will ce taKcn, since specific prececures P. ave eeen prepared and drills performed to verify the acecuacy of the procecures anc to train :ne plant cperatcrs. 1 one'c a w Su2N4ME h

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/ In t'..:i r analysii, diW States tt a 0.13 f: cisc.arge lire tr:m,,sith the aforS entioned vperatcr aC'. ', i:, t,0.lcSt,6initinG caSC. .ao arri VO at thi', Ccr.cl usicr, 32..e has 't:rt brr.t;d arlal,f Ses at :raar. 512e5 n. r., /, -'. 'i. u.1 3, v'. 'i., 0.1. ',, '1. w*, - 'v. '.+.'9 'e it u' r c. u '. *. c, - h i c "> 'a r".- .s e <1 o. c., v u s a... c e v. t.,c t a i 'c0 using an apprcvec Apper:cix t' r cdal icr Olcuccun, incicata cc e n u n c o.> e r'. fcr abcut hA seconcs for the 'v.13 f t' creck. For t'.is break size 205 nas CGnSeryctival,' calculcted the feik claG tcr.cer2 turn 0 to be approximately 1551 F, well telcw the li-'.its af le CFR Part 50.sS(a), for a pc,,er l avel or u, e c ega.;ctts tneraal. a Based en review of the 3 ",. S nrcry '..e finc that t.se calculaticvs 3 G;crt the cor.clusica :nat a U.13 ft2 cischarg2 line break is the rest limitir Neuever, ti'e -!.'. $ttr'ary cces not cercnstrate that the assurrtions J case. at:ployec in supplying heat ingcts ta the F09.7 portien uf the calcule,:icn: wera con:crvative. we are also revie,ing uretber use of sirpliti ac i ' cut in the FOAM calculetiens satisfies the recLirenent for calculation using an :;rrever: nccel. Accorcingly, we cannct cenclude at this tir.e that operation of THI-2 at 2568 negawatts *her al woulc ce fu'ly in ccnfer-ar.cc ill!D lC t'FE Part 50.M. 03 the Gther hand. the ren' t cf CCICulati0ns ret

• CVaila0le 2h;h:. !!iat fcr OperatiGn of thiS f 2Cill!y at, GLOP lt:VeIS u' !G.55Cc '"egCbdtts thenOI. ~.CCS ' erf ur"~anCO calcul atiGDS f O r L E.:2 J

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j#. 3 c e, a. 4. u. *.3 n -.4 .a p r,.. L. e 1.,.,.. - s-c.,,. (.,. c. n. e. -j :.=. o.., .s. a.,w s v-3 ATE D m s N2C FORM 318 (?-?6) NRCM J 40 W u. s. sovs m.ne s ev eninvino orricts n ote e:St ' - 7 Therercrc, until ',,e nave hac the oppcetunity to fully assass the e ' calculations, the staff cannot cetermine that operaticn cf T:tI-a: full pcwer uncer the ccnciticos of :ne revised calculations by J23 epplicable to this facility conferes fully to the re;uirements of 10 CFR Part 30.46. However, cparation of Tal-2 et pcuer levels up to EscJ Tegaudtts thcrrai aoc in accorcance witn apcropriate operatinc creccdures will ensure that the FCCS will ccnfera to the perferaar.ce criteria of 10 CFR Part b0.46. Therefore, until Cro calculations applicable to this f acility are cctrleted to assure full cccp7iance with 10 CFR Part 50.46, the peak clad terperature cargins provice recsonable cssurance that operation of the facility at pcner levels up to 2568 negawatts thernal with appropriate operating proccdures specified nerein will not encanger life er preserty or the cert:cn defense anc security. Witn the procccures described in the licensee's letters of May 5 and 11,1973, the staff believes that the licensee's acticas are apprcpriate and that these acticns shculd be confinled by hRC Order. In the ccurse of cur review of this matter, a related issue arcse:

he neec to ucaly creater uncertainties to tre easured values of neutron .ux in each quacrant of tne reactor ccre.

recently rescr:cd to:et Ed that on the casts of aperational ex;erier.c3 ': u.. i ,,,jijg. a reevaluaticr cf i easurece it errcr sta : stics and error propagatic., su==au r

