ML17276A449

Suppression Pool Temp Analysis.

ML17276A449
Person / Time
Site:	Columbia
Issue date:	10/31/1981
From:	Oh I, Quinn G, Eric Thomas STONE & WEBSTER, INC.
To:
Shared Package
ML17276A445	List: ML17276A438 ML17276A449
References
14057-U(D)-1, NUDOCS 8112180291
Download: ML17276A449 (51)

Text

14057-U(D)-1 SUPPRESSION POOL TEMPERATURE ANALYSIS PREPARED FOR WASHINGTON PUBLIC POWER SUPPLY SYSTEM By E. A. Thomas October 1981 Approved by: G. T. Quinn I. Oh Project Manage Coordinator, Nuclear Technology Division E. R. Scott Engineering Manage ent Sponsor Copyright 1981 Stone & Webster Engineering Corporation Denver, Colorado 80217 8llggg0~ ~'>>~~s

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SUMMARY

Calculations have been 'performed for six cases for MPPSS Nuclear Project No.

2 as required by Contract C-0699 using the CONTORT code.

The results for all of the requested cases are presented in the following table. The suppression pool temperature, reactor vessel pressure and coolant temperature transient response curves are shown in Figures 1 thr ough 18.

Peak Pool Case No. Accident Si le Failure 1a Stuck open safety 1 RHR system 190 Relief Valve (SORV)

SOR V Spurious isolation 178 2a Isolation/scram 1 RHR system 196 2b Isolation/scram SORV at scram 171 3b Small break accident Shutdown cooling ]7/

3c Small break accident Electrical division 198

TABLE OF CONTENTS P e Number Introduction Conclusions Analysis Method Discussion Assump tions Pertinent Input Parameters References Figures (1-12) 10

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INTRODUCTION The purpose of this analysis is to predict the transient temperature response of the WNP-2 suppression pool for six accident conditions selected by WPPSS.

CONCLUSIONS

1. The peak predicted bulk temperatures for four of the six cases investigated were equal to or less than the 190 guideline for maximum suppression pool bulk temperatures given by NUREG-0487. The other two cases (2A and 3C) fall within the limits of NUREG-0783. In applying the NUREG-0783 criteria, a maximum bulk temperature of 205oF (corresponding to a quencher submergence of 17 ft) is allowable.

2. The most important single failure is loss of one RHR system. The cases which yield the highest peak temperature all have one RHR system as the single failure.

ANALYSIS METHOD The CONTORT code was used for the analysis in accordance with the assumptions contained herein, which were agreed upon in the September 25, 1981 coordination meeting held in Stone E Webster (SWEC) offices at the Denver Operations Center.

CONTORT is designed to analyze the reactor and containment system transient during normal and abnormal shutdown of the reactor system. The program uses a finite difference technique using input specified time steps to solve the transient equations. In each time step, the program determines the mass and energy flow across all control volumes and performs thermodynamic state calculations for the reactor vessel and suppression pool assuming saturated equilibrium conditions.

Major output from CONTORT is the suppression pool temperatur e and reactor vessel pressure response.

DISCUSSION This analysis shows the strong dependence on RHR pool cooling. All of the cases in which only one RHR is available for pool cooling yield the highest pool temperatures.

Case 1a is strongly dependent on how long the main condenser is available as a heat sink.

For SORV cases, the pool was assumed to be heated from 90 F to 110 F before the operator manually scrams the reactor as required by the technical specifications. The initial pool temperature, therefore, was input as 110 F. The initial pool volume was adjusted for the mass of steam, transferred to the suppression pool through the stuck open relief valve, that is required to raise the pool temperature from 90 F to 110 F. Final pool mass and the time for the pool temperature to go from 90 F to 110 F were solved from two simultaneous equations. All of the computer runs start at t=O, the time of the scram.

'll This analysis was performed with the assumption that reactor water level is maintained by hot feedwater and CRD flow only. Feedwater is terminated and reactor core makeup is supplied by LPCI drawing suction from the suppression pool when either the enthalpy of the feedwater becomes less than the enthalpy of the pool, or the feedwater supply is exhausted, whichever comes first.

All other ECCS systems were not utilized.

The effective flow area of 0. 1 185 ft2 for SRV was determined in the following manner:

1 ~ SRV flow at rated reactor vessel pressure (1020 psia) and 122.5 percent of rated ASIDE flow is calculated.

2. The maximum steam mass flux, ibm/ft2 sec, (Hoody) assuming saturated steam at rated reactor vessel pressure (1020 psia) is determined from Reference 4.

3. The effective flow area of the safety relief valve is calculated by dividing the SRV flow rate by the Hoody mass flux.

