ML19261B637

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ECCS Reanalysis for Large Break LOCA Analysis Based on Feb 1978 Evaluation Model
ML19261B637
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Site: Prairie Island  Xcel Energy icon.png
Issue date: 02/21/1979
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NORTHERN STATES POWER CO.
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ML19261B638 List:
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NUDOCS 7902280311
Download: ML19261B637 (88)


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NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT TO THE UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NOS. 50-282 LICENSE NOS. DPR-42 306 DPR-60 LARGE BREAK LOSS OF COOLANT ACCIDENT ANALYSIS FEBRUARY 1978 EVALUATION MODEL 4

Date:

Februa ry 21, 1979

TABLE OF CONTENTS SUBJECT PACE

1.0 INTRODUCTION

1 2.0 THERMAL ANALYSIS...............................

1 2.1 Westinghouse Performance Crf teria for Emergency Core Cooling System.............

1-2 2.2 Method of Thermal Analysis................ 2 2.3 Results...................................

2 2.4 C on c lu s i o ns............................... 3-4 2.5 Generic Sensitivity Studies...............

4 3.0 R E F E RE N C E S..................................... 5 - 6 4.0 LIST OF TABLES.................................

7 Table 1........................................

8 Table 2........................................

9 Table 3........................................

10-11 Table 4........................................

12 Table 5........................................

13 5.0 LIST OF FIGURES................................

14 Figures 1-17................................... 15-31 APPENDIX A GENERIC SENSITIVITY STUDY RESULTS FOR TWO-LOOP PLANT WITH 14 x 14 FUEL......... A-1

1.0 INTRODUCTION

....................................A-2 2.0 T A BL E S.......................................... A -A-8 3.0 FIGURES.........................................A-9--A-55

1.0 INTRODUCTION

The analysis specified by 10CFR50.46III " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Power Reactors",

is presented in this report.

The results of the loss of coolant accident analyses are shown in Table 2 and show compliance with the Acceptance Criteria.

The analytical techniques used are in compliance with Appendix K of 10CFR50, and are described in Reference (2]. This information amends the Saf ety Analysis Report on Major Reactor Coolant System Pipe Ruptures.

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:

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.

2.

Injection of borated water provide- cat transfer from the core and prevents excessive clad temperatures.

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. Af ter the break develops, the time to departure f rom nucleate boiling is calculated, consistent with Appendix K of 10CFR50.

Thereaf ter the core heat transfer is based on local conditions with boiling and forced convection to steam as the major heat transfer transition mechanisms. During the refill period, rod-to-rod radiation is the only heat transfer mechanism.

When the Reactor Coolant System pressure f alls below 700 psia, the accumulators begin to inject borated water. The conservative assumption is made that the accumulator water injected bypasses the core and goes out through the break until the termination of bypass. This conservatism is again consistent with Appendix K of 10CFR50.

2.0 THERMAL ANALYSIS 2.1 Westinghouse Performance Criteria for Emergency Core Cooling System The reactor is designed to withstand thermal ef f ects 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 (ECCS) are designed so that the reactor can be safely shutdown and the essential heat transfer geometry of the core preserved following the accident.

The ECCS, even when operating during the injection mode with themostseveresingleggyivefailure, is designed to meet the Acceptance Criteria 2.0 THERMAL AMLYSIS (continued) 2.2 Method of Thermal Analysis The description of the various aspects of the loss of coolant accident analysis is given in Reference (2]. This document describes the major phenomena modeled, the interf aces 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 (3) th rough [6].

The analyses presented here were performed using the February 1978 version of the Westinghouse Evaluation Model. This version includes the r.udifications to the models, referenced above, specified in References (7, 10 and 11]. The February 1978 Westinghouse Evaluation Model is documented in Ref erences (12, 13, 14 and 15]. Containment data used to calculate ECCS backpressure is presented in Table 3.

2.3 Results The analysis of the loss of coolant accident is performed at 102 percent of the licensed core power rating. The peak linear power and total core power used in the analysis are given in Table 2.

