ML18149A546

Proposed Tech Specs,Increasing Allowable Operating Band for Accumulator Water Vol

ML18149A546
Person / Time
Site:	Surry
Issue date:	02/23/1987
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
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ML18149A545	List: ML18149A544 ML18149A546
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NUDOCS 8703030024
Download: ML18149A546 (63)

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I e e ATTACHMENT 1 TECHNICAL SPECIFICATIONS CHANGES FOR INCREASED FQ(Z) AND ACCUMIJLATOR WATER VOLUME OPERATING BAND SURRY UNITS 1 AND 2

~03030024 870223 .

PDR ADOCK 05000280 I p PDR

TS 3.3-1 3.3 SAFETY INJECTION SYSTEM Applicability Applies to the operating status of the Safety Injection System.

Objective To define those limiting conditions for operation that are necessary to provide sufficient borated cooling water to remove decay heat from the core in emergency situations.

Specifications A. A reactor shall not be made critical unless the following conditions are met:

1. The refueling water storage tank contains not less than 387,100 gallons of borated water. The boron concentration shall be at least 2000 ppm and not greater than 2200 ppm.

2. Each accumulator system is pressurized to at least 600 psia and 3 3 contains a minimum of 975 ft and a maximum of 1025 ft of borated water with a boron concentration of at least 1950 ppm.

e TS 3.3-8 Time After Shutdown Decay Heat,% of Rated Power 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.3 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.75 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 0.48 Thus, the requirement for core cooling, in case of a postulated loss-of-coolant accident while in the hot shutdown condition is reduced by orders of magnitude below the requirements for handling a postulated loss-of-coolant accident occurring during power operation. Placing and maintaining the reactor in the hot shutdown condition significantly reduces the potential consequences of a loss-of-coolant accident, allows access to some of the Safety Injection System components in order to effect repairs, and minimizes the exposure to thermal cycling.

Failure to complete repairs within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of going to hot shutdown condition is considered indicative of unforeseen problems, i.e., possibly the need of major maintenance. In such a case the reactor is to be put into the cold shutdown condition.

The accumulators are able to accept leakage from the Reactor Coolant System without any effect on their availability. Allowable inleakage is based on the volume of water that can be added to the initial amount without exceeding the volume given in Specification 3.3.A.2. The maximum acceptable inleakage is 50 cubic feet per tank.

• TS 3.12-3 B. Power Distribution Limits

1. At all times except during low power physics tests, the hot channel factors defined in the basis must meet the following limits:

FQ(Z) ~ 2.32/P x K(Z) for p > 0.5 FQ(Z) ~ 4.64 x K(Z) for P ~ 0.5 N

F8H ~ 1.55 [1 + 0.3 (1-P)] for three loop operation

~ 1.55 [1 + 0.2 (1-P)] for two loop operation where Pis the fraction of rated power at which the core is operating, K(Z) is the function given in TS Figure 3.12-8, and Z is the core height location of FQ.

2. Prior to exceeding 75% power following each core loading and during each effective full power month of operation thereafter, power distribution maps using the movable detector system shall be made to confirm that the hot channel factor limits of this specification are satisfied. For the purpose of this confirmation:

_Meas

a. The measurement of total peaking factor YQ shall be increased by eight percent to account for manufacturing tolerances, measurement error and the effects of rod bow.

The measurement of enthalpy rise hot channel factor F~H shall be increased by four percent to account for measurement error. If any measured hot channel factor exceeds its limit specified under Specification 3.12.B.1, the reactor power and high neutron flux trip setpoint shall be reduced until the limits under Specification 3.12.B.1 are met. If the hot channel factors cannot be brought to within the limits of FQ(Z) ~ 2.32 x K(Z) and N

F H ~ 1.55 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the Overpower 8T and 8

Overtemperature 8T trip setpoints shall be similarly reduced.

- TS 3.12-15 It should be noted that the enthalpy rise factors are based on integrals and are used as such in the DNB and LOCA calculations. Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in radial (x-y) power shapes throughout the core.

Thus, the radial power shape at the point of maximum heat flux is not necessarily directly related to the enthalpy rise factors. The results of the loss of coolant accident analyses are conservative with respect to the ECCS acceptance criteria as specified in 10 CFR 50.46 using the upper bound FQ(Z) times the hot channel factor normalized operating envelope given by TS Figure 3.12-8.

