ML20149M736

Design Basis Accident Containment Pressure & Temp Response for USAR Update

ML20149M736
Person / Time
Site:	Monticello
Issue date:	12/31/1994
From:	Mintz S, Ranganath S GENERAL ELECTRIC CO.
To:
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ML20149M719	List: ML20149M717 ML20149M722 ML20149M728 ML20149M736
References
NEDO-32418, NUDOCS 9701270134
Download: ML20149M736 (28)

Text

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. s Exhibit D Monticello Nuclear Generating Plant i

j License Amendment Reauest Dated January 23,1997 i General Electric Report NEDO-32418, December,1994 Monticello Nuclear Generation PIar,t Design Basis Accident Cont &hment Pressure and Temperature Response for USAR Update 1

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l NSP Errata: Paae 3-2, 2nd paraaraoh, first line {

! I l Change to read, "The wetwell expenences its maximum temperature later in the l l transient.. ."

l l

1 b?

' D-1 9701270134 970123 PDR ADOCK 05000263:

P PDR l

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GENuclear Energy I75 Cunner Avenue ,

San Jose. CA 95125 NEDO-32418 CLASSI DRF T23-00723 DECEhBER 1994 Monticello Nuclear Generating Plant Design Basis Accident Containment Pressure and Temperature Response for USAR Update Prepared by: M/[M S. Mintz #

Plant Upgrade Projects 4

Approved by: "M M d S. Ranganath" Projects Manager Enginming & Licensing Consulting Services Projects s

v

NEDO-32418 IMPORTANT INFORMATION REGARDING CONTENTS OF TIIIS REPORT The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between Northern States Power Company (NSP) and GE, as identified in Purchase Order Number PG2796SQ SPLM 3755 as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than NSP, or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy or usefulness of the information contained in this document, or that it use may not infringe privately owned rights.

1 i

, NEDO-32418 ABSTRACT This report provides the results for an analysis of the Monticello containment response during -

a design basis loss-of-coolant accident (DBA-LOCA) to update the analytical design basis of the Monticello DBA-LOCA containment pressure and temperature response. The results of the containment pressure and temperature response analyses described in this report can be '

used to update the containment analyses in Section 5.2 of the USAR.

The DBA-LOCA containment response is determined for the limiting single active failure, which is the loss of one standby diesel generator.

f) il

, NEDO-32418 TABLE OF CONTENTS ABSTRACT ii

1.0 INTRODUCTION

1-1 1.1 Background 1-1 1.2 Models 1-1 1.3 Additional Analyses 1-2 2.0 ASSUMPTIONS 2-1 3.0 CONTAINMENT PRESSURE AND TEMPERATURE RESPONSE 3-1 3.1 DBA-LOCA Long-Term Pressure and Temperature 3-1 3.2 DBA-LOCA Short-Term Pressure and Temperature 3-2

4.0 REFERENCES

4-1 APPENDICES A. Additional Analyses of DBA-LOCA A-1 B. Core Heat Data B-1 l

iii

NEDO-32418

1.0 INTRODUCTION

1.1 Background

The purpose of the analyses described in this report is to provide an updated analytical design basis for the Monticello DBA-LOCA containment pressure and temperature response. The results of these analyses can be used to update the licensing basis documents, including the USAR.

GE previously provided containment pressure and temperature curves for a DBA-LOCA in Reference 1. The results ofReference I were used in Section 5.2 of the Monticello USAR to show the containment pressure and temperature response for the design basis loss-of-coolant accident (DBA-LOCA). The analyses in Reference 1 used input assumptions that were s

consistent with those used for the Mark I Containment Long-Term Program (LTP). One assumption used in Reference 1 is the single active failure of one residual heat removal (RHR) cooling loop. With this assumption, one low pressure core spray (LPCS) pump is available for long-term core inventory makeup and one RHR loop is available for long-term containment cooling with one RHR heat exchanger, two RHR pumps and two sersice water (SW) pumps.