An analysis was made by SWEC at the Cherry Hill, N. J. office to benchmark the CONTORT results to the G. E. Code HEX, and was reported to Pennsylvania Power & Light Company for Susquehanna Steam Electric Units 1 & 2 (Docket Nos.

50-387 and 50-388, respectively). The CONTORT results agree'very well with the HEX results, with CONTORT predicting peak temperature about 2 F lower than HEX.

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ASSUHPTIONS The following are assumptions which have been utilized in the analysis.

These assumptions are in agreement with Reference 2.

I. Assum tions For All Cases Reactor power at the time of the scram is 102 percent of rated core power .

2~ Service water temperature is maximum and identical to value used in FSAR for containment analysis.

3~ The initial supppression pool temperatur e is the maximum technical specification limit for power operation without pool cooling.

4. The initial suppression pool volume is the minimum technical specification limit for power operation.

5. The safety relief valve flow is assumed to be 122.5 percent of the ASHE rated flow rate.

6. The RHR pool cooling mode is initiated 10 minutes after exceeding the technical specification limit for continuous power operation, TS1 (90oF).

7~ Hain steam line isolation valve (HSIV) closure time is assumed to be 3 seconds following an 0.5 second delay.

The reactor cooldown rate for manual depressurization by safety relief valves (SRV) is 100 F/hr, and begins after the suppression pool temperature exceeds TS4 (120 F), unless the depressurization rate for the event itself (e.g. SORY) exceeds the required rate at that time. To maintain the 100 F/hr cooldown rate requir es the actuation from one to seventeen SRV's, depending on reactor vessel pressure.

9 The mass of water contained inside the reactor vessel pedestal is neglected.

10. The energy absorbed by the containment heat sinks is neglected.

Reactor water level is maintained by feedwater and CRD flow.

Feedwater flow is terminated when the enthalphy of the feedwater is less than or equal to the enthalpy of the suppression pool. The feedwater is then replaced by LPCI taking suction from the suppression pool.

12. The RHR pool cooling pumps have 100 percent of their horsepower rating added directly to the suppression pool.

13. The generic G. E. decay heat curve, which is combined with power coastdown, is used in the analysis.

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II. Case-S ecific Assum tions

a. Case 1a SORV at ower

- One RHR heat exchanger is lost.

- Shutdown cooling is not available.

- The safety r elief valve sticks open at a pool tempature of 90oF(TS1) .

- Manual scram occurs at a pool tempatur e of 110 F (TS3); Time 0 for the analysis.

- The turbine stop valves are closed 20 seconds after scram.

MSIV's do not close.

- One RHR is in pool cooling mode 10 minutes after high pool temperature alarm (90 F).

- The bypass flow to the main condenser is re-established 20 minutes after SORV accident with full bypass (25$ of rated steam flow) capability.

- The use of main condenser is terminated when steam jet air ejector (SJAE) design pressure is reached. Depressurization continued using SRV's for 100 F/hr manual depressurization.

b. Case 1b SORV at Power Two RHR heat exchangers are available for pool cooling.

- Shutdown cooling is not available.

- The SRV sticks open at a pool temperature of 90 F (TS1).

- The operator scrams reactor at a pool temperature of 110oF (TS3). Due to a spurious signal, main steam isolation valves closure is initiated, with 3.5 seconds closure time. Time = 0 for the analysis.

- Two RHR's are in the pool cooling mode 10 minutes after high pool temperature alarm (90 F).

- Manual depressurization is initiated at a cooldown rate of 100 F/hr when pool temperatur e is 120 F.

c. Case 2a Isolation/Scram

- One RHR heat exchanger is lost.

- Shutdown cooling is not available.

- Isolation/scram occurs at a pool, temperature of 90 F. Time = 0 for the analysis.

The non-mechanistic closure of the HSIV's is in 3.5 seconds.

- One RHR is in pool cooling mode 10 minutes after high pool temperature alarm (TS1=90 F).

- manual depressurization is initiated at a cooldown rate of 100 F/hr when pool temperature is 120 F.

d. Case 2b Isolation/Scram

- Two RHR heat exchangers are available for pool cooling.

- Shutdown cooling is not available.

- Isolation/scram occurs at a pool temperature of 90 F. Time =

0 for the analysis.

- The non-mechanistic closure of the HSIV's is in 3.5 seconds.

- Safety relief valve sticks open at the time of the scram and cannot be closed.

- Two RHR heat exhangers are in the pool cooling mode 10 minutes after the pool high temperature alarm (90 F).

- Hanual depressurization is initiated at a cooldown rate of 100oF/hr when pool temperature is 120oF, unless SORV gives a cooldown rate higher than 100 F/hr.

e. Case 3b Small Break Accident

- Two RHR heat exchangers are available for pool cooling.

