Text

Tech Specs Page Revisions to License DPR-26 Re Accumulator Water Volume & New Limit for Total Nuclear Peaking Factor. W/Encl Safety Evaluation
	ML19259A599
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Site:	Indian Point
Issue date:	01/03/1979
From:	; CONSOLIDATED EDISON CO. OF NEW YORK, INC.
To:	;
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ANALYSIS OF THE EMERGENCY CORE COOLING SYSTEM IN ACCORDANCE WITH THE ACCEPTANCE CRITERIA OF 10CFR50.46 AND APPENDIX K OF 10CFR50 CONSOLIDATED EDISON COMPANY OF NEW YORK, INC.

INDIAN POINT UNIT NO. 2 DOCKET NO. 50-247 FACILITY OPERATING LICENSE NO. DPR-26 December, 1978 7901080(S%

ANALYSIS OF THE EMERGENCY CORE COOLING SYSTEM IN ACCORDANCE WITH THE ACCEPTANCE CRITERIA OF 10CFR50.46 AND APPENDIX K OF 10CFR50 The analysis specified by the Nuclear Regulatory Commission in the " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors" per 10CFR50.46 and Appendix K of 10CFR50 (Reference 1), as published in the Federal Register, January 4, 1974, for small breaks in the reactor coolant piping is presented in Reference 2 for Indian Point Unit No. 2 (IP2).

For major Reactor Coolant System pipe ruptures (Loss-of-Coolant Accident) the analysis specified by 10CFR50.46 (Reference 1) is presented herein.

The analytical techniques used are in compliance with Appendix K of 10CFR50 and are described in Reference 3.

The individual computer codes which comprise the Westinghouse Emergency Core Cooling System (ECCS) evaluation model are described in detail in References 4,5,6 and 7 along with the code modifications specified in References 8, 9,

and 10.

The analysis presented here was performed with the February 1978 version of the evaluation model-which includes modifications delineated in References 11, 12, 13 and 14.

The boundary considered for loss of coolant accidents as related to connecting pipings is defined in Section 4.1.3 of the IP3 FSAR.

Should a major break occur, depressurization of the Reactor Coolant System results in a pressure decrease in the pressurizer. Reactor trip signal occurs when the pressurizer low pressure trip setpoint is reached.

A Safety Injection System signal is actuated when the appropriate setpoint is reached.

These countermeasures will limit the consequences of the accident in two ways:

a.

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.

b.

Injection of borated water provides heat transfer from the core and prevents excessive clad temperature.

At the beginning of the blowdown phase, the entire Reactor Coolant System 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 10CFR50.

Thereafter, the core heat transfer is based on local conditions with transition boiling and forced convection tc steam as the major mechanisms.

During the refill period, rod-to rod radiation is the only mechanism.

When the Reactor Coolant System pressure falls below 600 psia, the accumulators begin to inject borated water.

The conservative assumption is made that accumulator water injection bypasses the core and goes out through the break until the termination of bypass.

This conservatism is consistent with Appendix K of 10CFR50.

Thermal Analysis Westinghouse Performance Criteria for Emergency Core Cooling System The reactor is designed to withstand thermal effects caused by a loss-of-coolant accident including the double ended severance of the largest Reactor Coolant System pipe.

The reactor core and internals together with the Emergency Core Cooling System are designed so that the reactor can be safely shut down and the essential heat transfer geometry of the core preserved following tha accident.

The Emergency Core Cooling System even when operating with the most severe single active failure is designed to meet the Acceptance Criteria.

Method of Thermal Analysis The description of the various aspects of the LOCA analysis is given in Reference 3.

This document describes the major phenomena modeled, the interface among the computer codes and features of the codes which maintain compliance with the Acceptance Criteria.

The individual codes are described in detail in References 4,5,6 and 7.

The analysis presented in this submittal is performed for the DECLG break with CD = 1.0, 0.8, 0.6 and 0.4 and with a reactor vessel fluid inlet temperature of 521.4 F (vessel 0

average temperature of 549 F) and consistent secondary-side initial conditions.