According to Northern States Power (NSP), the limiting single active failure is the loss of one standby diesel generator, which can reduce the number of pumps available for core makeup or containment cooling to three pumps. This will result in the availability of only one RHR pump and one SW pump for long-term post-LOCA containment cooling. The analysis in this report re-calculates the Monticello DBA-LOCA containment pressure and temperature response with the assumption of a single active failuie of one standby diesel generator.

1.2 Models The same models used for the Reference 1 analysis are used to perform the analysis described in this report. This includes the use of the vessel LOCA blowdown and short-term containment response model described in References 2, 3 and 4 and the LOCA long-term containment response model of Reference 5. For the long-term analysis of the DBA-LOCA, the ANS/ANS 5.1 decay heat model is used to calculate the fission product decay heat. This decay heat was added to the fission power for delayed neutron after scram, metal-water 1-1

NEDO-32418 reaction energy and fuel relaxation energy to determine the total core power after scram.

Other input values such as initial suppression pool volume and initial suppression pool temperature were selected to maximize the DBA-LOCA long-term suppression pool temperature response.

1.3 Additional Analyses The analysis of the long-term DBA-LOCA uses the minimum Technical Specification (TS) value of the suppression pool volume and vent submergence to maximize the long-term suppression pool temperature. Since these assumptions minmuze the peak DBA-LOCA drywell pressure, an additional analysis of the shon-term DBA-LOCA containment response was performed with maximum TS values of suppression pool volume and submergence. This additional analysis is used to determine the peak drywell pressure during a DBA-LOCA.

Sensitivity studies included as Appendix A to this report show (1) the effect of using a nominal value of the initial suppression pool water level on peak DBA-LOCA drywell pressure and suppression pool temperature and (2) the effect of using May-Witt decay heat on the peak suppression pool temperature.

1-2

, NEDO-32418 2.0 ASSUMPTIONS Input assumptic:ns for the DBA-LOCA analysis maintain the overall conservatism in 11 e evaluation by maximizing the suppression pool temperature for the long-term DBA-LOCA analysis and by maximizing the peak drywell pressure for the short-term DBA-LOCA analysis.

Table 2-1 provides values of key containment parameters (confirmed by NSP in Reference 6) used in the DBA-LOCA analysis. The core shutdown power used in the analyses is provided in Appendix B. The following key input assumptions are used in performing the Monticello containment DBA-LOCA pressure and temperature response analysis.

1. Normal operation of the plant is assumed except for the single active failure, which is the loss of a single standby diesel generator. This makes one RHR loop inoperative and reduces the number of pumps available for either core injection or pool cooling to three for any pump combination.

2. The reactor is assumed to be operating at is2% of the rated thermal power.

3. Vessel blowdown flow rates are based on the Homogeneous Equilibrium Model(Reference 2).

4. The core decay heat is based on ANSI /ANS-5.1-1979 decay heat (Reference 7). [For Cases 2 and A.2 in Appendix A, May-Witt decay heat (Reference 8) is used. ]

5. For the long-term DBA-LOCA analysis, the portion of the feedwater which maxunizes the suppression pool temperature is injected into the vessel.

6. Thermodynamic equilibrium exists between the liquids and gases in the drywell and between the suppression pool and the suppression chamber airspace to maximize long-term pressure.

2-1

NEDO-32418

7. The instantaneous guillotine break of the recirculation suction line is used as the worst case break.

8. The RHR intertie is present during the break and is included in the DBA-LOCA break size calculation.

9. For the long-term DBA-LOCA analysis, the initial suppression pool volume is at the minimum Technical Specification (TS) limit to maximize the calculated suppression pool temperature.

The corresponding subme.gence is 3.0 ft.

10. For the short-term DBA-LOCA analysis, the initial suppression pool volume is at the maximum TS limit to maximize the calculated peak drywell pressure. The corresponding submergence is 3.58 ft.

11. The drywell airspace is initially at the nominal temperature of 135'F and 20% relative humidity.

12. The wetwell airspace is initially equal to the initial suppression pool temperature of 90 F and at the maximum relative humidity of100%

13. The service water temperature is at the maximum value 90 F to maximize the suppression pool temperature.

14. The initial drywell and suppression chamber pressure are at the maximum expected operating values to maximize the containment pressure.

