: ..

C.omm.ealth Edison e R£GlllATOR~ ooCKET .flll COPY One First National Plaza, Chicago. Illinois

Address Reply to: Post Office Box 767 Chicago, Illinois 60690

September 13, 1978

Director of Nuclear Reactor Regulation u.s. Nuclear Regulatory Commission Washington, D.C. 20555

Subject: Dresden Station Units 2 & 3 Quad-Cities Station Units 1 & 2 Proposed Amendment to Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30 Associated with Reanalysis of the Loss-of­Coolant Accident NRC Docket Nos. 50-237/249/254/265

Reference- (a): R. L. Bolger letter to E. G. Case dated October 3, 1977

Dear Sir:_

Pursuant to 10 CFR 50.59, Commonwealth Edison proposes to amend the Dresden Station Units 2, 3 and Quad­cities Station Units 1, 2 Technical. Specifications regarding revised Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits associated with reanalysis of the Loss-of­Coolant Accident. Reference (a) transmitted the General Electric Report _NED0-24046, dated August 1977, entitled "Loss-of-coolant Accident Analysis Report for Dresden Units 2, 3 and Quad-Cities Units 1, 2 Nuclear Power Stations (Lead Plant)." NED0-24046 documents the completion of the ECCS reevaluation and forms the basis for the revised MAPLHGR limits.

During September 1976, General Electric initiated an input reverification program to remove known conservatisms in the 1975 Appendix K ECCS Evaluations. By December 1976, the reverification program had identified significant input errors in the reflood calculations. Additionally, for the first four BWR/3 plant analysis performed in 1975, a double credit for structural absorption was discovered.

-- Commonwealth Edison

Director of Nuclear Reactor Regulation September 13, 1978 Page 2

• Docket Nos. 50-237/249 50-254/265

On December 15, 1976, commonwealth Edison derated Dresden 2, 3 and Quad-Cities 1, 2 limits (MAPLHGR) by 4% to eliminate the double credit for structural absorption. In January 1977, General Electric estimated the net effect of input errors to be a 6.5% decrease in MAPLHGRs. commonwealth Edison implemented the estimated reduction, on an interim basis, pending a final ECCS reevaluation.

Report. NED0-24046 documents the completion of the ' required ECCS reevaluation. The analysis was performed using recent NRC approved model changes in the General Electric computer codes. These changes remove unnecessary conservatism and tend to counter the effects of the input errors.

Section 5 of the analysis report presents the input and model changes to the ECCS reevaluation. The input changes identified, with the exception of a more accurate DBA break size and peripheral bypass area, represent corrections to er~oneous input values used in previous analysis. Generally, the input changes increase· the delay to core reflood (more severe results).

Evaluation model changes, also presented in Section 5, are twofold. The first model change was required by NRC as a penalty in calculating the counter current flow limiting (CCFL) effect on additional delay in ultimate reflood. The second model change, in the assembly heatup calculation, removes con­servatism in the radiation and conduction equations.

Table 6 presents the single-failure analysis. These results are unchanged from the previous analysis.

Attacmnent I contains the proposed changes for Dresden Units 2 & 3 and Attachment II contains the proposed changes for Quad-Cities Units 1 & 2. The referenced NED0-24046 document and the attached Technical Specification changes have received onsite and offsite review and approval.

Commonwealth Edison

Director of Nuclear Reactor Regulation September 13, 1978 Page 3

• Docket Nos. 50-237/249 50-254/265

Pursuant to 10 CFR 170, Commonwealth Edison has determined that the proposed amendment is a combined Class III and Class I for eacb site. As such, we have enclosed a fee remittance in the amount of $8,800.00.

Three {3) signed originals and fifty-seven. (57) copies of this transmittal are provided for your. use.

attachment

SUBSCRIBED a~d ~~~ to befoz me this ~ day of ~ptofmM/L,;, 1978.

llartJt,~ m_ 11.iJlf ayf . r otary Public

Very truly yours,

Cordell Reed Assistant Vice-President

,. •

ATTACHMENT I

DRESDEN STATION UNITS' · 2 & 3

NRC DOCKET NOS •. 50-237/249

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3.5 ~Uniting Conditions for Operation Bases

A. Core Spray and LPCI Mode of the RHR System - This specification assure3 that adequate eme~gency cool~ng capability is available.

DPR-19

Based on the loss of coolant an~ly~~~ included in References (i) and (~) in accordance with 10CFR50,46 ~nd Appen­dix K, car~ cooling systems ~rovide sufficient cooling to the. core to dissipat~ the energy associat~d with the loss pf coolant accident, to i:µnit. the calculated peak ~lad tempe+atur~ to less than 2~00°F, to assure tpat core geometry remains intact, to limit the core wide clad m.etal-wate~ reactio" to less than 1%, an9 to limit the c~i~ culated local metal-w~ter ~eac~ion to less than i7%.