When an FQ measurement is taken, measurement error, manufacturing tolerances, and the effects of rod bow must be allowed for. Five percent is the appropriate allowance for measurement error for a full core map (greater than or equal to 38 thimbles, including a minimum of 2 thimbles per core quandrant, monitored) taken with the movable incore detector flux mapping system, three percent is the appropriate allowance for manufacturing tolerances, and five percent is appropriate allowance for rod bow. These uncertainties are statistically combined and result in a net increase of 1.08 that is applied to the measured value of FQ.

N In the specified limit of F~H' there is an eight percent allowance for uncertainties, which means that normal operation of the core is expected to result in F:H ~ 1.55 [l + 0.3 (l-P)]/1.08. The logic behind the larger uncertainty in this case is that (a) normal perturbations in the radial power N

shape (e.g., rod misalignment) affect F~H' in most cases without necessarily affecting FQ, (b) the operator has a direct influence on FQ through movement of rods and can limit it to the desired value; he has no direct control over N

F~H' and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests and which may influence FQ, can

-HOT CHANNEL FACTOR NORMALIZED e

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ATTACHMENT 2 LOCA-ECCS SAFETY EVALUATION FOR SURRY UNITS 1 AND 2

1.0 INTRODUCTION

A reanalysis of the Emergency Core Cooling System (ECCS) performance for the postulated large-break LOCA has been performed in compliance with Appendix K to 10 CFR 50. The results of this reanalysis are presented here, and are in compliance with 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors." This analysis was performed with the NRC-approved 1981 model with BART version of the Westinghouse LOCA-ECCS evaluation model (Ref. 1 and 2). The analysis includes the evaluation model revisions described in Reference 16 and approved by the NRC in Reference 17. The analytical techniques used are in full compliance with 10 CFR 50, Appendix K.

As required by Appendix K of 10 CFR 50, certain conservative assumptions were made for the LOCA-ECCS analysis. The assumptions pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA is assumed to occur, and include such items as the core peaking factors, the containment pressure, and the performance of the Emergency Core Cooling System. All assumptions and initial operating conditions used in this reanalysis were the same as those used in the previous LOCA-ECCS analysis (Ref. 3), with the following exceptions:

1. Data appropriate for the model 51F Replacement Steam generators were used.

2. The steam generator plugging level was increased from 3% to 7%.

3. The value of the limiting enthalpy rise hot channel factor (including uncertainties) was increased to 1.62. Even though a value of 1.62 was used in this analysis, the Technical Specification limit for F~H will remain at 1.55.

4. Fuel performance data using the new PAD thermal model (WCAP 8720 Addendum 2) were used.

5. 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.

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With the above changes incorporated into the analysis, it was found that the maximum assumed heat flux hot-channel factor could be 2.32 and still ensure compliance with the 10 CFR 50.46 acceptance criteria.

In addition to the base analysis, two sensitivity analyses were run to justify increasing the allowable operating band for accumulator water volume as defined in the Technical Specifications. One sensitivity analysis was run at a 3

high accumulator water volume of 1025 ft while the other was run at a low 3

accumulator water volume of 975 ft . In both cases it was found that the assumed heat flux hot-channel factor could be 2.32 and ensure compliance with the 10 CFR 50.46 acceptance criteria.

2.0 ACCIDENT DESCRIPTION A LOCA is the result of a rupture of the reactor coolant system (RCS) piping or of any line connected to the system. The system boundaries considered in the LOCA analysis are defined in the UFSAR, Sensitivity studies (Ref. 7) have indicated that a double-ended cold-leg guillotine (DECLG) pipe break is limiting. 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 signal, assumes the most limiting single-failure event. These countermeasures will limit the consequences of the accident in two ways:

1. Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat. No credit is taken in the analysis for 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.

Before 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 2

blowdown, heat from decay, hot internals, and the vessel continue to be trans-ferred 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, consistent with Appendix K of 10 CFR 50. Therefore, the core heat transfer is based on local conditions, with transition boiling and forced convection to steam as the major heat transfer mechanisms. During the refill period, it is assumed that rod-to-rod radiation is the only core heat transfer mechanism. The heat transfer between the reactor coolant system and the secondary system may be in either direction, depending on the relative temperatures. For the case of continued heat addition to the secondary side, secondary-side pressure increases and the main safety valves may actuate to reduce the pressure.

