ML19318A345

Large Break Loca/Eccs Performance Results.

ML19318A345
Person / Time
Site:	Millstone
Issue date:	06/30/1980
From:	NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:
Shared Package
ML19318A340	List: ML19318A339 ML19318A345
References
TAC-11348, TAC-11561, TAC-12505, TAC-42846, NUDOCS 8006190610
Download: ML19318A345 (38)

Text

DOCKET NO. 50-336 ATTACIDIENT 1 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 LARGE BREAK LOCA/ECCS PERFORMANCE RESULTS JUNE, 1980 8006190f(g

LOSS OF COOLANT ACCIDENTS RESULTING FROM PIPlhG BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY Introduction The Acceptance Criteria for LOCA analysis is aescribeo in 10CFR50.40 [lj as follows:

1. The calculated fuel element peak claa temperature is Delow tne 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 maount of Zircaloy in the reactor.

3. The clad temperature transient is terminateo at a time wnen tne core geometry is still amenable to cooling. The localized claacing oxi-dation limits of 17 percent are not exceecea auring or after quenching.

.4 The core remains amenable to cooling curing and af ter tne break.

5. The core temperature is reduced ano decay neat is removea for an extended period of time, as requirea by the long livea racioactivity remaining in the core.

These criteria were established to provide significant margin in E.mer-gency Core Cooling System (ECCS) Derformance following a LOCA.

Mathematical Model The requirements of an acceptable ECCS evaluation model are presentea in AppendixKof10CFR50[1]. .

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large Break LOCA Evaluation Model The analysis of a large break LOCA Transient is divided into three phases: 1) blowdown, 2) refill, and 3) reflood. There are three dis-tinct transients analyzed in each phase, namely the thermal-hydraulic transient in the RCS, the pressure and tenperature transient witnin the Containment, and the pellet and clad temperature transient of the hot-test fuel rod in the core. Based on these considerations, a system of interrelated computer codes has been developed for the analysis of the LOCA.

The description of the various aspects of the Westinghouse LOCA analysis methodology is given in Reference [2]. This document describes the major pheromena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with the Acceptance Criteria. The SATAN-VI, WREFLOOD, C0CO, and LOCTA-IV codes which are used in the LOCA analysis are described in detail in References [3]

through [6]; code modifications are specified in References [7] through (11). These codes are used to assess the core heat transfer geometry and to determine if the core remains amenable to cooling througnout ana subsequent to the blowdown, refill, and reflood phases of the LOCA. The SATAN-VI computer code analyzes the thermal-hydraulic transient in the RCS & ring blowdown, and the WREFLOOD computer code is used to calculate this transient iring the refill and reflood phases of the accident.

The C0C0 computer code is used to calculate the Containment pressure transient throughout the LOCA analysis. Similarly, tne LOCTA-IV compu-ter code is used to compute the thermal transient of the hottest fuel rod & ring the entire ana'.ysis.

SATAN-VI is used to calculate the RCS pressure, enthalpy, density, and the mass and energy flow rates in the RCS, as well as steam generator energy transfer between the primary ano secondary systems as a function of time during the blowdown phase of the LOCA. SATAN-VI also calculates the accunulator water flow rates and internal pressure ano the pipe ,

break mass and energy flow rates that are assumed to be vented to tne Containment during blowdown. At the end of the blewcown phase, these

' data are transferred to the WREFLOOD code. The mass and energy release rates d; ring blowdown are utilized in the C0C0 code for use in the determination of the Containment' pressure response daring this first phase of the LOCA. Additional SATAN-VI output cata including the core flow rates and enthalpy, the core pressure, and the core power decay transient, are transferred to the LOCTA-IV code.

With initial information from the SATAN-VI code, WREFLOOD uses a system thermal-hydraulic model to determine the core flooding rate (i.e., the rate at whicn coolant enters the bottam of the core), the coolant pres-sure and temperature, and the core water level daring the refill anc reflood phases of the LOCA. WREFLOOD also calculates the mass ano energy flow addition to tne Containment through the creas. Since tne mass flow rate to the Containment depenas upon the core flooding rate and the local core pressure, which is a function of the Containment backpressure, the WREFLOOD and C0C0 codes are interactively linkea.

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 tne fuel temperature. LOCTA-IV is used throughout the analysis of tne LOCA transient to calculate the fuel clad temperature and metal-water reat-tion of the hottest rod in the core.

The large break analysis was performed with the Westingnouse evaluation model which includes modifications delineated in References (7, 8,10 and 12]. Reactor Coolant p;mps are assumea to continue to run 0; ring blowdown unless otherwise noted.

l l

Results Large Break Results Based on the results of the LOCA sensitivity studies, (References [7]

and [13]) the limiting large break will be the double enoea colo leg guillotine (DECLG). This conclusion is confirmeo for the hillstone 2 Tnerefore, only plant specifically by docketed analyses (Reference 14).

the DECLG break need be considereo in the large Drea< ECCS perf ormance ois-analysis. Calculations were performeo for a range of Moooy creak charge coefficients. The results of these calculations are sumarizea in Tables 1 and 2. Containment parameters utilizeo in tne analyses arte provided in Table 3.

