ML20127N754

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Loss-of-Coolant Accident Analysis Conformance w/10CFR50 App K (Jet Pump Plant), Dtd June 1975
ML20127N754
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 06/30/1975
From:
NORTHERN STATES POWER CO.
To:
Shared Package
ML20127N739 List:
References
NUDOCS 9212010407
Download: ML20127N754 (45)
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{{#Wiki_filter:r-1 W FONTICELLO hUCLEAR P0h"ER SEATION LOSS-OF-COOLVT ACCIDENT AL\\ LYSES COTOR'tV:E FITH 10 CFR 50 APPECIX K (JET IUIP PLA'sT) JLNE 1975 9212010407 750709 PDR ADOCK 05000263 P PDR

N . DISCUSSION. Presented in the following dccument are the results of the loss-of-coolant - accident analysis of the '!ont icello Nucitar Power Station. _ The analysis was performed-using General'flectric calculational models which are consistent.with the requirements of Appendix K:of 10 CTR part 50. A-complete discussion of each code employed in the analysis is presentcd.in . Reference l. Between August and December,1974, General Electric and the USAEC worked together to-resolve differences in interpretation of Appendix K and to consider additionci phenomena in the evaluation models. As a result,,the models used in the present analysis diff =r from those used in previous . submittals in the following respects: (1) The new analysis assumes a fuel assembly planar power consistent with 102; of the MAPLHGR shown in the Figures; (2) Fission product decay is computed assuming an energy release rate of 200 MeV/ Fission; (3) Pool film boiling is assumed after nucleate boiling is lost during_ .the flow stagnation period; (4) The effects of' core spray entrainment and counter-current: flew limiting arc includad ir, the rcficcaing calculcticn. In addition, there have been a few other minor improvements to the computer codes which-ihdividually -and jointly hcve a small effect on the calculated-j. resul ts. The figures in tnis submittal reflect these changes, as well as the four major changes enumerated above. INPUT TO THE ANALYSIS A list of_the significant plant input parameters to the loss-of-coolant accident analysis is presented in Table 1. p l 4 1,

K D t* TABLE 1-SIGNIFICANT-INPUTS PARAMETERS TO THE j LOSS-OF-COOLANT ACCIDENT ANALYSIS j FOR !!ONTICELLO l PLANT PARAMETERS: Core Thermal Power....................... 1703 MWt which corresponds to l 102% of licensed core never* 6 Vessel Steam Output............. s.ola x 10 Lbm/h which corresponds to 102 % of rated steam flow Vessel Steam Dome Pressure............ 1040 psia Design Basis Recirculation Line 2 Break Area for Larae Breaks 3.4 ft 1.0 ft Recirculation Line Break Area for Small Breaks 1.0 -ft2 0.07 ft2-FUEL PARAMFTERS: PEAK TECHNICAL ' INITIAL SPECIFICATION DESIGN MINIMUM LINEAR. HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING PC'JER FUEL TYPE GEOMETRY (kw/ft) FACTOR RATIO Initia l tr.re - ( 7D225 - 7x7 17.s 1.g7 1,1 g ReloaC 1 70230 7x7 17.5 1.57 -1.13 Reload 2 80262 8x8-13.4 1,57 1.18-Reload 3 8D259-8x9 13.4 1.57 1.18 Reload 4 8D219 8x8 13.4 1.57 1.18 A more detailed list of. input to each model and its source-is presented in Section II of Reference 1.

  • This power level equals the Aopendix K requirerent of 1025. The core heatup calcu-lation assumes a bundle power consistent with oce' ration of the highest-powered rod at 102% of its rexinum (technical specification) linear heat generation rate.