The allowable repair time~ are es~ tablished so tpat the average risk ~~te for repair would be no greater th~n the basic risk rate~ The method aQ4 concept are described in Reference (3). Using the results

(1) "Loss-of-Coolant Accident Analysis Report for Dresden Units 2,3 and Quad-Cities Units 1, 2 Nuclear Power Static>ns (Lead Plant)", NED0-24046, August 1977.

dc,·e lnpl'd In this rc-for­'nce, •h, r~palr p~rJod is f9und to he le!'~ lh:tn V? •he t~s~ Jnten·at, This ~::i-umes th:tt the core spray ;incl LPCJ subsyi-1cms coniotitut~ :i

l QUt Of 3 system, J\OW(>\'Cr, th<.' COl"!lbinC'd cf-. reel or th~ lWQ ~.\·stems l() li:nit <.'.XCC'~!=in• Cl:td · temperatures mui-t al~o hC' co:i~idcrC'd. The iei;t lnlerrnl 1=r.C'cificll in ~rwcifica1inn -1. ;; w:is 3 months. ThC'rdorC', nn :tlh\\.~~J!c 1·qnir· perio<1 which npin!:Jin:: the h.'.l~ic ·ri~k con:;idcr­•ng single fai!urr~ should h(· I~~:: th:'ln ·l.-, d.ws · ;ind this SpC'~ific:1:ion is with:n thi~ f'('riocl. · fpr multiple f::iibn·::. a :;t:•.'r:cr ;:i!Cr\':tl 15 sp"'cificd and lo impro·;c !!w .u:.:ur:rncc >bt th~ remaining ~~·~tl.·m~ \\ill f~nct-in~ • .'.l 11:\ih···. test Is cnlle'! for. ·\lt!vn:l!h it i::: rccn~:iii'cd t.~:lt the. infor!n:i.Uo11 biH'n i:i :·l'ft·n·nc<' 3 pro­\ides a quant1t:iun• •~~c·t!iotl l•l C':>li~nl<' :i.llnw­:tble repair time~. the bck of opt'r:llin~ cbt:i to support the an:i lytic:t I app1·r.i:u·h pr<'\ <'nt ~ com­plete acccpJancc of a his 1~~c:J;r1d :u 1his lime. Ther~fot·c, the :ia:C':; ~~.'.lied ia the ~ri<'cific ~leans were t>~!auli~hC'<l "ith due rq:;rd to Jl!dgnwni. ·

Should on~ corc ~pqy ~9.h=-r~•<'-!ll ~ccome in­O[X'.rable, ~he re:naining core Fp:"l.\" :lncl th<' entH·~ LPCJ s,p>lc·m are a\·ail:\blc :::hould th~

(2) NED0-20~66, General ~lectric Company Analytical Model for Loss­pf-coolant Analysis in Accordance with l0CFR50 Appendix K.

(3) APED-"Guidelincs for Determining Safe Test Intervals and Repair TjJnes for Engineered Safeguarcls" ~ April 1969, ~-M~ Jacobs and P~W. MaJ;riott.

82

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I. J.veraf.18 Planar UfCJ!

Thia spec1f1cat1on assure~ that the peak cladd1n~ teBperature f ollo~1ng a post~1ate4 design basis loss-of-cooln.nt acc1d~nt wll1 not exceed the 2200"F limlt si:ec1f1ed 1~ 1ocrn.50 ~ppeJ'ldix K conslt1er1ng the postu1at~~ effects of fuel pellet dens!flcation,

Tho peak cl.addl.nc te111per~tur~ follo\llr.g ~ postulated loss-of-ccolnnt acc.1derit is. frl~arlly n function of the ~verage LMCR of all the reds 1n a fuel asi:;e111bly a~ any axial location and le only ~~pendent second~ arily on the rod ~o rod powe~ d1s~r1bwt1on 111th1n a fuel assembly,. Since e:q:.acted local. var1a.t1ons in power d1st~l\Jut1cn n1th1:1 a fuel assembly affect the calcn4t.ed peak o~ te:r.perature 'by lesG than ±~o°F relali ve to the peak temperature f9r n t:yplc~l fue 1 de~1~, the limit on th~ average r~anar U!C~ 1o £uff1ciflnt to assure that calcul~ted temp­eratures are belew th$ 1 QCFH.50, /\pr:ow.Hx l{ liralt. · ·

The maximum average plan~r lHCna cho'ttn 1 n F1~a J,5,1 are based on calculat1oos employ~ 1ng tho models described ,n Refta~~n·~e (1). Pm•cr opera.t1on wlth JJfCRs a.t or be lcw thos& shown1nFlg. 3.5.1 e~;sure!l that.the peak. cladding temperature follow1ns a pcsl~lated, lcs3-of-coolant accident wtll not e::-:ceed the 2200°F limit, Those values reprecont l1m1to

. for operation to ensure conf orman;:e wi lh 10CfR.50 and Appendix K ohly 3.f they al·o moro 111r.1 ting thsa ·other design pararr.eter!l.

