ML17333A289
| ML17333A289 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/22/1995 |
| From: | Rosenbaum E HOLTEC INTERNATIONAL |
| To: | |
| Shared Package | |
| ML17333A288 | List: |
| References | |
| HI-951389, HI-951389-R01, HI-951389-R1, NUDOCS 9602080080 | |
| Download: ML17333A289 (36) | |
Text
0 waeaa HOLTEC I N T E R N AT I 0 N A L UPDATED THERMAL-HYDRAULICANALYSIS of SPENT NUCLEAR FUEL POOL DONALD C. COOK NUCLEARPLANT INDIANAMICHIGANPOWER COMPANY by HOLTEC INTERNATIONAL HOLTEC PROJECT 51121 HOLTEC REPORT HI-951389 REPORT CATEGORY: I
> >>'.>> e>>'::.rr<<:::i.,j)% gee@<<<<e><<a e>>ee.g.
'. >>::??'("q*v<<c>>><<?? '>>I~@
<<><<>>> gwp
> <<<<>?,"??+?>><<>v&@>Ver<<>i... <<;:?? 9+>>.::. <<>>>>.
<<<<??
l 9a020sooso 96020<
PDR ADQCK 050003'L5'I P
PDR'
55555 HOLTEC I N T E R N AT I 0 N A L REVIEW AND CERTIFICATIONLOG DOCUMENT NAME:
HOLTEC DOCUMENT I.D. NUMBER:
HOLTEC PROJECT NUMBER:
CUSTOMER/CLIENT:
Updated Thermal-Hydraulic Analysis of Spent Nuclear Fuel Pool HI-951389 51121 American Electric Power Service Corporation REVISION BLOCK REVISIONt NUMBER ORIGINAL REVISION 1 REVISION 2 AUTHOR&,
DATE Cp g~san uw tl F
REVIEWER &
DATE a reGef-tf QA MANAGER
& DATE
'Va G~
lt-iO -
8'U, Q
~ Su t h sx-u.-
5 APPROVED
& DATE
< <<Cd((
~l /P) 9~
)>i~ifM REVISION 3 REVISION 4 REVISION 5 REVISION 6 This document conforms to the requirements of the design specification and the applicable sections of the governing codes.
Note:
Signatures and printed names are required in the review block.
A revision of this document willbe ordered by the Project Manager and carried out ifany of its contents is materially affected during evolution of this project. The determination as to the need for revision willbe made by the Project Manager with input from others, as deemed necessary by him.
Must be Project Manager or his designee.
THE REVISION CONTROL OF THIS DOCUMENT IS BY A "
SUMMARY
OF REVISIONS LOG" PLACED BEFORE THE TEXT OF THE REPORT.
Form: RCL.02
r 0
C'
SUMMARY
OF REVISIONS LOG HOLTEC REPORT HI-951389 Revision 1 contains several editorial changes and clarifications throughout the text.
Title Page Review and Certification Log Summary of Revisions Log Table of Contents Section 1
Section 2 Section 3 Section 4 Section 5 Section 6 Section 7 Tables Figures 10 Title Page Review and Certification Log Summary of Revisions Log Table of Contents Section 1
Section 2 Section 3 Section 4 Section 5
Section 6 Section 7 Tables Figures
$EE.'TI~N
1.0 INTRODUCTION
TABLE OF CONTENTS PA 2.0 ANALYSISPROCEDURES 3.0 DISCHARGE SCHEDULES AND SCENAMOS 4.0 ANALYSISRESULTS
5.0 CONCLUSION
6.0 REFERENCES
7.0 ANSWERS TO RAI BY NRR TABLES FIGURES
1.0 D
The Donald C. Cook Nuclear Plant, operated by American Electric Power (AEP) for Indiana Michigan Power, is a dual unit pressurized water reactor (PWR) generating station.
The rated thermal power of the Unit 1 reactor is 3250 MW(t). The rated thermal power of the Unit 2 reactor is 3588 MW(t). Spent nuclear fuel (SNF) assemblies discharged from both reactors are stored in a common spent fuel pool (SFP).
The SFP contains storage racks which provide sufficient storage locations for 3613 SNF assemblies.
Heat is generated by the continuing radioactive decay of the stored SNF assemblies.
Recent modification to the SNF discharge schedules for the two reactors willalter the decay heat generated by SNF assemblies in the SFP.
