ML17332A414

Rev 0 to HI-941183, Spent Nuclear Fuel Pool Thermal- Hydraulic Analysis Rept for DC Cook Nuclear Plant

ML17332A414
Person / Time
Site:	Cook
Issue date:	08/25/1994
From:	HOLTEC INTERNATIONAL
To:
Shared Package
ML17332A412	List: ML17332A411 ML17332A413 ML17332A414
References
HI-941183, HI-941183-R02, HI-941183-R2, NUDOCS 9411220367
Download: ML17332A414 (46)

Text

IRISH H 0 LTEC SPENT NUCLEARFUEL POOL THERMAI HYDRAULICANALYSISREPORT for DONALD C. COOK NUCLFARPLANT INDIANAMICHIGANPOWZR COMPANY by HOLTEC INI'ERNATIONAL HOLTEC PROJECT 40224 HOLTEC REPORT HI-941183-REPORT CATEGORY: I AUGUST, 1994 9411220367 941116 PDR ADOCK 050003i5 P

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SUMMARY

OF REVISIONS LOG HOL'IZCREPORT HIM1183 TitlePage Review and Cetti6cation Log Sumnuuy of Revisions Log Section 1 Section 2 Section 3 Section 4 3

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SUMMARY

OF REVISIONS LOG HOLTEC REPORT HI~1183

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~E555 H 0 LTEC REVE@ AND CERTIHCATIONLOG DOCUMENTNAME:

HOLTEC DOCUMENTLD. NUMBER:

HOLTEC PROJECT NUMBER:

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SPENT NUCLEARHJEL POOL TIIERMAL-HYDRAULICANALYSISREPORT for DONALD C. COOK NUCLEARPLANT HI-941183

.40224 INDIANAMICHIGANPOWER COMPANY ISSUE NUMBER ORIGINAL REVISION 1 REVISION 2 REVISION 3 AUTHOR 8r, DATE su~ 7/~ 7/1$

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~~ a~A REVISION 4 REVISION 5 REVISION 6 This document conforms to the requirements of the design speciTication and the applicable sections of the goverrung codes Note:

Signatures and printed names are required in the review block.

~ Must be Project Manager or his designee.

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1.0 In 1992, Donald C. Cook Nuclear Plant received an operating license amendment allowing the twin reactor pool to be reracked with "poisoned" high density racks to store fuel in a Mixed Zone Three Region arrangement.

Under a turnkey contract withHoltec International, Cook Nuclear Plant's owner, Indiana Michigan Power Company, xeradzd the Cook Nuclear Plant spent fuel pool with 23 Bee-standing modules containing a total of 3613 storage cells.

The object of this submittal is to darify.certain ambiguities in the original Licensing Report

'submitted in support of the 1992 license amendment request (Amendments 169 for Unit 1 and 152 for Unit 2) and to provide additional flexibilityin the plant's abBity to discharge fuel into the pool subsequent to a planned (or unplanned) shutdown of a reactor unit.

At the present time, Technical SpeciGcation 3/4.9.3 stipulates a nmumum incore decay after core subcxiticality of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> before any transfer offuel assemblies Rom the reactor to the spent fuel pool.

Considerations of ef5cient outage management waxxant that the plant staff initiate, at its option, fuel transfer 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after core subcriticality. This submittal provides a summa ofthe analyses carried out to demonstrate the acceptability ofreduction ofincore decay time Rom 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

Reducing the incore decay time prior to discharging the spent fuel to the spent fuel pool entails a potential change in the pool bulk temperature.

Inasmuch as the pool bulk temperature affects the thermal moment and shear in the reinforced concrete structure, itis necessary to determine the impact of the proposed. change on the pool structure as welL Computations to establish continued compliance of the'pool structure to the applicable regulatory requirements are also sununarized herein.

e The minor changes to the Licensing Report pertain to clarifying the Boral in-service inspection program, and editorial changes to the number ofcells ascribed to Regions 1, 2 and 3 of the Licensing Report [1] ate also included in this report.

2.0 THERMA HYDRAULICEVALUATION The thermal-hydraulic considerations documented in Section 5.0 ofRef, P] are repeated in this submittal to reflect the changes in (1) the minimum incore decay time and (2) minor revision of the refueling discharge schedule for both units at Cook Nuclear Plant. The methodology and computer codes used in this submittal are identical to those of Ref. [1].