(1) "Loss-of-Coolant Accident Analysis Report for Dresden Units 2, 3 and Quad-Cities Units 1, 2 Nµclear Powe+ Stations (Lead Plant)", NED0-2,1-046 August 1977 ~

DPR-19

The lflllimum p.v~~g~ planar u~GJis plot tea 1n f'1g. J. 5. t ~t higher expo~ure& reswlt in a 9, calculAted pe~k cla.4 te~peraturf,l of less

· than 2200°F ~ However tho maximum avemgo pllln~r lHGHs p.r~ &Shown en Fie. J. 5.1 a~ i1m1 ts becau~e conformance caicula Uons have· not been performed to ju3t1fy operation at UfGR:i Jll el(cess of thos~ nhown. ·

J, J,ocal lHGn

Th ta specif !cation ~aures tha.t thi> QaXlmun linear hea.t generation rato in any rod lo l,.eaa tl}8.ll th~ de!llgn llneu.r

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A. core Spray and ~PCI Mode of.the RHR ·system - Th~s ~pecificatiOJl csssu~es that adequate emer9ency <:Qoltng capability ~s available.

D~R-25

Based on the loss ot coolant analy~es included in References (i) and (i) l.q accordance with .lOCFRS0.46 and AppeQ­dix K, core cooiing systems provide sufficient cooling to the cor~ to di~sipate the energy associated with the loss of coolant accident~ to titni..~ the calculated peak clad temper~ture to less than 2200°F, to assur~ th~t· core geometry remains ~ntact~ to J,.imi..t the core wide clad metal-water reaction to less than 1%, and to limit the cal­culated local metal.~wat~r reaction to less than ~7%~

The allowable repai~ time~ ~~e ~s­tablished so that the .. av~rage rl~~ ~ilte for repair would bo no grea.te~ toan the basic risk rate. The ~ethod an~· concept are descr~bed in Reference (3). Using the result~

(1) ''Loss-of-Coal.ant Accident Analysis Repc:>rt for Dresden Units 2,3 and Quad-Citi~~ Units 1, 2 Nuclear Power Stations (Lead Plant)'', NED0-24046, August 1977.

. . · dc,·elopcd In this ttfcr-ence, '"' repnlr period Is found t<> Ix.- lc~s th.'n t/i lfle l~~t Jnten·al. Thli; a::i-ume$ th:tt the con~ fipray ?.n'1 LPCI su.bf.yl'tcms conflitute n 1 out of 3 sylitcm, howe\·cr; the comhinC'd C'f­f~ct Of lhc l\\'O S.\"SICO\S to limit CXCC'S~in• Cl:td

· l~mper~turc~ mu!'t :tlt'o hC' coa!"icl"r<'d. Thc · tei;t intcrrnl t:p<'cific:tl it1 ~1wrifica(ion .a.;; w:ts . 3 rno11ths •. Thl'rC'(orc:. nn :tfl . .,·.L-:~J!t' JTfl.iir . period which m:iin!:tin.:0 lhc h:t=ic ri!'~ con:;idcr­ln~ single failure~ sho~ld hl· I('~~ th:tn -1.i <l:ws and this Sp<'c!fic:1:io:i is \•:ith!n thi!" ('<'riod. • · for multiple f:.iil: . .ir(·::, :t ::h.•r:e1· i:l!c·n·:tl 1s specificd :ind to impro..-c· t!I(' :t~.:;ur:tnc<' th:ll the rcrnaininf: >'Y~!l.'J)lf ~'ill f~:nctiCln, :t cbih·-· test ts calktl fo1·. ·\lt!F"!t:!!h it i:::; rccf\~:til'cd '.~at the infor~n.1!.ion ~i\·c11 i:i 1·E'fl•rc·nc<' 3 ;>ro­~HI~~ :t quanll!Jll\.(' n~d!iod l•l <'~li:11:H<' :tllow­nl>lc rrp~ir lime~. the brk or or,~r;atini: cbt:t to Support the :tn:tlytic:\I 3ppl'O:tch Jll"<'H'Ol~ cnm­plc-le accC'plance of 1his n:C':hr1d :\l lhis lime. Therefore, th<.- :im<'s !-'!:tlcc! ia lh<' ~pccific ih.'ins were rs!abl!fhC'rl \\ilh due rq::trcl lo J~1.l~n1epl! ·

Sho4ld D!'l~ core spp~· f4h~~·!"tt'm bt'con1e in­Opl'.rable, the re:n:~!ning COrl' ~p:":t~· :10cl lht' e11l&J"c:: ._.PCJ f:!Y:>l(lm. :ar~ a,·:ailablc ::ho4 td the

(~) NEno-iosGG, General ~lectric ·· · Company Analytical Model for Los$• of-Coolant Analysis in Accordance wit~ 10CFR50 ~ppendix K.

(3) APED-"Guidelines for Determining Safe rest Intervals and Repair ~bnes foi Engineered Safeguards" April 1969, I.M. Jacobs and P.w~ Marriott.

82

---......,,.-~----!-'!--··-·- ··~·-··----··--·-·.._... ............. ..-.-.. .... --~.

3.5 L1ft!t1ng Condition tor ()peration Dnses (Cont'd)

I.' ft. veraee Planar WC~ .. ,

Thie spec1f1cat1on asoures th!lt the peak claddlnf\ teniperatu~ folloulng a por.tulate4

. daclgn basis lose-of-cool:mt nccldont wt». not exceed the 2200'1' ll·mtt -or..oc1f1c'1 in tocrn.50 Appendix K con~1ar.rlng the postu~t~4 ef fee ts of fuel pelle~ dens1n.ca.tion. .

