ML19268B991

Forwards Low Pressure ECCS Evaluation for 18% Steam Generator Tube Plugging

ML19268B991
Person / Time
Site:	Point Beach
Issue date:	11/27/1979
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	Harold Denton, Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 7911300505
Download: ML19268B991 (33)

Text

,

O wisconsin Electnc m com 231 W. MICHICAN, P 0. BOX 2046. WILWAL'KEt, WI $3201 November 27, 1979 Mr. Hamid R. Denton, Director Office of Nuclear Reactor Regulation V. S. NUCLEAR REGULATORY COMMISSION Washington, D. C.

20555 Attention: Mr. A. Schwencer, Chief Operating Reactor Branch No.1 Gentlemen:

DOCKET NOS. 50-266 AND 50-301 LOW PRESSURE ECCS EVALUATION FOR

_18% STEAM GENERATOR TUBE PLUGGING POINT BEACH NUCLEAR PLANT UNITS 1 AND 2 Enclosed herewith are the results of an ECCS reanalysis for operation of the Point Beach Nuclear Plant Units 1 and 2 at a reduced primary system pressure of 2000 psia with up to 18% of the U-tubes plugged in each steam generator. On the basis of generic sensitivity studies, wnich showed that the limiting break for Westinghouse two-loop plants with 14x14 fuel is a double-ended, cold-leg, guillotine pipe break with a discharge coefficient of 0.4, only the analysis for this specific limiting break for the Point Beach Nuclear Plant is provided. The results of these generic sensitivity studies were provided to you previously with our letter dated March 20, 1979.

The enclosed reanalysis was conducted at a reduced primary system pressure of 2000 psia using a core inlet temperature of a nominal 544*F. As the Point Beach Nuclear Plant utilizes upper plenum injection, a 60'F increase in temperature should be added to the calculated peak clad temperature to account for explicit modeling of upper plenum injection. Westinghouse has determined for the Point Beach Nuclear Plant that fuel rod bursting and blockage is appropriately accounted for in the ECCS analysis results. These results demonstrate that the Energency Core Cooling System continued to meet the Acceptance Criteria as presented in 10 CFR Part 50.46 of the Commission's Regulations.

As you know, an ECCS reanalysis for the 18% steam generator tube plugging limit and operation at 2250 psia was ubmitted to you by letter dated November 19, 1979. At that time, we s

.ed that Westinghouse ECCS sensitivity studies have shown that operatic. at reduced pressure of 2000 psia will have an insignifice.nt effect on the analysis results. The attached reanalysis confirms this statement.

Aco/

3

///

1 79133005c'S

i Mr. Harold R. Denton, Director November 27,1979 Should you have any questions regarding this analysis, please notify us as soon as possible so that we may provide you any necessary clarifications.

Very truly yours, C. W. Fay, Director Nuclear Power Department

_j -

5 The loss of Coolant Accident (LOCA) has been reanalyzed for Point Ccacn Units 1 and 2.

The following in formation amends the Sa fety Analysis The re-Report section on Major Reactor Coolant System Pipe Ruptures.

suits are consistent with acceptance criteria provided in re ference (1).

The description of the vcrious aspects of the LOCA onalysis is given in UCAP-E239.E23 The individual computer codes which comprise the Westingho>sa Emergancy Cere Cooling System (ECCS) evaluation made) are describcd in detail in separcte reports [3-6] along uith code modifi-cations speci fied in references 7, 9 and 10.

The analysis presented here was per formed with the Febeuery 1978 version of the evaluation r.odel which includes modificat'ic s delineated in re ferences 11, 12, 13

~: -

.. ~

A and 14.

Resug.,

The analysis of the loss of coolant accident is perforned at 102 percent'

.'o f the licensed c0re 'peacr rating.

The peak linear power and total core powcr used in the-analysis are given in Ttble 2.

'Since there is margin bettean the value of peak linecr pcwcr density used in this analysis and the value of the paak linc;r rowcr. dent.ity expected during plant opera-tien, the peak citd temperature calculated in this analysis is greater thot, the maxieum clad tenperaturc expected to exist.