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'T C "l e,Aa NRC EORM 318 ( 9.?6 ) NRCM 0240 W us s. aovenmuswv perintino orricas to7s -eas422 -u-r;reater uncartai:: ties should be applied to the r:easurec values of cuccran; is greater uncertainty was necessary to assure that the actual flux tilt. flux tilt dia not exceec :::e liaiting value assured in the evaluation cf p.;stulatac accidents includino a LOCA. A descripticn of the reevaluation and rcc;; rec.ced reduced linits en alle..able :aeast. rec flux tilt were presentcc in c CW,. reper subnitted to the staff en :ay 11,1970. My letter catec .ay 'O,197U.1et Ed requested Jr.encnent of the I:il-2 Technical $cecifications to reflect the ecre conservative lini;s. be nave reviewed the.C.. repor* and tne.et Ec request relative to this ::etter and have conclucec that the limit', recuestec f cr LI-2 are acceptable. Use or these limits is being authorized by Fr.:eno:en No. 4 to the TMI-? uperating License No. CPd-73, issued cn ::ay 19, 1970. III. Copies of the fcilowing cccunents are atailable for inspectice at N o blic Doc'z:ent Kcco a*. 1717 F Stree

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Cumissien's u 20:.55, and are beir:g placea in the Comissicn's Iccal pu::lic cccur.:ent rcen at the State Library of Pennsylvar.ia, Cc:rens,calth and Walnut Streets, Harrisburg, ?ennsylvania, 1712c: 1. Letter frer: J. G. herecin ;c S. A. 'lar a, Chief, Lich; ater Reac:crs branch No. 4, catac ::a'/ s,197E. e, I

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sv==au t * ,. s v. _ ( +, wt* .i. NRC FOR.M 318 (9 76).NICM 0240 W u. s. sovs mau sur painnmo opricri teve - emu -U s IV. Acerr ingly, cursuant to the Atenic Energy Act cf 19:4, as acended, end the Ccmission's Rules and Rcgulations in 10 CTR Parts :' eno 50, IT IS UC ERED THAT Facility Operating License !.c. 0FR-73 is hereby ccnditioned Oy adding the felicwing new provisicns: (1) As seen as pessible, the licensee shall sutnit a reevaluation (wi:olly in confomance riith 10 CFR Part 50.4c) of ECC cccling per#or acce calculated in occorcance with the B&'4 Evaluation : ccel for operatien with operating precedures described in its letter of.May 5,197;., (2) Until further authorization by the Cemission, the power level sball not exceec 2563 regawatts themal, and (2) until further authorization by the Ccmission, the licensee shall operate in accercance with the procedures cescribed in its letters of iay 5 anc riay 11, 1071. FOR THE SUCLEAR REGL'LATCRY CC:-:'ISSIO:

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Rcger S. Soyc, Director Division of Pr"cject Managerent Office of Nucieer Isaacter Pegulation Octed at tietnesda, Marylanc, shis 26th cay of ;ay 1978. OELD STreby* / /78

• See previous yellcw for concurrence DPM: LWR :(

DPM: LWR \\kDPM: LWR #d DPM: LWR /AD DPM: DEP DIR DPM:DIR c.,,c MService HSilver:M b* SVarga* DVassallo* RDeYouna RSoyd Fr+- s 5/ /73 5/ /78 5/_ /78 cs/-Rf / / '8 5/26/78 i NRC FORM 313 (9 76) NRCM 0240 W un s. movsawu rnt cannvino orricts is7e - ese422 i.*.

ccorcinil;, pursuant to tne Atenic Cnercy Act of 1954, as arenc

ee, ai c tre Ccmission's Rules and Regulations in 10 CFP. Pcrts 2 and 60, IT IS SCU.EL THAT Facility Operating License iso. CPR-73 is hereby conditicr.ed by <:cding the folicwing new prcvisicos:

(1) .is soon as pesticle, tne licensee si;all st:rit a rcevaluation (s. holly in confomance with 10 CFR 50.06) of ECCS ccolinq perfomar.ce calculatec in accurQoce with the 52.! Evaluation 'ecci for cperation :iG operatine prccedures cescribec in its letter of T'ay 5.1978, (2) Ur..il furtner authoriaatien by the Ccmission, the pc :er IcVel shall not exceed 2568 rega,,atts themal, ard (3) botil further authorization oy the Ccmission, the licensee shall crerate in accordance with the procedures described in its letters cf Say 5 and Fay 11, 1970. .c t s r, tu.,, c, c Y u, o. run ok,.. u, C L r,.- ..ccr.c. ama .~c o Peter 5. Scye, uirector Divisien of Project.Managerent Office cf.uclear.seactor Jeculaticn Mtad at Cethesca,*arylcnd, ,)' d-i s cc cf 'av 1973. MA i ~ DPM: LWR 44 DP'1: LWR d4 DPM:LA'R da DPM: LWR /AD DPM:DEP DIR DPM: DIR o,,,c s, MService .7, HSilver:tib ,SVarca DVassallo RCeYoung RBoyd ,u,,,,,,,, 5/24/78 STP/7M 5/1-f73 5/ /73 l 5/. /78 5/ /78 un, n_ y' q " NRC TORM 31s (9.*6) NRCM 0240 W un s. moven eassnr ensavino orricas is7e-eaus4 a UNITED STATES OF Ai! ERICA NUCLEAR REGULATORY C'JillISSION In the Matter of ) ) f!ETROPOLITAN EDISON C0!!P ANY, ET AL ) Docket tio. 50-320 ) Three flile Island Nuclear Station, Unit 2 ) ORDER FOR MODIFICATION OF LICENSE I. The Metropolitan Edison Company, et al (the licensee or Met Ed), is the holder of Facility Operating License No. DPR-73 which authorizes the operation of the nuclear power reactor known as Three Mile Island Nuclear Station, Unit 2 (the f acility or TMI-2), at reactor core pcwer levels not in excess of 2772 megawatts thermal (rated power). The facility, using a Babcock & Wilcox Company designed pressurized water reactor (PWR), is located at-the licensee's site in Dauphin County, Pennsylvania. II. In accordance with the requirenents of the Cornission's ECCS,1cceptance Criteria,10 CFR Part 50.46, the licensee submitted on March 31, 1976, an ELuS evaluation for the facility. The ECCS evaluation subnitted by the licensee was based upon an ECCS Evaluation flodel developed by the Babcock & Wilcox Comcany (B&W), the designer of the nuclear steam supply systen for this facility. The B&',i ECCS Evaluation Model had been previously found to confor, 6'i OL9 / _2 to the requirenents of the Connission's ECCS Acceptance Criteria,10 CFR Part 50.46 and Appendix K. The evaluation indicated that with the limits set forth in the facility's Technical Specifications, the ECCS cooling performance for the f acility vould conforn uith the criteria contained in 10 CFR Part 50 46(b) which govern calculated peak clad temperature, naximun cladding oxidation, naximum hydrogen generation, coolable geonetry and long-tern cooling. On April 12, 1978, B&W infonned the 'IRC that it had ieternined that in the event of a small break loss-of-coolant accident (LOCA) on the discharge side of a reactor coolant pump, high pressure injection (HPI) ficw to the core could he reduced somewhat. Subsequent calculations indicated that in such a case the calculated peak clad temperature might exceed 0 223G F. Previous small break analysas for 33W 177 fuel assenblv (FA) lowered loop plants had identified the limiting small break to be in the suction line of the reactor coolant pura. Recent analyses have shown that the discharge line break is more limiting than the suction line break. Tite Three 'lile Island Nuclear Station, Unit 2, has an EI.'.S configuration shich consists of t'to high pressure injection trains. Each train has an HPI pump and the tr in injects intu two of the four enci.nr coolant ,-<+. (34 n h V o 3_ system (RCS) cold legs on the discharge side of the PCS ourp. (Thare is also a third HPI punp installed.) The two parallel HPI trains are connectei but ar? iept isolateri by manual valves (kncan as the cross-econect valves) that are nornally closed. Upon receiving a safe v inject:on signal the HPI pump; are itirted an<l valves in the Four injetion lines are opened. Assuming loss of of fsite power and the viorst sin 9 e f ailt.re ? '#ailure of diesel to start) only one HPI pump would be available and two of the four injection valves would fail to ocen. If a small break is postulated to occur in the RCS piping between the RCS pump discharge and the reactor v 1, the high pressure injection fl oti injected into this line (a' nut half of the outnut of ona high pres-sure injection pump) could flow out the break. Therefore, for the worst conbination of break location and single f ailure, only one-hal f of the flow ra.e of a single high pressure injection punp would contribute to maintaining the coolant inventory in the re3ctor vessel. This situation had not been previously analyzed and R&W hau indicated that the limits speci fied in 11 CP ?)rt 10.16 ;ay he exceeded. 