Tho penk cl.a<ldln~ tepporature follo~lne; ~ po~tulated los5-of-ccplnnt ~cc l deni 1 ~ .. primarily A function of tho nvor~c1'> J !lr.O or all the reds in a ft:ol n~r.e~hly nt nny axial location end 19 ouly d·~rc,:dent socond• arl ly on the rod to rod pc~ro:r I:! 16 lrl twt!ol") ll1th1n a fuel assembly, Slnco cxinct~d lo~ var1a\1ons in pow~r d1&tr1Lut1on utth1a ~ fuel asi;ambly affect the calct1l:Lted rcak olad temperature by le.sG thon i~o°F rel..i.t~ Vt' to the pen.k tempcrat\!re fer n t:yplcal :fual <Sc~J.gn• the 111111 t on the JlVOl"a~o rl:m~r UrGH 1o Ellffichint to a.!Jsurc that calculated terir­ernturcs are bolew tha 10CfH)O, Apro111Hx K limit. · · .

The maximum averaee pwnnr JJ1CniJ chmtn 1n Fl~o J.5.1 e.:ro based on c1t.lculut1u1-j c~ploy"!" in~ tho models descrlbt.t) 1n Jl<Jf.Jro11~n (1L · -Power operation vlth l.!!1~ifo at· or t.: le~ thoc~

. 1>hown in Flg. J. 5.1. u~;r.t•l'\)!l lh~ t. ·th~ :pc:ik cladding temperature followtri~ ~ pc:Jlttbtcd lc5!J-of-coolanl 11ccl<lent will not exceed th~ 2200°F llml t. Those valuos re pro:.on t. l.1 mHB for operation to ensure ccnfonn2nc0 w1 lh 10CfR50 and Appendix K 0:'.ly tf they n1·0 rr.ore llmi Ung than other design pur<lrr.ctor-:.i.

(1) "Loss-of-Coolant Accident J\nalysis Report for Dre~den Units 2, 3 and Quad-cities Units 1, 2 Nuclear Power StAtions (Le~d

Pla.nt) "~ NED0-24046 Au9ust: 1977.

DPR-25

J~

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The 11..J.J(!muin rlv~1-atQ planar UIGJts piot ieci 1n F1t;. ), 5.1 nt h!ghe~ expogure& l'\?S\llt in ~ e cnlcu)Itt..:cl r..o.:tk clad temperature of lc:aa than 22ooor.· ~~ However tho rna:d1t1uJ11 averag-o plJrnar PIGJ\a P.f@ ahown cm FiB. J. 5.1 as llmi t.~ l:i'.!cnu;.a con"for~nce calculaUons havo

· ~ot b~c:1 performed to ju5Ufy operation at UIGHQ in e~ccqs of those ohown.

J,ocal lJlr.Jl --...-

Thie spcc1f 1,cat1on '"8Bures that th@ na:<lmun linear heat generation ~te in

. nn)' rod lo loo~ ~han the de5lgn ·uneur

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PLANAR AVERAGE EXPOSURE (MWO/T)

FIGURE 3.5-1

MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. PLANAR AVERAGE EXPOSURE

.. (Sheet 1 of 3)

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0 10,600. 20,000 30,000

PLANAR AVERAGE EXPOSURE (MWD/T)

FIGURE 3.5-1 (Sheet 2 of 3)

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o. 10.000 30,000

PLANAR AVERAGE EXPOSURE (MWD/T)

FIGURE 3.5-1

MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR VS. PLANAR AVERAGE EXPOSURE

(Sheet 3 of 3)

'-·

QUAD-CITIES DPR-29

3.5 LIMITING CONDITION FOR OPERATION BASES

A Core Spray and LPa Mode of the RHR System

This specification assures that adequate emergency cooling capability is available whenever irraJi ... ted fuel is in the reactor Vellscl.

Ba•ed on the loaa-ot-coolant analytical methods described in General Electric Topical Report NE00-20566 and the apcci!ic analysis in NED0-24046. "Loss-of-Cool~nt Analysis Rcpgrt. for · Drellden Units. 2. 3 and Quad-Cities Units 1. 2 Nuclear Power Stations (Lead. Plant)•. August 1977. core. coolinq systems· provide sufficient coolinq to.the· core to dissipate the energy associated with the loss-of-coolant accident. to limit ca.l.c•l•:\gd

. fuel cladding temperature to less than' 2200°F •· to assure that core qeometry remains intact •. to limit claddinq.metal-vater reaction to leas. than l"1 and. to. limit. the: calculated local metal-water reaction to less than 17%.. ·

·The limiting conditions of operation in Specifications 3.SA I through 3.5.A.6 specify the combinations . of operable s4bsystems: to as.sure the availability of the minimum cooling systems noted above. No single failure of ECCS equipment occurring during a .loss-of-coolant accident under these limiting conditions ofoperation will result in inad~uate cooling of the reactor core..

