ML20002C392

Facility Design Basis Accident LOCA Analysis for NFS Deomonstration Fuel Assemblies

ML20002C392
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	05/18/1972
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML20002C389	List: ML20002C388 ML20002C392
References
NUDOCS 8101100242
Download: ML20002C392 (17)

Text

{{#Wiki_filter:,- jy v BIG ROCK POI!ff DESIGN BASIS ACCIDENT IDSS OF (DOLANT ANALYSIS IVR N15 IMUNS11MT10N FUEL ASSEM3LIiiS 4 I 8101100SN1

INTRODl!Cfl0N 1his analysis of the design basis loss of coolant accident in the Big Rock Point Reactor Plant is provided by the General Electric Company at the request, of the Consumers Power Corporation. The pumosc of the analysis is to detennine the acceptability of Nuclear Fucl Services (NFS) demonstra-t ion fuel assemblics for use in the Big Rock Plant. The analysis is accom-plished in accordance with the Atomic Energy Commission's Interim Acceptance Criteria. In the few cases where the criteria are not applicabic because of the plant and fuci design, appropriate conservative assumptions are ~ nade. i Y

ty ' 4 - SU>f4\\RY The consequences of the design basis accident in the Big Rock Point Reactor plant have been calculated using the Atomic Energy Conmissien's Interim Acceptance Criteria. 'Ihc analysis was conducted for the MP3 demonstration fuel. A peak cladding temperature of 4129T was calcula.ed for the accident and it is concluded that if reloaded with NFS fuel the plant will meet the Interim Acceptance Criteria for the design basis accident. The current fuel results in a peak cladding temperature of 2740F. 1 a J i i i r

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• l FDDEl,S AND ASSINPTIONS

'Ihe following assumptions were made regarding the course of the design basis accident:

1. A doubic ended break occurs in the recirculation piping so that a 2

2.54 ft2 break emptics the steam drum and a 1.75 ft break emptics the pressure vessel.

2. All feed flow and recirculation flow is lost at the time of the accident.

3. The analysis was conducted for the Nuclear Fuel Services demonstration fuel assembly. This assembly includes eight hollow, unpowered rods (the 4 corner rods and the 4 rods on the diagonals next to the center rods) and 113 powered rods.

The models used and the assumptions made in the an lysis are as described in NEDO-10379 (as amended by the AEC Interim Criteria). HowcVer, since Big Rock Point is a non-jet pump plant, the model was modified as follows:

1. Heat transfer during the blowdown was calculated using the standard General Electric dryout correlation.

Previous submittal for non jet pump plants have used the same dryout heat transfer model.

2. The break flow from one of the vessels (the reactor pressure vessel or tije steam drum) is limited by friction. A no friction assumption is appropriate (or nearly so) for a jet pump plant where the break could be assumed to occur in the recirculation piping imediately adjacent to the reactor vessel. Thus, friction limits only that small portion of the total break flow which exits the vessel through the jet pumps.

Ilowever, no matter where the break is assumed to occur in the Big Rock Point recirculation piping, the flow from the reactor vessel or the i r

i 4 4_ steam drum or both will be limited by friction. Ec worst break location (i.e., that resulting in thc' fastest blowdown) was asstuned. - 3. The bundic decay power is.96 times the required decay power (ANS standard.+20%) to account for gama absorbtion (" gamma smearing") in components. (channels and control blades) surrounding the bundle. No power reduction was assumed to result from rod-to-rod, bundic-to-bundic or axial gamma smearing. ~

4. ne wetting time of hollow, unpowered rods in the NFS fuel assembly

'is calculated in a conservative manner, consistent with the Bh'R FLEOrr data. %c calculation is outlined in appendix A of this report. Such cniculations have not been included in previous submittals because other fuel designs do not include hellow rods. Including this effect is considered te be entirely consistent with the spirit of the Interin Acceptance Criteria. 6 9 I Y

~. DBA ANMNSIS IESULTS Figures 1, 2 and 3 describe the design basis accident and the heatup of the NFS fuel assenbly. The level loss is very rapid for this large break (figure 1) and the standard General Electric dryout heat transfer correlation predicts that nucleate boiling will be maintained until about 1.8 seconds and that the blowdown is over at about 8 seconds (figure 2). The resulting cladding heatup is shown in figurc 3. Rated spray flow is attained at 16 secords and the channel wets at 92 seconds. %c hollow ruds wet at 78 seconds. liighest temperature calculated is 2129F at 270 seconds. H e maximum local Zirconium oxide thickness is approximately 3 mils on the hottest rod at the end of the transient. This corresponds to a maximum of about 9% of the 0.034 inch thick cladding. He core average metal water reaction is 0.3%. Figure 4 shows the DBA heatup for the 9x9 "F" fuel at 10,000 mfd /T exposure (determined to be the worst case of the present fuel in the January 1 submittal). Peak cladding temperature is 2740 F at 330. seconds. The maximum local Zirconium oxide thickness is approximately 10 mils on the hottest rod at the end of the transient. H is corresponds to a maximum of about 25% of the 0.040 inch thick cladding. W e core average metal water reaction is about 0.8%. I ,~

