ML20204F494

Forwards Evaluation of Large Break LOCA Analysis for Increased Steam Generator Tube Plugging,Per 10CFR50.59. Plant Operation W/Steam Generator Tube Plugging Levels Up to 12% Does Not Involve Any Unreviewed Safety Questions

ML20204F494
Person / Time
Site:	North Anna
Issue date:	03/16/1987
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
87-121, NUDOCS 8703260205
Download: ML20204F494 (56)

Text

{{#Wiki_filter:- VINGINIA ELucrnic Axn Powen Coxiwxy Ricitwoxn,Vinoxx A eilent W.L. STEWART Vaca Passinawr Nucs. sam oramarnows March 16, 1987 U.S. Nuclear Regulatory Commission Serial No. 87-121 Attn: Document Control Desk E&C/J0E/cdk Washington, D.C. 20555 Docket No. 50-338 50-339 License Nos. NPF-4 NPF-7 Gentlemen: VIRGINIA ELECTRIC AND F0WER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 LARGE BREAK LOCA ANALYSIS FOR INCREASED STEAM GENERATOR TUBE PLUCCING Virginia Electric and Power Company has performed a LOCA-ECCS reanalysis for North Anna Units 1 and 2 using the NRC-approved Westinghouse 1981 ECCS Large Break Evaluation Model with BART. The evaluation model changes described in Reference 1 were included in this analysis. The analysis was performed by our Nuclear Engineering staff in accordance with the requirements of Appendix K to 10 CFR 50. The results satisfy the acceptance criteria delineated in 10 CFR 50.46 and support continued full power operation for both North Anna Units at steam generator tube plugging levels of up to 12 percent. The reanalysis was performed using the current Technical gpecificationsFQlimitof2.15and accumulator nominal water volume of 1025 ft per accumulator. Therefore no Technical Specification changes are required to permit operation at the increased plugging level. The detailed analysis results are provided in. The results of this reanalysis have been reviewed by the Station Nuclear Safety and Operating Committee and our Safety Evaluation and Control staff. It has been determined based on this analysis that plant operation with steam generator tube plugging levels up to 12 percent does not involve any unreviewed safety questions as defined in 10CFR50.59. In accordance with 10 CFR 50.59(b), this letter is being submitted to furnish the NRC with a description of a change to the plant as described in the UFSAR, including a summary of the safety evaluation which provides the basis for the determination that the change does not involve an unreviewed safety question. Very truly yours, SW W . Stewart g 8703260205 870316 PDR ADOCK 05000338 P pyg l l

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Attachment:

1. LOCA-ECCS Safety Evaluation-for North Anna Unit Nos. I and 2.

Reference:

1. M. Y. Young, " Addendum to BART-A1: A Computer Code for the Best ~ Estimate Analysis of Reload Transients" (Special Report: Thimble Modeling in Westinghouse ECCS Evaluation Model): WCAP-9561-P, Addendum 3, Revision 1, July,1986. cc: U.S. Nuclear Regulatory Commission 101 Marietta St., N.W. Suite 2900 Atlanta, GA 30323 Mr. J. L. Caldwell NRC Senior Resident Inspector North Anna Power Station i l l l- + k 4 i 1 ,I l l I

ATTACIMENT LOCA-ECCS SAFETY EVALUATION FOR NORTH ANNA UNIT NOS. 1 AND 2

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1.0 INTRODUCTION

5 s e /' 0 ( \\ A re-analysis of the Emergency Core CoolintJ. System (ECCS) perfomance for the postulated large-breaP LO'CA has' been performed b c5?pliance wit $ Appel. dix K top i t 10 CFR 50. The resulis of this r,e-ealysis arehreiented here, and are in compliance with 10 CFR 50.46, "Acieptance Crite'ria for Emergency Core Coolip; Systems for Light yatar Reactori." Thisaba)fsikwasperformedwiththe ~ NRC-approved 1981 inudel with BART version or the' Westinghouse LOCA-ECCS evaluation model (Pef.1 and 2). The'imlysis includes the' evaluation model i ( revisions described in Reference 16 and approved by O.e NRC in Reference 17. 4 The analytical techniques us3d are in full cocp) lance with 1 rF1 50, AppendiA g s 1 i i i \\ As required by Appendix K of 10 CFR 50, certair. conservative cssumptions were made for the LOCA-ECCS analysis. The assumpti6n's pertain to to ccnditions ofi the reactor and associated safety system equipment at the time that +.he LOCA is i\\ assumed to occur, and include,such items asfthe core peaking fa:: tors, the containment pressure, and the'performar,ce of the Emergency Core Cooling System. g All assumptions and initial operating conditions used in thi<, reanalysis were the same as those used in the previous LOCA-ECCS' analysis.(Ref. 3), with the following exceptions: 1. ThegteamgeWatcrplugdaglevelwasincreasedfrom7%to12%. s I ( 2. Fuel perfotmance data' d ing the c.ew PfD thermal model (WCAP 8720 Ad%ndM. ) were, used. ' rhe fuel calculitMrs assumed a chamfered fuel 2 pelletidesign,' wf th length to diameter rai.io (L/0) of 1.2. 3. The 1981 LOCA-ECCS evaluation model with the BART code (Ref. 4, 5 and-

16) with models for enhanced convection and grid rewetting (Reference

6) was used to perform this analysis.