The new discharge schedules call for increasing both the cycle length and the number of assemblies discharged per cycle.
While the longer cycle length increases the fuel decay time between discharges, reducing the decay heat contribution ofpreviously discharged fuel, italso increases the reactor "operating" time offreshly discharged fuel, which increases the decay heat contribution of the freshly discharged assemblies.
The purpose of this report is to characterize the thermal-hydraulic response of the SFP in lightof the new discharge schedule.
For all scenarios in this evaluation the in-core decay time (also referred to as the reactor hold time) for fuel from both reactors is assumed to be 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
The analyses documented herein parallel those contained in Holtec report HI-941183 f1]. The results contained in this report supersede the results of the earlier report.
Previous Holtec page 1
analyses of the Donald C. Cook SFP determined the fuel decay heat using the methods outlined in USNRC Branch Technical Position ASB 9-2 [2]. This analysis uses the ORIGEN-2 computer code [3] to determine the decay heat.
The ORIGEN-2 decay heat calculations are more rigorous
'nd more accurate than the approximate correlations of ASB 9-2.
page 2
2.0 AL I PR ED The foQowing analyses are performed in the course of this thermal-hydraulic evaluation:
(i)
Long-Term Decay Heat Calculation.
This analysis is performed to calculate the accumulated decay heat of all previously discharged SNF assemblies in the SFP.
(ii)
SFP Transient Thermal Response Determination. This analysis is performed to determine the SFP bulk temperature and decay heat profiles for the postulated final discharge into the po'ol.
(iii)
Time-to-Boil Calculation.
This analysis is performed to calculate the time required before boiling, in the wake of a postulated total loss of forced cooling of the SFP.
(v)
Maximum Local Temperature Calculations.
The maximum local water temperature and the maximum local fuel cladding temperature are determined, both with and without partial cell blockage, to evaluate the possibility of nucleate boiling on the surface of the fuel assemblies.
pagB 3
~ 4
~
~
3.0 DT HAR E HED ENARI
. The fuel discharge schedules (including both historic and future fuel discharges) for both the Unit 1 and Unit 2 reactors are presented in Table 1 and Table 2, respectively.
For postulated future f'uel discharges, the fuel enrichments for Unit 1 and Unit 2 are assumed to be 3.5 % and 4.0 %, respectively.
These enrichments are expected to provide a lower bound for future fuel enrichments, and therefore yield an upper bound for thermal power output. Because the analyses are performed for the last discharge scenario in which the SFP storage capacity is not exceeded, these two discharge schedules only include discharges up to the point where the SFP becomes full(cycle 25A) and not to the end-of-licensed reactor life.
A total of five discharge scenarios are evaluated in these analyses.
These discharge scenarios are identical to those presented in previous licensing submittals.
These five scenarios are summarized as follows:
Case 1A:
Normal Discharge, 1 Cooling Train design flow Case 1B:
During the cycle 25A discharge from the Unit 1 reactor, a total of 84 fuel assemblies are placed in the SFP.
All future fuel assemblies discharged are assumed to have been exposed to the maximum burnup of 52,200 hGVD/MTU.
The SFP water flowrate through the SFP cooling system (SFPCS) is assumed to be the design point value of 2,300 gal/min.
Normal Discharge, 1 Cooling Train maximum (as measured) flow This case is identical to Case 1A, except that the SFP water flowrate through the SFPCS is assumed to be the maximum (as measured) value of 2,800 gal/min.
Case 2:
Normal Discharge, 2 Cooling Trains This case is identical to Case 1A, except that two cooling trains are operating.
page 4
Case 3:
Back-to-Back Full Core Discharge, 2 Cooling Trains The Unit 2 reactor has an unplanned shutdown 30 days after the cycle 25A shutdown of the Unit 1 reactor.
A fullcore of 193 assemblies is removed from the Unit 2 reactor and placed in the fuel pool.
There are two cooling trains operating at design flow rates.
The fuel assemblies from the Unit 2 reactor are discharged in three groups.
The first group contains 65 assemblies with 64,800 MVD/MHJburnup, the second group contains 64 assemblies with 43,200 MWD/MTUburnup, 'and the third group contains 64 assemblies with 21,600 MVD/MTU.