Makeup to the secondary side is automatically provided by the auxiliary feedwater system. Coincident with the safety injection signal, normal feedwater flow is stopped by closing the main feedwater control valves and tripping the main feedwater pumps. Emergency feedwater flow is initiated by starting the auxiliary feedwater pumps. The secondary-side flow aids in the reduction of RCS pressure. When the reactor coolant system depressurizes to 600 psia, the accumulators begin to inject borated water into the reactor coolant loops. The conservative assumption is then made that injected accumulator water bypasses the core and goes out through the break until the termination of bypass. This conservatism is again consistent with Appendix K of 10 CFR 50. In addition, the reactor coolant pumps are assumed to be tripped at the initiation of the accident, and effects of pump coastdown are included in the blowdown analysis.

The water injected by the accumulators cools the core, and subsequent opera-tion of the low-head safety injection pumps supplies water for long-term cooling. When the refueling water storage tank (RWST) is nearly empty, long-term cooling of the core is accomplished by switching to the recirculation mode of core cooling, in which the spilled borated water is drawn from the containment sump by the low-head safety injection pumps and returned to the reactor vessel.

The containment spray system and the recirculation spray system operate to return the containment environment to subatmospheric pressure.

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3.0 ANALYSIS 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.

The description of the various aspects of the LOCA analysis methodology is 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 that ensure compliance with 10 CFR 50, Appendix K. The SATAN-VI, COCO, WREFLOOD, BART, and LOCTA-IV codes, which are used in the LOCA analysis, are described in detail in WCAP-8306 (Ref. 9), WCAP-8326 (Ref. 10), WCAP-8171 (Ref. 11), WCAP-9695 (Ref. 4), WCAP-10062 (Ref. 5), and WCAP-8305 (Ref. 12),

respectively. These codes assess whether sufficient 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 COCO computer code calculates the containment pressure transient during all three phases of the LOCA analysis. Similarly, the LOCTA-IV computer code is used to compute the thermal transient of the hottest fuel rod during the three phases.

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 COCO code for use in the determination of the containment pressure response during this first phase of the LOCA. Additional SATAN-VI output data from the end of the blowdown, including the core inlet flow rate and enthalpy, the core pressure, and the core power decay transient, are input to the LOCTA-IV code.

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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. WRE-FLOOD also calculates the mass and energy flow rates that are assumed to be vented to the containment. Since the mass flow rate to the containment depends upon the core pressure, which is a function of the containment back-pressure, the WREFLOOD and COCO codes are interactively linked. With the input and boundary conditions from WREFLOOD, the mechanistic core heat trans-fer 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, and conservatively selected initial RCS operating condi-tions. These initial conditions are summarized in Table 1 and Figure 1. The axial power shape of Figure 1 assumed 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 COCO code, which is also used throughout the LOCA analysis, calculates the containment pressure. Input to COCO is obtained from the mass and energy flow rates 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 COCO. These initial containment conditions and assumed modes of operation are provided in Table 2.

4.0 NON-LOCA SAFETY EVALUATION FOR 7% STEAM GENERATOR TUBE PLUGGING This Surry Power Station LOCA-ECCS reanalysis has evaluated plant operation at steam generator tube plugging levels of up to 7% based on the acceptance criteria delineated in 10 CFR 50.46. 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 suffi-cient quantity can potentially affect non-LOCA safety analysis due to reduced 5

primary system flow, more severe pump coastdown characteristics, and the reduction of the reactor primary coolant system volume. Primary flow rate becomes a key parameter in DNB limited events (e.g., Uncontrolled RCCA Bank Withdrawal at Power) when it falls below the thermal design flow rate. 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). (l 4 )

Flow measurements have been taken at the Surry Power Station for several levels of steam generator tube plugging prior to the steam generator replacements. The results of these measurements and supporting analyses are documented in Reference 14. The curve of flow versus plugging level presented in Reference 14 indicates that the conservatively estimated flow rate at the proposed 7% plugging level is still considerably larger than the Surry thermal design flow. Using the conservative Reference 14 assumptions for a 7%

plugging level, it is concluded that the replacement steam generators have greater than 2% margin to the thermal design flow. Therefore, the current docketed licensing analyses remain valid for those events in which flow rate is an important concern.

