ML19260C533

Forwards Documentation of LOCA-ECCS Analysis Conducted W/Nrc Approved Feb 1978 Version of Westinghouse ECCS Evaluation Model.No Unreviewed Safety Questions Involved

ML19260C533
Person / Time
Site:	North Anna
Issue date:	01/02/1980
From:	Stallings C VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, Schwencer A Office of Nuclear Reactor Regulation
References
986C, NUDOCS 8001070366
Download: ML19260C533 (37)

Text

{{#Wiki_filter:* VIRGINIA ELecTaic un Powan COMPANY Rzcamown.V:morazA 20261 January 2, 1980 Mr. Harold R. Denton, Director Serial No.: 986C Office of Nuclear Reactor Regulation FR/MLB: mvc Attn: Mr. A. Schwencer, Chief Operating Reactors Branch No. 1 Docket No.: 50-338 Division of Operating Reactors 50-339 U. S. Nuclear Regulatory Commission Washington, DC 20555 License No.: NPF-4

Dear Mr. Denton:

SUPPLEMENTAL INFORMATION TO AMENDMENT TO OPERATING LICENSE NPF-4 NORTH ANNA POWER STATION UNITS NO.1 AND 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGE NO. 27 Attached is the complete documentation of a LOCA-ECCS analysis conducted with the NRC approved February,1978 version of the Westinghouse ECCS Evaluation Model for North Anna Units Nos.1 and 2. The analysis has been conducted in compliance with Appendix K to 10CFR50 and meets the criteria delineated in 10CFR50.46. The significant results from this analysis were documented in my letter of December 19, 1979 (Serial No. 986B) and discussed by our Mr. M. L. Bowling with your Mr. J. E. Rosenthall on December 20, 1979. The analysis can be used as the basis for Technical Specification Change No. 27 which was submitted by my letter of November 29,1979 (Serial No. 986). The aaalysis provided in the Attachment has been reviewed and approved by both the Station Nuclear Safety and Operating Committee and the System Nuclear Safety and Operating Committee. It has been determined that no unreviewed safety questions (as defined in 10CFR50.59) are involved. 1695 252 8001070 % C,

e vinotwas ELECTRIC Axo Powra Couraxv to Mr. Harold R. Denton 2 Should you have questions, please contact our Mr. M. L. Bowling at (804) 771-3183 at your earliest convenience. Very truly yours, 5 ,/ G ? :. u. 9 ' C. M. Stallings Vice President-Power Supply and Production Operations Attachment LOCA-ECCS Safety Evaluation for North Anna Unit 1 cc: Mr. James P. O'Reilly, Director Office of Inspection and Enforcement, Region II Mr. O. D. Parr, Chief Light Water Reactors Branch No. 3 Division of Project Management Mr. Jack Rosenthall, Reactor Safety Branch Division of Operating Reactors 1695 253

ATTI.CENE:;T 1, 1695 254

PAGE 1

1.0 INTRODUCTION

A reanalysis of the ECCS performance for the postulated large break Loss of Coolant Accident (LOCA)* has been performed which is in compliance with Appendix K to 10 CFR 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 performed with the HRC approved (Ref.

2) February 1978 version of the Westinghouse LOCA-ECCS evaluation model.

The analytical techniqces 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 (ECCS).

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) the limiting value or 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'T based on The reanalysis of the small break LOCA is not necessary and therefore the analysis of this accident submitted by Reference 1 remains appli-I cable. 1695 25r3 i

PAGE 2 operational data was used versus 555'F; 4) more accurate data for several containment parameters were used; 5) 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 COOLANT 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 depressuriration 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 wayst 1. Reactor trip and borated water injection complement void formation in causing rapid reduction of p9wer 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 1695 256 i a

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 transferred to the reactor coolant system. At the beginning of the blowdown

phase, the entire RCS contains subcooled liquid which transfers heat from the core 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 uith 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 temperntures. 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 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 break until the termination of bypass. This conservatism is again consistent with Appendix K of 10 CFR 50. In addition, the reactor coolant I695 257

3 PAGE 4 pumps are assumed to be tripped at the initiation of tne 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 containr ent 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 a subatmospheric pressure. 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

RCS, 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 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 interfaces among the computer codes, and the features of the codes which ensure compliance with 10 CFR 50, Appendix K.

The SATAN-VI,

WREFLOOD, COCO, and LOCTA-IV codes, which are used in the LOCA analysis, 1695 258

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 SATAH-VI computer code analy=es the thermal-hydraulic transient in the RCS during blowdown and the COCO computer code is used to eniculate 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 RCS 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 tha+ 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 SATAH-VI output data from the end of

blowdoun, 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, WREFLOOD uses a

system thermal-hydraulic model to determine the core flooding rate (i.e., the rate at 1695 259

b 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. WREFLOOD 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 WREFLOOD and COCO codes are interactively linked. WREFLOOD is also linked to the LOCTA-IV code in that thermal-hydraulic parameters from WREFLOOD 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 SATAM-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 Figure 1 assumed for LOCTA-IV is a cosine curve which has been previously verified (Ref.

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 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. 1695 260

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 no allowance for tube plugging. The plugging level assumed for this analysis is 5%. The assumption of a small amount of steam generator tube plugging .n the analysis also affects the assumed core inlet temperature by requiring a decrease in this parameter. Consequently, a core inlet temperature of 550*r 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 acceptance 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 1695 201

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 37. 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 Westinghouse LOCA-ECCS Evaluation Model (Refs.