While the fuel pool willnot actually have enough storage locations to hold a full core offload after the normal cycle 25A discharge, this hypothetical case does provide a decay heat load that is guaranteed to bound any actual scenario (3800 assemblies).
Case 4 Back-to-Back Full Core Discharge, Single Cooling Train This case is identical to Case 3, except that only one cooling train is operational.
This case is not a design basis scenario for either the Cook Nuclear Plant or USNRC guidelines t'4], and is presented for reference only.
page 5
4.0 NALY I R 4.1 Long-Term Decay Heat Calculation The decay heat contribution of all SNF in the SFP as of June 24, 2010 isMetermined using the proprietary Holtec computer program LONGOR [5]. Based on the discharge schedules in Table 1 and Table 2, the long-term decay heat is 14,117,944 Btu/hr. All subsequent analyses are performed with the long-term decay heat contribution held constant at this value.
4.2 SFP Transient Thermal Response Determination The thermal response of the SFP and SFPCS is determined using the proprietary Holtec computer program BULKZEM[6]. The temperature and heat load profiles for each discharge scenario are presented graphically in Figures 1 through 10. Numerical results are summarized in Table 3.
The initial drop in temperature shown in Figures 1 through 5 is caused by an elevated initial temperature condition. The elevated initial temperatures have no effect on the analyses because all scenarios reach a steady-state temperature before commencement of fuel transfer.
The analysis shows that the limiting normal discharge scenario is Case 1A. The results of the analysis for this scenario demonstrates that the SFPCS can provides sufficient cooling to page 6
maintain the temperature of the SFP below 155 'F. The coincident decay heat load (excluding evaporative heat losses) is 27.2 MBtu/hr. If two SFPCS trains are available, the peak temperature and coincident decay heat load become 129 'F and 29.3 MBtu/hr.
For a back-to-back full core discharge scenario (Case 3) as prescribed in SRP 9.1.3 [4], the maximum temperature is less than 147 'F and the coincident decay heat load is 54.3 MBtu/hr.
In accordance with SRP 9.1.3, no single active failure need be associated with the full core discharge scenario.
For reference, however, during a back-to-back fullcore discharge with only 1 SFP train operating (Case 4) the maximum temperature is maintained below 181 'F and the coincident heat load is 47.3 MBtu/hr.
4.3 Time-to-Boil Calculation For each discharge scenario, the effects of a total loss of forced cooling is evaluated using the proprietary Holtec computer program TBOIL P].
For each scenario, it is conservatively assumed that the loss of cooling occurs at the instant of peak bulk temperature and that no makeup water is available to the SFP.
The evaluation results are summarized in Table 4. For the design basis cases, the minimum time between SFPCS Mure and pool boiling is 6.08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (Case 3). The maximum boiloffrate for this scenario is 103.68 gal/min. For the design basis normal discharge (Case 1A), the time between SFPCS failure and pool boiling is 9.45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> and the maximum boiloffrate is 63.35 gal/min.
page 7
4.4 Maximum Local Temperature Calculations The maximum local water temperature and maximum local fuel cladding temperature are determined for the point in time where the bulk SFP temperature reaches its maximum value using the proprietary Holtec computer program TKBU?OOL [8]. Both unblocked and 50%
blocked scenarios are evaluated.
The results of this analysis are summarized in Table 5. Por the limitingdischarge scenario (Case 1), the maximum local water temperature is calculated as 163.6 'F and the local maximum fuel cladding temperature is 214.5 'P. Ifthe limitingrack cells become blocked by 50%, the maximum water temperature increases to 223.5 'P and the maximum fuel cladding temperature increases to 254.3 'P.
The local boiling point is dependent on the local pressure.
At a water depth of23 feet, the local boiling point is 238 'P. The maximum local water temperature (50% blockage, above) is only 223.5 F. Thus, nucleate boiling willnot occur in the SFP, even under conditions of maximum heat flux and maximum bulk temperature.
page 8
i
~
~ ~
~ 2r il ~
5.0 CQHCUIRIIN The results presented in this report demonstrate that the, maximum bulk spent fuel pool water temperature is conservatively bounded by a temperature of 155'P, for normal discharge scenarios.
