ML20125A625
| ML20125A625 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/09/1975 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20125A623 | List: |
| References | |
| NUDOCS 9212080421 | |
| Download: ML20125A625 (46) | |
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l MONTIGLLO NUCLEAR POhTR STATION LOSS-OF-COOLANT ACCIDENT NMLYSES CONFOMW'E VITil 10 CFR 50 APPENDIX K (JET IUMP PLANT) JUNE 1975 l I 9212080421 750709 fDR ADOCK 05000263 PDR
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- gDISCUSSION Presented in the following document are the results 'of the loss-of-coolant-accident analysis of the !!onticello fluclear Power Station. - .The analysis was performed using General , Electric calculational models which are consistent with the requirements of Appendix K of 10 CFR part 50 ~ A complete discussion of each code employed in the analysis is presented in-Reference 1. ,
Between August and December, 1974, General Electric and the USAEC worked-together to resolve differences in interpretation of Appendix- K and to consider additional phenomena in the evaluation models. As a result, the-snodels used in the present analysis diff w 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 filn boiling is assumed after nucleate boiling is lost during the flow stagnation period; i (4) The effects of core spray entrainnent and counter-current flow limiting are included in the rcficoding calculcticn. ; In addition, there have been a few other minor irrprovements to the computer i codes which ihdividually and jointly hcVe a small effect on the calculated resul ts. The figures in this submittal reflect these changes, as well as , the four major changes enumerated above. INPUT TO THE AI;ALYSIS A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Table 1. 1 1 g t y lC p
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'T TABLE 1 SIGNIFICANT INPUTS PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR MONTICELLO PLANT PARAMETERS:
1703 MWt which corresponds to Core Thermal Power....................~. 102% of licensed core nower* 6 Vessel Steam Output............. s.913 x 10 lbm/h which corresponds to 102 y sf rated steam flow Vessel Steam Dome Pressure. . . . . . . . . . . . 1040__ psia Design Basis Recirculation Line 2 2 Break Area for Laroe Breaks 3.9 ft 1.0 ft Recirculation Line Break Area for Small Breaks 'l . 0 ft2 0.07 ft2 , 1 FUEL PARAMETERS: PEAK TECHNICAL ' INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL GENERATION RATE PEAKING POWER-FUEL BuMDLE GE0 METRY (kw/ft) FACTOR RATIO FUEL TYPE Initial u.re ( 7D225 7x7 17.5 1.57 1.18 Reload 1 j 7D230 7x7 17.5 1.57 1.18 i Reload 2 8D262 8x8 13.4 1.57 1.18 Reload 3 8x8 13.4 1.57 1.18 8D250 Reload 4 80219 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 Appendix K requirement of 102% The core heatup calcu-lation assumes a bundle power consistent with operation of the hiqhest powered rod at 102% of its maximum (technical specification) linear heat , generation rate.
- s. . .
m. 4 RESULTS OF THE AllALYSIS The results of the analysis are presented in the order in which 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 Thermal Hydraulics Model (LAMB)
- 8. Calculated by the Transient Critical Power Model (SCAT)
C. Calculated by the Long-Term Thermal Hydraulics Model (SAFE) D. Calculated by the Core Heatup l'odel (CHASTE) A sunmary of the results is presented in Table 2. At the MAPLHGR* employed in the-erPlyris, the mnst severe pipe break yleias a caiculatec peak ciadding temperature less than or equal to2200 F, a calculated maximum local metat-water reaction less than or equal to 17; and a calculated core-wide metal-water reaction less than or equal to 1%. Compliance with the 10CFR 0.46 criteria for coolable geometry and long-tern cooling has been shown in Reference 1. The reactor is, therefore, fully in conformance with 10CFR50.46 and Appendix K with operation at the MAPLHGR used in the anclysis. These values, if more limiting than other design parameters, represent limits for operation to ensure conformance with 10CFR50.46 and Appendix K. The peak cladding temperatures 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 main steam line, a feedwater line, and a core spray line are presented in Reference 2.