The analysis considers the reactor vessel upper head fluid temperature to be equal to the Reactor Coolant System hot leg fluid temperature.

The effect of using the hot leg temperature in the reactor vessel upper head region is described in Reference 15.

Reference 16 presents a break spectrum sensitivity study using the increased upper head fluid temperature.

In addition this analysis has included the effects of 6% uniform steam generator tube plugging.

Results Table IP2-1 presents the peak clad temperature and hot spot metal water reaction.

The time sequence of events during the large break is shown in Table IP2-2.

The SATAN-VI blowdown and reflood analysin of the loss-of-coo?. ant accident, and the peak linear power used in the rod heat up calculations (LOCTA-IV code) is based on 102 percent of the licensed core power rating.

Since there is margin between the value of the peak linear power density used in this analysis and the value expected in operation, a

lower peak clad temperature would be expected by using the peak linear power density expected during operation.

For the results discussed below, the hot spot is defined to be the location of maximum peak clad temperature.

This location is given in Table IP2-1.

1 Figures IP2-la through IP2-17 present the transient behavior of the principal parameters, as follows:

Figures IP2-la These figures show the fluid quality, through IP2-3d

.he mass velocity and the heat transfer coefficient (as calculated by the LOCTA-IV code) at the hot spot (location of maximum clad temperature) and burst location, on the hottest fuel rod (hot rod).

Figures IP2-4a These figures show the core pressure, the through IP2-6d flow rate out of the break

'*he sum of both ends for the guillotine break) and the core pressure drop (from the lower plenum near the core, to the upper plenum at the core outlet).

Figures IP2-7a These figures show the clad temperature through IP2-9d transient at the hot spot and the burst location, the fluid temperature (also for the hot spot and burst location), and the core flow (top and bottom).

Figures IP2-10a These figures show the core reflood through IP2-lld transient parameters (water level and flooding rate).

Figures IP2-12a These figures show the Emergency through IP2-13d Core Cooling System flow.

The accumulator flow assumed is the sum of that injected in the intact cold legs.

Figure IP2-14a These figures show the containment through IP2-14d pressure transient.

Figure IP2-15a These figures show the core power through IP2-15d transient.

Figure IP2-16 This figure shows the break energy released to the containment, for the worst break.

Figure IP2-17 This figure shows the containment wall condensing heat transfer coefficient, for the worst break.

In addition to the above, Tables IP2-4 and IP2-5 present the reload mass and energy released to the containment, and the broken loop accumulator mass and energy flow rate of the containment for the CD = 0.6 case (worst break).

The clad temperature analysis is based on a total peaking factor of 2.31.

The hot spot metal-water reaction reached is well below the embrittlement limit of 17% as required by 10CFR50.46.

In t.ddition, the total core metal-water reaction is less than 0.3% as compared to the 1%

criterion of 10CFR50.46.

The results of several sensitivity studies are reported in Reference 17.

These results are for conditions which are not limiting in nature and hence are reported on a generic basis.

Conclusions-Thermal Anaysis For breaks up to and including the double ended severance of a reactor coolant pipe, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10CFR50.46.

That is:

1.

The calculated peak fuel element clad temperature does not exceed 22000F based on a total core peaking factor of 2.31.

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 cladding oxidation limits of 17% are not exceeded during or after quenching.

4.

The core temperature is reduced and decay heat is removed for an extended period of time, as required by the longlived radioactivity remaining in the core.

References 1.

" Acceptance Criteria for Dnergency Core Cooling Systms for Light Water Cooled Nuclear Power Reactors", 10CFR50.46 and Appendix K of 10CFR50.46.' Federal Register, Volume 39, Number 3, January 4, 1974.

2.

Intter frun William J. Cahill, Jr. of Consolidated Edison Co. of New York, Inc. to Ibbert W. Reid of the Nuclear Regulatory Catmission, dated July 13, 1976. Indian Point Unit tb. 2 Small Break IDCA Analysis.

3.

Bordelon, F. M., et al., "The Westinghouse ECCS Evaluation Model -

Sunmary", ICAP-8339; July 1974.