15. The initial suppression pool temperature is at the maximum TS value (90 F) to maximize the calculated suppression pool temperature.

16. The Emergency-Core-Cooling System (ECCS) starts injecting at 30 seconds.

2-2

NEDO-32418

17. For the first 600 seconds, ECCS consists of one core spray pump and two LPCI pumps. After 600 seconds ECCS consists ofone core spray pump.

18. At 600 seconds suppression pool cooling is initiated. This is accomplished by switching one LPCI pump to suppression pool cooling and by turning off one LPCI pump and initiating one SW pump for suppression pool cooling.

19. Pool cooling consists of one RHR pool cooling loop with one LPCI pump and one SW pump.

20. Passive heat sinks in the drywell, suppression chamber airspace and suppression pool are conservatively neglected to maximize the suppression pool temperature. Heat transfer from the primary containment to the reactor building is also conservatively neglected.

21. Drywell fan coolers are inactive (non-safety equipment).

22. Control Rod drive flow is zero (non-safety equipment).

23. All core spray and LPCI/ RHR pumps have 100% of their horsepower rating converted to a pump heat input which is added either to the RPV liquid or suppression pool water.

. NEDO-32418 TABLE 2-1 INPUT PARAMETERS FOR CONTAINMENT ANALYSIS Value Used Parameter Units In Analysis Core Thermal Power MWt 1703 (102% or Rated Thermal Power)

Vessel Dome Pressure psig 1025 DBA-LOCA Break Area ft2 4.095 (includes RHR Intertie flow area)

Drywell Free (Airspace) Volume ft3 134,200 (including vent system)

Initial Suppression Chamber Free (Airspace) Volume Low Water Level (LWL) fl3 108,250 High Water Level (HWL) ft3 103,340 Initial Suppression Pool Volume Min. Water Level f13 68,000 Max WaterLevel ft3 72,910 No. of Downcomers 96 Total Downcomer Flow Area ft2 289.65 Initial Downcomer Submergence Min. Water Level ft 3.0 Max Water Level ft 3.58 Downcomer I.D. ft 1.96 Vent System Flow Path Loss Coefficient (includes exit loss) 5.17 Supp. Chamber (Toms) Major Radius ft 49.0 Supp. Chamber (Torus) Minor Radius ft 13.88 2-4

r NEDO-32418 TABLE 2-1 INPUT PARAhETERS FOR CONTAINhENT ANALYSIS (Continued)

Value Used Parameter ILnits in Analysis Suppression Chamber-to-Drywell Vacuum Breaker Opening Diff. Press.

- start psid 0.096

- full open psid 0.242 Supp. Chamber-to-Drywell Vacuum Breaker Valve Opening Time sec 1.0  ;

Supp. Chamber-to-Drywell Vacuum Breaker Flow Area (per valve ft2 1.65 System)

Supp. Chamber-to-Drywell Vacuum Breaker Flow Loss CoefEcient (including exit loss) 3.804 No. of Supp. Chamber-to-Drywell Vacuum Breaker Valve Systems (2 valves per assembly) 8 LPCI Pump Flow in Vessel gpm 7740 (2 pumps)

Injection Mode (< 600 sec)

Core Spray Pump Flow in Vessel gpm 270n (1 pump)

LPCI Pump Flow in Suppression gpm 4000 (1 pump)

Pool Cooling Mode (> 600 sec)

Service Water Flow to RHR gpm 3500 (1 pump)

Heat Exchanger LPCI/ Containment Cooling Heat Exchanger Kin Suppression Pool Cooling Mode (1 RHR /1 SW pump) Btu /sec- F 143.1 LPCI/ Containment Cooling Service Water Temperature F 90 LPCI/ Containment Cooling Pump Heat (per pump) hp 600 Core Spray Pump Heat (per pump) hp 800 2-5 l

NEDO-32418 TABLE 2-1 INPUT PARAMETERS FOR CONTAINMENT ANALYSIS (Continued)