4. The models used for this anclysis' took advantage of Westinghouse modeling developments since the Reference 3 analysis was performed, \\s such as those in Reference 4. Inputs were revised to be consistent with the Reference 16 ;,odeling requirements. In addition, review of [ the Reference 3 annlysis res'ults identified several excessive s conservatisms that have been appropriately adjusted for thi; analysis. I - 40-J0E-2119S ,1 A~

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-With the-above changes incorporated into the analysis, it was found that the assumed heat flux hot-channel factor could be maintained at 2.15 and still ensure compilance with the 10 CFR 50.46 acceptance criteria. 4 s .p v[ 2.0 ACCIDENT DESCRIPTION e m A LOCA'is the result of a rupture of the reactor coolant system (RCS) piping or ofanyfineconnectedtothesystem. The system boundaries considered in the LOCA af.alysis are defined in the UFSAR. Sensitivity studies (Ref. 7) have a'<t indicated that a double-ended cold-leg guillotine (DECLG) pipe break is ,Nimiting. Should a DECLG break occur, rapid depressurization of the reactor ( coolant system occurs. The reactor trip signal subsequently occurs when the pressurizer low-pressure trip setpoint is reached. A safety injection system (SIS) signal is actuated when the appropriate setpoint is reached and the high-head safety injection pumps are activated. The actuation and subsequent activation ~of the Emergency Core Cooling System, which occurs with the SIS ? signallassumesthemostlimitingsingle-failureevent. These countermeasures will limit the consequences of the accident in two ways:

1. ' Reactor trip and borated water injection complement void formation in y3 causing rapid reduction of power to a residual level corresponding to

' fission product decay heat. No credit is taken in the analysis for y the insertion of control rods to shut down the reactor. ( 2., Injection of borated water provides heat transfer from the core and prevents excessive clad temperature. u v. Befor{ the break occurs, the unit is in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. During blowdona, heat from decay, hot internals, and the vessel continue to be ~ transferred to the reactor coolant system. At the beginning of the blowdown phase, the entire reactor coolant system contains subcooled liquid that , transfers heat from the core by forced convection with some fully developed nucleate, boiling. After the break develops, the time to DNB is calculated, \\. consis'.ent with Appendix K of 10 CFR 50. Thereafter, the core heat transfer is j . o q ~., 40-J0E-2119S 2 =- _______________.___________________________._________._________________J

y k..a v e, y ;- based on local conditions, with transition boiling and forced convection to steam as the, major heat transfer mechanisms. Duririg the refill period, it is assumed that r,od-to-rod radiation is the only core heat transfer mechanism. The' heat transfer between the reactor coolant system and the secondary system J ,.riay be in either dir'ection, depending on the relaf,ive temperatures. For the y 2 case of continued heat a'd'dition to the secondary side, secondary-side pressure increases and the main safety valves may actuate to reduce the pressure. M akeup to the secondary side is automatically provided by the auxiliary 3 feed'/ater system. Coincident with the safety injection signal, normal 'i feedwater flow is stopped by closing the main'feedwater control valves ana ' tripping the main feedwater pumps. Emergency feedwater flow is initiated by starting the auxiliary feedwater pumps. The secondary-side flow aids in the freduction of RCS pressure. When the reactor coolant system depressurizes to 7g 600 psia,the}ccumulatorsbegintoinjectboratedwaterintothereactor coolant loops.,The conservative assumption is then made that injected. y accumulator water bypasses the core and goes out through the break until the b k {g, termination of b'ypass. This conservatism is again consistent with Appendix K a 4 " of 10 CFR 50. In addition, the reactor coolant pumps are assumed to be trfpped r k at the initiation of the accident, and effects of pump coastdown are included in the blowdown analysis. ] l,. ( 'The water injected by the accumulators cools the core, and subsequent _ operation of the low-head safety injection pumps supplies water for long-term,c;ooling. When the refueling w:ater storage tank (RWST) is nearlysempty, long-term cooling of t'he core is accomplished by switching to the recirculation mode' of. core cooling, in which theyspjlled borated water is drawn fr'om the containment sump by the low-head safety injection pumps and returned to'Uye. reactor vessel. 6 4 Thezc' ntainment spray system and thelrecirculation spray' system operate to e return. the containment environment jo'subatmospheric pressure. I t 40-J0E-2119S 3 i S' - w.

3 x 3.0 ANALYSIS f. The large-break LOCA transient is divided, for analytical purposes, into three phases:. blowdown, refill, and reflood. There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the reactor coolant system, the pressure and temperature transient within the containment and the fuel clad temperature transient of the hottest fuel rod in the core. Based on these considerations, a system of interrelated computer codes has been developed for the analysis, 20 The description of the various aspects of the LOCA analysis methodology is s y given in WCAP-8339 (Ref. 8). This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the codes i that ensure compliance with 10 CFR 50, Appendix K. The SATAN-VI, C0CO, WREFLOOD, BART, and LOCTA-IV codes, which are used in the LOCA analysis, are vg fI described in detail in WCAP-8306 (Ref. 9), WCAP-8326 (Ref. 10), WCAP-8171 (Ref. 11), WCAP-9695 (Ref. 4) and WCAP-10062 (Ref. 5), and WCAP-8305 (Ref. 12), respectively. These codes assess whethec _fficient heat transfer geometry and core amenability to cooling are preserved during the time spans applicable to the blowdown, refill, and reflood phases of the LOCA. The SATAN-VI computer ,< code analyzes the thermal-hydraulic transient in the reactor coolant system during blowdown, and the C0C0 computer code calculates the containment pressure t; transient during all three phases of tha LOCA analysis. The thermal-hydraulic .j response of the. reactor coolant system during refill and reflood is calculated by the WREFLOOD computer code. A mechanistic estimate of the heat transfer coefficient in the core during reflood is provided by the BART computer code. .For the three phases of the LOCA, the LOCTA-IV computer code is used to compute the thermal transient of the hottest fuel rod. .\\ SATAN-VI is used to determine the RCS pressure, enthalpy, and density, as well as the mass and energy flow rates in the reactor coolant system and steam-generator secondary, as a function of time during the blowdown phase of the LOCA. SATAN-VI also calculates the accumulator mass and pressure and the pipe break mass and energy flow rates that are assumed to be vented to the containment during blowdown. At the end of the blowdown, the mass and energy release rates during blowdown are transferred to the C0C0 code for use in the 40-J0E-2119S 4 )