Tdle 1 pre!.cn's the occurrenc$ tina for various events threughout the tci.id:ai transi.:n t.

t.

4 1

e Table 2 presents selected input values and results from the hot fuel rod

~

thermal tr,ansient calculation.

For these res'ults, the hot spot is de-fined as the location of maximum peak clad temperatures.

That location is specified in Table 2 for cach break analyzed.

The location is indi-cated in feet which presents elevation above the bottom o f the active fuel stack.

Table 3 presents a summary of the various containment systems parameters cnd structural param2ters which were used as input to the C0C0 computer codo[6] used in this analysis.

i Tcb1c; 4 and 5 present reflood, mass and'encrcy releases to the contcin-ment, cnd the broken loop accumulator mass and energy relecse to the containment, respectively. '

The results of several sensitivity studies are reported.E83 These resul.ts are fcr ccaditions which are not limiting,in nature and hence cre reported en a generic basis.

Ficurcr.1 through 17 present the transients for the principal partmeters for the break sizes cnclyzed.

The follouir.g itcms are not J:

3:.

Qu:?ity.. mass velocity and clad hcot transfer Figures 1 cocificient for the hotrpot and burst locations 6:

Coro presst re, break floc.', and core pressure drop.

figures 4 The break ficu is the sum o f the flourates from both

ends of the guillotine break. The core pressure drop is taken bs the pressure just be fore the core e

inlet to the pressure just beyond the core outlet Clad temperature, fluid temperature and core flow.

9:

Figures 7 The clad and fluid temperatures are for the hot spot and burst locations Downcemer and core wcter level during reficed, and Figures 10 - 11:

flooding rate Figures 12

~13:

Emergancy core cooling s.ystem flowrates, fer both accumulater and pumped safety injection Figures 14 - 15:

Containment pressure and core power transicn'ts Figy. E,16 - 17:

Brcck energy release during blewtown and the con-tainment w.'11 condensing hotst transfer coef ficient for tLo ucrat break d b O

G e WM e t@4e s= etw gynesee e seeme $4

• seeeG9 ##M ** % 498thanger egeoem N e op. ghet **w compyw us em este Meaght

--may* Ne ebmPenm m..

Conclut ione 1. Thermil An'ai ytit for breakt up to and including the double ended severance of a reactor

~

coolant pipe, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10CTP.50.45.II)

Thac is:

1.

The calculated peak clad temperature does not exceed 2200*F based on.

a totc1 core peaking factor of 2.32.

2.

The amount of fuel element cladding that reccts chemically with n!ater or sicam does not exceed 1 pet cent of the total imcur.t o f P.

p Zircalley iri the reactor.

zj'lj.

_'f.

f>,

3.

The clad temp:raidre transient is termir.atcd at a time when the core geomatry is still amencble to cooling.

The cladding oxidatic.)

limits of 17 perr.t are n6t exceede'.' during or after quenching.

4.

The core tempcature is reduccd and decay heat is removed for an extended period of time, as required by the long.. lived radioactivity remaining in the core.

The ef fects of upp;r plenum injection for l'qstinGouse-designed 2 loop plants has beca discussed with the sta f f.[15,16.17,b.19]

Based on interi: calce'.atiorf., a 60*F increase in calculated resh clad tem;::r-atures results from c::plicit modelling of uppcr pienem injection in the Point B.'ach power plent.

In ordtr to use the presc..t L'estinghourc ECCS cvaluation modelEI3d,15] to ;,nalyza a postulated LOCA in the

........,....s.....,,

,p.

.....,,m..

Point Beach plants and remain in compliance with 10CfR50.46,a limit o f It can be 2140*F on calcula':cd pcck clad temperatares tnust be observed.

seen from the results contained herein that this ECCS analysis for the Point Beach power plants remains in compitance with 10CTR50.46.

.~ ;W s o

i L.'-

j Re ferences for Section 15.4.1 e

1.

/seceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled tiuclear Power Reactors," 10CFR50.46 and Appendix K' of 10CFR50. /.6.

Federal Register, Volume 39, t: umber 3, January 4,1974

2., Dordelon, F.

M.', Massie, H. W., and Zordan, T. A.,

"Wes tinghouse ECCS Evaluation Model-Summary," WCAP-C339, July 1974.

3.

Bordelen, F. M., et al., " SATAN-VI Program: Comprehensiye Space-Tima D pendent Anclysis of L0ss-of-Coolant," WCAP-8302 (Prop-ri dary Yr.:r; ion), June )974.

4.

Bordelon, F. M., et al., "LOCTA-IV Program: Lost-o f-Coolant Tran-sier.t Analysis," WCAP-B;01 (Proprietary ycrsica), WCAP-3305 (Iton-Proprietary Version), June 1974..

5.

Kelly, R. D., et al., " Calculational Model for Core Re flooding After c Loss-of-Coolent Accident (UREFLOOD Code)." WCAP-8170 (Proprict'ry Version), !! CAP-8171 (lion-Prs prietary Version), June 1974.

6.

Gardelon, F.

.',' and l'erp'ay, E.

T., "bntainmr : Pres sure Anclysis Cada (C0CO)," !! CAP-07.N (P opriciary Versicn), WCAP-3326 (f:en-Proprintary V(rsion), June 1974.

7.

Pordelon, r. M., et al., "The !!cc.tinghouse ECCS Evalua tion Model:

'Supl.lcmantary In forraation." WCAP-0471 (Proprietary Versio*,),

l! CAP-01 <!? (Un-Propr iL t 'ry Yere irn), January 1975'.

e Salvatori, R., " Westinghouse ECCS - Plant Sensitivity Studies,"

8.

WCAP-8340 (Proprietary Version) WCAP-8355 (Ucn-Proprietary Ver-sion), July 1974.

" Westinghouse ECCS Evaluation 14odel, October,1975 Versions,"

9.

UCAP-8622 (Proprietary Version), WCAP-8523 (Non-Proprietary Ver-lsion),thvember,1975.

Letter from C. ticheldir.ger of Wcstin:hou's,c Electric Corpcration to 10.

D. B. Vr.r.salo o f the !!ucle:r Regulatory Commissien,' letter nunber ifs-CE-924, Janucry 23, 1976.

Kc11y, R. D., Thomps.on, C. M., et. al. '" West:nghouse Emergancy 11.

Core Cooling System Evaluatica 14odel for Analyzing Large LOCA's During Operatien ',!ith One Loep Out of Service for Plants Without Loop Isolation Valves," WCAP-9106 February,1978.

12 Eicheidinger, C., " Westinghouse ECCS Evaluation Model, february 1978 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-A (l;on-Proprictary Version), February,1078.

13.

Lette

' rom T..lj. Anderson o f Wes tinghouse Electric Corpcratinn to Jchn Slo 17. of the Nucicac Regulatory Commission, lettcr number t3-111A.1031, Nov.1,1978.

14. Letter from 1. M. Ander soa o f' Wes tinghous: E1cetric Corpor.; tion to R.1. Tedeco o f the Nucicar Regulatory Commission, letter nuraber IG-illa-2014,. Decenter 11, 1978.

" Safety Evaluation Report on ECCS Evaluation !Iocci for Westingho 15.

Two-Loop Plants," November,1977'.

16.* Letter from V. J. Esposito to 11. W. Gutzman of NSD Operations Sup port, both from Westinghouse Electric Corporation, letter number SE-SAI-2267, January 30, 1978.

"HRC Questions Regarding TAC 1/16/78 Submittal by Westinghouse 17.

Design;J Two-Loop Plant Operators," Februa.ry 1,1973.,

10.* Letter from '. J. Esposito to H. W. Gutzman of NSD Operations Sup-number port, both from 1.':stinghouse Electric Corporation, letter SE-0 Al-2290, February 17, 1972.