The impact of 7% tube plugging on dilution times in the uncontrolled boron dilution events is bounded by the analyses documented in References 14 and 15.

These analyses documented the impact of 40% steam generator tube plugging on non-LOCA safety analyses. Relative to the boron dilution events, the analyses indicated:

~For uncontrolled dilution during startup, time to criticality remains at least 82 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 while under automatic control, rod insertion limit alarms (low and low-low settings) provide the operator with adequate time (24 minutes) to determine cause, isolate the primary grade water source, and initiate reboration before total shutdown margin is lost due to dilution.

6 L

~For uncontrolled dilution at power while under e

manual control, the operator has ample time (22 minutes) after the over-temperature ~T alarm or trip to find and 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 7 percent, the currently licensed analyses remain valid and no reanalysis of the non-LOCA safety events is necessary.

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 CD= 0.4 and 0.6 discharge coefficients.

The 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 1942.5°F for the CD= 0.4 case, a maximum local cladding oxidation level of 4.44%, and a total core metal-water reaction of less than 0.3%. The corresponding results of the two accumulator volume sensitivity analyses, which were run using the CD= 0.4 discharge coefficient, are also presented in Tables 3 and 4. The low accumulator water volume analysis resulted in a peak clad temperature of 1968.1°F and a maximum local cladding oxidation level of 4.22%, while the high water volume analysis resulted in a peak clad temperature of 1932.0°F and a maximum local cladding oxidation of 3.35%. The detailed results of the base LOCA reanalysis 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.

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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 Flow Rate - Figures Sa and Sb show the calculated flow rate out of the break for the discharge coefficient used. The flow rate 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 7b 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 1942.5°F at the 8.00 ft elevation in the core.

8. Fluid Temperature - Figures Sa and Sb show the calculated fluid temperature for the hot spot and burst locations for the discharge coefficient used.

9. Core Flow - Figures 9a and 9b show the calculated core flow, both top and bottom, for the discharge coefficient used.

8

10. Reflood Transient Figures 10a and 10b show the reactor pressure vessel downcomer and core water levels for the discharge coefficient used. Figures Ila and llb 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 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 show 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 15b 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 17b 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.

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6.0 CONCLUSION

S 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.

3 . 3

6. Accumulator water volumes in the range of 975 ft to 1025 ft per accumulator have no significant impact on the results of the analysis and therefore, the emergency core cooling system continues to meet the 10 CFR 50.46 acceptance criteria for water volumes in this range.

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10 CFR 50.59 SAFETY EVALUATION The proposed changes have been reviewed against the criteria of 10 GFR 50.59 and does 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 LOGA analysis verifies that operation under the revised specifications would also not result in any increase in accident consequences.

2. No new accident types or equipment malfunction scenarios will be introduced as a result of operating in accordance with the revised specifications. The changes which potentially affect physical components in the plant systems (accumulator level and steam generator tube plugging) were 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 affected Technical Specifications, is not reduced. The revised EGGS analysis meets the requirements of 10 GFR 50.46. Additionally, since evaluation of non-LOGA accidents concluded that they bound expected results when considering the proposed changes, the existing analyses remain applicable. The current margin of safety is therefore maintained for LOGA and non-LOGA accidents.

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10 CFR 50.92 NO SIGNIFICANT HAZARDS DETERMINATION The proposed changes do not involve a significant hazards consideration because operation of Surry Units 1 and 2 in accordance with these changes would not:

(1) involve a significant increase in the probability or consequence of an accident previously evaluated. The revised analysis which supports these changes demonstrated that the ECCS acceptance criteria of 10 CFR 50.46 were met. The proposed changes have also been shown to have no impact upon the non-LOCA transients. Therefore, the probability or consequences of an accident previously evaluated will not increase.

(2) create the possibility of a new or different kind of accident from any accident previously identified. The proposed changes involve changes in assumptions made for previously evaluated LOCA accidents.

The revised analysis included these parameter changes and demonstrated that they would not cause a new accident. In addition, the increase in steam generator tube plugging was evaluated for impact upon RCS flow and RCS coolant volume. It has been demonstrated that the non-LOCA accidents for which these parameters are significant remain bounded by the existing analyses. Thus, the proposed changes will not create the possibility of a new or different kind of accident.