determination of the conta ment pressure response during this first phase of the LOCA. Additional SATAN-VI output data from the end of the blowdown, including the core inlet flowrate and enthalpy, the core pressure, and the core power decay transient, are input to the LOCTA-IV code. With input from the SATAN-VI code, WREFLOOD uses a system thermal-hydraulic model to determine the core flooding rate (i.e., the rate at which coolant enters the bottom of the core), the coolant pressure and temperature, and the quench front height during the refill and reflood phases of the LOCA. WREFLOOD also calculates the mass and energy flow rates that are assumed to be vented to the containment. Since the mass flowrate to the containment depends upon the core pressure, which is a function of the containment backpressure, the WREFLOOD and C0C0 codes are interactively linked. With the input and boundary conditions from WREFLOOD, the mechanistic core heat transfer model in BART calculates the fluid and heat transfer conditions in the core during reflood. LOCTA-IV is used throughout the analysis of the LOCA transient to calculate the fuel and clad temperature of the hottest rod in the core. The input to LOCTA-IV consists of appropriate thermal-hydraulic outputs from SATAN-VI, WREFLOOD and BART, ar.d conservatively selected initial RCS operating conditions. These initial conditions are summarized in Table 1 and Figure 1. The axial power shape of Figure 1 cssumed for LOCTA-IV is a chopped cosine curve that has been previously verified (Ref. 13) to be the shape that produces the maximum peak clad temperature. The C0C0 code, which is also used throughout the LOCA analysis, calculates the containment pressure. Input to C0C0 is obtained from the mass and energy flowrates assumed to be vented to the containment, as calculated by the SATAN-VI and WREFLOOD codes. In addition, conservatively chosen initial containment conditions and an assumed mode of operation for the containment cooling system are input to C0C0. These initial containment conditions and assumed modes of operation are provided in Table 2. 40-J0E-2119S 5

4.0 NON-LOCA SAFETY EVALUATION FOR 12% STEAM GENERATOR TUBE PLUGGING This North Anna Power Station LOCA-ECCS reanalysis has evaluated plant operation at steam generator tube plugging levels of up to 12% based on the acceptance criteria delineated in 10CFR50.46. An evaluation has been performed which concluded that reanalysis of non-LOCA accidents is not required to support this increased tube plugging level. Steam generator tube plugging in sufficient quantity can potentially affect non-LOCA safety analysis due to reduced primary system flow, more severe pump coastdown characteristics, and the reduction of the reactor primary coolant system volume. Primary flowrate becomes a key parameter in DNB limited events (e.g., Uncontrolled RCCA Bank Withdrawal at Power) when it falls below the thermal design flowrate. Pump coastdown characteristics impact analysis results when they become more severe than the conservative values used in the loss-of-flow related analyses. The reduced primary coolant system volume affects dilution times in uncontrolled boron dilution events. We have evaluated these concerns for the Surry Power Station Units in the past, when tube plugging levels became significant (greater than approximately 20 percent).(14) Flow measurements have been taken at the North Anna Power Station for several levels of steam generator tube plugging. These data were employed to obtain a conservative estimate of RCS flow versus tube plugging in Reference 18. The curve of flow versus plugging level presented in Reference 18 indicates that the conservatively estimated flow rate at the proposed 12% plugging level is still considerably larger (by approximately 2%) than the North Anna thermal design flow. Therefore, the current docketed licensing analyses remain valid for those events in which flowrate is an important concern. The loss-of-flow related analyses in Reference 3 used a limiting flow coastdown characteristic with the limiting ITDP initial flow rate. Since the conservatively estimated system flow rate is greater than the ITDP value, the coastdown flows for the 12% plugging level will be bounded by the coastdown flows in the Reference 3 analyses. The impact of 12% tube plugging on dilution times in the uncontrolled boron dilution events is bounded by the analyses documented in Reference 15. Relative to the boron dilution events, the analyses indicated: 40-J0E-2119S 6

p.,. For uncontrolled dilution during startup, time to criticality is at least 61 minutes. This is more than adequate time for the operator to recognize the high count rate signal and terminate the dilution flow. For uncontrolled dilution at power, the operator has ample time (57 minutes) after the over-temperature AT alarm or trip to determine the cause of dilution, isolate the water source, and initiate reboration before total shutdown margin is lost due to dilution. Tube plugging levels exhibit no influence on dilution times for the refueling mode of operation, since the steam generator volumes are not a part of the active system. This evaluation shows that for steam generator tube plugging levels of up to 12 percent, no reanalysis of the non-LOCA safety events is necessary and that the currently licensed analyses remain valid. 5.0 LARGE BREAK LOCA RESULTS Tables 1 and 2, and Figure 1 present the initial conditions and modes of operation that were assumed in the analysis. Table 3 presents the time sequence of events, and Table 4 presents the results for the double-ended cold-leg guillotine break for the C = 0.4 and 0.6 discharge coefficients. The D double-ended cold-leg guillotine break has been determined to be the limiting break size and location based on the sensitivity studies reported in Reference 7. The analysis resulted in a limiting peak clad temperature of 2080.0 F for the C = 0.4 case, a maximum local cladding oxidation level of 4.44%, and a D total core metal-water reaction of less than 0.3%. The detailed results of the LOCA re-analysis are provided in Tables 3 through 6 and Figures 2A through 18B. The figures show the following: 1. Peaking Factor vs. Core Height - Figure 1 shows the chopped cosine power shape used in the analysis. 40-J0E-2119S 7

