ML19260C701

LOCA-ECCS Safety Evaluation for Unit 1.

ML19260C701
Person / Time
Site:	North Anna
Issue date:	01/02/1980
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML19260C700	List: ML19260C699 ML19260C701
References
REF-PT21-79-182-001 NUDOCS 8001080473
Download: ML19260C701 (33)

Text

. _ _ _ . _ _ _ _ . . _ . . _ _ . _ _ _ .

Mr. Harold R. Denton 2 Should you havs questions, please contact our Mr. M. L. Sowling at (804) 771-3183 at your earliest convenience.

Very truly yours, 9

C. M. S tallings Vice President-?ower Supply and Production Cperations Attachment LOCA-ECCS Safety Evaluation for North Anna Unit 1 cc: Mr. Jaxtes P. O'Reilly, Director Office of Inspection and Enforcement, Ec3 on i II Mr. O. D. Parr, Chief Light Water Reactors Sranch Mo. 3 Division of Project Managetaent Mr. Jack Rosenthall, Reactor Safety Branch Division of Operating Reactors b b 1703 140.

8001080

• 493 .

• • O e.U/.GNE;T 1 1703 141

PAGE 1

1.0 INTRODUCTION

A reanalysis of the ECCS performance for the postulated large break Loss of Coolant Accident (LOCA)* has been pe: formed which is in compliance with Appendix K to 10 CTR 50. The results of this reanalysis are presented herein and are in compliance with 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors. This analysis was pe:fo:med with the NRC approved (Ref. 2) February 1978 version of the Westinghouse LOCA-ECCS evaluation model. The analytical techniques used are in full compliance with 10 CFR 50, Appendix K.

As required by Appendix X of 10 CFR 50, ce:tain conservative assumptions were made for the LOCA-ECCS analysis. The assumptions pertain to the conditions of the :eactor 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 (ECCS), All assumptions and initial ope:ating conditions used in this reanalysis we:e the same as those used in the previous LOCA-ECCS analysis (Ref. 3) with the following exceptions: 1) the limiting value of the heat flux hot channel factor was decreased from 2.21 to 2.10; 23 nominal design power rating was used versus engineered safeguards rating; 3) RCS cold leg temperature of 550*r based on

• The reanalysis of the small break LOCA is not necessary and therefore the analysis of this accident submitted by Reference 1 :emains appli-

~

cable.

1703 142 I

PAGE 2 operational data was used versus 555'F; 4

') more accurate data for several containment parameters were used; 51 5% of the steam generator tubes were assumed to be plugged; 6) 17x17 generic fuel parameters were used instead of plant specific fuel parameters.

2.0 DESCRIPTION

OF POSTULATED MAJOR REACTOR COOLAMT PIPE RUPTURE (LOSS OF COOLANT ACCIDENT - LOCA)

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 FSAR. Sensitivity studies (Reference 4) have indicated that a double-end leg guillotine (DECLG) pipe break is limiting. Should a DECLG break occur, rapid depressuri=ation of the RCS occurs. The reactor trip signal subsequently occurs when the pressurirer 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 ECCS, 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.

(It should be noted, however, that 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 temperatures.

Before the break occurs, the unit is in an equilibrium condition, i.e., the 1703 143 1

e

PAGE 3 heat generated in the core is being removed via the secondary system.

During blowdown, heat from decay, hot internals and the vessel continues to be transfer:ed to the reactor coolant system. At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the co:e by forced convection with some fully developed nucleate boiling. After the break develops, the time to departure from nucleate boiling is calculated, consistent with Appendix K of 10 CFR 50. Thereafter, the core heat transfer is based on local conditions with transition belling 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  : educe 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 feedwate: control valves and tripping the main feedwater pumps. Emergency feedwater flow is initiated by starting the auxiliary feeduater pumps. The seconda:y side flou aids in the reduction of reactor coolant system pressure. When the reactor coolant system depressuri=es 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 termination of bypass. This conservaF,sm is again break until the consistent with Appendix K of 10 CFR 50. In addition, the reacto coolant

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1703 144 1

PAGE 4 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 operation of the low head safety injection pumps supplies water for long term cooling. When the 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 retu:ned to the reactor vessel.

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

The large break LOCA transient is divided, rer analytical purposes, into th:ee phases: blowdown, refill, and reflood. There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the RCS, the pressure and temperature transient uithin the containment, and the fuel clad temperatu:e transient of the hottost fuel rod in the core. Based on these censiderations, a system of inter-related computer codes has been developed for the analysis of the LOCA.

The description of the various aspects of the LOCA analysis methodology is given in WCAp-8339(Ref. 5). This document describes the major phenomena modeled, the inte: faces among the computer codes, and the features of the codes which ensure compliance with 10 CFR 50, Appendix K. The SATAM-VI, WREyLOOD, COCO, and LOCTA-IV codes, which are used in the LOCA analysis, 1703 145 I

s PAGE , 5 are described in detail in WCAp-8306(Ref. 6), WCAp-8326(Ref. 7),

WCAp-8171(Ref. 8), and WCAp-8305(Ref. 9), respectively. These codes are able to assess whether sufficient heat transfer geometry and core amenablity to cooling are preserved during the time spans applicable to the blowdown, refill, and reflood phases of the LOCA. The SATAN-VI computer code analy es the thermal-hydraulic transient in the RCS during blowdown and the C0C0 computer code is used to calculate 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.

