ML17276A449
| 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: |
| 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 Project Manage I.
Oh Coordinator, Nuclear Technology Division E.
R. Scott Engineering Manage ent Sponsor Copyright 1981 Stone
& Webster Engineering Corporation
- Denver, Colorado 80217
"~~ ~so ~'>>~~s 8llggg0~
A CK 0500035,7 DR
~
g
~
)
~
'r
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.
Case No.
Accident Si le Failure Peak Pool 1a Stuck open safety Relief Valve (SORV) 1 RHR system 190 2a 2b 3b 3c SOR V Isolation/scram Isolation/scram Small break accident Small break accident Spurious isolation 1
RHR system SORV at scram Shutdown cooling Electrical division 178 196 171
]7/
198
TABLE OF CONTENTS P
e Number Introduction Conclusions Analysis Method Discussion Assump tions Pertinent Input Parameters References Figures (1-12) 10
P I
J
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/ft
- sec, (Hoody) assuming saturated 2
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.
N
, ~
I
~
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.
r
~
J
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
Initial reactor vessel liquid volume Total reactor vessel volume Initial pool temperature (Isolation/scram, SBA)
Initial pool temperature (SORV cases)
Initial pool volume (Isolation/scram, SBA)
Initial pool volume (SORV cases)
Feedwater flow rate CRD flow rate CRD enthalpy Feedwater "on" volume Feedwater "off" volume 1,020 psia 13, 189 ft3 23 625 ft3 90 F
110oF 127,007 ft3 129)774 ft3 3,997 ibm/sec 16.8 ibm/sec 68 Btu/ibm 13,090 ft3 13)290 ft3 Fee'dwater Inventor and Associated Enthal Mass
~(ibm 134,226 187, 826 370, 952 200,000 Enthalpy (pipe 4 fluid)
(Btu/ibm) 386 326 208 152 LPCI rated flow LPCI pump heat RHR heat exchanger R. factor RHR pump heat per pump Minimum reactor vessel pressure required for discharging flow through SRV (SORY, Isolation/scram)
Minimum reactor vessel pressure required for discharging flow through SRV (SBA cases)
Safety relief valve effective flow area Safety relief valve open to close band width 7,450 gpm 566 Btu/sec 289 Btu/sec F
566 Btu/sec 22.8 psia 20.7 psia
- 0. >>85 ft2 50 psi Input for Automatic SRV's No.
0 en Pressure (psia) 1 >091 1, 101 11111 1, 121 1, 131 Close Pressure (psia) 1,041 1,051 1,061 1,071 1,081
Reactor thermal power (102 percent of rated)
Hass of reactor vessel and internals Specific heat of reactor vessel
& internals Service water temperature Steam ejector design pressure (lower limit of RVP for discharge of steam to main=condenser)
RHR pump shutoff discharge pressure (LPCI injection valve pressure permissive to take RHR out of pool cooling) 3,389 HW 2.89 x 1061bm 0.12 Btu/ibm oF 85oF 140 psia 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 Mixture", 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).
9
C)
M 10 10 HOURS 10 10 CL 0
Z f
IIJ I
G n
CL G=42 1 m/ft sec 2
C3 CI (0
UJ CL CL (Q
G=94 ibm/ t sec 2
cD '0 10 10 TINE AFTER SCRRN SECONDS SUPPRESSION POOL TEMP. ANALYSIS OWNERS GROUP CASE iA SORV HRSHIHOIOH PUSLIC POKER SUPPLY SYSIEH HHP-2
a 4,
l'I I
I 'l T
10 10 HOURS 10 oo CO Q
o CD Co CO UJ lY o
CD UJ CO CO o
QJ o
)
C3 oo CC M
UJ 10 10 TINE AFTER SCRAH SECONDS SUPPRESSION POOL TEMP RNRLYSIS
.ONNER AS GROUP CRSE IR SORV HASHIHOTOH PUSL1C POHER SUPPLY SYSTHH HHP-R
DD 8) 10 10 HOURS 10 10 10 Q
D LIJ lQ I
D z
u)
CD D
C3 UJ D
(Q ID 0D W
(oD CD I
D N
DD 01 04 0s TIME RFTER SCRAM SECONDS SlJPPRESSION POOL TEMP. ANALYSIS ONNER'S GROUP CRSE 1A SORY HIISHLHOTSH PUBLiC PSHER SUPPLY SYSIEH HHP-2 F1.g.
3
~ l I
I
10 10 HOURS 10 10 0
o CD cn C7 (0
UJ IK Q
Q O
(0 10 TINE AFTER SCRAM SECONDS 10 SUPPRESSION POOL TEMP - ANALYSIS ONNER'S GROUP CASE IB SORV M4SHIHOTUH PUSLTC POMEM SUPPLY SYSTEM MHP-2
I
O CD OJ 10 10 HOURS 10 10 CDO CO CL UJ O
CO CO CO UJ o
CO UJ CO CO UJ CL tD I
O UJ 5!
CL 10 10 10 10 10 TINE FIFTER SCRRN SECONDS SUPPRESSION POOL TEMP. ANALYSIS ONNER'S GROUP CASE IB SORV NASKlKOTON PUSLTC PSNER SUPPLT STSTEK NKP-I Flg.