2. Mass Velocity - Figures 2A and 2B show the mass velocity at-the clad burst and hot-spot locations on the hottest fuel rod for the discharge coefficient used. 3. Heat Transfer Coefficient - Figures 3A and 3B show the heat transfer - coefficient at the clad burst and hot-spot locations on the hottest rod for the discharge coefficient used. -The values of heat transfer coefficient that are shown were calculated by the LOCTA-IV code based on equations for heat transfer in the nucleate boiling, transition boiling, film boiling, and steam cooling regimes. 4. Core Pressure - Figures 4A and 4B show the calculated pressure in the core for the discharge coefficient used. 5. Break Flowrate - Figures 5A and 5B show the calculated flowrate out of the break for the discharge coefficient used. The flowrate out of the break is plotted as the sum of flow at both the pressure vessel end and the reactor coolant pump end of the guillotine break. 6. Core Pressure Drop - Figures 6A and 6B show the calculated core pressure drop for the discharge coefficient used. The core pressure drop is interpreted as the pressure immediately before entering the core inlet to the pressure just outside the core outlet. 7. Peak Clad Temperature - Figures 7A and 78 show the calculated hot-spot clad temperature transient and the clad temperature transient at the burst location for the discharge coefficient used. The peak clad temperature for the limiting discharge coefficient of 0.4 is 2080.0 F at the 8.00 ft elevation in the core. 8. Fluid Temperature - Figures 8A and 8B show the calculated fluid temperature for the hot spot and burst locations for the discharge coefficient used. 9. Core Flow - Figures 9A and 98 show the calculated core flow, both top and bottom, for the discharge coefficient used. 40-J0E-2119S 8

10. Reflood Transient - Figures 10A and 10B show the reactor pressure vessel downcomer and core water levels for the discharge coefficient used.

Figures 11A and 11B show the core inlet velocity for the discharge coefficient used.

11. Accumulator Flow - Figures 12A and 12B show the calculated flow for the discharge coefficient used. The accumulator delivery during blowdown is discarded until the end of bypass is calculated.

Accumulator flow, however, is established in the refill-reflood e calculations. The accumulator flow assumed is the sum of that injected in the intact cold legs.

12. Pumped ECCS Flow (Reflood) - Figures 13A and 13B shcw the calculated flow of the emergency core cooling system for the discharge coefficient used.

13. Containment Pressure - Figures 14A and 14B show the calculated pressure transient for the discharge coefficient used. The analysis of this pressure transient is based on the data given in Tables 2, 5, and 6.

14. Core Power Transient - Figures 15A and ISB show the core power transient calculated by the SATAN-VI code for the discharge coefficient used.

15. Break Energy Release - Figure 16A and 16B show the break energy released to the containment for the discharge coefficient used.

16. Containment Wall Heat Transfer - Figure 17A and 178 show the containment wall heat transfer coefficient for the discharge coefficient used.

17. Fluid Quality - Figures 18A and 18B show the fluid quality at the clad burst and hot-spot locations (location of maximum clad temperature) on the hottest fuel rod (hot rod) for the limiting breaks. 40-J0E-2119S 9

6.0. CONCLUSIONS For breaks up to and including the double-ended rupture of a reactor coolant pipe, and for the operating conditions specified in Tables 1 and 2, the emergency core cooling system will meet the acceptance criteria as presented in 10 CFR 50.46, as follows: 1. The calculated peak fuel rod clad temperature is below the requirement of 2200 F. 2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1% of the total amount of Zircaloy in the reactor. 3. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The localized cladding oxidation limits of 17% are not exceeded during or after quenching. 4. The core remains amenable to cooling during and after the break. 5. The core temperature is reduced and the long-term decay heat is removed for an extended period of time. 40-J0E-2119S 10

10 CFR 50.59 SAFETY EVALUATI0ft The proposed changes have been reviewed against the criteria of 10 CFR 50.59 and do not involve any unreviewed safety questions. The specific bases for this determination are as follows: 1. Since the proposed changes involve parameters which are not accident initiators they will not increase the probability of occurrence of any malfunction or accident previously addressed. The reanalyzed large break LOCA analysis verifies that operation with the increased allowable steam generator tube plugging limit would also not result in any increase in accident consequences. 2. No nett accident types or equipment malfunction scenarios will be introduced as a result of operating with the increased steam generator tube plugging. The change which potentially affects physical components in the plant ystems (steam generator tube plugging) was explicitly included in the analysis and shown not to produce any new or unique accident precursors. 3. The margin of safety, as defined in the basis for the plant Technical Specifications, is not reduced. Operation with the increased steam generator. tube plugging level has been demonstrated not to reduce the margin to the LOCA acceptance limits. 40-J0E-2119S 11

8.0 REFEREllCES

1. Letter'from J. R. Miller, NRC, to E. P. Rahe, Westinghouse, " Acceptance for Referencing of the 1981 Version of the Westinghouse Large Break ECCS Evaluation Model," December 1, 1981.

2. Letter from C. O. Thomas, NRC to E. P. Rahe, Westinghouse, " Acceptance for Referencing of Licensing Topical Report WCAP-9561, BART A-1: A-Computer Code for Best Estimate Analyses of Reflood Transients," December 21, 1983, and Addenda 1 and 2.

(

3. Letter from W. L. Stewart, Vepco, to H. R. Denton, NRC, Serial No. 85-077, dated May 2, 1985.

4. Young, M. Y, et al., BART-A1: A Computer Code for the Best Estimate l

Analysis of Reflood Transients, WCAP-9695, January 1980. I

5. Chiou, J. S. et al., Models for PWR Reflood Calculations using the BART Code, WCAP-10062, December 1981.

6. Letter from C. O. Thomas, NRC, to E. P. Rahe, Westinghouse, " Acceptance for Referencing of Licensing Topical Report WCAP-10484(P), Spacer Grid Heat Transfer Effects During Reflood," June 21, 1984.

-7. R. Salvatori, Westinghouse-ECCS Sensitivity Studies, WCAP-8356, July 1974.