(3) involve a significant reduction in a margin of safety. The revised ECCS analysis meets the requirements of 10 CFR 50.46. Additionally, since the non-LOCA accidents are unaffected by the proposed changes, the existing analyses remain applicable. The current margin of safety as established by meeting regulatory requirements (e.g., 10 CFR 50.46) is therefore maintained for LOCA and non-LOCA accidents.

Furthermore, the calculated LOCA peak clad temperature is reduced for the analysis using the new assumptions.

Therefore, pursuant to 10 CFR 50.92 based on the above consideration it has been determined that this change does not involve a significant safety hazards consideration.

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8.0 REFERENCES

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. 0. 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.

3. Letter from C. M. Stallings, Virginia Electric and Power Company, to H. R.

Denton, NRC, dated May 31, 1979 (Serial No. 388).

4. Young, M. Y. et al., BART-Al: A Computer Code for the Best Estimate Analysis of Reflood Transients, WCAP-9695, January 1980.

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

6. Letter from C. 0. 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 (COCO), WCAP-8326, June 1974.

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

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

13. Letter from C. M. Stallings, Virginia Electric and Power Company, to E.G.

Case, NRC, Serial No. 092, dated February 17, 1978.

14. Letter from C. M. Stallings, Virginia Electric and Power Company, to E.G.

Case, NRC, Serial No. 344, August 9, 1977.

15. Updated Final Safety Analysis Report - Surry Power Station Units 1 and 2, Virginia Electric and Power Company, Rev. 3, June 1985.

16. M. Y. Young, "Addendum to BART-Al: 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 l," August 25, 1986.

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TABLE 1 INITIAL CORE CONDITIONS ASSUMED FOR THE DOUBLE-ENDED COLD-LEG GUILLOTINE BREAK (DECLG)

Calculational Input Core Power (MWt) 102% of 2441 Peak linear power (kW/ft) 102% of 14.39 Heat flux hot-channel factor (F) 2.32 Q N Enthalpy rise hot-channel factor (F ~H) 1.62 3

Accumulator water volume (ft, each) 1000 Reactor vessel upper head temperature equal to Thot Limiting Fuel Region and Cycle Cycle Region Unit 1 All All regions Unit 2 All All regions

e TABLE 2 CONTAINMENT DATA (DRY CONTAINMENT)

Net Free Volume 6 3

1. 863 X 10 £t Initial Conditions Pressure (total), psia 9.35 Temperature, °F 80 RWST temperature, °F 40 Service water temperature, °F 32.5a Outside temperature, °F 9 Spray System I - Containment Spray System Number of pumps operating 2 Runout flow rate, gpm 3200 Actuation time, sec 59 Spray System II - Inside Recirculation Spray Subsystem Number of pumps operating 2 Runout flow rate (each), gpm 3500 Actuation time, sec 190 6

Heat exchanger (UA (per pump)), Btu/hr-°F 5.18 X 10 Service water flow (per exchanger), gpm 6900 Spray System II - Outside Recirculation Spray Subsystem Number of pumps operating 2 Runout flow rate (each), gpm 3500 Actuation time, sec 365 Heat exchanger (UA (per pump)), Btu/hr-°F 6 5.18 X 10 Service water flow (per exchanger), gpm 6900 aSensitivity analyses provided in Reference 14 demonstrate that service water temperature levels as low as 25°F will have a negligible impact on the limiting results of the LOCA-ECCS analyses.

e TABLE 2 (Continued)

CONTAINMENT DATA (DRY CONTAINMENT)

Structural Heat Sinks Type/thickness (in.) 2 Area (ft ), with uncertainty Concrete/6 8,393 Concrete/12 62,271 Concrete/18 55,365 Concrete/24 11,591 Concrete/27 9,404 Concrete/36 3,636 Carbon steel/0.375 Concrete/54 46,489b Carbon steel/0.50 Concrete/30 25,652b Concrete/26 (floor) 12,110 Carbon steel/0.239 158,059b Stainless Steel/0.306 17,519 Aluminum/0.0091 3,911 bCredit for painted surfaces was taken only for the nominal surface area.