Since there is margin between the value of peak linear power density used in this analysis and the value of the peak linear power density expected during plant operation, the peak clad temperature calculated in this analysis is greater than the maximum clad temperature expected to exist.

Table 1 presents the occurrence time for various events th roughou t the accident transient.

Table 2 presents selected input values and results from the hot fuel rod thermal transient calculation.

For these results, the hot spot is defined as the location of maximum peak clad temperatures. That location is specified in Table 2 for the break analyzed. The location is indicated in feet, which presents elevation above the bottom of the ac;ive fuel stack.

Table 3 presents a summary of the various containment systems parameters and structural p9gymeters which ware used as input to the COCO computer code used in this analysis.

Tables 4 and 5 present reflood mass and energy releases to the containment, and the broken loop accumulator mass and energy release to the containment, respectively.

The results of several sensitivity studies are reported These results are for conditions which are not limiting in nature and hence are reported on a generic basis.

THEa d ANALYSIS (continued) 2.0 2.3 Results (continued)

Figures 1 through 17 present the transients for the principal parameters for the break sizes analyzed. The following items are noted:

Figures 1-3:

Quality, mass velocity and clad heat transfer coefficient for the hot spot and burst locations.

Figures 4-6:

Core pressure, break flow, and core pressure drop.

The break flow is the sum of the flowrates f rom both ends of the guillotine ',reak.

The core pressure drop is taken :.. the pressure just before the core inlet to the pressure just beyond the core outlet.

Figures 7-9:

Clad temperature, fluid temperature and core flow. The clad and fluid temperatures are for the hot spot and burst locations.

Figures 10-11:

Downcomer and core water level during reflood, and flooding rate.

Figures 12-13:

Emergency core cooling system flowrates, for both accumulator and pumped safety injection.

Figures 14-10:

Containment pressure and core power transients.

Figures 16-17:

Break energy release during blowdown and the containment wall condensing heat transfer coef ficient for the worst break.

2.4 Conclusions Fcr breaks up to and including the double ended severance of a reactor coolant pipe, the Emergency Core Cooling System ggj will meet the Acceptance Criteria as presented in 10CFR50.46 That is:

1-ihe calculated peak clad temperature does not exceed 2200 F based on a total core peaking factor of 2.28.

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.

2.0 THEILMAL ANALYSIS (continued) 2.4 Conclusions (continued) 3.

The clad temperature trenstent 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 af ter quenching.

4.

The core temperature is reduced and decay heat is removed f or an extended period of time, as required by the long-lived radioactivity remaining in the core.

Theeffectjgg[2DyPerplenuminjectionwerediscussedwith the staf f The modeling of upper plenum injection has been shown to be a benefit for the Prairie Island Plant per the above ref erences.

Westinghouse is currently developing a model f or two-loop plants, however, this model has not been completed. Thus, the results presented herein represent a conservative analysis without explicit modeling of upper plenum injection.

2.5 Generic Sensitivity Studies Appendix A contains the results of a generic sensitivity study for a typical two-loop plant with 14x14 fuel.

This sensitivity study was performed to demonstrate that th e limiting break does not change due to a correction in the metal-water heat of reaction calculation which is included in the February 1978 version of the Westinghouse ECCS evaluation model. The results of this generic sensitivity study show that the limiting break for Westinghouse plants of this type is a double-ended cold leg guillotine with a dischargecoefficienI8h]0.4. This agrees with past sensitivity studies

3.0 REFERENCES

1.

" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors", 10CFR50.46 and Appendix K of 10CFR50.46. Federal Register, Volume 39, Number 3, January 4, 19 74.

2.

Bordelon, F M, Massic, H W, and Zordan, T A " Westinghouse ECCS 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 Version), WCAP-8306 (Non-Proprietary Version),

June, 19 74.

4.

Bordelon, F M., et al., "LOCTA-IV Program:

Los s -of-Coola nt Transient Analysis", WCAP-8301 (Proprietary Version),

WCAP-8305 (Non-Proprietary Version), June, 1974.

5.