8. F. M. Bordelon et al., Westinghouse ECCS Evaluation Model - Summary, WCAP-8339, July 1974.

9. F. M. Bordelon et al., SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant, WCAP-8306, June 1974.

10. F. M. Bordelon and E. T. Murphy, Containment Pressure Analysis Code

.C0 ), WCAP-8326, June 1974. 40-J0E-21195 12 )

'11. R. D. ' Kelly et al., Calculational Model for' Core Reflooding After a Loss-of-Coolant Accident (WREFLOOD Code), WCAP-8171, June 1974.

12. F. M. Bordelon et al., LOCTA-IV Program: Loss-of-Coolant Transient Analysis, WCAP-8305, June 1974.

13. Letter from C. M. Stallings, Vepco, to E. G. Case, NRC, Serial No. 092, dated February 17, 1978.

14. Letter from C. M. Stallings, Vepco, to E. G. Case, NRC, Serial No. 344, August 9, 1977.

15. Updated Final Safety Analysis Report - North Anna Power Station Units 1 and 2, Virginia Electric and Power Company, Rev. 4, June 1985.

16. M. Y. Young, " Addendum to BART-A1: A Computer Code for the Best Estimate Analysis of Reload' Transients " (Special Report: Thimble Modeling in Westinghouse ECCS Evaluation Model): WCAP-9561-P, Addendum 3, Revision 1, July, 1986.

17. Letter from Charles E. Rossi, NRC, to E. P. Rahe, Westinghouse, " Acceptance for Referencing of Licensing Topical Report WCAP-9561, Addendum 3, Revision 1," August 25, 1986.

18. Letter from R. H. Leasburg, Vepco, to H. R. Denton, NRC, Serial No. 080, February 12, 1982.

40-J0E-2119S 13

TABLE 1 INITIAL CORE CONDITIONS ASSUMED FOR THE DOUBLE-ENDED COLD-LEG GUILLOTINE BREAK (DECLG) Calculational Input Core Power (MWt) 102% of 2898 Peak linear power (kW/ft) 102% of 12.220. Heat flux hot-channel factor (F ) 2.15 g Enthalpy rise hot-channel factor (F I) 1.55 H 3 Accumulator water volume 'ft, each) 1025 Reactor vessel upper herd temperature equal to Thot Limiting Fuel Region and Cycle Cycle Region Unit 1 All All regions Unit 2 All All regions 40-J0E-2119S 14

w-.- l. l_ TABLE 2 CONTAllOENT DATA (DRY CONTAllOENT) F 6 3 Net Free Volume l'.916 x 10 ft Initial Conditions Pressure (total)', psia 9.50 Temperature, F 90 RWST temperature, F-35 -Outside temperature, F -10 Containment Quench Spray System Number of pumps operating 2 Runout flowrate (each), gpm 2000 Actuation time, sec 59 i Structural Heat Sinks 2 Type / thickness (in.) Area (ft ), with uncertainty Concrete /6 8,393 Concrete /12 62,271 Concrete /to 55,365 Concrete /24 11,591 Concrete /27 9,404 Concrete /36 3,636 Carbon steel /0.375, Concrete /54 22,039' Carbon Steel /0.375, Concrete /54 28,933 Carbon steel /0.50, Concrete /30 25,673 Concrete /26.4 (floor), Carbon Steel /0.25, Concrete /120 12,110 Carbon steel /0.371 160,328 Stainless Steel /0.407 10.527 Carbon Steel /0.882 9,894 Carbon Steel /0.059 60,875 40-J0E-2119S 15

TABLE 3 TIE SEQUENCE OF EVENTS FOR DECLG C = 0.4 C = 0.6 D D (sec) (sec) Start 0.0 0.0 Reactor trip 0.630 0.615-Safety injection signal 2.60 2.07 Accumulator injection 17.10 13.00 Pump injection 27.60 27.07 ( End of bypass 32.57 27.10 End of blowdown 32.57 27.10 Bottom of core recovery 46.39 40.70 } Accumulator empty 57.02 51.82 40-J0E-21195 16

TABLE 4 RESULTS FOR DECLG C = 0.4 C = 0.6 D D Peak clad temperature, F 2080.0 1873.8 Peak clad location, ft 8.00 7.25 Local Zr/H O reaction 2 (max),% 4.44 2.61 Local Zr/H 0 location, f t 8.00 6.50 2 Total Zr/H O reaction, % <0.3 <0.3 2 Hot-rod burst time, sec 43.00 67.40 Hot-rod burst location, ft 5.75 6.50 40-J0E-2119S 17

I E TABLE 5 REFLOOD MASS AND ENERGY RELEASES DECLG (C = 0.4) D Total Mass TotalEgergy Time (sec) Flow Rate (lb/sec) Flow Rate (10 Btu /sec) 46.386 0.0 0.0 47.611 0.67 0.009 57.147 86.90 1.080 71.672 118.16 1.214 89.572 228.51 1.471 109.272 253.93 1.474 130.572 260.90 1.427 162.572 303.77 1.458 TABLE 6 BROKEN LOOP ACCUMULATOR FLOW TO CONTAINMENT DECLG (C = 0.4) D a Time (sec) Mass Flow Rate (lbm/sec) 0.00 4095.55 1.01 3691.53 3.01 3155.48 5.01 2801.81 7.01 2542.15 10.01 2250.29 15.01 1912.00 20.01 1678.42 25.01 1514.79 30.01 1546.18 aFor energy flowrate, multiply mass flow rate by a constant of 59.62 Btu /lbm. i 40-J0E-2119S 18

1 i AXIAL POWER SHAPE 12 PCT SGTP 1981 MODEL WITH BART 2.2 e w l / -_ N p l 2* / \\ t 1.8 p 1.7 r l U } \\ 2* 2 / i \\ 2* ) e / \\ i B / \\ l 2' q 1.3 I / i \\ A 1.1 1 / ~l \\ e4 M 1 g_ i / \\ l / l \\ j / \\ j 0.5 3 r 0.4 i 0 2 4 8 8 10 12 CORE HEIGHT (FT) I FIGURE 1 PEAKING FACTOR VERSUS CORE HEIGHT - FQ = 2.15 1

VRA LOCTA DECK CD:0.4 12 PCTTP 02-09-87 BURST F0:2.15 TAVG:586.8F NEW PAD-FDH:1.55 CHAMFERED FUEL L/D=1.2 TiilMBLE FIX INCLUDED MASS VELOCITY

BURST, 5.75 FTC )

PEAK, 8.00 FT(*) Symbol (*) is offset a constant distance above plotted values. 2e.

u. is.

