ML17261A104
| ML17261A104 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/02/1979 |
| From: | White L ROCHESTER GAS & ELECTRIC CORP. |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7902090327 | |
| Download: ML17261A104 (116) | |
Text
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REGULAT(Q INFORMATION DISTRIBUTION+STEM <BIDS)
ACCESS I (+0 NB Rt 7902090327 DOC. DATEt 79/02/02 NOTARI ZED t NO DOCKET 0 FACIAL:50-244 Robert Emmet Ginna Nuc lear, Power, Plant, Unit 1, Roches,05000244 AUZM.NAME AUTHOR AFF ILIATI.ON WHITE,L.D.
Rochester Gas
- 8. Electric Corp.
RECI.P.NAME RECIPIENT AFFILIATION ZIEMANN,,D;L.
Operating Rectors Branch 2
SUBJECT:
Forwards ECCS Reevaluation for sub)ect facility.Corrects
.for.zirconium-Water reactor error 8, fulfills all Q
. requirements stated in 780501 exemption from requirements of
.10CF R50.46(a) (1).
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ROCHESTER GAS AND ELECTRIC CORPORATION e
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E ~ IAIDO TOAA I
'IA11 A '
/'9 EAST AVENUE, ROCHESTER, N.Y. 14649 LEON D. WHITE. JR.
VICE PRESIDENT February 2, 1979 TC I. C P H0 R C ARCA COOE TIE 546 2700 Director of Nuclear React;or Regulations Attention:
Nr. Dennis L. Ziemann, Chief Operating Reactor Branch No.
2 U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
ECCS Reevaluation With Correction for Zr-H20 Error R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Nr. Ziemann:
On Nay 1, 1978 the Nuclear Regulatory Commission issed an Exemption to the license of the R.E.
Ginna Nuclear Power Plant related to the requirements of 10 CFR 50.46(a)(l) that ECCS performance be calculated in accordance with an acceptable calcu-lational model which conforms to the provisions in Appendix K, 10 CFR 50.
The previously approved Westinghouse model was found to contain an error associated with the zirconium-water reactor heat generation.
Attachment A to this letter contains an ECCS Reevaluation.
for the R.E. Ginna Nuclear Power Plant, which corrects for the zirconium-water reaction error and fulfills all the requirements stated in the Nay 1, 1978 exemption.
Very truly yours, LDW:cern Attachment L.D. White, Jr.
REGULATORY DOCKET FILE COPY
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Attachment A -
ECCS Reevaluation Metal - Water Correction The Loss of Coolant Accident (LOCA) has been re-analyzed for R.E.
Ginna.
The description of the various aspects of the LOCA analysis is given in WCAP-8339 The individual computer codes which comprise the Westinghouse Emergency Core Cooling System (ECCS) evaluation model are described in detail in separate reports along with code modifications specified in references 7, 10, and ll.
The analysis presented here was performed with the February 1978 version of the evaluation model which in-cludes modifications delineated in references 12, 13, 14, and 15.
Results The analysis of the loss oi 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 densi ty 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 occurence time for various events throughout 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'hat location is specified in Table 2 for the break analyzed.
The location is indicated in feet, which presents elevation above the bottom of the active fuel stack.
Table 3 presents a summary of the various containment systems parameters and structural parameters which were 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.
I t7902090327I
I:8)
~
~
~,The nesults 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.
" Figures 1 through 17 present the transients'or the principal parameters for the break sizes analyzed.
The following items are noted:
Figures 1 - 3:
- equality, mass velocity and clad heat transfer coefficient for the hotspot and burst locations.
Figures 4 - 6:
Core pressure, break flow, and core pressure drop.
The break flow is the sum of the flowrates from both ends of the guillotine break.
The core pressure drop is taken as 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 flood-ing rate.
Figures 12 - 13:
Emergency core cooling system flowrates, for both accumulator and pumped safety injection.
Figures 14 - 15:
Containment pressure and core power transients.
Figures 16 - 17:
Break energy release during blowdown and the contain-ment wall condensing heat transfer coefficient for the worst break.
gh
- 4 pi'~(", Conclusions
- Thermal Anal sis J
()
For breaks up to and including the double ended severance of a reactor coolant pipe, the Emergency Core Cooling System will meet the Acceptanct, Criteria as presented in 10CFR50.46
~ j.