Core spray distribution has. been shown~, in full-scale· tests of systems similar in design· to that of Quad-Cities I and 2. to exceed the· minimum. rc"quirements by at least 25%. In addition. cooling efrcctivenes5 has becn·dcmonlitrated at less th;in half the rated· flow in c;imulated fuel a~.~cmhlics with heater· rod~ to duplkarc the li<"cay hc.-at char;Ktcri~•io of irradiatC'J fuC'I. Th<" an·ilknt ;a11;1ly~•~ •~

additionally conservative in that: nu crc:Ji1 is talcn for spray cooling. o~ the reactor core ' before the internal pressure has. fallen to 90 psig. . .

The LPCI mode of the RHR system is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident. This system functions in combination wuh the core spray system · to prevent excessive fuel cladding temperature. The LPCI mode of. the RHR system in combination with the core spray subsystem provides adequate cooling for break areas of approximately 0.2 ft2 up to and I including 4el8 ft2, the· latter being the double-ended recirculation line break with the equalizer line • between the recirculation loops open without assistance from the high-pressure emergency core cooling subsystems.

The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The· method and concept are descri~ in Reference I. Using the results de·•eloped in this reference; the repair period is found to t'il! le!is than half the test interval. This assumeli that the core spray sub.o;ystems and l.P('l l."<'nstitutc a. lme-out-of-tW\\ system; however. the combined eftect of the two systems, to limit eu·essi\·c daJJing lClllJ'ICr.uun: mu."t also ~ consiJcn:d. The test interval sp'-'<:ilieJ in Spccifl1.:ation 4.5 was .l months. l"herefore. an allowable n:pair pcrio-J which maintains the b.tsic risk considering single failures sh\luld be less than 30 Jays. and this specification is within this period. For multiple failures, a shorter inten·al is specith:J; to improve the assurance that the remaining sylitems will function, a daily test is called for. Although it is recognized that. the information given in Reference I provides a quantitati\·c method to estimate allowable n:pair times, the lack of ope;:ating data to support the analytical approach prevents compkte ac~cptance of this method at this time. Therefore, the times stated in the specific items were established with due.regard to judgmenL

Should one core spray subsystem become inoperable, the remaining core spray subsystem and the entire. LPCI mode of the RHR system are available;: should the need for core cooling arise. To assure that the · remaining core spray. the LPCI mode of the RHR system, ·md the diesel generators are available, they are demonstrated to be operable immediately. This demonstration includes a manual initiation of the pumps and associated valves and diesel generators. Based on judgments of the reliability of the remaining systems, i.e., the core spray and LPCI. a 7-day repair period was ob~ined.

3.514.5-11

9· QUAD-QTIES DPR-29

Should the loss of one RHR rump occur, a nearly full complement of co.-e and containment cooling equipment is a.vailahlc:. Thn.-c RH R pumps in conjunction with the core: srray suhsystem w_ill rierform the core cooling function. Because of the: availahility <if the majority of the core cooling equipment. wh!-:h will be demons.trated to he: opc:rahle, a 30-day rerair period is justified. If the: LPCI mode of the RH R system is not available, at least two RHR pumps :must be available to fulfill the containment cooling function. The 7-day repair period is set on this basis. · .

B. RHR · Senic:e Wat er

The containment cooling mode of the RHR system is provided to remove heat energy from the containment in the event of a loss•of-coolant accident. For the flow specified. the containment long-:errr. pressure is limited to less than 8 psig and is therefore more than ample to provide the required beat-removal capability (reference SAR Section 5.2.3.~ ).

The containment cooling mode of the· RHR syst~m consists of two loops. each containing two RHR service water pumps. one heat exchanger, two RHR pumps, and the associated valves. piping. elec.:trical equipment~ and· instrumentation. Either set of equipment is capable of performing the containme~t cooling function: loss of one RHR service water pump does not seriously jeor:udize the containment cooling. capability, as any one of the remaining three pumps c;in. s;1tisfy the cooling rc4uircments. Since there is some rcdundmH.j' left~ a 30-day repair pl.'riod is adequate. Loss of one loop of the wntainment cooling mode of the RHR system leaves one rcrnaining system to perform the containment c0oling function. The operable system is demonstrated to be operable each day when the above condition_ occurs. Based on the fact that when one· loop of the coniainmen1 cooling :node· of 1he RHR system becomes inoperable, only one system remains, which is tested daily. a 7~day repair period was specified.

. . C. Hlgh-Pr~surc Coolant· lnj"-ction.

The high-pressure coolant injection subsys1em is provided to adequately cool the core for all pipe breaks smaller than. those for whiCh the LPCI moJe of the: RH R system or core spray subsystems can protect the· core;

The HPCI meets this requirement without the use of offsite electrical power. For the pipe. hreaks for which the HPCI is intended to function, the core never uncovers and is continuously cooled, thus no cladding damage occurs (reference SAR Section 6.2.5.3 ). T~e repair times for the limiting conditions of operation were set considering the use of the HPCl as part ?f the isolation cooling sys1em.