4.

Bordelon, F. M., et al., " SATAN-VI Program: Cmprehensive Space-Time Dependent Analysis of Ioss-of-Coolant", FCAP-8306 (Non-Proprietary), June 1974.

5.

Bordelon, F. M., et al, "IDCTA-IV Program: Ioss-of-Coolant Transient Analysis", NCAP-8301 (Proprietary), VCAP-8305 (Non-Proprietary), June 1974.

6.

Kelly, R.D., et al., " Calculational bbdel for Core Reflooding after a Ioss-of-Coolant Accident (WREFIDOD Code)", PCAP-8171 (Non-Properietary),

June 1974.

7.

Bordelon, F. M., and Murphy, E. T., " Containment Pressure Analysis Code (COCO)", NCAP-8327 (Proprietary), NCAP-8326 (Non-Proprietary),

June 1974.

8.

Bordelon, F. M., et al., "The Westinchouse EDCS Evaluation Model -

Supplementary Infonnation",FCAP-8471 (Proprietary), VCAP-8472 (Non-Proprietary), January 1975.

9.

" Westinghouse DCCS Evaluation Model, October 1975 Version", NCAP-8622 (Proprietary), and VCAP-8623 (Non-Proprietary),

Novmber 1975.

10.

Ictter frm C. Eicheldinger of Nestinghouse Electric Corporation to D. B. Vassalo of the Nuclear Regulatory Conmission, Intter Nt'.nber NS-CE-924, dated January 23, 1976.

11.

Kelly, R. D., Tlungson, C. M., et al., "Westinglouse Dnergency Core Cooling Systm Evaluation Model for Analyzing Large IDCA's During Operation With One Icop Out of Service for Plants Without Icop Isolation Valves," NCAP-9166, February,1978.

12.

Eicheldinger C., " Westinghouse ECCS Evaluation Model, February 1978 Version", NCAP-9220-P-A (Propriecry), VCAP-9221-A (Non-Proprietary),

February, 1978.

13. Ictter frm T. M. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Ccmnission, Ictter Number NS-

'INA-1981, Novmeber 1,1978.

References (Cont'd)

14. Intter frcm T. M. Anderson of Westinghouse Electric Corporation to Tedesco of the Nuclear Regulatory Ccmnission, Ietter NtInber NS 'IM-2014, dated Decenber 11, 1978.

15.

Letter frca C. Eicheldinger of Westinglxxise Electric Corporation to V. Stello of the Nuclear Regulatory Ccmnission, Intter Ntaber NS-CE-ll63, dated August 13, 1976.

16. Beck, II.

S., et al., " Westinghouse ECCS - ruar Icop Plant (15x15)

Sensitivity Studies with Upper Head Fluid Tenperature at T Hot",

NCAP-8855-A(Non-Proprietary), Eby 1977.

17. Salvatori, R., " Westinghouse ECCS - Plant Sensitivity Studies",

NCAP-8340 (Proprietary), NCAP-8356 (Non-Proprietary), July,1974.

TABLE IP2-1 IARGE BREAK - E.W'LTS AND ANALYSIS _I_t[P,U_T DECLG DECLG DECLG DECLG (C =0.6)

(C =0.4)

(C =0. 8 (C =1.0)

D D

Results OF 2137 2078 2172.5 1684 Peak Clad Temp.

Peak Clad Location Ft.

6.0 7.5 7.5 7.5 Local Zr/H O Rxn (max)%

6.6 4.5 6.14 1.03 2

Local Zr/H O Location Ft.

6.0 6.25 6.0 7.5 2

Total Zr/H O Rxn %

40.3 40.3 40.3 40.3 2

Hot Rod Burst Time sec 29.5 31.2 32.8 N

q)

Hot Rod Burst Location Ft.