That is:
1.
The calculated peak clad temperature does not exceed 2200'F based on a total core peaking factor of 2.32 2.
The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircalloy in the reactor.
.3.'he '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 long-lived radioactivity remaining in the core.
The effects of upper plenum injection for W designed 2-loop plants has been discussed with the staff
~
Based on interium calculations, a
15'F increase in calculated peak clad temperatures results from explicit modeling of upper plenum injection in the R.E.
Ginna Plant:
The calculated peak clad temperature, including this increase, meets the 2200'F limit.
Attached in Appendix A, are the results of a generic sensitivity study for a typical 2-loop plant with 14 x 14 fuel.
This sensitivity study was performed to demonstrate that the limiting break does not change due to a correction in the metal-water heat of reaction-calculation which is included in the February I
1978 version of the Westinghouse ECCS evaluation model.
1'he results of. this generic sensitivity study show that the liri>iting break for Hestinghouse plants of this type is a double-ended cold leg'uillotine with a discharge coefficient of 0.4.
Since this agrees with past sensitivi ty studies only the limiting LS,9j break for R.E.
Ginna is printed here.
J V'
References for Section 15.4.1 1
l.
"Acceptance Criteria for Emergency Core Cooling Systems for Light Hater Cooled Nuclear Power Reactors",
10CFR50.46 and Appendix K of 10CFR50.46.
Federal
- Register, Volume 39, Number 3, January 4,
1974.
2, Bordelon, F.H., Massie, H.H., and Zordan, T.A., "Westinghouse ECCS Evaluation t<odel-Summary",
WCAP-8339, July 1974.
3.
Bordelon, F.tl., et al.,
"SATAN-VI Program:
Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant",
HCAP-8302 (Proprietary Version),
WCAP-8306 (Non-Proprietary Version), June 1974.
4.
Bordelon, F.H., et al.,
"LOCTA-IV Program:
Loss-of-Coolant Transient Analysis", HCAP-8301 (Proprietary Version),
HCAP-8305 (Non-Proprietary Version), June 1974.
5.
Kelly, R.D., et al., "Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code)".
WCAP-8170 (Proprietary Version),
HCAP-8171 (Non-Proprietary Version), June 1974.
6.
Bordelon, F.H., and Murphy E.T., "Containment Pressure Analysis Code (COCO)", HCAP-8327 (Proprietary Version),
WCAP-8326 (Non-Proprietary Version),
June 1974.
7.
Bordelon, F.M., et al.,
"The Westinghouse ECCS Evaulation Model:
Supple-mentary Information", WCAP-8471 (Proprietary Version),
HCAP-8472 (Non-Proprietary Version), January 1975.
8.
Salvatori, R., "Westinghouse ECCS - Plant Sensitivity Studies",
HCAP-8340 (Proprietary Version),
HCAP-8356 (Non-Proprietary Version), July 1974.
9.
Delsignore T., et al.,
"Westinghouse ECCS Two-Loop Sensitivity Studies (14 x 14)",
HCAP 8854, (Non-Proprietary, Version),
September 1976.
10.
"Westinghouse ECCS Evaluation Model, October, 1975 Versions",-WCAP-8622 (Proprietary Version),
HCAP-8623 (Non-Proprietary Version), November, 1975.
ll.
Letter from. C. Eicheldinger of Westinghouse Electric Corporation to D.B.
Yassalo of the Nuclear Regulatory Commission, letter number NS-CE-924, January 23, 1976.
12.
Kelly, R.D., Thompson, C.M., et al., "Hestinghouse Emergency Core Cooling System Evaluation Model for Analyzing Large LOCA's During Operation with One Loop Out of Service for Plants without Loop Isolation Yalves",
WCAP-9166, February, 1978.
13.
Eicheldinger, C., "Westinghouse ECCS Evaluation Model, February 1978 Yersion",
HCAP-9220 (Proprietary Version),
HCAP-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
'5.
Letter from T.M. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Comnission, letter number NS-TMA-1834, June 20, 1978.