Kelly, R D, et al., " Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (UREFLOOD Code)".

WCAP-8170 (Proprietary Version), WCAP-8171 (Non-Proprietary Version),

June, 1974.

6.

Bordelon, F M, and Murphy, E T, " Containment Pressure Analysis Code (COCO)", WCAP-8327 (Proprietary Version),

WCAP-8326 (Non-Proprietary Version), June, 1974.

7.

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

Supplementary Information", WCAP-8471 (Proprietary Version), WCAP-8472 (Non-Proprietary Version), January, 1975.

8.

Salvatori, R, " Westinghouse ECCS - Plant Scncitivity Studies",

WCAP-8340 (Proprietary Version), WCAP-8356 (Non-Proprietary Version), July, 1974.

9 Delsignore, T, et al., " Westinghouse ECCS Two-Loop Sensitivity Studies (14x14)", WCAP 8854, (Non-Proprietary Version),

September, 1976.

10.

" Westinghouse ECCS Evaluation Model, October, 19 75 Versions",

WCAP-8622 (Proprietary Version), WCAP-8623 (Non-Proprietary Version), November, 1975.

11.

Letter from C Eicheldinger of Westinghouse Electric Corporation to D B Vassalo of the Nuclear Regulatory Commission, letter number NS-CE-924, January 23, 1976.

12.

Kelly, R D, Thompson, C M, et al., " Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing Large LOCA's During Operation with One Loop Out of Service for Plants without Loop Isolation Valves", WCAP-9166, February, 1978.

3.0 REFERENCES

(continued) 13.

Eicheldinger, C, " Westinghouse ECCS Evaluation Model, February 1978 Version", WCAP-9220 (Proprietary Version),

WCAP-9221 (Non-Proprietary Version), February 1978.

14.

Letter from T M Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, letter number NS-TMA-1830, June, 1978.

15.

Letter from T M Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, letter number NS-TMA-1834, June 20, 1978.

16.

" Safety Evaluation Report on ECCS Evaluation Model for Westinghouse Two-Loop Plants", November, 1977.

17.

Letter from L 0 Mayer, (NSP) to Director of Nuclear Reactor Regulation (NRC), February 24, 1978.

18.

Letter f rom D K Davis (NRC) to L 0 Mayer (NSP), " Request for Additional Information in regard to January 16, 19 78 letters from the two-loop plant owner eperators", February 10, 1978.

19.

Letter from L 0 Mayer (NSP) to Director of Nuclear Reactor Regulation (NRC), "ECCS Evaluation Model", March 17, 1978.

20.

" Safety Evaluation Report on Interim ECCS Evaluation Model for Westinghouse Two-Loop Plants", March, 1978.

4.0 TABLES List of Tables

  1. E" Table 1

Large Break - Time Sequence of Events 8

2 Large Break - Analysis input and Results 9

3 Dry Contair. ment Data 10-11 4

Reflood Mass and Energy Releases 12 5

Broken Loop Accumulator Mass and 13 Energy Release

TABLE 1 LARGE BREAK - TIME SEQUENCE OF EVENTS EVENT OCCURRENCE TIME (qFCO*n91 PZCLC, C = 0.4 D

Accident Initiation 0.0 Reactor Trip Signal 0.55 Safety Injection Signal

.65 Start Accumulator Injection 10.0 End of ECC Bypass 19.85 End of Blowdown 19.85 Start Pumped ECC Injection 25.65 Bottom of Core Recovery 32.7 Accumulators Empty 42.1

TABLE 2 LARGE BREAK - ANALYSIS INPUT AND RESULTS Quantities in the calculations:

Licensed core power rating 102% of 1650 MWt Total core peaking factor 2.28 Peak linear power 102% of 14.12 kw/ft Accumulator water volume 1250 cubic feet per tank Accumulator pressure 700 PSIA Number of Safety Injection Pumps Operating 2

Steam Generator Tube Plugging Level 1

percent (uniform)

)