The analysis procedures are summarized in Section 2.1; the discharge scenarios are shown in Section 22, and the results are presented in Section 23.

2.1 Anal s Procedures The thermal-hydraulic evaluation for the spent fuel pool and the rack array consist of the the following discrete steps:

Evaluation of long term decay heat load, which is the accumulating spent fuel decay heat generation based on the existing and the predicted operating cycles at the time instant of the final refueling cycle according to the storage capacity ofthe fuel pool. The heat load is treated as constant to combine with the transient decay heat generated by the final discharge.

Evaluation of the total transient decay heat load including the long term decay heat determined in (i) and the pool bulk temperature as a function of time during the final postulated discharge scenarios.

Evaluation of the time-to-boil ifall forced heat rejection paths from the pool are lost.

(iv)

Determination of the maximum pool local temperature at the instant when the bulk temperature reaches its maximum value.

(v)

Evaluation ofthe maximum fuel cladding temperature to establish that bulk nucleate boiling at any location around the fuel is not possible with cooling available.

Compute the effect of a blocked fuel cell opening on the local water and maximum cladding temperature.

2-1

2.2 Dischar e Scenario The revised existing and projected spent fuel discharge schedules for D. C. Cook spent fuel pool from both units are shown m Table 2.1. The decay heat generation rate in the pool is computed using this data Alldischarge scenaxios considered herein are intended to be predicated on the maximum residual heat load fxompreviously discharged fuel. Accordingly, all four discharge scenarios (Case 1 through 4 below) are considered during a refueling outage close to the end ofthe licensed storage capacity of 3613 cells, when the pool has the highest decay heat generation rate Rom-the'old'fuel stored in the pool. Since the decay heat generation generally depends on both the total number of assemblies in the pool and the decay time of the last discharged batch, three candidate instances of maximum decay heat load exist. Calculations are performed for the decay heat during the refueling of cycle 20B (Unit 2 cycle 20), 25A (Unit 1 Cycle 25), and 21b because they feature different'ombinations of the total number assemblies and the time duration between the outages.

The results indicate that the pool has slightly higher decay heat generation rate from the previously discharged fuel during cycle 20B refueling in December, 2009, compared to the two other candidate cases, and therefore, the discharge scenarios willbe considered during this outage". Please note that this analysis'bounds the conditions up to Cycle 21b, when a hypothetical maximum 3824 spent fuel assembHes willbe in the pool after a back-to-back full core offload.

In this manner, this analysis provides conservative thermal-hydraulic calculation for the entire storage life.

The size of the normal discharge batch is assumed to be 80 assemblies, as was the case in

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the rerack licensing submittal.

CASE 1-Normal Dischar e Sin le Train In cycle 20B refueling (from Unit 2), a total of 80 assemblies are discharged to the pool.

The fuel transfer starts 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown and transfers to the pool at the rate of 4 assemblies per hour. Allthe fuel discharged are assumed to have 1260 EPPD of operation at a rated power of 3411 MW in the reactor. One of the two spent fuel pool 2-2

cooling trains is running to cool the pool. The case is also analyzed for actual measured SFP flowof2800 gpm. The results correspoadiag to design basis Qow (2300 gpm) aad 2800 gpm (actual measures) are labeled as Case 1A and 1B, respectively. The design basis Qow rates are used for all other cases. A maximum of 3399 assemblies (assmne 80 instead of 76 assemblies dischaiged in this batch 20B) are considered in this case.

CASE 2 - Normal Dischar e Both Trains Same,.as. Case

1. except for that two cooling trains are available. Figure 2.1 schematically shows the normal discharge.

CASE 3 - Back-To-Back Full Core 08load Both Trains The Unit 1 reactor has an unplanned shutdown 30 days after the Unit 2 shutdown. ARll core of 193 assemblies are discharged to the pool after the Unit 2 normal discharge. The Rllcore ofQoad starts 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown and transfers fuel assemblies to the pool at the rate of4 assemblies per hour. The average burnup ofthe core is assumed to be that 80, assemblies have 420 EFPD of operatioa in the reactor, and the remaining 113 assemblies are assigned to have 1260 EFPD ofoperation. Two spent fuel pool cooling trains are running to cool the pool. Figure 22 schematically shows the discharge. Amaximum of 3592 assemblies are considered in this scenario.