16.
Letter from E.G.
Case of iVRC to L.D. White of RGGE - December 16, 1977.
17.
Letter from L. D. White of RGEE to A. Schwencer of NRC - January 16, 1978.
18.
Letter from A. Schwencer of NRC to L.D. White of RGEE - February 10, 1978.
19.
Letter from K.H. Amish of RGSE to A. Schwencer of NRC - February 15, 1978.
20'.
"Safety Evaluation Report on ?nterim ECCS Evaluation model f'r Ilestinghouse
~
~
~
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Two-Loop Plants",
March 1'978 transmitted by letter from D.L. Ziemann of NRC to L.D. White of RG&E - May 1, 1978.
EVENT Accident Initiati on Reactor Trip S ignal Safety Injection S ignal Start Accumulator Injecti on End of ECC Bypass End of Blovidov~n Bottom of Core Recovery Accumulators Empty Start Pumped ECC Injection TABLE 1 LARGE BREAK - TI/lE SE UENCE OF EVENTS OCCDREIICE TIIIE (~SECONDS DECLG,Cp =0.4 0.0 0.67 0.75 10.3 21.95 21.95
- 42. 4 58.
25.75
uantities in the calculations:
Licensed core power rating Total core peaking factor Peak linear power Accumulator water volume Accumulator pressure TABLE 2 LARGE BREAK - ANALYSIS INPUT AND RESULTS 1025 of 1520 HNt 2.32 102%of 13.57 kw/ft 1100 cubic feet per tank 700 PSIA Number of Safety Injection Pumps Operating Steam Generator Tube Plugging Level-Fuel Parameters-Cycle 7
~10 ercent (uniform)
Region All Results DECLG CD 0'4 Peak clad temperature (oF)
Location (feet) 2057 7,5 ifIaximum local clad/water reaction (X)
Location (feet) 4.7 7.5 Total core clad/water reaction
('A)
Hot rod burst time (seconds)
Location (feet)
<0. 3 31.4 6.0
TABLE 3 Dr Containment Data Net Free Volume Initial Conditions Pressure Temperature e
. RMST Temperature Service 'ilater Temperature Outside Temperature 1066000 FT3 14.7 PSIA 90 oF 60 oF 37 oF
-10 oF SPRAY SYSTEt'1 Number of Pumps Operating Runout Floiv Rate Actuation Time 2
1800GPH 2
SEC SAFEGUARDS FAN COOLERS Number of Fan Coolers Operating 4
Fastest Post Accident Initiation of Fan Coolers 20 SEC
STRUCTURAL l(EAT SIHKS TABLE 3 (cont'd)
THICKNESS 1.3 I
.375 30.0
.375
- 30. 0 24,0
.375 24.0 60.0
.375
- 42. 0.
24.0
.375 24.0
.25 30 '
- 30. 0 30,0 6.0 1.5 1.0
.5
.75 24.0
.125 (in) nsulation Steel Concrete Steel Concrete Concrete Steel Concrete Concrete Steel Concrete Concrete Steel Concrete Steel Concrete Concrete Concrete Concrete Steel Steel Steel Steel Concrete Steel AREA (ft2) 36945 12370 7955 2342.
297 6400 6900 14900 6170 9174 56020 8568 5756 9162 7000
Reflood Viass And Energy Releases
- TABLE 4 Time (Sec. )
Viass Flowrate (g
)
Energy Flowrate
(~B ")
- 43. 0 48.8 59.0
- 71. 0
'85. 3 101.3 118.6 157. 0 202. 0 257.3
.012
- 30. 2 65.8 124.4 199.
210. 2 214. 3 221. 1 227.8 235.9 15.2 39108.
83997.
98618 113470
- 111998, 108754.
101680.
94315.
87218.