n. 4 RESULTS OF THE /M LYSIS The results of the analysis are presented in the order in thich they are calculated. The presentation of the results is divided into four major portions according to the model from which the output is obtained. These portions are: A. Calculated by the Short-Term Thernal Hydraulics tiodel (LAMS) B. Calculated by the Transient Critical Power Model (SCAT) C. Calculated by the Long-Term Thermal Hydraulics Model (SAFE) D. Calculated by the Core Heatup Mod::1 (CMASTE) A surnary of the results is prescnted in Table 2. At the MAPLl!GR* cmployed in the and ycis, the n., t cevere pipe break yleias a calculated peak cladding temperature less than or equal to2203 F, a calculcted caximun local neta t-water reaction less than or equal to 17:: and a calculated core-wide metal-water reacticn less than or equal to 1%. Compliance with the 100F350.46 criteria for coolable gecmetry and long-tern cooling has been shovin in Reference 1. The reactor is, therefore, fully in conformance with 100FR53.46 and Appendix K with operation at the MAPLHGR used in the an; lysis. These values, if more limiting than other design paraccters, represent limits for cperation to ensure conformance with-10CFRSO.46 and Appendix X. The peak cladding te.:peratures as a function of time are shown in Figure D-la and D-lb. Other parameters relevant to the analysis are shown in the attached figures and are described in subsequent paragraphs. Results for guillotine severances of a nain stean line, a feeduater line, and a core spray line are presented in Reference 2. l

  • Maxirm (Bundle) Average Planar Linear Heat Generation Rate i

l, l

c TABLE 2 APPENDIX K RESULTS FOR MONTICELLO Break Size Core-Wide Location Peak Local -Metal-Water Single Failure PCT (OF) Oxidation ': Reaction DBAANALYSIS(I} 3.9 ft2(DBA) Recire Suction 2200()) 8.7% 0.5 LPCI Injection Valve BREAKSPECTRU!! ANALYSIS () 3.9 ft2 (DBA) Recirc Suction 2200(I) 8.7% 0.5 LPCI Injection Valve 2 1.0 ft Large Recirc Suction Break 1670()) <1 LPCI Injection Methods Valve Small Break 1690(2) <1 Methods 2 0.07 ft 1430(2) <1 Recirc Suction HPCI Notes: CHASTE - large break methods Non-DBA reflood For other breaks in spectrum see lead plant analysis, Reference 2. For justification-of selection of lead plant, see Reference 3. . l

. ~. _ _.. _ _ =_. - - A.. Appendix K Shbrt-Tc'ra Thermal Hydraulics Analysis General Descriction of the LA"B Code F The LAli3 code is a model which is used to analyze the short-term thermodynamics and thermal hydraulics behavior of the coolant in the vessel during a postulated loss-of-coolant accident. In particular, LA"3 predicts the core flou, core inlet enthalpy, and core pressure during the bloudown prior to.the end of_ lower plenum flashing (ne20 seconds). For a detailed description of the codel and a-discussion regarding-sources of input to the model refer to the "LNIB Code Documentation" portion'of. Reference 1. Resuits of ton iA"6 Annivsis Presented in the section are results of the loss-of-coolant accident analysis which are calculated by LAl's. Table 3 lists the figures provided for all the analyses. These results include the fc110wi:;;: Parameter Figure Core Average Inlet Flow Rate (Normalized to unity at the beginning of the accident) -- Following a Design Basis Accident A-la Following a 1.0 Sq. Ft. Break A-Id Core Inlet Enthalpy -- Following a Design Basis Accident A-2a Following a 1.0 Sq. Ft. Break A-2d Core Average Pressure Following.a Design Basis Accident A-3a -- Following a 1.0 Sq. Ft. Break A-3d These results are input to the SCAT code discussed in Section B. - c n