CASE 4 - Back-To-Back Full Core 081oad Sin e Train Same as Case 3 except only one cooling train is in operation. This case is not a design basis scenario for Cook Nuclear Plant or the USNRC guidelines (NUREG-0800). Itis presented for reference purposes only.

2-3

The calculated maximum accumulating long term decay heat during the outages close to the end ofthe fuel pool storage capacity is 18.15 x 10~ Btu/hr based on the discharge projections shown in Table 2.1. The maximum number of cycles considered is based on the maximum storage capacity of 3613 ceHs. The maximum bulk pool temperature results and the heat loads at the instant ofmaximum temperature are presented in Table 22. The time varying bulk pool temperatures and heat loads in the pool are plotted vs. time-after-shutdown in Figures 2.3 to 2;12. Itis shown from the analyses that the spent fuel pool cooling.system has suf6cient cooling capacity to maintain the spent fuel pool bulk water temperature at or below 161'F (Case 1A) during a normal refueling discharge (80 assemblies), with one or two cooling trains operating, and the net normal heat load, coincident to the maximum water temperature, is 30.8 x 10'tu/hr(excluding evaporation heat losses). Two trains ofthe spent fuel pool cooling system have sufEcient heat removal capacity to maintain the spent fuel pool bulkwater temperature below 151'F (Case 3) during an assumed back-to-back fullcore oQload and the coincident abnormal heat load is 58.7 x 10'tu/hr (excluding evaporation heat losses).

As shown in Table Z2, the previous licensing basis analysis indicated that the maximum normal water temperature was 16(PP. The previous net normal heat load coincident to the maximum water temperature was 30.2 x. 10~ Btu/hr(excluding evaporation heat losses).

Comparison withthe previous rerack submittal analysis bulkpool temperature results (also provided in Table 22) shows that the proposed thermal-hydraulic changes have insigniGcant thermal consequences.

The previous maximum abnormal water temperature was 144'F during an assumed back-to-back full core oQload. The previous coincident abnormal heat load was 50.7 x 10~ Btu/hr (excluding evaporation heat losses).

The losswf~ling events have also been considered for the speci6ed discharge scenarios.

The loss ofall forced cooling is conservatively assumed to occur at the instant of peak pool temperature. Table 2.3 summarizes the results ofthe time-to-boil and maximum evaporation rate under the conservative assumption that no makeup water is provided to the pool. The 2-4

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calculated minimum time f'rom the loss-of-pool cooling until the pool boils for the design bases case is 451 hours0.00522 days <br />0.125 hours <br />7.457011e-4 weeks <br />1.716055e-4 months <br /> (Case 3) and the maximum boiloffrate is 129.23 gpm during the hll core oEoad. The time-to-boil is 728 hrs and maximum boiloffrate is 7222 gpm during the design basis normal discharge.

Consistent with our approach to make the most conservative assessments of temperature, the local water temperature calculations are performed assuming that the pool is at its peak 0

bulk temperature.

Thus, the local water temperature evaluation is, in essence, calculation of the temperature increment over the theoretical spatially uniform value due to local hot spots (due to the presence of a highly heat emissive fuel bundle).

The maximum local water temperature for the limiting case (Case IA) is calculated to be 171.9'F and the maximum local fuel cladding temperature is 224.4'P. Ifthe limiting cells are 50% blocked on the top, the maximum local water temperature becomes 2315'F and the maximum fuel cladding temperature is 264.2'P (see Table 2.4). The local boBing point at the depth of 23 ft of water 8 238'P. Therefore, nucleate boiTing willnot occur even

'round the fuel rods, even under conditions of maximum postulated heat Qux.

2.4 6'ect on Pool Structure Itis recalled from'the rerack licensing submittal that the structural evaluation of the spent

&elpool reinforced concrete structure was based on a temperature differential, AT, of85'P between the inside and outside faces of the pool structure.

A thermal heat Qow path analysis across the reinforced concrete sections forthe highest peak pool bulk temperature case shows 6T to be 69'F. Therefore, the margins ofsafety for the pool structure reported in the rerack submittal continue to bound the actual conditions.