Fuel Parameters -

Cycle Generic Region Generic Results DECLG, CD" Peak clad temperature ( F) 2179 Location (feet) 7.5 Maximum local clad / water reaction (%)

7.8 Location (feet) 7.5 Total core clad / water reaction (%)

<0. 3 Hot rod burst time (seconds) 25.8 Location (feet) 5.75 TABLE 3 DRY CONTAINMENT DATA 6

3 NET FREE VOLUME 1.37 x 10 ft INITIAL CONDITIONS Pressure 14.7 psia Temperature 90 F RWST Temperature 70 F Service Water Temperature 32 F Outside Temperature

-20 F SPRAY SYSTEM Number of Pumps Operating 2

Runout Flow Rate 1600 gpm/each Actuation Time 15 see SAFEGUARDS FAN COOLERS Number of Fan Coolers Operating 4

Fastest Post-Accident Initiation of Fan Coolers 15 sec STRUCTURAL llEAT SINKS Thickness (in.)

Area (ft )

1.5 steel 41300 0.75 steel 32000 0.25 steel 12 concrete 7860 0.375 steel 6800 0.25 steel 32000 0.5 steel 44000 0.145 stcel 1695 0.09 steel 12400 0.1 steel 6000 0.1875 steel 35125 1.44 steel 2200 12 concrete 40800 6 concrete 25070 3 concrete 7570 TABLE 3 (cont'd)

PAINTED SURFACES Steel Minimum Paintgd Paint Thickness (in)

Area (Ft )

Thickness (mils) 1.5 41300 11

.75 32000 11

.5 44000 11

.375 6800 11

.1875 13125 11

.145 1695 11 1.44 2200 11 Concrete Minimum Painted p,

Thickness (in)

Area (Ft2)

Thickness (mils) 12 4080 18 6

25070 18 3

7570 18

. TABLE 4 Reflood flass and Energy Releases Time (sec)

(lb/sec) f"o[9#(BTU /sec) lass 33.3

.002 2.27 37.9 34.2 44140.

46.9 76.8 73533.

60.9 216.1 103410.

78.5 236.1 103034.

98.5 241.8 100172.

120.6 245.8 97408.

171.1 253.3 91544.

232.5 261.2 85339.

320.5 277.5 79575.

TABLE 5 Broken Loop Accumulator Mass and Energy Release Mass Energy Time (sec)

Flow (1b/ ec)

Flow (

sec) s 0.0 4876.

292045.

2.0

4248, 254419, 4.0 3822.

228897.

6.0 3507.

210036.

8.0 3257.

195040.

10.0 3051.

182751.

12.0 2882.

172584.

14.0 2737.

163933.

16.0 2615.

156588.

18.0 2512.

150428.

20.0

2431, 145610.

22.0 2352.

140868.

24.0 2274.

136165 25.6 2219.

132899.

5.0 FIGURES The list below summarizes the figures contained in this section with title and page number.

List of Figures Page 1.

Quality of Fluid 15 2.

Mass Velocity 16 3.

Heat Transfer Coefficient 17 4.

Core Pressure (Top) lc 5.

Break Flow 19 6.

Core Pressure Drop 20 7.

Hot Rod Clad Average Temperature 21 8.

Fluid Temperature 22 9.

Core Flow Rate (Top) 23 10.

Water Level (Downcomer and Core) 24 11.

Flood Rate 25 12.

Accumulator Flow 26 13.

Pumped Safety Injection Flow Rate 27 14.

Containment Pressure 28 15.

Core Power 29 16.

Break Energy 30 17.

Condensing Wall Heat Transfer Coefficient 31

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INTRODUCTION For a given parameter, the generie table and figure numbers ar.: th e same as in the Prairie Island analysis.

In this report, the figures are labeled A for 1.0DECLG, B for 0.6DECLG, and C for 0.4DECLC.

Figures 13 A, B, C, 14 A, B, C and 17 were redrawn from the graphs supplied by Westinghouse in order to provide a more legible figure for review.