Value Used Parameter LLnits in Analysis Time for Operator to Turn On Suppression Pool Cooling Mode (after LOCA signal) sec 600 Feedwater Addition (to RPV after start of event; mass and energy)*

Feedwater Mass Enthalpy Node (thm) (Btullbm) 1 39063 346.1 2 27344 308.1 3 74594 275.9 4 37361 201.4

• The feedwater table shown above gives the feedwater mass added and associated feedwater enthalpy. This table reflects feedwater temperature conditions in the feedwater train prior to the DBA-LOCA. Each node corresponds to a section of the feedwater train with feedwater at a lumped temperature. Only the portion of the feedwater in the feedwater train with a temperature higher than the peak suppression pool temperature was added.

, 2-6 l

1

NEDO-32418 3.0 CONTAINMENT PRESSURE AND TEMPERATURE RESPONSE -

3.1 DBA-LOCA Long-Term Pressure and Temperature (Case 1)

The long-term containment response to a postulated recirculation line beak is illustrated in Figures 3-1 to 3-3. To cover the peak pool temperature, the analyses are extended to one million seconds (about 11.6 days).

For this transient, the reactor quickly blows down due to the large break size. The vessel pressure drops too rapidly for the High Pressure Coolant Injection system to supply any makeup water. At the time of the break, it is assumed that the offsite AC power is lost with the limiting single active failure of one standby diesel generator. The Emergency Core Cooling System (ECCS) power must therefore be supplied by the remaining diesel generator.

This causes a delay of about 30 seconds before rated ECCS flow is available.

Figure 3-1 shows the expected pressure in the drywell and pressure suppression chamber (wetwell). After the vessel is flooded, ECCS flow cascades from the break and quenches the steam in the drywell. The wetwell-to-drywell vacuum breakers open and allow air to return to .

the drywell. At approximately 114 seconds into the accident, the pressures in the two chambers are essentially equal and remain equal for the remainder of the transiem.

The drywell temperature response is plotted in Figure 3-2. Figure 3-3 gives the suppression pool temperature response. The wetwell airspace is assumed to be in thermal equilibrium with the suppression pool.

Table 3-1 lists the maximum pressures and temperatures expected in the containment for this accident. For the drywell compartment, the maxunum pressure and temperature occur early in the transient during the blowdown phase. At this time, the inertial effects of the flow in the vent system cause the drywell to remain at a much greater pressure than the wetwell. These high pressures are quickly dissipated as shown in Figure 3-1.

It should be noted that the value of the peak drywell pressure in Table 3-1 for Case 1 is slightly lower than that found in NEDO-30477 (Reference 7). This is because the input assumptions used for Case 1 (including the use of minimum water suppression pool water j volume) were to maxunize the peak suppression pool temperature, while the purpose of 4

3- 1

NEDO-32418 NEDO-30477 is to maximize the peak drywell nressure. The effect of using maximum water level on peak drywell pressure is discussed in Sei ta 3.2.

The wetwell experiences its maximum pressure and temperature later in the transient (at approximately 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />). The suppression pool continues to heat up due to decay heat from the reactor. The heat is transferred to the pool by the steam and ECCS water coming out of the reactor and flowing through the vent system. Eventually, the Residual Heat Removal (RHR) System is able to tum the transient around and remove the heat from the pool faster than it is added. This turn-around point corresponds to the maximum suppression pool temperature.

One option that the operator has is to align the RHR System in the containment spray mode.

This would quench the steam in the containment airspace and rapidly drop the temperature and pressure. It is conservative to neglect this option when maximizing the containment pressure response.

3.2 DBA-LOCA Short-Term Pressure and Temperature (Case 2)

Figures 3-4 and 3-5 show the short-term (here defmed as 0- 30 seconds) DBA-LOCA pressure and temperature response, assuming the suppression pool level and vent submergence are initially at the TS value for high water level (HWL).

With the assumption of HWL, the pressure required to clear the water initially in the submerged portion of the downcomer increases. This results in an increase in the drywell pressure response. Use of HWL for determining peak drywell pressure is consistent with the assumptions used during the Mark I Containment Long-Term Program (Reference 9) and the input assumptions for the containment analysis of Reference 10. May-Witt decay heat is also assumed for this case, to be consistent with the analyses of References 9 and 10.