2-5

Z5 Conclusion The foregoing results indicate that the maximum bulkspent fuel pool water temperature is increased by 1'P buxom the previous 160'P to 161'P, Therefore, the margin of safety established in the original rerack license submittal [1] has not been signiGcantly reduced.

2-6

Table 2.1 FUEL CYCLE AND SPENT FUEL DISCHARGE

SUMMARY

UNlT 1 Cyclo BOC Dato EOC Date Cyclo EFPD Discharge Assemblies Cumulative Discharge Into Pool from Unit 1 Total

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Pool Inventory 1A 2A 4A 5A '.

18-Jan-75 20-Feb-77 18-Jun-78 08-Jul-79 04-Aug-80 23-Dec-76 06-Apr-78 06-Apr-79 30-May-80 29-May-81 463 268 217 65 64 65 64 193 322 193 338 494 65 65

. 129 6A 01-Aug-81

'04-Jul-82 64 386 558 7A 16-Sept-82

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17-Jul-83 265 80 466 710 SA 9A

,10A 11A 21-Oct-83 17-Nov-85 05-Oct-87 30-Jun-89 23-Jan-91 06-Apr-85 22-Jun-87.

19-Mar-89 11-Oct-90 22-Jun-92 410 428.5 437 459 80 80 80

)80 80 546 626 786 866 882 1050 1210 1367 1523 2-7

Table 2.1 (continued)

FUEL CYCLE AND SPBNT FUEL DISCHARGE

SUMMARY

UNlT 1 Cyclo BOC Date EOC Date Cycle EFPD Discharge Assemblies Cumulative Discharge Into Pool from Unit 1 Total Pool Inventory 13A 14A 28-OctM 12-Feb-94 11-May-94 05-Jul-95 445 420 80 80 1026 1603 1759 15A 16A 17A 1&A 19A 20A 21A 02-Nov-95 21-Mar-97 08-Aug-98 26-Dec-99 14-May1 0]-OcWQ 24-Mar44

.26-Dec-96 15-May-98 02-Oct-99 18-Feb%1 08-Ju142 25-Nov43 18-May45 420 420 420 420 420 420 420 80 80 80 80 80 80 80 1106 1186 1266 1346 1426 1506 1586 1915 2071 2383 2539 2695 2851 11-Aug45 05~t46 29-Dec46

'2-Feb48 17-May4&

. 11-Ju149 28-Nov-10 420 420 420 420 80 80 t 80 80 1666 1746 1826 1906 3163 3319 3475

. 2-8

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Table 2.1 (continued)

PUBL CYCLE AND SPENT FUEL DISCHARGE

SUMMARY

Cycle BOC Date UNIT2 BOC Date Cyclo BFPD Discharge Assemblies Cumulative Discharge Into Pool

&om Unit 2 Total Pool Inventory 1B 2B 3B 4B 5B 6B 7B SB 9B 10B 11B 12B 13/

14B 15B

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16B 10-Mar-78 18-Jan-80 19-May-81 21-Jan-83 07-Jul-S4 11-Jul-S6 17-Mar-89 10-Nov-90 17-Dec-92 26-Nov-94 14-Apr-96 01-Sap-97 19-Jan89 12-Jul40 29-Nov41 18-Apr43 20-Oct-79 15-Mar-81 22-Nov-82 10-Mar-84 28-Feb-86 01-May-SS 30-Jun-90 20-Feb-92 02-Sep-94 20-Jan-96 0&-Jun-97 26-Oct-98 14-Mar40 05-Sep1 23-Jan43 11-Jun44 396 335 453

.337 428 407

. 420 428 420 420 420 420 420 420 420 80 72 92 88 80 77 76 76 76 76 76 76 76 76 76 80 172 336 424 581 657 733 885 961 1037 1113 1189 1265 273 430 630 970 1130 1287 1443 1679 1835 1991 2147 2303 2459 2615 2771 2-9

Table 2.1 (continued)

FUEL CYCLE AND SPENT FUEL DISCHARGE

SUMMARY

UNIT2 Cycle BOC Date EOC Date Cycle EFPD Discharge Assemblies Cumulative Discharge Into Pool from Unit 2 Total Pool Inventory 17B 18B 19B 20B 21B 22-Jan46 11-Jun47 28-Oct48 21-Apr-10 29-Oct45 18-Mar47 04-Aug48 22-Dec49 15-Jun-11 420 420 420 420 420 76 76 76 76 76 1341 1417 1493 1569 1645 2927 3083 3395 3551 2-10

Table 2.2 MMGMUMSFP BULKPOOL TBMPBRATURBAND COINCIDENTTIME.