J \\. \\bf %4 y t J l 3. \\ / \\ ,/ l l v g 2s. i l ies iel te2 is5 02/09/87 FIGURE 2A MASS VELOCITY VERSUS TIME DECLG (CD = 0.4)

l lj l F 8 6 8 5 7 G 8 V / A 1 1 D ) 1 E + / 5 T D ( e 20 R U T m l A L F B C N 5 2 T I S ^ R X 7 N U l B F K t 5 E A f 1 L E v . B P 2 M i s I e r 0 H ) E u M F T l p I C av ,~ T 7 2 T S) 8 . F d U6 e 1 S t 8R0 1 0 t 2E 1 D 5 o V= lp E / c RYD 2 L 6 UTC 1, 0 e ts GI( v ( IC i L o P E , b 1 FOG LL a I U T EC i VE T F S e t D E R c S n j ~- S P D U a A E B t M s 2 R i L 1 E d F t 6 M n A a t 0 H s j C n o D c C 5 a 5 K t C 1 Y es E T f f D H 1 o D C A F O s i T L C D E ) D A V ( L P 3 S lo - A W S b2 e e a a 2 a e z s 4 s t t R E A m y . N M S _a3. E 3;! i' Iljl

VRA LOCTA DECK CD:0.4 12 PCTTP 02-09-37 BURST F0:2.15 TAVG:586.8F l NEW PAD-FDH:1.55 CHAMFERED FUEL L/D=1.2 THIMBLE FIX INCLUDED HEAT TRANS. COEFFICIENT

BURST, 5.75 FTC )

PEAK, 8.00 FT(=) Symbol (*) is offset a constant distance above plotted values. ] 185 ~ y 1 i i I i A / I [ f S A IV e Ili h 5 s G t VsD%~ ~ t,J j ist i E Y l l l I t a". s 25. sa. 75. too. 125. Isa. 17s. 2es. 225. 25s. 27s. see. 02/09/87 i FIGURE 3A HEAT TRANSFER COEFFICIENT VERSUS TIllE DECLG (CD = 0.4)

VRA LOCTA DE CK CD:0.6 12 PCTTP 02-11-87 FO:2.15 BURST BART TAVG:586.8F f.EW PAD-FDH:1.55 CHAMFERED FUEL L/D 1.2 THIMBLE FIX INCLUDED HEAT TRANS. COEFFICIENT

BURST, 6.50 FTC )

PEAK, 7.25 FT(*) Symbol (*) is offset a constant distance above plotted values. te5 W i 4 ~ l i I i l I a 1E' 4 } \\ \\ l f m = i T q E i M 32-9 c 1 2 i c i Y i I c. 2s. se. 7s.

ez.

r2s. iss. 27s. 2aa. 22s. 2ss. 27s. see. ~ t in ist c ' 02/11/87 m FIGURE 3B HEAT TRANSFER C0 EFFICIENT VERSUS TIME DECLG (CD = 0.6) l

1.I t l 7 8 T / U 3 P 0 N / I 2 0 5 D 3 E TADPU 03 6 X 3 AM T N 0 EM O 2 I C ) 5 T P 2

• 2

( S) U4 7 0 S 8 F AR0 4E / V - 3 ERED 0 G P 0 URC / N O 2 GU( 2 I T c' IS A n EL FSG 0 G RC G i N PE 4 U ) D a E . L i R O P ( sv O i C \\ 'D E C B U M \\ E T O S T A G T o C S O 3 t B A G Z VA 2 E T 1 RO 3 M C O G N L C N E \\ A D E T R g A 4 U S . S 0 0 0 0 0 o 0 0 0 0 0 0 S 5 0 5 0 5 1 1 A : E 2 2 R D R V C P 5 I. llllIll l l fI lI l u

i ) VRA SATAN NOM TAVG CASE CD 0.6 02/06/87 PCONT=36 UPDATED INPUT CD:0.6 DECLG 12 % SG TUBE PLUGGING F0-2.20 MAX PRESSURE CORE BOTTOM ( ) TOP, (* ) 2500 i I E 3 E 2000 e 1500 i N 1000- % \\ I N 500 T \\ i N Os. 2. 4 6. e. 18. 12. 14 1s. no. as. 22. 24 as. Je. TipE IEC 02/06/87 1 FICURE 4B CORE PRESSURE VERSUS TIME j DECLG (CD = 0.6) l

VRA SATAN NOM TAVG CASE CD 0.4 02/03/87 PCONT=36 UPDATED INPUT CD:0.4 DECLG 12 X SG TUBE PLUGGING F0:2.20 MAX BREAK FLOW .6E + 05 \\ .5E + 05 R d.4E + 05 E \\ 5.3E+05 \\ \\ .2E + 05 h \\ .1E + 05 \\ \\ -Q ~ I .lE + 05 S. J.E 5. 7.5

18. 12.5 IE. 17.5 28. 22.5 25. 27.5 SS. 52.5 55.