The results for Case 2 are summarized in Table 3-1. An increase in the peak drywell pressure of 0.3 psi relative to the value for Case 1 is indicated due to the higher initial water level and submergence.

3- 2

NEDO-32418 TABLE 3-1 MONTICELLO MAXIMUM CONTAINMENT CONDITIONS FOR ADESIGNBASIS ACCIDENT Case 1 - DBA-LOCA Long-Term Containment Pressure and Temperature Maximum Drywell Pressure = 41.9 psig at 1.2 see Maximum Drywell Temperature = 282 F at 2.6 see Maximum Wetwell Pressure = 30.6 psig at 1080 see Maximum Wetwell and Suppression Pool Temperature = 184.0 F at 31200 sec Case 2 - DBA-LOCA Shon-Term Containment Pressure and Temperature Maximum Drywell Pressure = 42.2 psig at 1.2 see Maximum Drywell Temperature = 282 F st 2.6 sec 3- 3

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. NEDO-32418 450.

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NEDO-32418

4.0 REFERENCES

1) NEDO-30485, "Monticello Design Basis Accident Containment Pressure and Temperature Response for FSAR Update," December 1983.

2) NEDO-21052, " Maximum Discharge Rate of Liquid-Vapor Mixtures from Vessels,"

General Electric Company, September 1975.

3) NEDO-10320, "The GE Pressure Suppression Contaimnent System Analytical Model," April 1971.

4) NEDO-2053'a , "The General Electric Mark III Pressure Suppression Containment System Analytical Model," June 1974.

5) NEDO-20533-1, "The General Electric Mark III Suppression Containment System Analytical Model Supplement 1," September 1975.

6) Letter, S. Shirey (NSP) to S. b atz (GE), " Input Parameters for Monticello Containment Analysis to Update 11 Long-Term DBA-LOCA Containment Pressure and Temperature Curves of NEDO-30485," October 3,1994.

7) " Decay Heat Power in Light Water Reactors," ANSI /ANS - 5.1 - 1979, Approved by American National Standards Institute, August 29,1979.

8) NEDO-10625, " Power Generation in a BWR Following Normal Shutdown or Loss-of-Coolant Accident Conditions," March 1973.

9) NEDO-24576, Rev.1, " Mark I Containment Program Plant Unique Load Definit.on, Monticello Nuclear Power Plant,' October 1981.

10) NEDO-30477, " Safety Analysis of the RHR Intertie Line, Monticello Nuclear Generating Plant, " 1983 4-1

g *

• 2 NEDO-32418 APPENDICES A. ADDITIONAL ANALYSES OF DBA-LOCA B. CORE HEAT DATA d

s'

p ._ _ _ - __ _ _ . - .

y,...

~

NEDO-32418 i

APPENDIX A ADDITIONAL ANALYSES OF DBA-LOCA i

The results of sensitivity analyses of the DBA-LOCA are presented here. Two cases are included which determined: (1) the DBA-LOCA long-term containment response with the suppression pool initially at normal water level (NWL) and (2) the DBA-LOCA long-term l 1

containment response with the use ofMay-Witt decay heat.

Table A-1 summarizes the key inputs and analyses results for Cases A.1 and A.2.

i The results of Case A.1 show that an increase in the initial water level to NWL produces a slight increase in the peak drywell pressure and a slight decrease in the peak suppression pool temperature.