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( 10 VARTATION OF lHAKING FAC1DRS The DBA temperature transient was calculated throughout the life of the t - NFS fuel assembly to determine the limiting peak cladding temperature. 1he limiting case (O Ms'D/ ton) is plotted in figure 3. For other exposures the calculated peak cladding temperatures are as follows: P Exposure Feak Clad Temperaturc 0 hMD / Metric ton 2129 F 5582-2124 F 10858 2113 F 15160 2123 F 9 i e 4 1

1 t 1). CONCI,USIONS As a result of this analysis, the following conclusions can be made regarding the Big Rock Point design basis accident with NFS fuel assemblies: 1. The peak cla'dding temperature is 2129 F, meeting the AEC limit of-2300 F. 2. The core average calculated metal water reaction percentage is 0.3 percent, well below the conservative AEC limit of one percent. Therefore, the Big Rock Point plant if loaded with NFS demoistration fuel meets the interim acceptance criteria and is adequately protected against the design basis accident. i t t I

\\ APPTRDIX A - IDLIIM R0n IEITING FDDl!L l

===. Background=== A procedure,for calculating the wetting film velocity on heated rods during core spray operation was presented in Appendix D of GEAP 13190. ' Die result was an expression for the inverse velocity (equation 10, Appendix D of GEAP 13190), l TTS q b ATu oTs whe c T local unwetted cladding temperature, F = T =c lant saturation, taken as 212F S Tg = Leidenfrost superheat = 70F (see page D-5 of GEAP 13190) T = temperature difference between wetted cladding g and coolant temperaturc = zero for unheated rods ' J' A- ,C hk. ~ f'=claddingdensity,- with a; C = cladding specific heat, h; [ = cladding thickness, inches; It = heat transfer coefficient on wetted cladding = 6 2 2.25 x 10 Btu /hr ft F (see page D-6 of GEAP-13190); f l R = cladding themal conductivity, Btu / hr ft Fp I

( hliTI'ING VlilDCflY Folt N1?S 111111. ASSIMBLTl!S -= - - = 'lhe inverse velocity expressed by equation (1) is applicab]c at one atmosphere for rods of any type. For a particular type of rod, it is only necessary to detennine _the constant A. In the case of the 0.100 thick stain 1'ess steel test rods, A = 0.147 (equation 10 in Appendix D of GW'13190). For the 0.034 inch thic:- hollow Zircaloy rods in the NFS demonstration fuel assembly, A = 0.0225. 'lhat is, given the tame tempera-ture, the wetting front will travel faster on the hollow Zircaicy rods, principallybecausethefCproductissmaller(athin,hollowtube opposed to a thick tube filled with insulating material). Thus, we expect the expression -1 Q = 0.0225 T-T S (2) aT t To be a "best estimate"% the wetting velncity: h'c realiza hs.cc..g, Giul in this case it is more appropriate to have a conservative estimate for the 'ocal wetting velocity. Figure D-5 of GEAP 13190 shows that the expression (1) predicted about half of a set or rod wetting data within 20% and 90% of the set within 10%. If we require our expression to conservatively predict 90~6 of a set of similar data, we adjust the con-stant A by using A= 1.4 A = 0.0.116. That is, we decrease the wetting i velocity by 40?,. A conservative estimate of the film wetting velocity is therefore from equation (1). -1 A' T-T g = S att 0.0316 = (T-T) or 3 70 (b 2220 ft = T 712 mTn (3) t { r ^

ROD hiiTTING TIME h'e next need to establish the local cladding temperature variation with tin.c on the hollow rods as they are heated by radiation from'their neighbors. For this particular case, nine DBA heatup calculations were made to specify the axial temperature distribution as a function of temperature. The wetting time was then determined by integrating equation (3) ntunerically. The worst case axial peaking factor distribution was used (end of cycle distribution witif peak power near the bottom of the bundles) to calculate the wetting time at the hot axial node. A " worst case" set of locil peaking facto s was used in the calculation. The maximum peaking factor on each rod throughout life was assumed to exist at one point in time. The calculated wetting time was 78 seconds after the accident. Ch:utnel wetting, using the usual methods (i.e. in accordance with the interim acceptance criteria) occurred at 92 seconds. e r }