5
J hl
~
~
I
~ f 4
0
~
10 10'OURS 10
~
I U
lO I
lQ Q
oo lO I
o z
o CD UJ o
(Q lD hI) o I
o CC UJ CL pl 04 0
TINE FlFTER SCRRN SECONOS SUPPRESSlON POOL TEMP. ANALYSlS OWNER'S GROUP CASE 18 SORY HASHIHOTOH PUPLTC POMER SUPPLI STSTOl HHP-2 Fig.
6
0-s 10 HOURS 10 10 10 G=4'bm/f t sec 2
G=9 ibm/ft sec 2
10 10 TIME AFTER SCRAM SECONDS SUPPRESSION POOL TEt1P.
ANALYSIS OWNER'S GROUP CASE 2A ISOL/SCRAM NASNINOTSN PUSL1C PONER SUPPLT STSTDI NNP>>R Fig.
7
~
~
10 10 HOURS 10 10 DO n
Q hl OD 67 M
M UJ O
Q hJM (0
O
)
D 10 10 TINE Rt=TER SCRAN SECONDS SUPPRESSION POOL TEMP-ANALYSES ONNER'$
OROUP CASF PA ISOL/SCRAM NRSHTHOTCH PQBLTC PONfR SUPPLY SYSTfH NHP-2
R I
I ~
O CD CO 10 10 HOURS 10 10 Q
0 CD tQ I
O z
u)
C3 O
C3 ILI O
ID F) hJ CD I
O bl lK 0
04 TINE FIFTER
$ CRAN $ ECON0$
SUPPRESSION POOL TEMP. ANALYSIS ONNER'S GROUP CASE 2A ISOL/SCRAN KRSKIKIIlOK PUQLIC POKER SUPPLY SYSTSII KKP-R Fig.
9
~ 1 4
g
10 10 HOURS 10 10 Q
D C
QJl o
Q C3 o
(0 CO lU Q
Q o
(Q cA 0P 10 10 TIME AFTER
$ CRRM
$ ECOND$
SUPPRESSION POOL TEMP ANALYSIS ONNER'S GROUP CASE 2B IS0I/SCR><
HASHIKOTOH PUPLTC POKER SUPPLT STSTEH HHP-2
8 V
I
OOM 10 HOURS 10 10 OO UJ D
C)
(0M UJ 0
(0 UJ (0
(0 QJ O
)
O I
O (Z
ON 10 10 TINE RFTER SCRRN SECONDS SUPPRESS1OH POOL TEMP. ANALYSIS ORNER'S GROUP CASE 2B ISOL/SCRAM HASHLKOZOH tUBLfC tONER SUttLT STSTEH HHt-R
IP
~
1 J
E I
I 4
~
C 8
10 10 HOURS 10 10 II I
lA Q
DD IJJ lO I
D CC D
G 01 04 TINE AFTER SCRAM SECONOS SUPPRESSION POOL TEMP ANALYSiS ONNER'S GROUP CASE 2B TSOL/SCRAN HRSHIHOZCH PUBLIC POXfR 8UPPLY 8Y8lfH HHP-2
10 10 HOURS 10 10 C
CD CO I
C3 CD Q
CD CD cn (0le CL Q
Q CD (0
10 10 10 lo TIME RFTER SCRRM SECONDS 10 SUPPRESSION POOL TEMP ANALYSIS OWNERS GROUP CASE 38 SBA NASNINOION tUBLIC tONER SUttLY SYSIEII NNt-2
~
~
h
Oo 10 10 HOURS 10 CC OO n
Q UJ o
C)
(0M hl o
II (0
IIJ (0M hl IK C7I OO 10 TINE AFTER SCRAN SECONDS SUPPRESSION POOL TENP.
ANALYSIS ONNER'S GROUP CASE 3B SBA NASNIKSTSN PUSLIC POKER SUPPLY STSTKII NNP-S Flg.
14
~
~
h
~
4.
~
hi h
OO (0
10 10 HOURS 10 O
Z Cl:
G C3 C3 ILI O
tQ 0) 1LI O
D (0
I O
lLI 01 10 TINE AFTER SCRFIN SECONQS SUPPRESSION POOL TEMP ANALYSIS OWNER'S GROUP CASE 38 SBA NR8HlHOTaH PUBLTC POHER 8UPPLY 8T8TEII HHP-2 Fi@.
15
e
~
I
10 10 HOURS 10 G=42 ibm/ft sec 2
G Q
G=94 ibm/f e sec 2
C7 co C)
UJ Q
Q Q
O (0
10 TINE AFTER
$ CRRN
$ ECDND$
SUPPRESSION POOL TEMP-ANALYSIS ONNER'S GROUP CASE 3C SBA HASHTHOTSH PUSLTC PSHEH SUPPLZ SZSTEH HHP-t
V
~
t
~
r
DD Cd 10 10 HOUR5 10 10
~\\
1 el 1
DD n
Q W
DD C)
M (0
W 0
W COM W
D 0
CDl D
W N
lE 10 10 TINE AFTER SCRAM SECONDS SUPPRESSiON POOL TEMP. ANALYSIS OHNER'S GROUP CASE 3C 3~A NASNTXOTSN PUSLTC PSNER SUPPLY SYSTEX NNP-R
oo Q) 10 10 HOURS 10 10 Q
oD LD I
no Z
CC oo ILI o
IQ ca ILI e)o C3I o ILI 01 0
TINE FlFTER SCRFIN SECONDS SUPPRESSION POOL TEMP. ANALYSIS OWNER'S GROUP CASE 3C SBA NSIIIIIOTalI PuBL1C P8IIER 8UPPLY 8Y8tol IIIIP-t