TABLE 5 Broken Loop Accumulator Hass and Energy Release Time (sec.)
f4ass Flow (lb' Energy Flow (BT" )
- 1. 010
'.010 3.010 4, 010 5 ~ 010
- 6. 010 7.'010
- 8. 010
- 9. 010
- 10. 010
- 11. 010
- 12. 010
- 13. 010
- 14. 0'l 0
- 15. 010 16, 010
- 17. 010
- 18. 010
'l9. 010
- 20. 010
- 21. 010
.24. 867 25.367 25.867 26.367 26.867 2?.367 27.867 28.367 28.867 29.367 29.867 30.367 30.867
- 31. 367
- 31. 867 32.367 32.867
- 33. 367
- 33. 867 34.367 34.867 35.367 35.867
" 36.367 36.867 37.367 37.867 38.367 38.867 39.367 39.867 40.367 40.867 41
~ 367
- 41. 867 42.367 42 407 2522. 382 2409; 835 2312. 506 2227.453 2151. 288 2082.020 2018. 485
)960.065 1906.637 1857.462 1812. 132 1770. 395 1?3'l.668 1695.646 1662.489 1631. 888 1603. 576 1577. 205 1552. 533 1529.328 150?.532 1432.522 1422.482 1412. 637 1Ii02.981 1393.507 1384.209 1375. 083 1366. 12'l 135?.319 1348.672 1340. 175 1331.824
'1323. 614 1315. 541 1307.600 1299.?89 1? 92. 103 1284.539 1277. 094 1269.763 1262.545 1255.436 124i8.43?
124i1. 532 1234. 733 1228. 031 1Z21. 425 1214. 912 1208. 489 1202. 156 1195.908 1189.745 1183.664 1177.663 1171, 740 1165.89'165.430 1S1065.432 144325. 009 138496. 001 133402. 162 128840. 609 124692 '07 120887.076 117388.301 114188. 462 111243.390 108528.602 106028.966 103?09.594 101552. 254 99566.480
'?733.755 96038. 191 94458.?81 92981. 197 91591.. 436 90286.074 85793.738 85192.430 84602.821 84024.513 83457. 128 82900. 305 82353.700 81816. 984 81289. 844 80771. 980 80263. 105 79762. 946 79271. 240 78787.736 78312. '193
?7844.381 77384.077 76931. 070 764i85. 154 76046. 135 7S613.824 75188.039 74768.607 74355.359 7394i8. 134 73546.777 73151. 136 72761. 069 72376.435 71997. 101 71622. 935 71253. 813 70889.615 70530.222 70175.521 69825.405 69797.590
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1.<000 1.2500 R.
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CINNA NUCLEAR PO'MER PLANT ZIRC MATER ECCS ANALYSIS DOUBLE ENDEO COLO LEG GUILLOTINE CO
=
0 ~ I OUALITY OF FLUIO
~'URST't
'6:00 FT( )" PEAKED '7':50 FT(<)'
lot LJ 0
1.0000 0.7500 0.5000 0.2500 0.0 00 0
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GINNA NUCLEAR POMER PLANT jIRC MATER ECCS ANALYSIS OOUBLE ENOEO COLO LEC GUILLOTINE CO
= O.l MASS VELOCITY BURST 6.00 FT(
1 PEAK 7.50 Ftl
)
T5.000 I
I 50.000 25.000 I
u 0.0 lsl
< -25.000 X
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ID CVl o.n a norm o 0 o norm o o Qoaon ID ~ O)OIO
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600.00 500.00
~i l00.00 300.00 200.00 R.
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GINNA NUCLEAR POWER PLANT 2IRC
'WATER ECCS ANALYSIS OOUBLE ENOEO COLO LEG GUILLOTINE --
CO
= O.I.
HEAT TRANS ~ COEFEICIENT BURST 6.00 FT(
)
PEAK. 7.50 FTl
)
50.000 cc0.000 30.000
" 20.000 i
O'.IIII 6.00UO 5.0000 i.0000 3.0000
~
2.0000 1.0000 CD E
CD CD CD CD CCJ FIGURE 3 TINE (SEC)
CD CD CD CD C
CD CD CD
g c>00.0 R.
E.
GINHA NUCLEAR POMER PLANT ZIRC MATER ECCS ANALYSIS DOUBLE ENDED COLD LEG GUILLOTINE --
CD
= O.I PRESSURE CONE BOTTOM
(
)
TOP e
lt) c 2000.0 nl X
1500. 0 Z
1000.0 500.00 0.0 CD CD CD CD CD I/l CD CD C7 CD TIME <SEC).