6.0 6.25 6.0 C

H CD Calculation 00 CD NSSS Power Mwt 102% of 2758 D]f Peak Linear Power kw/ft 102% of 13.36 Peaking Factor (At License Rating) 2.31 3

Accumulator Water Volume (ft per tank) 716 Accumulator Pressure (psia) 600 Number of Safety Injection Pumps operating 2

Steam generator tube Plugging Level (%)

6 (uniform)

TABLE IP2-2 LARGE BREAK - TIME SEQUENCE OF EVENTS OCCURENCE TIME (SECONDS)

DECLG DECLG DECLG DECLG (C =1.0)

(Cp=0.8)

(C =0.6)

(C =0.4)

D D

D Accident Initiation 0.0 0.0 0.0 0.0-Reactor Trip Signal 0.54 0.54 0.54 0.55 Safety Injection Signal 0.97 1.04 1.17 1.43 Start Accumulator Injection 14.2 14.4 16.1 21.4 Start Pumped ECC Injection 25.97 26.04 26.17 26.43 End of ECC Bypass 25.73 26.6 29.0 33.4 End of Blowdown 29.0 29.3 31.5 36.5 Bottora of Core Recovery 40.4 41.6 44.8 48.5 Accumulator Empty 46.96 47.2 49.2 55.1

TABLE IP2-3 LARGE BREAK CONTAINMER_DA_TA 6

3 NET FREE VOLUME 2.61 x 10 ft INITIAL CONDITIONS Pressure 14.7 psia 0

Temperature 90 F

RWST Temperature 40 OF Service Water Temperature 35 0F Outside Temperature

-20 OF SPRAY SYSTEM Number of Pumps Operating 2

Runout Flow Rate 3000 gpm Actuation Time 20 secs SAFEGUARDS FAN COOLERS Number of Fan Coolers Operating 5

Fastest Post Accident Initiation of Fan Coolers 30 secs STRUCTURAL TIEAT SINKS 2

Thickness (In)

Area (Ft )

1.

0.007 Paint, 0.375 steel,54.0 concrete 45,684 2.

0.007 Paint,0.5 steel, 42.0 concrete 28,613 3.

12.0 concrete 15,000 4

0.375 stainless steel, 12.0 concrete 10,000 5.

12.0 concrete 61,000 6.

0.5 steel 68,792 7.

0.007 Paint, 0.375 steel 81,704

TABLE IP2-3 (continued)

.I._ARGE BREAK _CONTA_ME DAT_A _

2 Thickness (In)

Area (Ft )

8.

O.25 steel 27,948 9.

0.007 Paint, 0.1875 steel 69,800 10.

0.125 steel 3,000 11.

0.138 steel 22,000 12.

0.0525 steel 10,000 13.

0.019 stainless steel, 1.25 insulation, 0.75 steel, 54.0 concrete 785 14.

0.019 stainless steel, 1.25 insulation, 6849 0.5 steel, 54.0 concrete

~

15.

0.025 stainless steel, 1.5 insulation 3816 0.5 steel, 54.0 concrete 16.

0.025 stainless steel, 1.5 insulation, 4362 0.375 steel, 54.0 concrete

TABLE IP2-4 REPLOOD !%SS AND ENERGY RELEASE TO THE CONTAINMENT 0.6 DECLG BREAK T_iyM_[ seq)

Mass _,,(1, m/,sec),

Enerqv [B,TU/,s,e_c) b 44.8 0.0 0.0 45.6 5.34 6902.

50.2 32.7 42198.

60.7 38.7 49335.

77.3 65.63 81625.

95.0 89.45 110172.

130.2 123.76 151134.

164.0 360.19 220762.

200.0 372.2 208660.

TABLE IP2-5 BROKEN LOOP ACCUMUIATOR MASS AND EIERGY

- - _-_._ _ _ _ _ _ __._ _._ _ _ _ __. _ _ _ _ _ 7 _ _.

RELEASE TO THE CONTAINMENT 0.6 DECLG BREAK Time (sec)

M_a s_s_(lbm/se_q),

Energy (BTU /sec) 0.0 4288.

255679.

2.0 3417.

206971.

4.0 2997.

178708.

6.0 2672.

159350.

8.0 2430.

144890.

10.0 2237.

133417.

12.0 2079.

123957.

14.0

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