i a. c B. Appendix K Transient Critical Power Analysis General Description of the SC\\T Code The SCAT code is used to evaluate the short-term themal hydraulics response of the coolant in the core during a postulated loss-of-coolant accident. In particular, the convective heat transfer process in the themilly limiting fuel bundle is analyzed during the transient. For a detailed descriptien of the model and a discussion regarding sources of input to the model refer to the " SCAT Code Ibcumentation" portion of Reference 1. Results of the SCAT Analvsis Presented in this section are results of the loss-of-coolant accident analysis which are calculated by SCAT. Table 3 lists the figures provided for all the analyses. These results include the following: Parancter Figure Minimum Critical Power Ratio Following a Design Basis Accident, 8x8 B-la-1 Following a Design Basis Accident, 7x7 B-la-2 -- Following a 1.0 Sq. Ft. Break, 8x8 B-1d Convective !! cat Transfer Coefficient- -- Following a Design Basis Accident B-2a --.Following a 1.0 Sq. Ft. Break B-2d These results are used as input to the O!ASTE code discussed in Section D. ( l l 1 C, AcmdixKi.ong-TermThernalHydraulics-Analysis General'Descrintion-of_ SAFE Co_de The SAFE code is a model which is used to analyze the long-term thermodynamic. behavior of the coolant in the vessel during both small and large breaks. LSince the calculational procedure of thcl loss-of-coolant accident analysis differs. depending on whether or not a break is classified as "snall" or "large," it is: appropriate to distinguish between tuo classifications of breaks. 4 snall break is defined as that size break for which nucleate, boiling heat transfer exists in th core until the heat fluxes are below the pool boiling critical power condition. This occurs approximately 20 to 25 seconds after the break. For small breaks,-core heatup is, therefore, based solely on the uncovery and recovery of the fuci 'and the duration of spray cooling all of which are predicted by the SAFE code. For the "large" break. analysis, the LAMB and SAFE codes are employed to determine ~th'e time-of boiling transition and the post-boiling transition convective heat transfer coefficient durino i.he blundunn. Tim 5AFC code calculotts thc uncovcry *:nd rc-fle: ding ef the fuel and thn duration of. spray cooling. The SAFE anaiytical model has been expanded and refined to consider explicitly the' following phenomena: (1) Counter-current flow limiting (CCFL-) in the fuel bundles and in the core bypass region, of ECCS water injected over the core; i= -(2) Entrainment and loss of ECCS water injected over the core; and-(3) Filling of discrete volumes _(control rod guide tubes, core bypass ~and lower plenum) which were previously:taken together. Calculation of these effects is. presently external to the SAFE _ code: -the calculati'ona-logic Will eventually be incorporated in the SAFE code. For a detailed description of the nodel and a discussion regarding sourcesLof input - to the.model refef to the " SAFE Code Occumentation" partion oflSection.II of Reference 1. - L w + w r. 3 14-+,

'{ Result,s of the SWE Analysis ~ ~ - Presented in this s' ction 'are results of the loss-of-coolant' accident analys. e which are calculated by. SVE. Table 3 lists the figures provided for all the analyses. These results include the following: Parameter Figure Water Level Inside Shroud Following a Design Basis Accident C-la-1 (LPCI Inj. Valve Failure) Following a 1.0 Sq. Ft. Large Break C-Id 1-(LPCI Inj. Valve Failure) Following a 1. 0 Sq. Ft. Smil Break C-2a -- (LPCI Inj'. Valve Failure) Following a 0.07 Sq. f t. Smil Break C-2b-1 (HPCI Failure) Reactor Vessel Pressure Following a Design Basis Accident C-la-2 (LPCI Inj. Valve Failure) Following a 1.0 Sq. Ft. Large Break C-Id-2 (LPCI Inj. Valve Failure) Following a-1.0 Sq, Ft. Smil Break .C-2a-2 (LPCI'Inj. Valve Failure) - - Following a 0.07 Sq. Ft. Smil Break -C-2b OIPCI Failure) o l I-l -_g_ i . ~. l