CIl CD CD CD CD Cll CD CD CD CCC FIGURE 4 '
1.DOE+05 1;50Ei0%
LJ
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GINNA NUCLEAR PO'WER PLANT ZIRC WATER ECCS ANALYSIS DOUBLE ENDED COLD LEG GUILLOTINE -" CO
= 0.1
~ BREAK FLOM Z.SOE+Oi 0.0
-2.50EK)i
-5.DOE+OS
-7.50EK)l
-I.COEDS C>
C3 C)
I/1 ED C)
C)
C7 TIHE (SEC)
<<l Cl C)
AJ I
"FIGURE 5
70.000 R: E.
GINNA NUCLEAR POMER PLANT
lSRC~ATER ECCS ANALYSIS OOUBLE ENOEO COLO LEG GUILLOTINE CO
= 0.4 CORE PR.OROP 50.000 cK 25.000 Cl CC 0.0
-25.000
-50.000
-10.000 CD C)
C)
C)
TIHE (SEC)
C)
Ifl C)
ED C)
O, AJ FIGURE 6
2500.0 R.
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GIHNA NUCLEAR POMER PLANT llIIC MATER-ECCS ANALYSIS OOUBLE EHOEO COLO LEG GUILLOTIHE --
CO -"O.i CLAO AVG.TEMP.HOT R90 BURST 6.00 FT(
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PEAK 7.50 FT( ~ )
cc 2000.0 o
CC 1500.0 x
X laJ I
o 1000.0 500.00 0.0 o
8 8.
o TINE (SEC) 8' oo o
o FIGURE 7
t
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2000.0
)750.0 R.
E.
GINNA NUCLEAR POMER PLAHT ZIRC MATER ECCS ANALYSIS OOUBLE ENOEO COLO LEG GUILLOTINE --
CO
= O.I FLUIO TEMPERATURE BURST>>
6.00 FT(
)
PEAK>> 7.50 FT)>)
o 1500.0 1250.0 w
1000.0 750.00
~e I
500.00 250.00 0.0 n
~ n 8
CV TIME (SEC) n n
Vl
'FIGURE 8
r f
l 1
7000.0 R.
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ClNNA NUCLEAR PAMER PLAHT ZIRC MATER ECCS ANALYSlS OouBLE EHOEo coLO LEC culLLOTlNE co
= O.i l-FLOVRATE CORE BOTTOM
(
)
TOP e
(i)
'J leI/l 5000.0 2500.0 I
0.0
-25i,0.0
-5000.0
-7000.0 C7 C)
CD CSg C7 ED Cl C)
TlHE (SEC)
C)
C)
C) l/l CD C)
CD CD I/l Cl/
FIGURE 9
20.000 l7.500 R.
E.
GINHA HUCLEAR POMER PLAHT 2IRC MATER ECCS ANALYSIS OOUOLE EHOEO COLO LEG GUILLOTINE CO
= O.I VATER LEVELtI'T)
I'5.000 12.500 I
~
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DOWNCOtlMER 10.000 7.5000.
5.0000 CORE 2.5000 0.0 CI Cl n
Cl n
Cl 8
AJ TIME (SEC) n if'IGURE 10
2.0000 1.7500 R.
E.
GIHNA NUCl.EAR POMER PLANT 2IRC MATER ECCS ANALYSIS DOUBLE EHOEO COLO LEG GUILLOTINE CO
= O.l FLOOO RATE 1lN/SEC) 1.5000 I
1.2500 CD CD 1.0000 0.7500 0.5000 0.2500 0.0 CD CD TIHE
<SEC)
CD CD 8
CD CD CD FIGURE 11
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2560.0 R.
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GINNA NUCLEAR POWER PLANT
=- - "
'2IRC HATER'CCS ANALYSIS OOUBLE ENOEO COLO LEG GUILLOTINE --
CO
= 0.4 ACCUH.