%ximum (Bundle) Average Planar Linear Heat Generation Rate d-
TABLE 2 APPENDIX K RESULTS FOR MONTICELLO Break Size Location ' Core-Wide Single Failure Peak Local Metal-Water PCT (OF) Oxidation Reaction t DBAANALYSIS(I) 3.9 ft2 (DBA) Recirc Suction 2200()) 8.7% LPCI Injection Valve 0.5 > l BREAK SPECTRUtt ANALYSIS (3) 3.9 ft2 (DBA) Recirc Suction 2200()) 8.7% LPCI Injection Valve 0.5 i 2 1.0 ft Large , Recirc Suction Break \ 1670)) <1 LPCI Injection Methods Valve Small Break 1690(2) y) , Methods , 2 l 0.07 ft Rectre Suction 1430(2) <) - HPCI a i Notes: i CHASTE - large break methods Non-DBA reflood
- 3) For other breaks in spectrum see lead plant analysis, Reference 2. ,
For justification of selection of lead plant, see Reference 3. ; A
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"A. Appendix V. Short-Tern Thermal Hydraulics Analysis General Description of the Lt.!'.B Code The LAlt3 code 13 a rtodel which is used to analyze the short-term thermodynamics and thermal hydraulics behavior '.f the coolant in the vessel during a postuiated loss-of-coolant accident. In particular, LA!'3 predicts the core flow, core inlet enthalpy and core pressure during the bloudoun prior to the end of lower plenum flashing (N20 seconds). For a detailed description of the r.odel and a discussion regarding sources of input to the rodel refer to the "LAt:0 Code Docurentation" portion of.
Reference 1. - kosta t' of tha 1 AN Arnlysis 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 follwin;;: , Parameter Figure Core Average Inlet Flow Rate (Nomalized to unity at the beginning of the accident)
-- Following a Design Ibsis 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. Pt. Break A-3d These results are input to the SCAT code discussed in Section B.
o e B. Appendix K Tnnsient Critical Power Analysis General Description of the SCAT Code The SCAT code is used to evaluate the short-term thernal 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 description of the model and a discussion regarding sources of input to the nodel refer to the " SCAT Code Doctnentation" portion of Reference 1. Results of the SC,iT Analysis 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: Parameter Figure Minimtn Critical Power Ratio .
-- Following a Design Basis Accident, 8x8 B-la-1
-- Following a Design Basis Accident, 'x7 , B-la-2 _
-- Following a 1.0 Sq. Ft. Break, Sx8 B-Id Convective Heat 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 DiASTE code discussed in Section D.
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M C. Accr. dix X Lons-Tern,Thernal Hydraulics Analysis i General Descrintion of SAFE Code The SAFE code is a model which is used to analyte the long-term thermodynamic , behavior of the coolant in the vessel during both small and large breaks. Since the calculational procedure of the 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. A sna11 break > is defined as that size break for which nucleate, boiling hut transfer exists in the core until the heat fluxes are below the pool boiling critical power condition. This occurs approximately 20 to 25 seconds af ter the break. For small breaks,_ core heatup is, therefore, based solely on the uncovery and recovery of the fuot 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 the time of boiling transition and the post-boiling transition convective heat transfer coefficient durinn the bivudunn. The SAFE ccJs calculatts the uncov:.ry cnd rc-f1:oding cf the fuel and tho 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; (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 sources. of input to the model refer to the " SAFE Code Documentation" portion of Section II of Reference 1.
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1' Results of the SAFE Analvsis ; Presented in this section are results of the loss-of-coolant accident analys: which are calculated by SAFE. 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 (LPCT Inj. Valve Failure) .
