ML19241B369
| ML19241B369 | |
| Person / Time | |
|---|---|
| Issue date: | 05/15/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| FOIA-79-98 PP-770515, NUDOCS 7907160249 | |
| Download: ML19241B369 (49) | |
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I 9/ C E l'ower Systems Tcl 203/Gta.1011 i 8.'.0 I U. 75 Com*,u;non Engm.:ering. Inc. Teles: 9 9297 1000 l>rorpect I hl! Ilomt y. Wir.d.;or. Connect cut OG035 4 I Q. fTlPOWER Nan 2,50 d T, Z!. ' a... ] SYSTEMS j i September 18 J* TD-CE-l h, 1975 d 9/-24 1~ ,,, 9 a ennessee' Valley Authority Y///A*Tg.//;*.*. 204 Union Building DI Knoxv'ille, Tennessee 37902 ]gi -- f---- i Attention: lir. D. R. ' Patterson l YMT' ~ Chief Mechanical Engineer J G " Centlemen:
Subject:
Nuclear Steam Supply Systems TVA'X-24 and X-25 Installation TVA Order No. 60-84840 3 CE Contracts 14074 and 14174 -.g,yn D,d,. i .-u., .m (~ - .E ',,f SMALL BREAK LOCA s r; '. k' ,s .[# M Ref. (A) : D.V. Craf to D.R. Patterson, I () p$ i ' ([(G -' ' -. TD-CE-117, 6/11/75, subject: L-Ifecting Minutes of 5/27 & 5/28 in Windsor I During the Energency Feedwater Meeting of 5/27 and 5/28,1975, C-E agreed to pro-vide core information concerning our analysis, of the subject accident. This commitment was documented in Item 23 of Reference (A). The nttached report de-scribes the results of our analysis and we expect that it should s'atisfy your l concerns. If you desire more information on this subject, please contact Mr. F. Z. Stiteler in Knoxville. N 191973 --DEP:AH .CC: C'J:'-le, 401 U3-K Very puly yours, PJiEo3;es,1G0 D3-K SEc lor. :07 P2-K ggg9 I:.:'on;;;c.'.n'/, 101 AD-K T.!!Pierco, 015 L3-X (2) R. Lumpkin, J W EC2a :'11, 423 lUJ-K Project Manager 7 ~ hM/22TashP:,433MI3-K j"B h % Fe, 507 iliT-K DRP:AH Jd N Att. 9,g',,.l!})hj.r.c3_. fad cc: W. B. Wade, TVA, u/o att . < t.4 4 'y/ J. Cinder, CE, v/o att C '2 g. "l~~~~~Il-~~il ~~ i ._! - ' -4 "i l-t"
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. 'T e7 2L / / SMALL BREIK LOCA Introduction In a meeting held in Windsor on 5/27-5/28/75, Mr. Sabin of TVA expressed a concern for a class of small break LOCA's whose depressurization rates 1 are relatively slow compared to those small breaks presented in CESSAR, Section 6.3.3.3. He worried about the possibility of inadequate HPSI flow for such breaks and also asked about the LOCA requirements for Em-ergency Feedwater and operator actions. i This transmittal constitutes a response to action item number 23 of Ref-erence 2. -r y Summa ry - l The results herein demonstrate that for any cold leg break less than or eq6al to 0.1 ft2 in size, adequate liquid inventory is maintained in the reactor vessel to keep the active fuel completely covered at all times. i The results shown in CESSAR demonstrate that for_those small breaks in which some degree of core uncovery is predicted, the ECCS criteria 3 are met and peak clad temperatures are less than 12000F, It has been assumed that for all breaks no operator actions are performed. At some later time following the LOCA, operator procedures for long term cooling will be implemented. This meco, however, does, not address the long term cooling phase of the accident. Long term cooling performance 1.*. and operator procedures will be discussed in a future submittal,;g, gg i var ~ h e
.j J ~ ~ 'Since NRC requires the assumption of loss-of-of fsite power, main feedwater flow is not available. F.or breaks larp,er t'han 0.1 ft2,'those presented in CESSAR, Emergency Feedwater (EFW)'was not represented and indeed was foun'd to have little influence on the performance of such-relatively large break sizes. However, for breaks equal to or smal.ler than 0.l'ft2,'EFW was found abs,olutely necessary. For tihese smaller' bicaks,.the steam generators must remove a significant part'of t'he. decay energy ginerated in the core'. If t' hey'do not, then primary side re-pressurizat. ion occurs since 'the break is too smail 'to emit all of the steam boiled off in the core., s. , 3 . Discussion Calculations of breaks 0.1 f t2 and smaller have been performed with the. CELEVEL code. This code was specifically written for application to such small breaks. ~ Q, h The. predicted two phase level transients for several breaks.0.1 ft 2 and smaller are shown in Figure 1. The corresponding pressure transients are ~ giv en in Figure 2. These results demonstrate that. although the depres-surization, rates for such small sized break sizes are gradual, adequate HPSI pump inflow is provided to keep the