0. Automatic Pressure Relief

The relief valves of the automatic pressure relief subsystem are a backup to the HPCI subsystem. They enable the·corc spray subsystem or LPCI moJe of the RHR system to provide protection agains.t the smJll pipe break in the event of HPCI failure hy Jerre'\surizing the reactor vessel rapidly enough to actuate the core spray subsystems or LPCl nwJe of the RH I{ syst\."m. The core spray suhsyMem and tht: LPCI ·1· mode of the: RHR system pw"·iuc ).Ulhcient llow of nivl:.tnt to limit fud cladding temperatures1t1less than

2200°F, to assure that core geometry remains intact, to limit the core wide clad metal-water reaction to less than 1%~and to limit the calculated. local metal-water reaction to less:. than 17°/o.

E.

Loss of l of the relief valves affects the pressure relieving capability and, therefore, a 7 day repair period is Sl;)ecified.

Loss of more than one relief v~lve significantly reduces the pressure relief capability~· thus a 24-:hour repair

· period is specified based· on tJ:ie HPCI system availability during this period.. ·

RCIC

The RCIC system is provided to suprly continuous makeup water to the: reactor core when the reactor is isolated from the: turhini.: and whl:n the lel:dwatcr sy)otcm is nm available. Under thc:sc: conditions tht: pumping capacity of the RCIC system is suttil:icnt tl> mainl;i.in the water level above thi:: 1.:orl: without :my other water sys1c:m in opera1ion. If the water lt~vcl;in the reactor vc~sd Ji.:1.:rc:a~es to the RCJC initiation leveJ, the sysu:m autornati1.:ally starts. The sy!>tem inay Jbo be manually initiated at any time.

I I

l

QUAD-CITIES DPR-29

For core ftow ~ates lcs.'> than rated. the steady ~late MCPR is increased hy the formula given in the · specification. This assures that the MCPR will be maintained greater than.that sp~ifi::d in Sp~~ification I.I.A even in the event that the motor-generator set speed controller causes the scoop tube pos1uoner for

. the fluid coupler to move to the maximum speed position.

References J. t M. Jacobs· and P; W. Marrin, GE Topical Report APF.D-5736, 'Guideline~ for Determining Safe :rest

Intervals and Repair Times for Engineered: S;1fcguanJs; April I %9.

2. "Loss:-of:-Coolant-Accident Analysis Report for uresden Units 2, 3 and Quad-Cities. Units. 1, 2· Nucle~r Power Stations (Lead Plant)", NED0-24046,

August 1977:.

3. GE Topical Report NEDM-10735, 'Fuel Densification Effects on General Electric Boiling Water Reactor Fuel: Section 3.2.l, Supplement 6, August 1973. .

4. J. A Hinds,. GE. letter to V. A Moore, USAEC. 'Plant E~aluation· with GE GEGAP~ITV December 12. 197):;

5. USAEC Repor4 'Supplement 110 the Technical Report on Densification· of General Eiectric Reactor Fuels.' December 14; 1973.

.·

3.514.S-IS

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PLANAR AVERAGE EXPOSURE (MWO/T)

FIGURE 3.5-1

MAXIMUM AVERAGE PLANAR LINEAR U[AT <;[N(RATION RATE (MAPLHGR)

. VS. PLANAR AVERAGE EXPOSURE

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.... c( w x a: <­'-' .... z~ ~- ........ ...J 3"

¥ a::­c(. zw· c( ...... ...Jc(· ~ a::. wz: C)O ce- II a: ...... UJc( >ex <w

z ·~w

;:) C) 2:

I --,· ··· ·- ·- - - - -··-·· 1---·. ---- -- .. - ·- -·- •. : j • l -------·-····-· 1· .. : . ...... ··- ······ i ·-·- .. .

.. ! --

: ------ . 1-. I

. . . --- -- ·-· --- -·· -------

J· I

I i

. ..

. I

I

- I

I I ! . I

i

r-------~-1.. . . . ; ..... .

I . '

' .. I i ! i

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______ 1_~ ____ .i ________ _t_ __ -· -·--- -- ··--· ----- ·-------~-- -------. 1' : . I I

. : : . ! - . . . ~. . ; 1 · . . i ' i . . .. . . -. -· . f .. ~- ... l.

-~----~.-~'. ___ -l _ _. __ ;·---:-~~~---:~-1 _··:·-_: -:~~~~---r~--.- --~ -·:·---~--1----

i i. :· I I .. ····· r- . i

-~-. r

10~.----------------...... --------------~----------------1 0 10,000 20,000 30,000

PLANAR AVERAGE EXPOSURE (MWO/T)

FIGURE 3. 5-1 (Sheet .2 of 2)

MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. PLANAR AVERAGE EXPOSURE

QUAD-CITIF.. DPR-JO

Based on the facl that when one loop or the containment cooling mode of the RHR systcm l'lc!rnmes inoperahle. only one system remains. which is 11.-sh:d daily. a 7-day n:pair period was specificd.