Following a 1.0 Sq. Ft. Smil Break C-2a-1 (LPCI Inj . Valve Failure)
-- Following a 0.07 Sq. f t. Smil Break C-2b-1 OIPCI Failure) l Reactor Vessel Pressure Following a Design Basis Accident C-la 2 (LPCI Inj. Valve Failure)
-- Following a 1.0 Sq. Pt. Large Break C-Id-2 (LPCI Inj. Yalve Failure)
-- Following a 1.0 Sq. Ft. Small Break C-2a-2
-(LPCI Inj. Valve Failure)
-- Following a 0.07 Sq. Ft. Smil. Break C-2b-2 OIPCI Failure)
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. . . . - _ 4 4 D,, Appendix K Core Heatun Analys_is, Goncral Description of CHASTE Code l The Transient thermal response of the core to a loss-of-coolant accident ! calculated by CESTE can generally be bro;en down into four stages; (1) l fuel pjn temcerature redistribution; (2) fuel reo bundle temperature redis-tribution; (3) metal-water reaction neatuo; :nd (4) core standby cooling i system effects. Phenomena occurring durir: .50sc stages that are considered in the analysis are described below. Tuel 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 temoer-ature profile in the fuel rod. Initially, the temperature profile is steep tccouse of the high ocuer generation rates during normal operatien. Following the snat-down, the sensible heat in the fuel will be redistributcd by then al conduction within the fuel and cladding and by convectica and radiation in the gap between fuel and cladding. with the amount of heat removed being dependent on surf ace conditions. At the erd of three or more fuel time constants (fuel thennal time constant is abcut 8 to 10 seconcs), the radial tcTperature profile in the fuel pin is almost flat, consistent with the low fission product decay power generation. Tuel Rod Bundle Temperature Redistribution As the cladding temperature increases and the core coolant void fraction approaches unity, radiant heat transnission between rods and the channel wall tends to equalize the tc,perature of all rods at a given axial position. The total energy in the core continues to increase during this period due to continuing fission product decay. Metal-Water Reaction Heatup The fuel pin cladding is made of Zircaloy, which reacts with steam at high temper-atures. The zircalcy-ste:n chemical reaction rate is exothermic and stongly dependent upon the reaction temperature. The temperature cependence is exoonential and the rate of reaction becs.es significant at cladding temperatures in tne
.ratige of 2200'T or higher.
Emergency Core Cooling System (ECCS) Effects Redundant cmcrgency core cooling systems performance for a given 1.0CA is depend-ent upon the conditions of the accident. The core cooling systems will provida sufficient cooling to prevent excessive cladding heatup. The primary purpose of the core heatup analysis is to determine the effectiveness of the emergency core cooling systcms. For a detailed description of the CHASTE model and a discussion regarding sources of input to the medel refer to the "CMSTE Code Documentation" portion of Sectien 11 of Reference 1.
---------------------------4 A break spectrum analysis has been performed using the CHASTE code showing that the most limiting (highest calculated) peak clad temperature is associated with the design basis accident. The conclusion of this analysis is applicable 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 permit, operating license, or reload license, the DBA peak cladding temperature, peak local oxidation, and a MAPLHGR is determined for each fuel type of interest. For calculational convenience iri some cases, tha rod-to-rod power distribution is assumed to be flat and the least favorable exposure is assumed in determining gap 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 LAMB calculation is performed. In mixed cores, full-core LAMB calc 11ations are performed for 7x7 and 8x8 fuel and the more restrictive of the two is used in the SCAT input. B. For jet-pump pi e. ;wlations are performed for 7x7 fuel and 8x8 fuel as ap,i 4?i ,. C. A SAFE and a DBA-ht a siculation is performed, 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 may 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 CHA3TE 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 Temperature Following a Design Basis Accident D-la
-- Following a 1.0 Sq. f t. Large Break D-id
-- Following a 1.0 Sq. Ft. Small Break D-2a
-- Following a 0.07 Sq. ft. Small Break D-2b Peak Cladding Temperature and Local Peak Oxidation versus D-3 Break Area Peak Cladding Temperature and Local Peak 0xidattor versus Planar Exposure Initial Core Fuel (7D225) D-4a Reload 1 Fuel (7D230) D-4b Reload 2 Fuel (8D262) 0-4c Reload 3 Fuel (8D250) D-4d Reload 4 Fuel (8D219) D-4e
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.: LARGE BREAK MET 103 thTERMEDIATI sarAK $ PALL BREAK . - .