SAT;N-VI is used to determine the RCS pressure, enthalpy, and density, as well as the mass and energy flow rates in the RCS and steam generato:

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 :elease :ates 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 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.

With input from the SATAN-VI code, WRETLOOD uses a system thermal-hydraulic model to determine the core flooding rate (i.e., the rate at i703 146 i

PAGE 6 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. WRETLOOD 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 backpressure, the WRITLOOD and C0C0 codes are interacti.vely linked. WRITLOOD is also linked to the LOCTA-IV code in that thermal-hydraulic parameters from WRETLOOD are used by LOCTA-IV in its calculation of the fuel temperature.

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 output from SATAN-VI and WREFLOOD and conservatively selected, initial RCS operating conditions.

These initial conditions are summari=ed in Table 1 and Figure 1, (The axial power shape of Tigure 1 assumed for LOCTA-IV is a cosine curve which has been previously verified (Red. 10) 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 C0CO is obtained from the mass and energy flou 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 Tabl,n *

,7~03 147

, , o PAGE 7 3.0 DISCUSSION OF SIGNIFICANT INPUT Significant differences in input between this analysis and the currently applicable analysis are delineated in Section 1.0 and discussed in more detail below. The changes made in the analysis reflect the operational conditions and limits necessary to allow full power operation at steam generator tube plugging levels of up to 5%.

The most notable input change for this analysis is the increase in assumed steam generator tube plugging. The currently applicable analysis made ac allowance for tube plugging. The plugging level assumed for this analyca:

is 5%.

The assumption of a small amount of steam generator tube plugging in the analysis also affects the assumed core inlet temperature by requiring a decrease in this parameter. Consequently, a core inlet temperature of 550*F was assumed. This value is the best estimate core inlet temperature as determined from operational data and is adequate to encompass the applicable steam generator tube plugging range.

In order to ensure compliance with the 10 CFR 50.46 acceptanca criteria, a change to the assumed value of the heat flux hot channel factor was made, Specifically, the assumed heat flux hot channel factor was decreased from 2.21 to 2.10.

Previous analyses had assumed an Engineered Safeguards power level in the blowdown portion of the analysis for future operational flexibility. The 1703 148 i

i PAGE 8 current analysis assumed a nominal design power level as the basic power input throughout the analysis.

Several changes were made to the containment parameters. The amounts of the various structural heat sinks provided in Table 2 were reviewed in detail.

Based on the as-built plant containment, the heat sinks were conservatively revised to include the additiion of a 3% uncertainty to all surface areas.

As allowed by the NRC, credit has been taken for paint on carbon steel surfaces.

The calculation was performed assuming conservative generic 17x17 fuel parameters. The previous analysis had assumed cycle specific 17x17 parameters.

Finally, the analysis was conducted with the February, 1978 version of the

'4e s tingh ous e LCCA-ECCS Evsluation Model (Refs. 11,12,13).

4.0 RESULTS Tables 1 and 2 and Tigure 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 (DECLG) for the CD=0.4 discharge coefficient. The DECLG has been determined to be the limiting break si=a and location based on the sensitivity studies reported in Reference 4 Further, all previous LOCA-ECCS submittals for the North Anna units have resulted in the CD=0.4 discharge coefficient being the limiting break sire. The applicability of this conclusion (i.e. CD=0.4 is the limiting break sire) for this analysis 1703 149

e . .

PAGE 9 was explicitly verified. Consequently, only the results of the most limiting break site are presented in the figures and remaining tables in this submittal. The current analysis resulted in a limiting peak clad temperature of 2088*T, a maximum local cladding oxidation level of 5.81%,

and a total core metal-water reaction of less than 0.3%. The detailed results of the LOCA reanalysis are provided in Tables 3 through 6 and Figures 2 through 18.

5.0 CONCLUSI0HS Tor breaks up to and including the double-ended severance of a reacter 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 CTR 50.46. That is:

1, The calculated peak fuel red clad temperature is below the requirement of 2200*T.

2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent 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.

1703 150 I

J PAGE 10

6.0 REFERENCES

1. Final Safety Analysis Report, North Anna Power Station, Units 1 and 2, Virginia Electric and Pouer Company.

2. Letter from J.F. Stol=(HRC) to T.M. Anderson (Westinghouse), dated

~

August 29, 1978.

3. Letter from C.M. Stallings(Vepco) to E.G. Case (NRC), Serial No.

258, May 5, 1978.