A-2

4 TABLES

TABLE 1 LARGE BREAK - TIME SEQUENCE OF EVENTS OCCURENCE TIME (SECONDS)

EVENT 0.6 DECLG, CD= 0.4 1.0 DECLG, C DECLG, CD=

=

D Accident Initiation 0.0 0.0 0.0 0.52 0.53 0.54 Reactor Trip Signal 0.48 0.56 0.67 Safety Injection Signal Start Accumulator Injection 5.7 7.5 9.8 15.15 16.75 19.96 End of ECC Bypass End of Blowdown 15.15 16.75 19.96 28.6 30.0 32.8 Bottom of Core Recovery 37.5 39.1 42.1 Accumulators Empty 25.48 25.56 25.6 Start Pumped ECC Injection T

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TABLE E LARGE BREAK - ANALYSIS INPUT AND RESULTS f

Quantities in the calculations:

i Licensed core power rating 102% of 1650 MWt 2.31 Total core peaking factor Peak linear power 102% of 14 3 kw/ft Accumulator water volume 1250 cubic feet per tank Accumulator pressure 700 PSIA 2

Number of Safety Injection Pumps Operating I

percent (uniform)

Steam Generator Tube Plugging Level Fuel Parameters -

Cycle Generic Region Generic Resul t; DECLG, CD=

1.0 DECLG,CD=

0.6 DECLG, CD=

0.4 Peak clad temperature (OF) 1936 1964 219:(_

Location (feet) 7.5 7.5 7.5 Maximum local clad / water reaction (%)

4.1 4.5 9.1 Location (feet) 7.5 7.5 7.5 5-Total core clad / water reaction (%)

<0.3

<0.3

<0.3 78.0 70.6 25.6 H6t rod burst time (seconds)

Location (feet) 7.0 6.75 5.75

TABLE 3 DRY C0tTAlf1MEtJT DATA 6

3 flET FREE VOLUME 1.37 x 10 ft IlilTIAL CO: DITI0riS Pressure 14.7 psia 0

Temperature 90 F RWST Temperature 700 F 0 f Service Water Temperature 32 Outside Temperature

- 200 F SPRAY SYSTEM fiumber of Pumps Operating 2

Runout Flow Rate 1600 gpm/each Actuation Tine 15 sec SAFEGUARDS FAft COOLERS J

tiumber of Fan Coolers Operating 4

Fastest Post-Accident Initiation of Fan Coolers 15 sec STRUCTURAL HEAT SI:tKS 2

Thickness (in.)

Area (ft )

1.5 steel 41300 0.75 steel 32000 0.25 steel 12 concrete 7860 0.375 steel 6800 0.25 steel 32000 0.5 steel 44000 1

0.145 steel 1695 0.09 steel 12400 0.1 steel 6000 0.1875 steel 35125 1.44 steel 2200 12 concrete 40800 6 concrete 25070 3 concrete 7570 A-5

TABLE 3 (cont'd)

PAI.'1TED SURFACES Steel Minimum Painted Paint Thickness (in)

Area (ft2)

Thickness (mil s) 1.5 41300 11

.75 32000 11

.5 44000 11

.375 6800 11

.1875 13125 11

.145 i 1695 11 1.44 2200 11 Concrete Minimum Painted Paint Thickness (in)

Area (ft )

Thickness (mils) 2 12 4080 18 6

25070 18 3

7570 18 A-6

i TABLE 4 Reflood Mass and Energy Release, C, = 0.4 DE D

Mass (lb/sec)

Energy (BTll/sec)

Time fl 0W Flow 2.43

.002 33.4 38.0 34.16 44025.1 47.0 137.97 85563.1 61.3 225.22 102143.2 79.2 237.1 100581.8 99.5 242.07 98133.1 121.7 245.93 95368.0 171.6 252.7 89344.0 230.0 259.35 82964.7 300.6 266.46 76280.0 395.7 276.64 69442.6 A-7

TABLE 5 Broken loop Accumulator 11 ass and Energy Release Tine toass Energy Flow (lb/sec)

Flo.v (BTU /sec) 0.0 4876.4 292045.

2.0 4249.

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