The results of Case A.2 quantify the effect ofusing May-Witt instead of ANS 5.1. The higher f decay heat with May-Witt relative to ANS 5 increases the integrated energy deposited to the suppression pool, which results in a higher peak suppression pool temperature. Use of May-Witt decay heat instead of ANS 5.1 decay heat has negligible impact on the DBA-LOCA' short-term peak drywell pressure and temperature. ,

t I

i A-1 l

\

NEDO-32418 TABLE A-1

SUMMARY

OF DBA-LOCA ADDITIONAL CASE RESULTS Parameter Case A.1 Case A.2 Case 1 (From Table 2)

Duration 0. - 1.0 E06 0. - 1.0 E06 0. - 1.0 E-6 (sec)

Suppression Pool NWL LWL LWL Water Level Decay Heat ANS 5.1 May-Witt ANS 5.1 Peak DW 42.0 41.9 41.9 Pressure (psig)

Peak DW Temperature 282 282 282 (F)

Peak Suppression 182 196 184 Pool Temperature ( F)

Peak W W 29.9 32.1 30.6 Pressure (psig) l I

1 l

l A-2

.a * ,

NEDO-32418 APPENDIX B CORE DECAY HEAT DATA Table B-1 provides the core heat (Btu /sec) based on the May-Witt (Reference B.1) decay heat model. The core heat includes decay heat (May-Witt), metal-water reaction energy, fission power and fuel relaxation energy. The core heat in Table B-1 is normalized to the initial core thermal power of 1703 MWt.

Table B-2 provides the core heat (Btu /sec) based on the ANS 5.1 (Reference B.2) decay hea*

model. The core heat includes decay heat (ANS 5.1-1979), metal-water reaction energy, fission power and fuel relaxation energy. The core heat in Table B-2 is normalized to the initial core thermal power of 1703 MWt.

References:

B.1) NEDO-10625, " Power Generation in a BWR Following Normal Shutdown or Loss-Of-Coolant Accident Conditions," March 1973.

B.2) " Decay Heat Power in Light Water Reactors," ANSI /ANS-5.1 - 1979, Approved by American National Standards Institute, August 29,1979.

1 B-1 l

l

NEDO-32418 TABLE B-1 CORE HEAT MAY-WITT Time (sec) Core Heat *

0. 1.002 0.1 1.007 0.2 0.9658 0.6 0.7111 0.8 0.6521 1.0 0.5328 2.0 0.4866 4.0 0.5477 6.0 0.5681 8.0 0.5391

10. 0.4825

20. 0.2069

40. 0.05693

60. 0.044

80. 0.0413 100. 0.03993 200. 0.03365 400. 0.02827 600. 0.02549 800. 0.02365 1000. 0.0223 2000. 0.0184 4000. 0.0151 6000. 0.0135 8000. 0.0126 10000. 0.0120 20000. 0.0101 40000. 0.008125 lE5 0.006245 2E5 0.005126 3E5 0.004096 4E5 0.003596 8E5 0.003196 1E6 0.002985 1E8 0.002985

• Core Heat (normalized to the initial core thermal power of 1703 MWt)

= decay heat + fission power + fuel relaxation energy + metal-water reaction energy B-2 W

,U J . .

NEDO-32418 TABLE B.2 CORE HEAT ANS 5.1 l Time (sec) Core Heat *

0. 1.006

1. 0.5634

4. 0.5319

10. 0.3479

20. 0.1092

40. 0.0563

60. 0.04050

80. 0.0385 120. 0.0363 120. *

• 0.0303 200. 0.0274 400. 0.0241 600. 0.0221 1000. 0.0196 2000, 0.0160 4000. 0.0127 6000. 0.0112 8000. 0.0103 10000. 0.00972 14400. 0.00928 18000. 0.00881 20000. 0.00859 28800. 0.00788 36000. 0.00748 60000. 0.00658 IES 0.00572 4E5 0.00353 8E5 0.00261 1E6 0.00237 2E6 0.00175

• Core Heat (normalized to the initial core thermal power of 1703 MWt)

= decay heat + fission power + fuel relaxation energy + metal-water reaction energy

" Metal-water reaction heat is assumed to end at 120 seconds.

B-3

- .. .. ~ .-

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. NEDO-32418 l

DISTRIBUTION LIST Northern States Power Company (10)

(c/o C. N. Gallt, M/C HME)

MLC J. Casillas 747 C. N. Gallt HME S. Mintz (10) 172 D. C. Pappone 172 S. Ranganath 747 P. T. Tran 172 C. T. Young 172 GENE Library (2) 528 l

i

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