C. High-Pra..ure CcNJlant Injection

The high-pressure coolant injc.-ction suh-;ystem is providcd to adcqu;irc:ly cool the core for all pipt.· h_rcaks smaller than those: for which the LPCI mode of the RHR sys1cm or cClfe spray suhsrcems <.:an protect the

·core. '

The HPCI mc.-cts this requirenu:nt without the use ofotf.iite ch:t·tril·;al p.1wer. h1r the pip.: hr1.·ab li1r which the HPCI is intended to function. the core never unco\'crs ;ind is continuously coolcd. thus no dadJin~ damage occurs< rcl~rent·e SAR Sc.-ction ti.:?.5.J ). The repair times for rhe limitii1g rnnditilins or" opcrati\1n were set '-'Onsidering the use of the H PCI as part of the isolation cooling. system.

0: Automatic Pressure. Rl'licfc

_The relief valves of the automatic pressure relief suhsystem are a had .. up to the H PCI suhystem. Thev ~nahle· the core spray suhsystem or l.PC'I mode ol' th1.· R If R s:-stem co·pro\'ide proll.'rtion ;1~•;1 i nsl Chi.' ,mall pipe hreal. in the l"\'l'lll of llPCI fa1hm.· hy lkpr1.·ssuri1.1111,'. llw real·tor Vt.'ssc! rapidly enough II• ;1l"111;11c !111.·

'core 5pray suhsystems.or. LPCI modt.: of the IUIR 'Ystem. The core spr:iy suhsysccm a.nd thc LPCI inoJe of.the RUR. system provide sutlicient !low of coolant to limit fuel daJding temperatures to I less-..t.~n 2200°F·, t<;> assure that core geometry remaining intact,. \ to ll.II1it the core.wide. cl.ad metal-water reaction to less than 1%,. and'·:to limit the calculated local metal-water reaction to less than 17% •.

Loss:· of l of the relief valves affe~"ts the 1?ressure relieving· capability and,. therefore,. a: 7 day repair period. is specified~. Loss· of more than one relief valve significantly reduces; the· pressure· r.e·lief. capability, thus a. 24-hour repair period. is specified. based on the HPCI system availability during this period.

E .. RCIC

Th(' RCIC s):stc:m is prnviJeJ to· supply conti111ll•t1s 111:11..eup wall'r en lhl' r1.·:1l·tor uire wh1.·n !11.: r1.·:1l'11 ir is isol;steJ from the turhinc and when the lccJwacer 'Ysh:m is not a\':iilahle. Undl·r these: l"1111Jitions the pumping capacity of the: RC'IC system is sutlicient to maintain the w;11er kvel ahlwc the: core wi1lwut :in~ other water system in operation. If the water kvcl in the. reacmr n.-s.;el dene:ises to tnc R('f(' 1111uau11n level. the system- automatically starts. The system may also he manually initiated at :my timc.

The HPCI' system provides an ~hcrnate methllJ 11f supplying makeup-water to the rc<ll"tor shnuld the normal feedwater become: unavailaolc:. Therefore: the s~ilication calls_ for an operaoility check nf the HPC'l:system shoulJ· the RC'IC system ti< found to ti< inl'pt:rablc: ·

F. f.111«,,Ct:~y Coolin~ An1ilabilit~·

The purpose llf Sr\\:l·i1i,·;1lil•ll .l:'.F j, w assure .1 111ini111u111 ,,r n're c1>1,lint! cyuipmcnt is :1vaibhlc at :.ill 1i11u:s. If. li•r e~ampll.:. l•111.· 1.·l•l'I.' ,pr;1:--· wl."n: 11ut l•l'sl·rv11.·1.· .111J Lhl· J11.""d w_h1d1'1'\\•W1."fl'J the ''l'l'\\'"lll' l11rc spray were \lUt 11f sel\·ir1.·, l•uly LWl• }{Ill{ pu111ps w1111IJ Ix: av:11lahlc. Lil..ew1,.l'. 11' tw•• KHK J'L:lll:'" ''..:r..: out or ~rvicc anJ tWll RH!{ 'ervil·e \\o;llcr pumps 1111 thl' opp.1s1Le ,..iJe wen: :1ls11 llUt 111' s..:rvi\.·c.: lh' containmenl ,·l~lli11g _w11ulJ tx· ;1\-.11lahk' .. It i~ during r1."f11ding •.•11L;1;;cs 1h;1t 111aj,1r 111;1in1.:n.11H.:c is performed :111\J dun11r "1d1 tilllC Lh.IL .ii( l,1\\"(lll.""'lll'I.' l·l're l."l•l•lin:; ,.\,..(l.'111" 111.ly he ,iu't lli' 'l'~\'il."<.:. rhi., spccitkatillll (ll'll\;1J1.·s 1h.11 slwuld 1111,. 11l·l·ur. ill> \\llrl.. will he p1."ri\•r111cd· 1111 thl." pnm.1r~ ,y,1..::11 \\ 1111.:h could lead to Jrainin!! the \'esscl. 1'11is WlHI.. w11ulJ induJ..: wurk llll lC.:rt;1i111.·l1111rlll n1J Jrivc l'iltnp-'111.·nts anJ recirculation svs~em. Thus. the sp..:cilic;llilln prcduJc:. thc C.:\'Cnts which wulJ r..:4uire wr..: l'• 111lin;;. Specification J. 9 n;ust also he consulted Ill J..:terminc.: 111her re4uirc:ments lor Lhc.: Ji..::.d gc.:neraturs.