This value is 6 'P less than that calculated in the previous evaluation [1]. Therefore, the modification to the fuel assembly discharge schedules has the effect ofincreasing the margin of safety over that established in the previous submittal.
page 9
6.0 SHHR~
[1]
Holtec Report HI-941183, "Spent Nuclear Fuel Pool Thermal-Hydraulic Analysis Report for Donald C. Cook Nuclear Plant", Rev. 2
[2]
USNRC Branch Technical Position ASB 9-2, "Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, 7/81
[3]
A.G. Croff, "ORNL Isotope generation and Depletion, A User's Manual for the ORIGEN-2 Computer Code", ORNL/TM-7175, RSIC/CCC-371, Oak Ridge National Laboratory, July, 1980.
[4]
NUIT-0800, Standard Review Plan, Section 9.1.3
[5]
Holtec Report HI-951390, "QADocumentation for LONGOR", Rev. 0
[6]
Holtec Report HI-951391, "QA Documentation for BULKI'EM",Rev. 0.
[7]
Holtec Report HI-92832, "QA Documentation for TBOIL",Rev. 2.
[8]
Holtec Report HI-92833, "QA Documentation for 'IKERPOOL", Rev. 0.
page l0
7.0 AN WER T RAI BY NR This section contains responses to questions posed by the USNRC in response to the previously submitted Holtec Report HI-941183. Any reference to tables or figures in this section refer to the corresponding articles in either the original licensing report (dated July, 26 1991) or Holtec Report HI-941183.
Q l.
Itis assumed that the combined SFP Hx heat load and evaporative heat losses (as shown in Table 2.2 of Holtec Report HI-941183) are equivalent to the total decay heat generation in each case, e.g., Case 1A SFP Hx load 30.84 E6 BTU/Hr. + evaporative losses 3. 14 E6 BTU/Hr., for a total of 33.98 E6 BTU/Hr. Ifthis is incorrect, explain what the correct decay heat load is in each case and justify any differences.
A 1.
Your assumption is correct as stated.
The total decay heat generation is equal to the sum of the SFP Hx heat load and the evaporative cooling loss.
Q2.
Apreliminary comparison was made of the decay heat generated by the 80 fuel elements deposited in the spent fuel pool in cases 1A, 1B and 2 for the decay times shown in Table 2.2 with similar cases in Table 5.5.1 of your previous submittal dated July 26, 1991 wherein the fuel was permitted to decay for 168 in lieu of the decay period of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> presently requested.
That comparison shows differences of2.6 to 2.7 E6 BTU/Hr.
in lieu of the differences you show of 0.71 to 0.86 E6 BTU/Hr.
Justify your calculations.
A 2.
The reduction in the reactor hold time will increase the decay heat of the freshly discharged fuel assemblies only, the decay heat from previously discharged fuels is not affected by the change in hold time.
However, the fresh fuel decay heat accounts for less than 50% of the total heat generation.
Additionally, Holtec Report HI-941183 incorporates changes to the refueling schedule which reduce the decay heat contribution of the previously discharged fuel.
The reduction in the decay heat of the previously discharged fuel serves to limitthe increase in the total decay heat generation rate.
Q3.
Explain whether you have deviated from Table 2.1 ofHoltec report HI-941183 in using the number of those discharged assemblies and dates ofdischarges in calculating the heat generation of the spent fuel assemblies-stored in the spent fuel pool. For example, you page 11
~
~ill
~
~
~
e ~
~
a y
s stated that you calculated the heat generation for 80 fuel assemblies in a normal discharge batch in lieu of 76 shown in Table 2.1.
A 3.
The decay heat calculations in Holtec Report HI-941183 are devised to provide an upper bound to any actual discharge scenarios.
The calculation of the decay heat generation from all previously discharged fuels is based on the refueling schedule of Table 2.1.
However, the normal discharge batch from Unit 1 contains more assemblies than does Unit 2.
To provide an analysis that bounds all normal discharge scenarios, the final discharge batch size was assumed to be the Unit 1 batch size of 80 assemblies.
This assumption serves to increase the conservatism of the analysis.
Q 4.
Provide the decay heat generation rate for the assemblies deposited in the pool for each discharge cycle used in your calculations for Cases 3 and 4. Ifyou do not use the discharges and cycle EFPD shown in Table 2.1 explain the method used and justify its application.
A4.
The decay heat generation rates for fuel from each previous discharge cycle are summarized below.
It was conservatively assumed that the EFPD for all previously discharged fuel assemblies was 1260 days.