core covered at all. times. ~ s The Safety Injection Tanks (SIT) become activated at approximately 900 ~ acconds for the 0.1 ft2'. case. Neither the.05 ft2 nor the.005 f t2 cases resulted in SIT activation before the level transient is turned around. For both breaks, the level transient bottoms out at an elevation above the top of th'e activei fuel and recovers from there via HPSI inflow alone. ~ For all of these breaks, steam generator heat transfer plays an important role. Since EW is supplied to each steam generator, an adequate secon. dary liquid inventory is maintained to support heat rem 6 val by condensa- '7 . tion of the steam in the U-tubes on the primary side. The secondary ~ ~ G522b4 i --eom.--
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inventory is assumed to be saturated at a temperature correspondirig / 'to the lower setpoint of the pfety valves (1270 psia). The con-2 densation film coef ficient used was approximately 200 BTU /hr-f t op, - This is a conservative value which properly accounts for the.in fluence of non-condensibles. ~ - To cicarly indicatc the influence.of EW, Figure 3 has been included. * * ~ Results are shown for the reactor vessel pressure. and. level transients for ~ cases with and without emergency feedwater. As shown, if-EFW is not available a_re-pr.essurization occurs W ich in turn reduces HPSI inflow, and the level transient is significantly aggravated. Neither case shown } in, Figure 3 should be. construed to specifically represent the performance of,. System 80. EW is available on all System 80 plants.. The case without EFW is',shown only for demonstrative purposes. For both cases shoun, the HPSI capacities assumed were much less than those of System 80..Thus,, the Icvel transient are nore severe. ~ Ma[orAssumotfons ~ As required by NRC,, the folloding conservative assuiaptions were ' applied in analyzing all small' breaks (< 1.3 f t ): 2 1. Loss of:offsite power assumed'upon a scram signal. 2. The worst single failure of emergency equipment was assumed (failure of one, emergency diesel generator.to start). 3. An uncertainty margin of-+20% was added to the calculated decay heat generated in the reactor.
- 4.. Conservativ'e setpoints and delays were assurled for scram, SIAS and equipment startup.
As,sumption (1) results'in a description of the accident in which the reactor coolant pumps begin ~coastdown,the main steam isolation valves shut, and the main feedwater pumps stop upon receipt of a sc; ram signal. Main steam and feedwter flows are ramped to zero in 0.5 seconds. ~ f DOG 66 1
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g ' Assumption (2) allous credit for only one train of ECCS pumps and one . h:. train of Emergency Feedwater (EFW). Thus, only one HPSI pump and one r EW pump are available, to satisfy LOCA requirements. It.was additionally . assumed that 25% of 'the HPSI inflow spills out through the, break. ~ The flow capacit'y re'quirements.for the llPSI and E W pumps are proportional t'o values of decay. heat. Assumption (3I, therefore, results in a prolm.r- ~ tionate increase in these requirements. The HPSI pump capac' ty assumed i 's shown in Figure 4. The EW capacity was assumed to be 875 gpm (minimum). i Assump' tion -(4) resulted in the following setpoint and delay assumptions: a) Reactor scram occurs upon a. low pressurizer pressure (LPP) indication of'1728 psia (minimum). Upon receipt of the LPP, ) signal there is a 0.9 second delay before the control rods .I .. ' begin motion followed'by -a 3.5 second interval to fully / Insert the ' rods. ~
- C b) SIAS occurs upon a pressurizer pressure indicatio.1 of
(#pA. 1550 psia (minimum). The SIAS automatically starts the caergen y diesel generators which provide p~ower to the . HPSI pumps and, the EW pumps. The total time from receipt of an SIAS to the time the HPSI pumps provide flow is 30 seconds- (maximum). The total time del'ay from SIAS until l EN reaches the steam generators is 45 seconds (maximum). References ~ 1. Combustion Engineering, Inc., Standard Safety Analysis Report (CESS'AR), Docket No. STN-50-470, Amendment No. 31, Section 6.3.3.3. 2.. -D. V. Graf to D. R. Patterson, TD-CE-ll7, subj ect: " Meeting Minutes of 5/27 and 5/28 in Uindsor." ,3. Acceptance Criteria'for Emergency Core Cooling Systems for Light- . Water-Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No. 3, 'Jriday, January 4,1974. ~ ~ e ,,qa h,; FLC/ age --%.,___~..
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