TIE ISECI 02/03/87 FIGURE 5A BREAK FLOW RATE VERSUS TITE DECLG (CD = 0.4) i i ? i l l ~

7 8 T / U 6 P 0 N / I 2 0 D E TA D P U 6 X 3 A = M T N 0 EM O 2 I C T P 2 S U) 7 0 S6 R 8 F t 0E0 / 5V = 60 G EE RTD t ' / N UAC GR( 2 I I FWG 0 G 4 OL w G t LC FE 6 U D . L K 0 P A \\ t ER D E B C B N 1 U E T S A G C S G I VA 2 T 1 4 \\ M O G 2 \\ N L C W N E O A D L 5 5 5 5 5 5 5 0 5 T F A 6 + + + + + + + + K S E E E E E E E E 0 A 7 6 5 4 3 2 l 1 A E R D R i C B $w

VRA SATAN NOM TAVG CASE ED 0.4 02/03/87 PCONT:36 UPDATED INPUT CD=0.4 DECLG 12 % SG TUBE PLUGGING FO-2.20 MAX CORE PR. DROP so. O se. E_

i. ~~

E as. l 5 e. r h -30. j -40. l l -ee. -se. l S. 2.5 5. 7.5 Ie. 12.5 15. 17.5 as. 22.5 25. 27.5 m. 52.5 55. TifE lEC3 02/03/87 1 FIGURE 6A CORE PRESSURE DROP VERSUS TIME DECLG (CD = 0.4)

VRA SATAN NOM TAVG CASE ED 0.6 02/06/87 PCONT:36 UPDATED INPUT,, CD=0.6 DECLG 12 % SG TUBE PLUGGING FO:2.20 MAX ~ CORE PR. DROP 'l 6, 1 l SC. ~ 2 68. E % 48. -m,.- 5 l g l = as. ,-), I u l 8 {~ .i s. m j -W l r* s \\ l ( "$ t 7 l .sa. v ^^ l -68. 1 i y 88. S. 2. 4 6. S. IS. 12. 14 !&< IS. 29. 22. 24 25. 26. T!st EffCI 1 02/06/87 s

. FIGURE 6B.

, CORE JRESSdnE DROP VE)$$3 -TI!*E. DECLG (CD = 0.6) / r e + y E +

l1j, J

F 8 7 8 6 / 8 9 5/D ) 0 E / G D ( .s 2 V U T a 0 A L F s N T,C N 0 .s 5 l 0 7 X '6 2 1 2 I

F a

0 K s2 F E A L E' T B f; N s S,M 2 T N 2 N R I E s U H ) e I S B T u .e N C l s A a N 7 2 T v 2 RT) 8 . F d 4 1 e E AR0 9 = 5 t s 7U t 0 D 7 7 w m T= o 1 EA / l C' RRD p 2 ,L _ 5 UEC s" 0 GP( e IM Pju L - v s' FEG o t g b i" TL - W C T T a DE AD TF S e s L / E R c r 2 C n P D U a 1 K E B t A s 2 R E i E. d s P 1 s r ./ t i 4 M D n . A 0 a t 0 H R s C n s o 7 D T c ..C 5 O ~. 5 H a K . t s A C 1 P e s s /v E Mf D H E f o D T. s A F s 2 i T G A C D V) O A A (* 1 L P s ~ Dl s. os s s A W Ab e e mk e, s, n s R E L y i V N CS _ ;s" e 9- " gd j1 l l!

,..)- 1 j x., j ~ VRA LOCTA DECK CD:0.6 12 PCTTP 02-10-87 FQ:2.15 N08tNtST 8 ART TAVG:586.8 NEW PAD-FDH:1.55 CHAMFERED FUEL LID 1.2 THIMBLE FIX IECLUDED I CLAD AVG. TEMP. HOT ROD BURST. 6.50 FT( ). PEAK. 7.M FT( * ) + i 2500 r~ n,r ^ l 2000 l ' r A [ N I 1 l A N N 1500 g i K N ~ N [ ~ 1000 i ~ \\ 500 \\ O 0 25 50 75 100 125 150 175 200 225 250 275 300 i 02/10/87 FIGURE 7B l PEAK CLAD TEMPERATURE TRN4SIEllT DECLG (CD = 0.6) d

1, VRA LOCTA DECK CD=0.4 12 PCTTP 02-09-87 BURST F0:2.15 TAVG:586.8F NEW PAD-FDH:1.55 CHAMFERED FUEL L/D=1.2 THIMBLE FIX INCLUDED FLUID TEMPERATURE

BURST, 5.75 FTC )

PEAK, 8.00 FT(*) Symbol (*) is offset a constant distance above plotted values. 2000.

1758.

- }IM N ^ g 1588 \\ g 1258. L r x x m x w f E 758. E h W. y 9 3 ' 25s. l

• e.

25. 58. 75. ISS. 125. ISS. 175. 298. 225. 250. 275. 588. 02/09/87 FIGURE 8A FLUID TEMPERATURE VERSUS TIME DECLG (CD = 0.4)

VRA LOCTA DE CK CD:0.6 12 PETTP 02-11-87 F0:2.15 BURST BART TAVG:586.8F NEW PAD-FDH:1.55 CHAMFERED FUEL L/D:1.2 THIMBLE FlX INCLUDED FLulD TEMPERATURE

BURST, 6.50 FT( )

PEAK, 7.25 FT(=) Symbol (*) is offset a constant distance above plotted values. 2033.