Cycle 10 12 13 Unit 1 Decay Heat (BTV/Hr) 182106 184518 189342 197181 198990 204417 261702 273159 287631 300294 311751 324414 337077 pag6 12 Unit 2 Decay Heat (BTU/Hr) 239391 285822 232155 232155 306928 293661 297882 305118 325017 335871 346725 358785 370845
~ I
~
~
~ ~
~
C
~
I 4I 14 15 16 17 18 19 20 21 22 23 24 348534 361197 373257 385920 399789 414261 434160 472149 566820 773046 2981835 384714 400392 422100 467928 598176 1078164 page 13
Table 1 - Discharge Schedule for Unit 1 Reactor Cycle EOC Date
¹ of Assys Pool Total Burnup Enrich.
Weight U 1A 23-Dec-1976 2A 6-Apr-1978 6-Apr-1979 4A 30-May-1980 SA 29-May-1981
. 65 65 65 129 193 338 494 19,100 29,100 34,200 31,900 31,300 2.25 2.80 3.29 2.93 2.92 452.6 454.9 451.7 429.1 428.4 6A 7A 8A 4-Jul-1982 17-Jul-1983 6-Apr-1985 80 80 558 710 882 31,600 31,400 30,100 2.90 2.91 2.91 427.4 427.4 427.4 9A 22-Jun-1987 10A 19-Mar-1989 11A 11-Oct-1990 12A 22-Jun-1992 13A 12-Feb-1994 14A 14-Jul-1995 15A 20-Dec-1996 16A 25-May-1998 17A 28-Oct-1999 18A 1-Apr-2001 19A 4-Sep-2002 20A 7-Feb-2004 21A 12-Jul-2005 22A 15-Dec-2006 23A 19-May-2008 24A 22-Oct-2009 25A 27-Mar-2011 80 80 80 80 80
'80 84 84 84 84 84 84 84 84 84 1050 1210 1367 1523 1603 1759 1927 2095 2263 2431 2599 2767 2935 3103 3271 3439 3607 35,300 37,800 35,600 39,400 38,400 36,700 52,200 52,200 52,200 52,200 52,200 52,200 52,200 52,200 52,200 52,200 52,200 3.19 3.47 3.43 3.37 3.45 3.40 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 446.4 459.8 460.7 460.2 460.8 461.1 461.0 461.0 461.0 461.0 461.0 461.0 461.0 461.0 461.0 461.0 461.0
~
~
a
~ y(,4
~ ~
I Table 2 - Discharge Schedule for Unit 2 Reactor Cycle 1B EOC Date 20-Oct-1979 80 273 16,600 0 of Assys Pool Total Burnup Enrich.
2.16 Weight U 459.3 2B 3B 4B 8B 9B 10B llB 12B 13B 14B 15B 17B 18B 19B 20B 15-Mar-1981 22-Nov-1982 10-Mar-1984 28-Fab-1986 1-May-1988 30-Jun-1990 20-Feb-1992 6-Sep-1994 15-Mar-1996 23-Aug-1997 26-Jan-1999 30-Jun-2000 3-Dec-2001 8-May-2003 10-Oct-2004 15-Mar-2006 18-Aug-2007 20-Jan-2009 25-Jun-2010 92 72 92 88 80 77 76 76 84 84 84 84 84 84 84 84 84 430 630 802 970 1130 1287 1443 1679 1843 2011 2179 2347 2515 2683 2851 3019 3187 3355 3523 28,200 32,100 34,300 36, 100 39,400 40,200 42,500 43,500 64,800 64,800 64,800 64,800 64,800 64,800 64,800 64,800 64,800 64,800 64,800 2.84 3.38 3.27 3.57 3.64 3.77 3.96 3.82 4.00 4.00 4.00 4.00 4.00 4.00 4.00 4.00 4.00 4.00 459.6 459.4 459.2 419.9 402.2 402.7 402.9 412.7 410.0 410.0 410.0 410.0 410.0 410.0 410.0 410.0 410.0 410.0 410.0
Table 3 - Maximum SPP Bulk Temperature and Coincident Heat Loads and Losses Case Number CaselA Case 1B Case 2 Maximum SFP Temperature ('F) 154.37 151.39 128.68 146.88 180.75 Coincident Time After Reactor Shutdown (hours) 138.0 137.0 131.0 155.0
- 157.0
- Coincident Heat Load to SPF HXs (MBtu/hr) 27.19 27.55 29.32 54.27 47.32 Coincident Evaporation Heat Losses (MBtu/hr) 2.35 2.04 0.57 1.63 8.40 Number of Cooling Trains
- Por Case 3 and Case 4, the coincident time is measured from the shutdown of the Unit 2 reactor (second discharge).