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l' 8 - ^ 1.0 IT 1.0 FT 5"ALL E2K. . . !
. ARCI BREAK SMLL BREAK l 0.6 !mLLIBRK.- P -- g .
MLT100S METIf005 CCRE STRAY FE'+MTERjMAINSTEAR l- . D B *. 7 DBA I .60 DBA ItPCI FAIL - LINE LIKE LTNE f Core Average Inlet i , e l Flow A-la A-lb* A-Ic* A-Id -- - ---- ----- --- ----- . s l l .
,Cere Inlet Enthalpy A-2a A-2b' A-2c* A-2d --- , ,.
I . ,
- Core Average Pressure A-3a A-3b* A-3c* A-3d --- ---- --- l ---- --- -- !
trinteum Critical '. . - l , Po.er Ratio B-la B-lb' B-1c' B-Id ----
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1 iConvective Heat Trans-
'fer Coef ficient B-Za ,- B-2b' S-Ze* B-2d-D-2a D-2h D-2c' s D-2d* D-2e* - ..
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. Water level Inside ishroud At.D C-la~ C-1b* jC-lc* C-Id C-2a i C-2h 8 C-2c' C-d* ,
C-2e* C-2f* *
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e w l Reactor Vessel Pressurc ,- I l 3
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-l Peak Cladding Te7pera- g [
s ture D-la D-lb* D-Ic* .D-Id * .D-2a .D-2b D-2c' ; D-2d
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2 BreJk Spectrum D-3 , , [
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i resh Cladding 0-4 + k' g . Teeperature and Mant- ' . rum Omi*ation vs. ' .,. Espose - i . . (
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i tcas Ti:ESE PLUTS SEE LEAD FIMrr AMALYSIS FOR twa/3 Quad Cities 2 . t . b 4 e g .
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$1NGLE FAILUTE STUDY ON ECC SYSTEM M1NUALLY CONTROLLED ELECTRICALLY OPERATED YALVES The effects of a single failure or ocerator error that causes any manually controlled, electrically ocerated valve in the ECC Systea to move to a position thit could ad/crsely affect tne ECCS has been stuJied. The purpose of this evolus cion is to deterline that any such malfunction does not affect the ECCS r.cce than the results of thu vorst single (111ure uhich is reported in the LOCA calculctions perionr.ed in accordance with ICCFR50 Appendix K.