- Shutdown cooling is not available.

- A small steamline rupture occurs at a pool temperature of 90 F. Time = 0 for analysis. Flow area from break = 0.01 ft2

- Scram at Time = 0 with non-mechanistic is 3.5 seconds MSIV closure time.

- Two RHR heat exchangers in the pool cooling mode 10 minutes after high pool temperature alarm (90 F).

- Hanual depressurization is initiated at a cooldown rate of 100 F/hr when pool temperatur e is 120oF.

- Both RHR's switch to LPCI mode when the reactor vessel pressure equals the RHR pump shutoff head (333 psia), with a 10-minute delay for the operator to convert manually back to pool cooling.

f. Case 3c Small Break Accident

- One RHR heat exchanger is available for pool cooling.

- Shutdown cooling is not available.

- A small steamline rupture occurs at a pool temperature of 90oF (Time = 0) Flow area from break = 0.01 ft2.

- The scram at Time = 0 on high drywell pressure.

- The isolation of MSIV's at Time = 0 with non-mechanistic 3.5 seconds closur e time.

- One RHR heat exchanger is in the pool cooling mode 10 minutes after the high pool temperature alarm.

Hanual depressurization is initiated at a cooldown rate of 100 F/hr when pool temperature e is 120 F.

- The RHR switches to LPCI mode when the reactor vessel pressure equals the RHR pump shutoff head (333 psia), with a 10-minute delay for the operator to conver t manually back to pool cooling.

PERTINENT INPUT PARAMETERS Initial reactor vessel pressur e 1,020 psia Initial reactor vessel liquid volume 13, 189 ft3 Total reactor vessel volume 23 625 ft3 Initial pool temperature (Isolation/scram, SBA) 90 F Initial pool temperature (SORV cases) 110oF Initial pool volume (Isolation/scram, SBA) 127,007 ft3 Initial pool volume (SORV cases) 129)774 ft3 Feedwater flow rate 3,997 ibm/sec CRD flow rate 16.8 ibm/sec CRD enthalpy 68 Btu/ibm Feedwater "on" volume 13,090 ft3 Feedwater "off" volume 13)290 ft3 Fee'dwater Inventor and Associated Enthal Mass Enthalpy (pipe 4 fluid)

~(ibm (Btu/ibm) 134,226 386 187, 826 326 370, 952 208 200,000 152 LPCI rated flow 7,450 gpm LPCI pump heat 566 Btu/sec RHR heat exchanger R. factor 289 Btu/sec F RHR pump heat per pump 566 Btu/sec Minimum reactor vessel pressure required for discharging flow through SRV (SORY, Isolation/scram) 22.8 psia Minimum reactor vessel pressure required for discharging flow through SRV (SBA cases) 20.7 psia Safety relief valve effective flow area 0. >>85 ft2 Safety relief valve open to close band width 50 psi Input for Automatic SRV's No. 0 en Pressure Close Pressure (psia) (psia) 1 >091 1,041 1, 101 1,051 11111 1,061 1, 121 1,071 1, 131 1,081

Reactor thermal power (102 percent of rated) 3,389 HW Hass of reactor vessel and internals 2.89 x 1061bm Specific heat of reactor vessel & internals 0.12 Btu/ibm oF Service water temperature 85oF Steam ejector design pressure (lower limit of RVP for discharge of steam to main=condenser) 140 psia RHR pump shutoff discharge pressure (LPCI injection valve pressure permissive to take RHR out of pool cooling) 333 psia

REFERENCES

1. Supply System Contract No. C-0699, Modification No. 1, Appendix A, "Statement of Work", Dated September 29, 1981.

2. "Assumptions for Use in Analyzing Mark II BWR Suppression Pool Temperature Response to Plant Transients Involving Safety/Relief Valve Discharge",

Rev. 1, December 1980 (White Paper Rev. 1).

3. ASME Steam Tables, Second Edition, 1967.

4. "Maximum Flow Rate of a Single Component, Two Phase Mixtur e", F. J. Moody, APED-4378, October 25, 1963.

5. Stone 4 Webster Computer Code CONTORT (containment and reactor vessel transient code), NU-163, Version 01, Level 00.

6. Pump Data Sheet, Residual Heat Removal Pump, Ingersoll-Rand Order No.

006-36045, Pump Serial Number 0473-111/112/113.

7. Letter from L. E. Ostrom to J. W. Yetter, dated October 13, 1981, "Supply System Contract No. C-0699, Plant Specific Data".

8. "Suppression Pool Temperature Limits for BWR Containments" USNRC NUREG-0783 (Draft).

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