Table 4 - Boiling Times and Maximum Evaporation Rates Case Number CaselA Case1B Case2 Time to Start of Boiling (hours) 9.45 9.89 13.37 6.08 2.95 Maximum Evaporation Rate (gal/min) 63.35 63.40 63.64 103.68 103.85
Table 5 - Maximum Local Pool Water and Fuel Cladding Temperature (Case1A)
No Blockage 50% Blockage Max Local Pool Water Temp
/0~
163.6 223.5 Max Local Fuel Cladding Temp
('F) 214.5 254.3
160 Spent Fuel Pool Bulk Water Temperature Profile Case 1A, Normal Discharge, 1 Cooling Train, Design Flow Rates 155 150 145 4
140 135 130 125 200 0
300 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> Time After Reactor Shutdown Figure 1: SFP Bulk Water Temperature Profile for Case.1A
160 Spent Fuel Pool Bulk Water Temperature Profile Case 1B, Normal Discharge, 1 Cooling Train, Maximum Flow Rates 155 150 145 M
g
~
140 135 130 125 120 200 0
100 300 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />
. Time After Reactor Shutdown Figure 2: SFP Bulk Water Temperature Profile for Case 1B 500
135 Spent Fuel Pool Bulk Water Temperature Profile Case 2, Normal Discharge, 2 Cooling Trains, Design Flow Rates 130
~
125 120 115 110 100 200 0
300 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> Time After Reactor Shutdown Figure 3: SFP Bulk Water Temperature Profile for Case 2
155 150 145 140 Q ~
135 g
~
130 125 Spent Fuel Pool Bulk Water Temperature Profile Case 3, Back-to-Back Discharge, 2 Cooling Trains, Design Flow Rates 115 110 105 800 1200 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> Time After Unit 1 Reactor Shutdown Figure 4: SFP Bulk Water Temperature Profile for Case 3
C
190 Spent Fuel Pool Bulk Water Temperature Profile Case 4, Back-to-Back Discharge, 1 Cooling Train, Design Plow Rates Q'80 170 160 e
M g
~
1SO 140 130 120 110 0
200 400 600 800 1000
. hours Time After Unit 1 Reactor Shutdown Figure 5: SFP Bulk Water Temperature Profile for Case 4 1200 1400
Spent Fuel Pool Decay Heat Load and Loss Profiles Case 1A, Normal Discharge, 1 Cooling Train, Design Flow Rates 35 30 25 g~ )20 ca A
10 200 400 0
100 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> Time After Reactor Shutdown Figure 6: SFP Decay Heat Load and Loss Profiles for Case 1A
Spent Fuel Pool Decay Heat Load and Loss Profiles
- Case 1B, Normal Discharge, 1 Cooling Train, Maximum Plow Rates 35 30 25
)
20 10 400 0
100 200 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> Time After Reactor Shutdown Figure 7: SFP Decay Heat Load and Loss Profiles for Case 1B
Spent Fuel Pool Decay Heat Load and.Loss Profiles Case 2, Normal Discharge, 2 Cooling Trains, Design Plow Rates 35 30 2S
)
20 10 100 400 0
200 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> Time After Reactor Shutdown Figure 8: SFP Decay Heat Load and Loss Profiles for Case 2
Spent Fuel Pool Decay Heat Load and Loss Profiles Case 3, Back-to-Back Discharge, 2 Cooling Trains, Design Flow Rates 60 50 20 10 0
hours Time After Unit 1 Reactor Shutdown Figure 9: SFP Decay Heat Load and Loss Profiles for Case 3 1200
Spent Fuel Pool Decay Heat Load and Loss Profiles Case 4, Back-to-Back Discharge,I Cooling Train, Design Flow Rates 60 50 40 m~30 o g 20 10 0
0 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> Time After Unit 1 Reactor Shutdown Figure 10: SFP Decay Heat Load and Loss Profiles for Case 4
0 A.sigf