11,12,13). 4.0 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 (DECLG) for the CD=0.4 discharge coefficient. The DECLG has been determined to be the limiting break si=e 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 si=e. The applicability of this conclusion (i.e. CD=0.4 is the limiting break si=e) for this analysis 1695 262

a PAGE 9 was explicitly verified. Consequently, only the results of the most limiting break si=e are presented in the figures and remaining tables in this submittal. The current analysis resulted in a liuiting peak clad temperature of 2088'F, 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 CONCLUSION

S For breaks up to and including the double-ended severance 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. That is: 1. The calculated peak fuel rod clad temperature is belou the requirement of 2200*F. 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 locali=ed cladding oxidation limits of 17% are not exceeded during or after quenching. 4. The core remains amenable to cooling during and aftcr the break. ~ 5. The core temperature is reduced and the long-term decay heat is removed for an extended period of time. ~ 1695 263 I

i PAGE 10

6.0 REFERENCES

1. Final Sa#-ty Analysis Report, North Anna Power Station, Units 1 and 2, Virginia Electric and Power Company. 2. Letter from J.F. Stol=(NRC) to T.M. Anderson (Westinghouse), dated August 29, 1978. 3. Letter from C.M. Stallings(Vepco) to E.G. Case (MRC), Serial No. 258, May 5, 1978. 4. Buterbaugh, T.L.,

Johnson, W.J.,

and Kopelic, S.D., " Westinghouse ECCS Plant Sensitivity Studies," WCAP-8356, July 1974. 5.

Bordelon, F.M.,

et. al., " Westinghouse ECCS Evaluation Model-Summary," WCAP-8339, July, 1974. 6. .Bordelon, F.M., et. al., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant," WCAP-8306, June 1974. 7.

Bordelon, F.M.,

and Murphy, E.T., " Containment Pressure Analysis Code (COCO)," WCAP-8326, June 1974. 8.

Kelly, R.D.,

et. al., " Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD 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.

Anderson (Westinghouse) to J.F. Stol=(NRC), Serial No. NS-TMA-1981, November 1, 1978.

13. Letter from T.M.

Anderson (Westinghouse) to R. Tedesco(NRC), Serial No. NS-TMA-2014, December 11, 1978. 1695 264 1

TAELE 1 I!!ITIAL CORE CONDITIC!IS ASSUMED FOR T1IE DOUELE-EllDED COLD LEC Ci ILLOTINE UREAK (DEC'.C) J Calculational Input Cora Power (FMt 102% of) 2775 Peak Linear Power (kw/ft, 102% of) 11.43 Heat Flux Hot Channel Factor (F ) q 2.10 Enthalpy Rise Hot Channel Factor (F{1 ) g 1.55 Accuculator Water Volume (ft each) 1025 Reactor Vessel l'pper Head Temperature Equal hot Limitins; Fuel Recion and Cycle Cycle Region Unit 1 ALL ALL Regions Unit 2 ALL ALL Regions 1695 265

TABLE 2 CONTAINMENT DATA NET FREE VOLUME 1.916 x 10 ft INITIAL CONDITIONS Pressure 9.6 psia Temperature 90 F L'ST Temperature 35 F Outside Temperature -10 F SPRAY SYSTEM Number of Pumps Operating 2 Runcat Flow Rate (per pump) 2000 gpm 63 Time in which spray is ef fective 59 secs STRUCTURAL HEAT SINKS Thickness (In) Area (Ft ), w/ uncertainty 6 Concrete 8,393 12 Concrete 62,271 18 Concrete 55,365 24 concretc 11,591 27 Concrete 9,404 36 Concrete 3,636 .375 Steel, 54 Concrete 22,039 .375 Steel, 54 Concrete 28,933 .500 Steel, 30 Concrete 25,673 26.4 Concrete,.25 Steel, 120 concrete 12,110 407 S tainicss Steel 10,527 .458 Steel 160,325 .8S2 Steci 9,894 .059 Steel 60,875 (1) See the res ponse to Comment S6.106 of the FSAR for a detailed breakdoun of the containment heat sinks and for jus tification of the other input parameters used to calculate containment pressure. 1695 266

TABLE 3 TIME SEQUENCE OF EVENTS DECLO CD=0.4 (Sec) START 0.0 Reactor Trip 0.72 S. 1. Signal 2.42 Acc. Injection 16.3 End of Bypass 26.33 Pump Injection 27.42 End of Blowdown 29.52 Bottom of Core Recovery 39.74 Acc. Empty 52.3 1695 267

TABLE 4 RESULTS FOR DECLG C =0.4 D Peak Clad Temp, OF 2088 Peak Clad Location Ft. 7.5 Local Zr/H O RXN (max), % 5.81 2 Local Zr/H O Location, Ft. 7.25 2 Total Zr/H O RXN, % <0.3 2 Hot Rod Burst Time, sec. 34.8 Hot Rod Burst Location, Ft. 6.0 1695 268

TABLE 5 REFLOOD MASS AND ENERGY RELEASES DECLC (C ~ D TOTAL MASS TOTAL ENERGY TIME (SEC) FLOWRATE (LB/SEC) FLOWRATE (10 BTU /SEC) 39.7 0.0 0.0 40.8 0.748 0.0098 46.0 35.45 0.4623 55.0 210.96 1.395 68.4 245.57 1.412 85.0 258.49 1.380 103.7 266.38 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 p

TABLE 6 EROKEli LOOP ACCUMULATOR FLOW TO C0!! TAI!! MINT DECLC, C D TIME (SEC) MASS FLOWRATE* (LBM/SEC) 0.0 4010 1.0 3622 3.0 3104 5.0 2761 7.0 2506 10.0 2225 15.0 1894 20.0 1673 25.0 1523 30.0 1420

• For energy flowrate multiply mass flowrate by a constant of 59.60 BTU / LUM.

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