The results of the break spectrun aralysis show the single failure which results in the naximum calculated pes' clad tcrperature (FCT). For any other single failure to to core significant, its effect on tne ECCS must be greater than tais single failure. Therefore, a study vas made to determine if the ralfunction of a ,anually controlled, electrically oper-ated valve by scme unkncun cause or oy an operator imprcocrly positioning a control switch could affect the ECCS more severely than this failure. In accordance with aporo,riate n IEEE standards, the ECC System valves are electrically assiCnad to difforcnt divisions of poter supply. The effect of an onorator lucroperiv oc waium a :, i ngl e ...i W . vr. U.c ce..t;*cl p: ncl is to ccuse cnly a siat'le vrive tb i nye to an incorrect Ecs1 tion. ror the operater error of actualitig a sinf.e s'.; itch cf t' c 195 Sycte,1, the system valvcs are not actuated. Ho. sever, the consecuences of a malfunction which causes one ADS valve to inadvertently open has been noted. The sumr.ary of the ECCS Valve Sirgle Failure Analysis is provided in the attache,d Tcble 4. Cc~ paring the effects of the single valve failure noted in Table I with the results o.f the Appendix K LOCA analysis, it can be seen that these failures are not more severe tnan those reported. The single failures considered for the ECCS analysis are presented in Table 5 . l -
TABLE 4 g j- MONTICELLO q ECCS SINGLE VALVE FAILURE ANALYSIS
!' POSITION FOR NORMAL PLANT - OPERATION CONSEQUENCES OF VALVE FAILURE ASSUMED TOGETHER WITH '
SYSTEM VALVE (S) 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
- Suppression Pool- X Uti_lize Condensate Storage Tank water Suction Valve
- 1 Suppression Pool X Partial loss of flow due to flow to suppression pool r'
4 Test Return-
- Injection (s) ,
X X Negate HPCI , Turbine Inlet (s) .X X Negate HPCI i: I Low Pressure Inj ection(s) ' X- X Negate use of LPCI {CoolantInjection - I Minimum' Flow X Partial flow loss in one loop due to flow to suppression [- pool Cross Tie. X No LPCI fix: Negate on LPCI toop (two pumps per loop)
- ' Test Return X No consequence
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i . . l HX Bypass X X Reduce Flow due to HX Pressure Drop L Purp Suction X Negate one loop h' ;JAutomatic Depressurization '
- . System One Relief Valve - X Vessel depressurizes faster, increases rate of HPCI injection (assuming the failure of a single ADS valve i; to open does not affect the results because the effects on small breaks is insignificant with HPCI in operation)'
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-- _ - - . - . - - - - - . - - . . - - - = - - - - - - - - - - - -
TABLE 5 SINGLE FAILURES CONSIDERED FOR ECCS ANALYSIS _. PLANT SINGLE FAILURE REMAINING ECCS BWR/3 LPCI Injection Valve 2 C5 + HPCI + ADS MONTICELLO HPCI 2 CS + CPCI + ADS (SuctionBreak) l 1 l l i l Reference Plant Analysis The lead plant for this product line BWR is Quad Cities 2. ( The 60% DBA, 80% DBA analyses, additional Small Break analyses. Core Spray line break, Feedwater line break, and Main Steam line break analyses for the lead niant are applicable to this plant and are hereby incorporated by reference (3). REFERENCES
- 1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K. NED-20566 (drif t), submitted August 1974, and General Electric Refill /Reflood Calculation (supplement to SAFE Code Description) transmitted to the USAEC by Ictter, 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 o 4 FIGURE A-1.a . - NORMALIZED CORE AVERAGE INLET FLOW FOLLOWING A DESIGN BASIS ACCIDENT l\ MONT-ICELLO 8X8 SUCTF1 BRK DSR i 1
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TIME AFTER BREAK SECONDS t
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FIGURE A-2a , ,
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CORE INLET ENTHALPHY FOLLOWING ~ A DESIGN BASIS ACCIDENT MONTICELLO 8X8 SUCTN BRX DBA 560. - t A CD __I N D H 5110 . - - - - - - - - - - - - - - - - . - 00 . I 1 rs t # l : . 3 M Z . UJ 520. ] _ _I - CL - (C UJ s O __J -
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FIGURE A-2d , CORE INLET ENTHAi.PY FOLLOWING A 1.0 50. FT. BREAK MONTICELLO 8X8 SUCTN BRK 1.SO.FT. 560. , . t aa N a F-- 540. i a3 , f- - 1 r o z LU 520. _J a_ _
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TIME AFTER BREAK, SECONDS 4
FIGURE A-3a - CORE AVERAGE PRESSURE FOLLOWING A DESIG1 BASIS ACCIDENT MONTICELLO 8X8 SUCTN BRK DBA 1.2
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FIGURE A-3 d , CORE AVEPAGE PRESSURE FOLLOWING A isq. ft. BRK
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