1750.

o U /1 f N' X ^ 1258. 3 3 \\ N ,iaee. \\ \\ \\ ~ [ /

258.

y x b \\ 530. " 258. ~ 8. 25. 58. 75. 102. 125. 158. 175. 200. 225. 253. 275. 588. 02/11/87 FIGURE 8B FLUID TEMPERATURE VERSUS TIME DECLG (CD = 0.6)

7 8 T / U 3 P 0 N / I 2 0 D 5 E 5 TA 5 D 2s PU .s 6 X e ) 3 A M = M 5 O T T 7 T N 0 2 OB O 2 D C ) 5 N P 2 2 A ( ) 7 O 5 P4 0 8 F 2 AT0 / 2 9( = 3 EE 0 G P R! D s UI C / N O al GT( T C 2 I I 0 G

5. E FSG S

UL G 7 SC I 4 U ) 1 RE E . L M ED .I V 0 P ( 5T W 1 O D E L C B S. F U M 2 E E T O 1 R ~ O S T C A G T 8 C S O 1 B 5 G % V 7 A 2 E T 1 R O l M C 5 e O G N L l 5 L C E 2 N E T A D A 4 T R J e A 4 W 0 0 0 0 0 0 0 0 0 S . O 0 0 0 0 0 0 0 0 0 L 0 0 0 0 0 0 0 0 8 6 4 2 2 4 6 8 A F R D V C Z C E k i I ll l l! .;j ll1 l l ll

2 VRA SATAN NOM TAVG CASE CD 0.6 02/06/87 PCONT 36 UPDATED INPUT' CD:0.6 DECLG 12 % SG TUBE PLUGGING F0:2.20 MAX 2-FLOWRATE CORE BOTTOM ( ) TOP, (* ) 8000 C D6000{-l 4000 5 } 2000 I n, _J O ' 'f N / l .zeee I 1 l -4000 II -6000 -8000 i S. 2. 4. 6. 8. 10. 12. 14. 16. 19. 28. 22. 24 26. 29. ] Tipt 15CCI 02/06/87 l FIGURE 98 CORE FLOW VERSUS TIME (TOP AND BOTTOM) DECLG (CD = 0.6)

VIRGINIA POWER VRA RCO 1981 MODEL - FOR BART ANALYSIS 0.4 DECLG 2.15 F0T 12 PC TUBE PLUGGING 275 PSIG BACKFILL 02-07-87 WATER LEVEL (FT) 29. 17.5 15.

r h

- i2.5 Y U i iS. Uu i 7.5 5. / 2.5 ( 48, 68. SS. ISS. 129. I48. 168. ISS. 298. 229. 248. 260. 298. 538. 528. TIE ISCCB 02/07/87 FIGURE 10A REFLOOD TRANSIENT - CORE AND DOWNCOMER WATER LEVELS DECLG (CD = 0.4)

7 8 9 i 0 7 2 8 0 / S 9 I L 0 S L / Y 2 I = L F 0 s A K N C A A e B es T S R G L .s E A I s V B S a E L P R R e E O 5 F 7 m TA 2 W R = E = M L O E G C D N N) = W6 O I O M G 2 BD0 0 G 1D= 1 U s N 8 L s EAD a* R C 9 P UE( n*' GR 1 IOG E FCL B C U i* E D T T O n' N E C C i I S R P ~ NA = R A 2, T i R 1 D V O T n O O L i F R F ER EW 5 s O 1 ) i T P 2 F A C 1 / L I N G E I L V a G C E / R E L c D I V R a 6 E 5 s. s. s. s s s . T M 7 i 2 i 7 2 0 A 1 1 W _3WM," I d ' i j:!!

^

,i.! !!Ii t I:l l!lii i i ii <

_i 7 8 7 0 7 2 8 0 / S 7 I L 0 S L / 2 Y I L F 0 A K N C A A B TR G A I B S P Y R T O 5 I e C F 7 2 O 2 L E V L 2 T E G E L D N .e N) O I i a I4 M G t AE0 G 1R 1O= 1 U 0 C 8 L 0 E D 2' R C 9 P C U ( 1 .E GT E eS I NG B e' FEL s IC U E SE ND T .l O ei AR s C C T n R P D O A 2 e O a L R 1 ~ l F V E R T e O a R F ) t E C m W 5 E ee O 1 S w s P . / 2 N A I e I C e N G E I L T G C A \\ e R E R s D I V D .e 4 O k O s. s. 2 i 0 L t F gM1= ~

VIRGINIA POWER VRA REO 1981 -HODEL - FOR BART ANALYSIS 0.6 DECLG 2.15 FOT 12 PC TUBE PLUGGING 275 PSIG BACKFILL 02-09-87 FLOOD RATE (IN/SEC) l i G 1.5 U 1: \\ N 1 N m g,- f m ? ~- .5 I 1 i 48. 68. SS. ISS. 129. 148. 168. leJ. 200. 228. 248. 268. 200. 588. $29. 111T SEC 02/09/87 l FIGURE 11B l REFLOOD TRANSIENT - CORE INLET VELOCITY DECLG (CD = 0.6)

Il' i l 7 8 T / U 3 P 0 N / 2 I 0 D 55 E TA 5 D 2 y 5 PU 8 6 X 5 ) 3 A iW l

M 1

5 0 T 7 D N G 2 W O O Z L C 5 B ( P 2 2 / E 7 0 / 5 l) l 4 I 8 F 2 T / 2 A 0 3 2S b 1U= 0 G S 1 N NI ERD REC 2 I C UV( E G 0 G

5. S IWG G

A 7 FOL I 4 U 1 LC E FE . L f D 0 P .i R 5 T O 1 T D E A L C B 5 U U 2 M E T U 1 C S C A G A 8 C S 1 G % V 5 A 2 7 T 1 M 5 1 O G N L W C O 5 N E L 2 A D F T A 4 .S M S 3 S m S 8 A

C M

8 0 U 0 8 85 1 R D C S V C A uE; >36 f5 {1ll f

VRA SATAN NOM TAVG CASE CD 0.6 02/06/87 PCONT=36 UPDATED INPUT CD:0.6 DECLG 12 % SG TUBE PLUGGING F0:2.20 MAX ACCUM. FLOH Se. s- d s iO V N O ,}}