4 Buterbaugh. T.L., Johnson, W.J., and Kopelic, S.D., "Westinghause ECCS Plant Sensitivity Studies," WCAP-8356, July 1974

5. Bordelen, F.M., et. al., " Westinghouse ECCS Evaluation Model-Summary," WCAP-8339, July, 1974.

6. 3erdelon, P.M., et. al., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant," WCAP-8306, June 1974

7. 3ordelon, F.M., and Murphy, E.T., " Containment Pressure Analysis Code (C0CO)," WCAP-8326, June 1974.

8. Kelly, R.D., et. al., " Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLCOD Code)," WCAP-8171, June 1974.

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

10. Letter from C.M. Stallings(Vepco) to E.G. Case (NRC), Serial No. 092, February 17, 1978.

11. " Westinghouse ECCS Evaluation Model- February 1978 Version,"

WCAP-9220.

12. Letter from T.M. Ande r s o n ( *e e s tingh ous e ) to J.F. Stol (NRC),

Serial No. MS-TMA-1981, .Mvember 1, 1978.

13. Letter from T.M. Anderson (Westinghouse) to R. Tedesco(NRC), Serial No. NS-TMA-2014, December 11, 1978.

1703 151 1

TABLE 1 I:!ITI AL CORE CO';DITIC:IS ASSUMED FOR IllE DOUBLE-ENDED COLD LEC CUILLOTINE JREA (DECLC)

Calculational Inout Core Power (MWt , 102': of) 2775 Peak Linear Power (kw/ft, 102*: of) 11.43 Heat Flux Hot Channel Factor (Fq ) 2.10 N

Enthalpy Rise Hot Channel Factor (FaH) 1.55 Accumulator Water volume (f t , each) 1025 Reacter Vessel Upper Head Temperature Equal to T hot Limiting Fuel 'tecien and Cvela Cvele Region Unit 1 ALL ALL Regions Unit 2 ALL ALL Regicas 1703 152

TABLE 2 CONTAINMENT DATA NET FREE VOLUME 1.916 x 10 ft INITIAL FONDITIONF Pressure 9.6 psia Temperature 90 ?

RWST Temperature 35 F Outside Temperature -10 F SPRXI SYSTEM Number of Pumps Operating 2 Runcut Flow Rate (per pump) 2000 gpm 63 Time in which spray is effective 59 secs STRUCTURAL HEAT SINF.S 'l)

Thickness (In) Area (Ft ), w/ uncertainty 6 Concrete 8,393 12 Concrete 62,271 18 Concrete 55,365 24 Concrete 11,591 27 Concrete 9,404 36 Concr'ete 3,636

.375 Steel, 54 Concrete 22,039

.375 Steel, 54 Concrete 28,933

.500 Steel, 30 Concrete 25,673 26.4 Concre te, .25 Steel, 120 Concrete 12,110

.407 S tainicss Steel 10,527

.458 Steel 160,32S

.SS2 Steci 9,S94

.059 Steel 60,875 (l) See the response to Comment S6.106 of the FSAR for a detailed breakdctin of the containment heat sinks and for justification of the other input parameters used to calculate containu nt pressure.

1703 153

TABLE 3 TIME SEQUENCE OF EVENTS DECLC

.' CD=0.4 (Sec)

START -

0.0 Reactor Trip 0.72 S. I. Signal 2.42 Acc. Injectica 16.3 End of Bypass 26.33 Pump Inje::ica 27.42 End of Blewdown 29.52 Bo . tom of Core Recovery 39.74 Acc. Empty 52.3

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TABLE 4 RESULTS FOR OECLG C3=0.4 Peak Clad Temp, OF 2088 Peak Clad Lccation Ft. 7.5 Local Zr/H 2O RXN (max), % 5.81 Local "r/H 2O Location, Ft. 7.25 Total Zr/H 2O PJCI, % <0.3 Hot Fed Burst Time, sec. 34.8 Mot Red Burst Location, Et. 6.0 r-1703 153

TABLE 5 REFLOOD MASS M1D ENERGY RELEASES DECLC (CD

= 0.4)

TOTAL MASS TOTAL EliERCY 5

TIME (SEC) FLCWRATE (L3/SEC) FLOWRATE (10 BTU /SEC) 39.7 0.0 0.0 40.8 0.743 0.0095 46.0 35.45 0.4623 55.0 210.96 1.395 68.4 245.57 1.412 85.0 258.49 1. 380 103.7 266.58 1.338 124.2 273.18 1.290 170.4 284.97 1.194 225.6 297.34 1.091 299.5 317.77 1.003 457.3 347.37 0.8037 1703 156

C TASLE 6 ER0 KEN LOOP ACCUMULATOR FLOW TO CCt" "ulNT DECLC, C '

• D TIME (SEC) MASS FLOWRATE* (LSM/SEC) 0.0 4010 1.0 3622 3.0 3104 5.0 2761 7.0 25C6 10.0 2225 15.0 1894 20.0 1673 25.0 1523 30.0 1420 eFor energy flowrate cultiply mass flowrate by a cons:an: of 59.60 3TU/LBM

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