t FIIJ11)RYOlfr CALGil.ATIONf llaving calculated the wetting time, we must next be sure that the wetted rods will not later be dried out as a result of radiation heat transfer from neighboring hot rods. A simpic, conservative calculation was selected. We first assume that the fraction of core spray water availabic to wet the rods at the top of the bundic is the smaller of two fractions: 1){p Rod Perimeter = Total Perimeter (ruds + channel) and 2)fa,calculatedbyassumingaflatspraydistributionoverthebundle and assitning that water drops incident at the top ' closer to the hollow rod than the channel will wet the rod in preference to the channel. For this case-fp = 0.0444 is less thm fa = 0.08 so we use 0.0444. The minimum bundle spray rate (at 10 psig reactor pressure) is 4.6 GPM (=38.3 lb/ min) so that a minimum of 0.0444 x 38.3 = 1.70 lb/ min is available to wet the center hollow rods. We next get an upper bound on the avaporation rate by asstating black body radiation over the entire icngth of the bundle with all neighboring rods at 2300F. This is c1carly conservative since (1) the peak cladding terperature is Icss than 2300F, (2) the temperature is much 1 css than 2300'F over the 1cngth of the bundle and (3) the cladding emissivity (0.67) is less than unity. - The resulting maximum evaporation rate is 1.17 lb/ min, Icss than the minimum 1.70 lb/ min availabic, and the film will not evaporate later in the transient. (

AEC DIS' ' ]UTION _FOR PART _.50. 2 DOC M E TERIAI, O'HiPORARY F0fei1 CONTROL NO: 2760 e FRCT: Consumers Power Company DATE OF DOC: DATE REC'D LTR MEMO RPT OTHER Jackson, Michigan 49201 Rn1ph P. Sewell 5-18-72 TO: Gd1G 5-22-72 X CC UTHER SENT AEC PDR X g Bloch 1 SDiT LOCAL PDR~ CLASS: PRO? 1:a70 lin'UT KO CYb X20'D DOCKET NO: 60 xd 50-155 DESCRII' TION: n;;cLosu;;.,3 : Ltr re our 1-17-72 ltr, trans the following: REPORT: Design Basis ' Accident Loss Of Coolani Analysis for NFS Demonstration Fuel Assemblie 30 NOT 30VF' ACU0WLEDGED I PTMP Nav?. Bic Rock PMnt ( 60 cys ree'd) rro. n ewpTd*t /TWOm D TTON 5-2'4-72 AB BUTLER (DRL) -DICKER (DREP) SCHI::.IEL(DRL) XZIEMAUN(DRL) W/ Copies W/ Copies W/ Copies W/9 Copics U/ Copics CIIITWOOD(DML) GOLLER(DEL) SCHWENCER(DEL) W/ Copies W/ Copies W/ Copies W/ Copies V/ Copies, CIARK(DRL) KNIEL(DEL) STOLZ(DRL) W/ Copics W/ Copics W/ Copies W/ Copies U/ Copics H. DEITTON KNIGHTOH(DREP) VASSALLO(DRL) W/ Copies W/ Copics W/ Copies W/ Copies U/ Copics TturoMT. T)Te,TUTT)T1"' TOM Reg Filch $ Vollmer(DRL) O H. Fitzpatrich(DREP) EO Lainas % AEC PDR O siit;cr(DEL) O Kas tncr(DR2?) Novak $ Compliance (2) CCOX gLong(DRL) QLic. Asst (D22P) gross 5 OGC, Rm ? $06A QCollins(DRL) O McCreless(D't3P) (Ir.Hanauer D $ Muntzing & Staff $ Thor.,pson(DRL) O G. Blanc (DREP) hLauben O Morris /Dubc/Uilt.on(D2L) O D. E, i:unn(DRL) O Project Leadc (DREP) S B. Colmar 'Schroeder(DRL) [,1 Lenaroyn(DRL) D DR3P rile Room W Ros n O Dobc/U'. lsca (DRL) p( E. C. Ccsa(DRS)4If G A. Draitman(SLR) KP. Morris D Soyd(DRL) K;isccary(DRS)gypt D Caltar..ar(SLR) Q M DcYouns,(DRL) $ S tello(DRS) O Tre;racl(DIP) O %i!.Denton(DRL)(b) M V. Moore (DRs) O T:iocnton(: <::4) O ,Q Klechcr(DRL) O Lan:;a(buC) M Rautii(DRS) O OCrimos(DRL)! O Smilcy(D:iL) O shco (ca OGcnill(DRL)k O L. no.ers(nREP) O?.m ca (a > $ Tedesco(DRL)///~4 O Muller (Da(?) M suc..w t. (3Rt o EXTERNAL DISTRTBirPION i s- / _ - Local PDR -) - sad)tA/NY/ PDR

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,#1 - NSIC(Buchanan) 1 - AEC $Q LIB J-004 1 - ASLB(Yore /Woodard 1 - R. Carroll OC, GT l-- Chief Water Reactors "H" St ) X16.- Holding for ACRS 2 - Warr n Nyer I 1 - C. Miles, C-459, GT 1 - R. Catlin,(A-170, GT 1-ConsultantNewmark/Blum/Agabian) 9 - National Laboratories. (ANL/ORNL/BNWL3}}