The maximum clad temperature calculated for a large break is 2111*F which is less than the Acceptance Criteria limit of 2200*F of 10CFR50.46. Maxinum local metal-water reaction is 5.5 percent whicn is well below the embrittlement limit of 17 percent as requirea Dy 10CFR50.46. Total core metal-water reaction is less than 0.3 percent for all breaks, as compared with the 1 percent criterion of 10CFR50.46, and the clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. As a result, the core tenpera-tore will continue to drop and the ability to remove decay heat genera-ted in the fuel for an extended period of time will be maintained.

Figures 1 through 26 present the parameters of principal interest from the large break ECCS analyses. For all cases analyzea transients of tne following parameters are presented:

1. Hot spot clad temperature.

2. Coolant pressure in the reactor core.

3. Water level in the core and downcuner during reflood.

4 Containment pressare transient

D For the limiting break analyzed, the following aooitional transient parameters are presented in the figures:

1. Core flow during blowdown (inlet and outlet).

2. Fuel roc heat transfer CDefficients.

3. Hot spot fluid temperature.

4. Mass released to Containment curing olowdown.

' 5. Energy released to containment auring olowdown.

6. :1uid quality in the hot assembly during Dlowdown.

7. Mass velocity during blowoown.

8. Safety injection tank water flow rate into RC5 auring olowao.n (per tank).

9. Osmpe: safety injection water-flow rete ouring retlooa.

10. :cre re'looding rate.

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REFERENCES

1. " Acceptance Criteria for Emergency Core Cooling Systems for Lignt.

Water Cooled Nuclear Power Reactors," 10CFR50.46 ano Appendix K of 10CFR50. Federal Register, Volume 39, Number 3, January 4, 1974.

"Westingnouse tLL)

2. Bordelon, F. M., Massie, H. W. and Zoroan T. A.,

Evaluation Model - Summary," WCAP-8339, July 1974.

3. Bordelon, F. M., et al., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss of Coolant," WCAP-8302 (Proprietary) ana WC AP-8306 (Non-Proprietary), June 1974.

4. Kelly, R. D., et al., " Calculational Model for Core Refloooing af ter a Loss of Coolant Accident (WREFLOOD Code)," WCAP-617u (Proprietary) and WCAP 6171 (Non-Proprietary), June 1974.

Bordelon, F. M. and Murphy, E. T., " Containment Pressure Analysis 5.

Coce (C0CO)," WCAP-8327 (Proprietary) ano WCAP-8326 (hon-Proprietary), June 1974.

6. Bordelon, F. M. , et al ., "LOCTA-IV Program: Loss of Coolant Tran-sient Analysis," WCAP-8301 (Proprietary) ana WCAP-8305 (Non-Prcprietary), June 1974

7. Ferguson, K. L. , and Kemper, R. M., ECCS Evaluation Mooel for Westinghouse Fuel Reloads of Combustion Engineering NSSS, WCxP-9526 (Proprietary) and WCAP-9529 (Non-Proprietary), June 1979.

6. Ferguson, K. L., ana Kemper, R. M., Accenoun to ECLS Evaluation Mocel for Westinghouse Fuel Reloads of Comoustion Engineering kdds, October 1979.

9. Borcelon, F. M., et al., " Westinghouse ECCS Evaluation Nooel - Sup-plementary Information," WCAP-8471 (Proprietary) anc ntnP-d*72 (non-Proprietary), April 1975.

• 10. " Westinghouse ECCS Evaluation Model - October 1975 Version,"

WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary).

November 1975.

11. Letter NS-CE-924, dated January 23, 1976, C. Eiche1dinger (Westing-house) to D. B. Vasta11o (NRC).

12. Eicheldinger, C., " Westinghouse ECCS Evaluation Model,, February 1978 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-P-A (Non-Proprietary Version), February 1973.

13. Salvatori, R., " Westinghouse ECCS - Plant Sensitivity Studies, WCAP-8340 (Proprietary) and WCAP-8356 (hon-Proprietary), July 1974

14. W. G. Counsil to R. Reid Docket ho. 50-336, March 30,1979.

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TABLE 1 LARGE BREAK.

TIME SEQUENCE OF EVENTS CD=0.6 DECLG-C =0.8 DECLG C =0.6 DECLG C =0.4 DECLG. RC Pumps Tripped D D D (Sec) (Sec) (Sec) (Sec)

START 0.0 0.0 0.0 0.0 S. 1. Signal

• 0.6 0.69 0.85 0.68 S. I. Tank Injection 13.3 15.7 21.6 16.4 End of Blowdown 20.19 21.65 29.07 22.43 Bottom of Core Recovery 33.2 34.6 43.0 36.4 S. 1. Tank Empty 64.2 66.8 73.2 67.5 4 End of Bypass 20.18 21.65 29.07 22.43

)

i

• from containment pressure sensor 4

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=

o. .