D. Appendix'M Core Heatun Analysis-Gencral Description of CHASTE Code The Transient thermal response of the core to a loss-of-coolant accident - calculated by CGSTE can cenerally be bro;en down into four stages; (1) fuci pjn temrerature redistribution; (2) f'di rea bundle te.perature redis-tribution; (3) metal water reaction heatur. ind (4) core standby ecoling system effects. Phenomena occurring durir;.hese stages that are considered in'the analysis are described below. Fuel Pin Temperature Redistribution following a reactor shutdown, a large heat source is still available within the core in the form of sensible heat in the fuel. This is represented by the temper-ature profile in the fuel rod. Initially, the temperature profile is steep because of the high oc;;er generatien ratcs during normal operaticn. Following the snut-down, the sensible heat in the fuel will be redistributed by themal conduction within the fuci and cladding and by ccnvecticn and radia*. ion in the gap ben cen fuel and cladding, with the amount of heat removed being dependent on-surface conditions. At the end of three or more fuel time constants (fuel thermal time constant is abcut 8 to 10 seconds), the adial temperature profile in the fuel pin is almost flat, consistent with ti.c lou fission product decay power generation. Fuel Rod Bundle Temperature Redistribution As the cladding tercerature incrcases and the core coolant voic fraction approaches unity, radiant heat transmission bcts:cen reds and the channel all tends to 1 equalize the te'cerature of all rcos at a given axial position. The total energy in the core contir.ucs to increase during this period due to continuing fission product decay. Metal-1.'a ter Reac tion Heatup The fuel pin cladding is made of Zircaloy, which reacts with steam at high temper-atures. The zircalcy-steam chemical reaction rate is exothermic and stongly dependent upon the reactien tc perature. The temperature cecendence is excenential and the rate of reacticn beccmes significant at cladding temperatures in the rahge of 2200'F or higher. Emergency Core Ccoling Systcm (ECCS) Effects Redur. dant c crgency core cooling systems performance for a given L.0CA is depend-ent upon the cenditions of the accident. The core cooling systems will provide sufficient coolin to prevent-excessive clecding heatup. The primary purpose of the core heatua analysis is to deternine the effectiveness of the emergency. core coolir,g systcms. For a detailed de cription of the CHASTE model and a discussion regarding scurces of. input to the m: del refer to the "CFASTE Code Documentation" portion of Section-11 of Reference 1..-

A-break. spectrum 4nalysis has-been performed using the CKASTE code showirig that the most lirliting -(highest calculated) peak clad temperature is associated with .the design basis accident. The conclusion'of this analysis is appluable to- .this plant. The analysis has been documented in the Quad Cities Station Special Report 15, Supplement.C (Docket No. 50-254). For each submittal of a construction pernit, operating 'icene, or reload license, the DBA peak cladding temperature, peak local oxidatur,, and a MAPLHGP,is determined fcr each fuel type of interest. For calculational convenience in some cases, the rod-to-rod power distribution is assumed to be flat and the least favorable exposure is assumed in determining cap cenductance. Calculation _ of the results under these conditions conservatively represents the results at all exposures. The code. application is described, briefly, as follows: A. For jet-pump plants a LAMS calculation is performed. In mixed cores, full-core LAMB calculations are performed for 7x7 and 8x8 fuel and the more restrictive of the two is used in the SCAT input. B. For jet-pump plants, SCAT calculations are performed for 7x7 fuel and 8x8 fuel, as appropriate. C. A SAFE and a DBA-REFLOOD calculation is perforted, assuming the fuel to be the most predominant type of bundle in the core (7x7 or 8x8). D. CHASTE calculations are performed for each fule type (which in a given reactor nay include several 7x7 fuel types and several 8x8 fuel types) at several exposure points. The MAPLHGR, peak cladding temperature and maximum local oxidation variations with exposure for each fuel type are the results of these calculations. Results of the CHASTE Analysis Presented in this section are results of the loss-of-coolant accident analysis which are calculated by CHASTE. These results include the following: Parameter Figure Peak Cladding Tenperature Following a Design Basis Accident D-la j Following a 1.0 Sq. f t. Large Break D-ld l Following a 1.0 Sq. Ft. Small Break D-2a Following a 0.07 Sq. Ft. Small Break D-2b L Peak Cladding Temperature and Local Peak 0xidation versus D-3 l Break Area Deak Cladd.ng Tenperature and Local-Peak 0xidatior versus-l.- Planar _ Exposure Initial Core fuel (7D225) D-4a Reload 1 Fuel (70233) D-4b Reload 2 Fuel ~(80262) D-4c-Reload 3 Fuel (8D250) D-4d Reload 4 Fuel (SD219) D-4e l- ,,-n

TABl4 3 KEY TU TICt! PES LAftCE BREAK KET1f03 INTEVEDI ATE B'tif r. SMALL EREAK I 1~C FT 1 *0 IT ! ?tM.L Ftt. S**ALL E2K.. ' ARGE DREAK 5'A3LL I" TEAK