801 has stated that they have analyzed a spectrum of small breaks in the pump discharge line and have detenained that to met the linits of 10 CFR Part 50.46, operator action is required to open the turn manually-operated cross-connect valves and to manually open the two motor-driven isolation nives which had f ailed to oper. and align all four isolation valves. ,-a,...r._4Q. e4 J_ This would allow the flow from the one HPI pump to feed all four reactor coolant legs. B&W has assumed that 30 percent of the flow would be lost through the break and 70 percent would contribute to recovering the core. 6&W has prepared a summary entitled " Analysis of Snall Breaks in the Reactor Coolant Purp Discharce Piping for the B&W Lowered Loca 177 FA Plants," May 1,1973 (the B&W Sumnary), which describes the nethods used and the results obtained in the above aralysis. The analysis models operator action by assuning a step increase in flow to the reactor vessel (with balanced flow in the three intact loops) ten minutes after the LOCA reactor protection system trip signal occurs. By letter dated May 5,1978, Met Ed submitted a copy cf the S&W Sumnary for our review. In their subnittal Met Ed stated that they had reviewed the B&W Sumary and determined that tne results were acclicable to TMI-2 and that operation of TMI-2 up to 2563 megawatts thernal would be in full conformance with 10 CFR Part 50.46. They also stated that additional analyses will be available to the Comissinn for pcwer level s up to 100 percent power (2772 megawatts thernal) by June 1,19'" In their submittal of May 5,1978, Met Ed also stated that they had modified certain plant procedures to provide the necessary ocerator actions on a time scale consistent with that assumed in the analysis, GE -C12 and that they had conducted drills to verify that the assured operator response time was achievable. The Commissien's Office of Inspection and Enforcement has confirred that appropriate procedures are in place and that drills were perfarned which verified operator response time. Met Ed also connitted to subrit as scon as possible a request for arendment of the TMI-2 Technical Specifications as approcriate to reflect adopt.on of these procedures, and connitted to submit a proposal for a perman nt solution to this problem by Augusc 5,1978. In their letter of flay 11, 1978, Met Ed provided additional information clarifying aspects of the prcposed manual actions. In the event of a small break and a limiting single failure, manual action will be taken to begin opening the crossconnect valves and the isolation valves within five ninutes and have them opened and an adecuate flow split obtained within 10 minutes. To facilitate this operaticn the licensee has ccmnitted to maintain one of the series-connected, nanually-operated cross-connect valves normally open. The analyses performed by B&W assumed that the flow split was established at 650 seconds by cperator action. We conclude that the analyses are a reasonable approxiuation of the cuerator action that actually will be taken, since specific precedures have been prepared and drilis performed to verify the adecuacy of the procedures and to train tha plant operators. 'f ~ C 0 . 7 In their analysis, BfW states that a 0.13 ft-discherge line break, with the aforenentioned operator actions, is the most liniting case. To arrive at this conclusion, B&W has perfarned analyses at break sizes 2 of 0.04, 0.07, 0.1, 0.13, 0.15, 0.17, 0.2, a nd 0.3 f t. The results, which were obtained using an appreved Appendix K model for blowdown, indicate core 2 uncovery for abcut 300 seconds for the 0.13 f t break. For this break size 88W has conservatively calculated the peak clad temperaturn 0 to be approximately 1551 F, well below the limits cf 10 CFR Part 50.46(b), for a pcuer level cf 2568 megawatts thermal. Basad on re iew of the B&W Sumnary we find that the calculations support 2 the conclusion that a 0.13 f t discharge line break is the most limiting case. However, the BZW Sumnary dces not denonstrate that tae assumption; employed in supplying heat inputs to the FOAM portion of the calculations were conservative. We are also reviewing whether use of simplifiec inout in the FOAli calculations satisfies the requirement for calculation using an approved nadel. Accordingly, we cannot conclude at this time that operation of TMI-2 at 2568 negawatts thernal would be fully in conformance with 10 CFR Part 50.46. Cn the other hand, the range of calculations r.cw available shows that for operation of this facility at power levels up to 2568 megawatts thermal, ECCS performance calculations for the limiting small break indicate that this break has a very substantial margin on peak clad temperature below the limits of 10 CFR Part 50.46(b) If accrcpriate operator action is properly taken (as described above). 