TABLE 2 LARGE EAEAK-Co=0.6 CD=0.8 DECLG. CD=0.6 DECLG CD=0.4 DECLG RC Pumps Tripped -

Results Peak Clad Temp *F 1985 2111 2000 1976 Peak Clad Location,Ft. 7.0 7.5 7.0 7.0 Local Zr/Hp0 Rxn(max) % 3.7 5.5 3.9 3.6 Local Zr/H2O Location,Ft. 7.5 7.5 7.0 7.0 Total Zr/H O <0.3 <0.3 <0.3 <0.3 2 Rxn, %

Hot Rod Burst Time,sec 36.9 31.6 53.2 41.2 Hot Rod Burst Location, Ft. 5.70 5.70 6.25 5.70 Calculation Assumptions NSSS Power,Mwt,102% of 2700 Peak Core Linear Power, kw/ft 15.6 S.I. Tank Actuation Pressure, psia 215 S.I. Tank Water Volume, f t 3 per tank 1107 i

1 Y

T' TABLE 3 Millstone Unit 2 Containment Physical Pargmeters_

1.938 x 106 ft3 het Free Volume Containment Initial Conditions: 99 %

Humidity 60*F Containment Temperature Enclosure Building Temperature 60*F

, 4u,F Ground Temperature 14.7 psia Initial Pressure Initial Time ~for: 26 seconos

' Spray Flow 0.0 seconos

Fans (3) 14.0 seconos Additional Fan Containment Spray Water

• 60 F Temperature Flow Rate (Total, 2 pumps) 3300 gpm Fan Cooling Capacity (Per Fan)

VaporTemperature(*Fl Capacity (BTU /5ec)

)

60 0.0 l

145 3360.0

165 5260.0 1- 28800.0 l

300 350 32400.0 Containment Heat Absorbing Surfaces

-1. Surface Areas ana Th cknesses

a. Shell and come - 71,870 Ft2 (1) Paint - 0.003 In. (one side exposee to containment atmosphere)

(2) Carbon steel - 0.25 In.

(3) Concrete - 3.0 Ft. (one side exposea to enclosure ouilaing atmosph::re)

b. Unlineo Concrete - 62,800 Ft2 (1) Concrete - 2.0 Ft. (one siae exposed to containment atmosphere, one-Side insulated)

c. Galvanized Steel - 120,000 Ft2

.(1) Zinc 0.0036 In. (one sice exposeo to containment

. atmosphere)

(2) Carbon steel.- 0.20 In. (one sioe insulateo)

- _ _ ~ _ _ , _ - . - . _ _ _ _. - _ - ._

TABLE 3 (Cont'o.)

Millstone Unit 2 Containment Physical Parameters i-

d. Painted Thin Steel - 56,850 Ft2 (1) Paint - 0.003 In. (one side exposed to containment atmosphere)

(2) Carbon steel -'0.2' In. (one siae insulatea)

e. Painted Steel - 32,600 Ft2 (1) Paint - 0.003 In. (one side exposed to containment atmopshere) 4 (2) Carbon steel - 0.26 In. (one siae insulatea)

f. Painted Steel - 22,425 ft2 (1) Paint - 0.003 In. (one side exposed to containment atmosphere)

(2) Carbon steel - 0.86 In. (one siae insulatea) 9 Painted Thick Steel 4,230 Ft2 (1) Paint - 0.003 In. (one side exposea to containment atmosphere)

(2) Carbon steel - 2.94 In. (one sice insulatea)

h. Containment Penetration Area - 3,000 Ft2 (1) Paint - 0.003 In. (one side exposea to containment atmosphere)

(2) Carbon steel - 0.75 In.

(3) Concrete - 3.75 Ft. (one siae exposed to enclosure builaing atmosphere)

i. Stainless Steel Line Concrete - 8,340 Ft2 (1) Stainless steel - 0.25 In. (one side exposed to containment atmosphere)

(2) Concrete - 2.0 Ft. (one side insulatea)

j. Base Slab'- 11,130 Ft2 (1) Concrete - 8.0 Ft. (one side exposed to containment sump, one side exposed to ground)

k. Neutron Shield - 1400 Ft2 (1) Stainless steel - 0.024 Ft. (botn sides exposea to containment atmosphere)

2. Thermal Properties Heat Capacity Material Conductivity (BTU /hr-ft- F ) (BT0/ f t3_* F )

a. Concrete 2.0 30

b. Carbon Steel 35.0 55

c. Stainless Steel 10.0 62

d. Paint 1.5 32

-e. Zinc 70.0 45

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nc... . I l . am m.n... in aan s.= nr i n u 7. = n w { C i Im... - w R - N i l in.. . , I a f ' w i.0.0. 1 5 ~ 1 3 1 4 me. . m i k 4 s e a a s '

3 8 i i i i i

VM suci i j FIGURE 19 - HOT SPOT CLAD TEMPERATURE, 0.8 DECLG

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t i l ' ' -4 lIi nd 4 S aN a P R ad T G c L t C s E i D t n 6 i T 0 _ E R U S S ai E R P T N A L O O C R O T C A E R 3g 4 2 E R U G I F l- _

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1 30 25' - 52 20 - - _if i m T 3 g 15 - m W o. 10 - j 5 - i l l 0 0 100 2% M TIME (SECONDS) Figure 26 Cor.tainment Pressure, CD = 0.6 DECLG, Pump Trb

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