FLOW J
e
.C, 2000.0 I ~
o 1500.0
\\J 1000.0 500.00 0.0 oo n
CD I/1 CD CD TIHE {SEC)
CD CD CD CD CD AJ CD I(
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TABLE 1 v
Lar e Break - Time Se uence of Events EVENT OCCURENCE TIHE SECONDS
- DECLG, CD =1.0
- DECLG, CD =0.6 DECLG,CD = 0.4 Accident Initiation Reactor Trip Signal Safety Injectio~ Signal Start Accumulator Injection End of ECC Bypass End of Blovido>vn Bottom of Core Recovery Accumulators Empty Start Pumped ECC Injection 0.0 0.52 0.48 5.7 15.15 15.15 28.6 37.5 25.48 0.0 0.53 0.56 7.5 16.75 16.75 30.0 39.1 25.56 0.0 0.54 0.67 9.8 19.96 19.96 32.8 42.1 25.6
TABLE 2 LARGE BREAK - ANALYSIS INPUT AND RESULTS uantities in the calculations:
Licensed core power rating Total core peaking factor Peak linear power Accumulator water volume Accumulator pressure Number of Safety Injection Pumps Operating Steam Generator Tube Plugging Level 102% of 1650 Yiltt 2.31 102% of 14.3 kw/ft 1250 cubic feet per tank 700 PSIA 1
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Fuel Parameters-Cycle Generic Region Generic Results
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0.4 Peak clad temperature
( F)
Location (feet) flaximum local clad/water reaction
(%)
Location (feet)
Total core clad/water reaction(/)
Hot rod burst time (seconds)
Location (feet) 1936 7.5 4.1 7.5
<0.3 78.0 7.0 1964 7.5 4.5 7.5
<0.3 70.6 6.75 2193 7.5 9.1 7.5
<0.3 25.6 5.75
TABLE 3 DRY CONTAINMENT DATA NET FREE VOLUME INITIAL CONDITIONS Pressure Temperature RllST Temperature Service Hater Temperature Outside Temperature SPRAY SYSTEH Number of Pumps Operating Runout Floe> Rate Actuation Time SAFEGUARDS FAN COOLERS Number of Fan Coolers Operating Fastest Post-Accident Initiati~~ g [an STRUCTURAL HEAT SINKS Thickness (in.)
1,5steel 0.75 steel 0.25 steel 12 concrete 0.375 steel 0.25 steel 0.5 steel 0.145 steel 0.09 steel
- 0. 1 steel 0.1875 steel 1.44 steel 12 concrete 6 concrete 3 concrete 1.37 x 10 ft 14.7 psia 90o F
70o F
32' 20o F
2 1600 gpm/each 15 sec 4
15 sec Area (ft2) 41300 32000 7860 6800 32000 44000 1695 12400 6000 13125 2200 40800 25070 7570
P I
TABLE 3 (cont'd)
PAINTED SURFACES Steel Thickness (in) 1.5
.75
.5
. 375
. 1875
. 145
- 1. 44 t1inimum Painted Area (ft2)
, 41300 32000 44000 6800 13125 1695 2200 Paint Thickness (mils) 11 11 11 11 11 11 11 Concrete Thickness (in) 12 6
3 Minimum Painted Area (ft )
4080 25070 7570 Paint Thickness (mils) 18 18 18
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1 TASLE 4 REFL009 MASS ANO ENERGY RELEASE Cp = 0.4 DCCLG Time Mass Flow (lb/sec)
Energy Flow (~ "/sec) 33.4 38.0 47.0 61.3 79.2 99.5 121.7 171. 6 230. 0 300. 6 395. 7
.002 34.16 137.97-225.22 237.1 242.07 245.93 252.7 259.35 266.46 276.64 2.43 44025.1 85563.1 102148.2 100581.8 98133.1 95368.0 89344.0 82964.7 76280.0
. 69442.6
TABLE 5 Broken Loop Accumulator Mass and Energy Release Time (1asg Floiv ('b/sec)
Energy Flow (B'Ulsec) 0.0 2.0 4 0 6.0 8.0 3.0. 0 12.0 3.4. 0 j.6. 0 3.8. 0
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254469.0 228957.0 210127.0 195166.0
.182888.0 172734.0 164089.0 156730.0 150557.0 140842. 0 132606,0 0.0
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