0. 0 7 l ~s -

MLT11005 71L1tKDS CODE SPRAY TC* WATER I MAIN STEN 1 DPA 1DM .60 DPA . I!PCI TAlt, LINE LINE LfNE l l Core Average Inlet I, e Flow A-la A-lb* A-Ic' A-Id Core Inlet Enthalpy A-2a A-Zb* A-2c' A-2J l Core Average Pressure A-3a A-3b* A-3c* A-3d l stinirum Eritical ic-er Patto 8-la B-lb' B-Ic* D-Id ~ jconvective ifcat Trans-l 5-2s-B-2b' 5-2c* B-2d D-2a D-2b D-2c' D-2d* D-2e* y

fer C; efficient i

l AShreud AhD C-la' C-lb* 3 C-Ic* C-Id C-2a C-2h I C-2c' C-d* C-2e* C-2f* l . Liter Icvel inside ) 4 I f e a l y3 Reactor Vessel Pressurc .I i - l. i I g l Peik Cladling Tempera-D-la D-lb* D-Ic* .D-I d ' .D-2a D-2b D-2c' D-2d* g D-2e* ture s-1 Bre.lk Spectrum D-3 l I'

Peak Cladding D-4 reercrsture and Stani.

l cum cuidation vs. l rwpose wAuica a-s. rea 21ntsE rters stE LEAD Plur ANALYSIS FOR Dwn/3 Quad Cities 2 i '- ~ l l e e

~, SINGLE FAILUT.E STUDY ON ECC SYSTEM M4NUALLY CONTROLLED ELECTRICALLY OPERATED VALVES The effects of a singla failure er crerator error that causes any manually controlled, electrically crerated valve in the ECC Systea to move to a position that could adtersely aff:ct tne ECCS has been studied. The purpose of this evalua; ion is to detemnine that any such malfunction does not affect tha ECCS n:re than the results of the worst single f 2ilure which is reported in the LCCA calculctions perfonned in accordance with ICCFR50 Appendix K. The results of the break spectrun analysis short the single failure which results in the maximur calculated per clad temperature (FCT). For any other single failure to te more significant, its effect on tne ECCS must be greater t::an tais sir 0le failure. Therefere, a. study was made to determine if the malfunction of a.inually controlled, electrically Oper-ated valve by scne unkncun cause or oy an operatcr imprcaerly positioning a control switch could affect the ECCS more severely than this failure. In accordance with aporopriate IEEE standards, the ECC Systen valves are electrically assign 3d to cifferent divisions of power supply. The effect or an onera!.or moroveriv ec uan um a s invie s..ii.c;. vn u.a ca.t;*ol ~ p;ncl is to cause caly a single veive ib inve to an incorrect pcs1 tion. ror the operater error of actua Lit.g a singic s'.; itch Of the EDS Syctem, the system valvcs are not actuated. However, the consecuances of a malfunction which causes one ADS valve to inadvertently open has been noted. The summary of the ECCS Valve Single Failure Analysis is provided in the attached Tcble

4. Comparing the effects of the single valve failure noted in Tabfe I with the results of the Appendix K LOCA analysis, it can be seen that these failurcs are not more severe tnan those reported.

The single failures considered for the ECCS analysis are presented in Taole 5. -.

TABLE 4 MONTICELLO ECCS SINGLE VALVE FAILURE ANALYSIS POSITION FOR NORMAL PLANT CONSEQUENCES OF VALVE FAILURE ASSUMED TOGETHER WITH OPERATION SYSTEM VALVE (5) CLOSED OPENED DESIGN BASIS LOCA Core Spray Suction X Negate use of one core spray loop Injection (s) X X Negate use of one core spray loop Test Return X Negate use of one core spray loop High Pressure Coolant Injection Condensate Suction X Utilize Suppression Pool Water Utilize Condensate Storage Tank water Suppression Pool X Suction Valve Partial loss of flow due to flow to suppression pool I Suppression Pool X Test Return Injection (s) X X Negate HPCI Turbine Inlet (s) X X Negate HPCI Low Pressure Coolant Injection Injection (s) X X Negate use of LPCI Partial flow loss in one loop due to flow to suppression Minimum Flow X pool Cross Tie. X No LPCI fix: Negate on LPCI Loop (two pumps per loop) Test Return X No consequence HX Bypass X X Reduce Flow due to HX Pressure Drop Automatic Pump Suction X Nenate one loop Depressurization Vessel depressurizes faster, increases rate of HPCI System One Relief Valve X injection (assuming the failure of a single ADS valve to open does not affect the results because the effects on small breaks is insignificant with HPCI in operation)' .t