7 ~ (:St - f' 4 Q - Therefore, until we have had the ocportunity to fully assess the S&W calculations, the staff cannot determine that operation of TMI-2 at full power under the conditions of the revised calculations by B&W applicable to this facility conferns fully to the requirenents of 10 CFR Part 50.46. However, operation of TMI-2 at power levels up to 2568 megawatts thermal and in accordance with appropriate operating procedures will ensure that the ECCS will conform to the cerformance criteria of 10 CFR Part 50.46. Therefore, until B&W calculations applicable to this f acility are ccmpleted to assure full compliance with 10 CFR Part 50.46, the peak clad temperature margins provide reasonable assurance that operation of the facility at power levels up to 2568 megawatts thermal with appropriate operating procedures specified herein will not endanger life or property or the corron defense and se:urity. With the procedures described in the licensee's l?tters of May 5 and 11,1970, the staff believes that the licensee's actions are appropriate and that these actions should be confirmed by NRC Order. In the course of our review of this matter, a related issue arose: the need to apply greater uncertainties to the measured values of neutron flux in each quadrant of the reactor core. B&W recently reported to Pet Ed that on the basis of operational experience and a reevaluation of measurement error statistics and error cropagation, c-. - _ c -- -S-a greater uncertainties should be applied to the neasured values of quadrant flux tilt. This greater uncertainty was necessary to assure that the actual flux tilt did not exceed the limiting value assumed in the evaluation of ~ custulated accidents including a LOCA. A description of the reevaluation and recconended reduced limits on allowable measured flux tiit were presented in a B&W report submitted to the staff on fiay 11, 1978. By letter dated May 10, 1978, Met Ed requested arendment of the TMI-2 Technical Specifications to reflect the more c.nservative limits. We have reviewed the Bf.W report and the Het Ed request relative to this natter and have concluded that the limits requested for THI-2 are acceptable. Use of these limits is being authorized by Amendnent No.'4 to the TMI-2 Operating License No. CPR-73, issued on May 19, 1978. III. Copies of the following documents are available for inspection at the Commission's Public Docunent Room at 1717 H Street N.w., Washington, D. C. 20555, and are being placed in the Comnission's local public document room at the Ctate Library of Pennsylvania, Commonwealth and Walnut Streets, Harrisburg, Pennsylvania, 17126: 1. Letter frca J. G. Herbein to S. A. Varga, Chief, Light Water Reactors Branch No. 4, dated May 5,1978. 2. Letter frcn J. G. Herbein to S. A. Varga, Chief, Light Water Reactors Branch No. 1, dated May 11, 1978. E,Z $ 6 h _g. I '/. Accordingly, cursuant to the Atenic Energy Act of 1954, as anended, and the Commissien's Rules and Regulations in 10 CFR Parts 2 and 50, IT IS ORDEPED THAT Facility Operating License uo. DPR-73 is hereby conditioned by adding the following new provisions: (1) As soon as possible, the licensee shall subnit a reevaluation (wholly in confoniance with 10 CFR Part 50.46) of ECCS cooling performance calcJlated in aCCordance with the B&W Evaluation Model for operation with operating, acedures described in its letter of May 5,1978, (2) Until further authorization by the Ccomission, tt._ power level shall rot exceed 2568 negawatts thermal, and (3) Until further authorization by the Ccrnission, the licensee shall operate in accordance with the procedures described in its letters of May 5 and May 11, 1978. FOR THE HUCLEAR REGULATORY CCFMISSION / s Roger S. Boyd, Director le Division of Project "anagement Of fice of nuclear P.eactor Regulaticn Dated at Bethesda, Maryland, this 26th aay of May 1978. c. c r,~ JN, ss... $