"Sa fety Evaluation Repar'. on :nterim ECCS Evaluation Model fo.-

19.

Wc:.tinghout : Two-Loop Picnts," March 197S.

• Wiscont.in Electric Power should supply the proper references by which these re ferences were formally transmitted to the NP.C.

e

TABLE 1 T-LARGE CREN' - TIME SEOUTIlCE OF EV!rlTS Occurrence Time (Seconds)

Event DECLG, CD - 0.4 Accident initiatien 0.0 Reactor Trip Signal 6,5. 3'7 Safety Injection Signal

.8 St.cet Accuculator Injection En 9,7 Er:d of ECC Bypass iT.*D 2i.'l End of Tileedoien C0 11.'1 Ecttom of Core Recovery C,a 42,1 Accer !1ator:, Empty C/:3 57.2 Ste : rumped ECC Injection 25.8 4

e 6

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e TABLE 2 e

LARGE tREAK - AfiALYSIS IllPUT A!:D RESULTS_

Ducn ti tice, in the Calculations _:

102 percent o f 1518.5 it4t Licensed core power rating 2 32 Totc) cope peching factor 102 p:rcent o f 13.23 kw/ f t Port lineer pouce 1200 cubic fcet per tent t.cce+ulator water volume 700 psia Acccmuistor tressu e ibt.ber o f sa foty injection pumps c;e ating 2

Steua Lencrcior tube pivggint; level 10 percent (uni form) foci pr.rctasters -

Cycle.:Generi,c P.cgior.t:wie DECLG, CA - 0. 4 f:esuM:

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Pcok clad tempcrature (*F)

CI3 2042.

7.5 '

I ocatien (feet) 14aximum Iccal clad / water reuttion (percent)

S'.'* 1

-3 7.5 Locction (feet)

Total core clad / water reaction (percent)

<0.3 T3

tb llot red berst time (seconds) c 1.orttic. ((cet) 5.7!i

.. p:

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TABLE 3 T

C0tiTAlt iEllT 0TTA (DRY Cor:TAlt iErlT) 6 f3 1.055 x 10 liet Frco Volume Initial Conditions 14.7 psia Pressure 90 *F Temperature 34 *F R1!ST Temp:rsturc 33 'F Servicc thter Temptrature

-25,*F Outside Temperatur>

Sprt.y Sys tr.m lumbcr of rump: Operit'in g 2

Runo.it F10 ' l'c te 1950 grm each 10 secs A:tuation Tiir.e e

Safcp.'erds Fan Coolers IM.6cr of Fcn Coolers Opsrating 4

Fastest Fos.t Acciant Initiation of f an 35 secs Coolers 4

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TABLE 3 (Cont) g.

STT!UCTU'tAL HEAT S117. DATA

• 2 Area (t _

Thici: ness (in)

Material 56020

.322 Stec1/ concrete, 12 6230

' Steel / concrete,12 f$(./. $,).'.*.)

.188 2400 Steel /cencrete, 12

.25 690 Steel / concrete,12,

.100 103724

.094 Steel 11710

.3?4 Stee) 4730

.443 Steci 5441

.SG)

Steel 4490

.712 Ste:1 1.0 Steel

,' 957 3667 2.034 Steci 10221

.125 Steel 16551

.20?

Steci 2707

.5 Steci 13825

.322 Steel 141100

.055 Steel 38200

.036

~ Steel

.1 Stenl / concrete, 3

,19500 19500 3.0 Concrete 6:500 30.0 Concretc

....s.

f TABLE 3 (Cont) s PAINTED STRUCTURAL HEAT SIMP, DATA Structural Heat Sink Structural Heat Sink Paint Thickness Suriccc Area (Ft )

Thickness (In)

(Mils) 2

.322 7.5 56020 2400

.25 7.5 103724

.094 7.5 11710

. 30 ?.

7.5 4730

.443 7.5 5441

.501 7.5 I

'I{'jp 4490

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Time (Sec)

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