^ --v TABLE 5 SINGLE FAILURES CONSIDERED FOR ECCS ANALYSIS PLANT SINGLE FAILURE REMAINING ECCS ~BWR/3 LPCI Injection Valve 2 CS + HPCI + ADS MONTICELLO HPCI 2 CS + CPCI + ADS .(SuctionBreak) .i j ;

u Reference Plant Analysis-The lead plant for this product line BWR is Quad Cities 2. (2) The 60% DBA, 805 DBA analyses, additional Small Break analyses, Core Spray line break, Feedwater line break, and Main Stean'line break analyses for the lead niant are applicable to this plant and are hereby' incorporatec by reference (3). REFERENCES 1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K, NED-20566 (draft), submitted August 1974, and General Electric Refill /Reflood Calculation (supplement to SAFE Code Description) transmitted to the USAEC by letter, G. L. Gyorey to Victor Stello, Jr., dated, December 20, 1974. 2. Quad Cities Station Special Report No.15, Supplement C, Unit 2 and Attachment A (Proprietary information). 3. Letter, G. L. Gyorey 'to V. Stello, " Compliance with Acceptance Criteria of 10CFR50.46," May 12,1975. l II l l l l I 7 FIGURE A-1 a fl0RMALIZED CORE AVERAGE IriLET FL0tt FOLLOWING A DESIG!{ BASIS ACCIDEtlT I I MONT1 CELLO OXO SUCTfLDRK DBR 1. 1 I l 1 l .g,0.6g r___.-_._., _J -LL JET PUMP UNC0VERY ls J cc i O [ L) \\ .l' a 3 1 trJ 0.2 ,,a ~ wR _J ~ C.I:- M y .a y_ A -0.2 8 i 't i !i, LOWER PLENUM FLASHING ~ 0. ti, ~ 0. 1, TIME AFTER BREAK SECON6S t

FIGURE A-1 d . NORMALIZED CORE AVERAGE INLET FLOW - FOLLOWIfiG A lsq. ft. BRK MONTICELLO BX8 SUCTN BRK 1.SO.FT. 1. 1 ~ i ~ x 0.6-O __J L1-g t window ] h ( Gin =100,000 lb/flR-FT2) I o ( O LtJ 0.2 g N m 1( .__J ~ Jp) o j' L0kER PLENUti FLASHIflG 'I'. -0.2 O. 8 IG. 2 11. 32. . TIME AFTER BREAK, SECONDS "n s. -_--_.._-.___-..-,_-..n. A

FIGURE A-Za CORE IflLET Et1THALPHY FOLLOWING A DESIGil BASIS ACCIDEffT I MONTICELLO 8X8 SUCTN BRM DBA ~ SGO. e aa J NoH 5'10. aa ~ g wq r4 i z 2 m. -t UJ 520. 1 S ct cc 93 ..o _J ~. . 500'. ~ t'I. D. 4 6. 12. 16.. TIME AFTER BREAK, SECONDS ..p. .J

4 ii TIGURE A-2d CORE If4LET EfiTHA'LPY FOLLOW 1 fig A 1.0 50. FT. BREAK 1 .MONTICELLO 8X8 SUCIN BRK 1.SC.FT. 4 560. 1 m i -J N 1 4 3 1-- 5'40. m f-j- I 1 a 5, - 2, Lt3 520. / -J \\ j O_. I g ' tri. ~ CJ 1 ~ l 500 I'U.'I'? 8. 16. 2 11. 32. TIME AFTER. BREAK, SECONDS

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