TABLE 3 TIME SEQUENCE OF EVENTS FOR DECLG CD=0.4 CD=0.4 CD= 0.4 CD= 0.6 Accum=975 Accum=l025 (sec) (sec) (sec) (sec)

Start 0.0 0.0 0.0 0.0 Reactor trip 0.426 0.420 0.426 0.426 Safety injection signal 2.52 2.02 2.52 2.52 Accumulator injection 15.70 11.80 15.70 15.70 Pump injection 27.52 27.02 27.52 27.52 End of bypass 30.95 25.40 30.71 31.08 End of blowdown 30.95 25.40 30.71 31.08 Bottom of core recovery 43.42 37.80 42.93 43.81 Accumulator empty 53.78 48.20 52.01 55.66

TABLE 4 RESULTS FOR DECLG CD= 0.4 CD= 0.4 Accum=975 Accum=1025 Peak clad temperature, OF 1942.5 1880.6 1968 .1 1932.0 Peak clad location, ft 8.00 8.00 8.00 8.00 Local Zr/H 0 reaction 2

(max), % 4.44 2.08 4.22 3.35 Local Zr/Hz° location, ft 6.25 6.00 6.00 6.00 Total Zr/H 0 reaction, %

2 < 0.3 < 0.3 < 0.3 < 0.3 Hot-rod burst time, sec 41.00 47.40 40.90 40.70 Hot-rod burst location, ft 6.25 6.00 6.00 6.00

e e TABLE 5 REFLOOD MASS AND ENERGY RELEASES DECLG (~ = 0.4)

Total Mass Total Energy Time (sec) 5 Flow Rate (lb/sec) Flow Rate (10 Btu/sec) 43.421 o.o 0.0 44.221 6.97 0.090 53.948 76.53 0.957 67.323 82.25 1.025 84.023 86.54 1.076 101. 823 240.72 1.472 120.473 293.08 1.553 140.173 298.65 1.508 254.823 350.92 1.384 TABLE 6 BROKEN LOOP ACCUMULATOR FLOW TO CONTAINMENT DECLG (CD= 0.4) a Time (sec) Mass Flow Rate (lbm/sec) 0.00 4214.45 1.01 3809.67 3.01 3267.18 5.01 2906.93 7.01 2642.70 10.01 2346.39 15.01 2003.57 20.01 1765.02 25.01 1590. 72 27.01 1535.55 a

For energy flow rate, multiply mass flow rate by a constant of 59.62 Btu/lbm.

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r-*---t----1 0.1 -+---+----+----1--- -r------  ! I ---+l----,1~---+-,- >---~-------

0 -+---+-----1----+---J,--+---1--+--~~---'-,- ~ , ~ , - + - - t - - - - t - - - - 1 1 - - - - - - 1 0 40 80 120 160 200 240 280 TIME (SEC)

FIGURE l7B CONTAINMENT WALL HEAT TRANSFER COEFFICIENT DECLG (CD= 0.6)

LOCTA-VPA 0.4 OECLG-7PCT SGTP-81 MOO W/8ART-574.4TAVG JW20I=14 NEW ?AO-FQ=2.32-15X15 CHAMFER L/0=1.2-THIMBLE FIX NTRODS=NROOS QUALITY OF FLUID BURST, 6.25 FT( ) PEAK, 8.00 FT(~)

Symbol (*) is offset a constant distance above plotted values.

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it: , ~

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1'JI 115 T lt1( I S(C l 08/18/86 FI.CURE 18A FLUID QUALITY VERSUS TIME DECLG (CD = 0.4)

LOCTA-VPA 0.6 DECLG-7PCT SGTP-81 MOD W/BART-574.4TAVG NEW PAD-FQ=2.32-15X15 CHAMFER L/0=1.2-THIMBLE FIX 2 NTRODA=NRODA QUALITY OF FLUID BURST, 6.00 FT( ) PEAK, 8.00 FT(*)

Symbol (*) is offset a constant distance above plotted values.

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u,a 111,111 u,5 Tlt1C: ISC:CI 08/20/86 FIGURE 18B FLUID QUALITY VERSUS TJME DECLG (CD= 0.6)

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ATTACHMENT 3 VOUCHER CHECK SURRY UNITS 1 AND 2