Quad-Cities Units I and 1 share ccrtain pml·ess systems such as th.: rn;1k.:up Jeminaalizc.:rs and the.: radwaste system :ind also slllllC ,..11Cty systems such as the s1anJhy g:is trcatmi:nt systcm. han..:ries . .:nJ

3.514.5-12

.. . ... "'··•· ~ ••

QUAD-CITIES DPR-29

H. Conden. .. ate Pump Room Flood Protection

See Specification 3.5.H.

L Average Planar LHGR

This specification assures that the peak c!adding_ temperature following the postulated design-basi~ loss-of-coolant. accident will not exceed the 2200'-'F limit specified in· the 10 CFR 50 Appendix K

considering the postulated effects of fuel pellet densification. . .

The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat-generation rate of all the rods of a fuel assembly at any axial location and is. only secondarily dependent on· the rod.do·rod p<lWl'r distrihuti1m within ;:n assl•mhly. Sinn· exrcctcd loc.11 variations in power JiMrihutiun within a ti1d ;assl·ml•lv :illi.·l·t tlw l·akul;1tl·,r pl·;1k rbddinr tcmrnaturc bydcss than t20 ·· r relative to the peak tempc:ratun: ti1r a typir;1I fuc:I design. th'-· limit on lhl· avcr;1r,c: planJ LHGR is sutlicicnt lO assure that calculated tcmpcraturc:s arc hclow the limit. The maximum average planar LHG R's shown in Figure 3.5-1 are. based on calculations employing the· models described in Reference 2 • ··

J~ Local LHGR

This specification as."iurcs that the. maximum linear heat-generation rate in any rod is less than the design linear heat-generation rate even if fuel relict dcnsitication is po~tulated. The power spike .penalty specified is based. on that presented in Reference 3 and assumes a linearly im:rcasing variation in axial gaps between core bottom and top and assures with a 95% confidence that no more than one fuel rod exceeds the design linear heat-generation rate due 10 power spiking. An irradiation' growth factor of 0.25% was used as the basis for determining !::.IP in accordance with References 4. and 5 •

K. Minimum. Critical Power Ratio. (MCPR)

The: steady state values for. MCPR specified in this specification were selected to provide margin to accommo­date transients and uncertainties in monitoring: the. core operating state as well as uncertainties in the· critical power correlation' itse.lf. These values. also assure that operation will be sud1 that the initial condition assumed for the LDCA analysis. an MCPR of LI 8. is satisfil'd. For any of the spedal set of transients or disturb:mces

·caused· by single operator ·error or single equipment malfunction, it is required that design analyses_ initialized. at this steady-state operating limit yield a MCPR of nut less than that. specified in Specification I. I .A at any time during the transient; assuming instrument trip settings given in Specification 2.1. For analysis of the thermal consequences of these transients;. the limiting value of MCPR stated in this specification is con­servatively assumed to exist prior to the initiation of the transients. The results apply with increased con· servatism while operating with MCPR's greater than specified.

Tho most limiting tran~nts with. res~ct to MCPRare generally:

a) Rod. withdrawal error

b) Turbine trip without bypass.

c) . Loss of feedwater heater

Several factors influence which of these transients results in the largest reduction ·in critical power ratio such as the specific fuel loading, exposure. and fuel type. The cum:nt cycles reload licensing submittai specifies the limiting transients for a given exposure increment for each fuel type. The ,·alues specified as the: Limiting Condition of Operation are cunservativc:ly chosi:n as thl.l most restrictive over the entire cycle for each fuel type.

3.S/4.S-14

'

•'/

~-··

,-···.,

I

QUAD-CITIES DPR-30

• For core flow rates less than rated, the steady state MCPR is increased by the fonnula given in the specifi­cation. This assures that the MCPR will be maintained greater than that specified in Specification I. I .A even in the event that the motor-generator set speed controller. causes the scoop tube positioner for the fluid ..:oupler to mow to the maximum speed position.

References

1. "Loss-of-Coolant Accident Analysis Report for Dresden Units 2,· 3 and Quad-Cities Units 1,. 2 Nuclear Power Stations (Lead Plants)", NED0-24046, August 1977' ..

., ....

3.

4.

GE Topical Report NED0-20566~ 'General Flectric Cempany Analytic-.d Modi:! for l0Shlf-Cu1..1lant Analysis

in Accordance with 10 CFR 50· Appendix K.'

L M~ Jacobs and P: w: Marriott; GE Topical Report APED 5736. 'Guidelines for Determining Safe Test

Intervals and Repair Times for Engineered Safeguards,' April 1.969. .

•fuel Densitication Effects on General. Electric Boiling Water Reactor Fuel.' Section. 3.2.1. Supplement 6.

August 1973: '.

J. A. Hinds. G.E. Letter to· V. A. Moore; USAEC 'Plant Evaluation with GE GEGAP-HI; December

1973:

USAEC Report, 'Supplement I to the Technical Report on Densification of General Electr.ic Reactor Fuels.'

Dettmber t:4; 1973'.

3.S/4.S-14a