Maximum Pool Temp,, 'F Case Number and

.Description 1A (normal discharge, Design Basis Flow) 1B (normal discharge, actual S.F. water Qow) 2 (normal discharge, Design Basis Qow) 3 (Back-to-back fullcore ofQoad) 4 (same as 3, reference case only)

Present Submittal 160.48 157.25 132.26 15057 185.07 Previous Value 15954 15631 13157 143.84

~ 176.91 Present Coincident Time After Reactor

Shutdown, hrs.

136 136 129 155 156 Present Coincident.

Heat Load to SFP HXs 10'tu/hr 30.84 31.28 33.62 58.66 49.87 Present Coincident Evaporation Heat Losses 10'tu/hr 3.14 2.70 0.72 1.96 10.65 Number of Cooling Trains 2-11

Table 2,3 RESULTS OF LOSS-OF-COOLING (No Makeup Water Assumed)

Case

.Number New Computed Value Existing Submittal Time Required for Operator action (hours)

New Maximum Evaporation Rate (GPM) 1A 1B 7.28 7.72 10.58 4,51 1.98 7.82 8,27 11.52 5.74 3.02 72.22

.72,27 72.56 129.23 129.55 2-12

Table 2.4 hGQDMUMLOCALPOOL WATERAND FUEL CLADDINGTEMPERATURE FOR THE LIMITINGCASE

~ (CASE 1A)

Maximum Local Pool Water Temp., 'F Maximum Local Fuel Cladding Temp,,

op No Blockage 50% Blockage 171.9 231.5 224.4 264.2 2-13

HOLTEC INTERNATIONAL NORMALREFUELING DISCHARGE SPENT FUEL INVENTORYBEFORE CYCLE208 OUTAGE 100 HOURS 0

80 ASSEMBIJES OFFLOAD IN20 HRS SCHEDULED REACTOR SHUTDOWN FOR OUTAGE 20B FIGURE 2.$ DONALDC. COOK SPENT FUEL POOL DISCHARGE SCENARIO CASES 1 &2

BATCHDISCHARGE FULL CORE OFFLOAD FROM THE OTHER UNIT 100 HOURS 100 HOURS OFFLOADAT4 ASSEMBUES/HR I

30 DAYS FULLCORE OFFLOADAT4 ASSEMBUES/HR REACTOR SHUTDOWN'EACTOR SHUTDOWN FIGURE 2.2 DONALDC. COOK SPENT FUEL POOL DISCHARGE SCENARIO CASES 3 8,4

HOLTEC INTERNATIONAL DONALD C.

COOK SPENT FUEL POOL NORNAL DISCHARGE (88 ASSENBLIES) 2388 GPN SFP FLOW ONE COOLING TRAIN, CASE 1A REACTOR SHUTDOWN 165 168 LLJ~ 158 OO Q 145 m

'35 8

-188 288 388 488 TINE AFTER REACTOR SHUTDOWN, HRS FIGURE 2.3 SFP BULK WATER TEMPERATURE PROFILE

HOLTEC INTERNATIONAL DONALD C.

COOK SPENT FUEL POOL NORMAL DISCHARGE (88 ASSEN3LIES) 2888 GPJ1 SFP FLOW ONE COOLING TRAIN, CASE 18 REACTOR SHUTDOWN 168

'165 OO 0- 146 148 8

188 288 388 488 TINE AFTER REACTOR SHUTDOWN, HRS FIGURE 2.d SFP BULK WATER TEMPERATURE PROFILE

HOLTEC INTERNATIONAL DONALD C.

COOK SPENT FUEL POOL NORMAL DISCHARGE ( 88 ASSEN3LIES) 2388 GPM SFP FLOW TWO COOLING TRAINS, CASE 2 REACTOR SHUTDOWN 13Ei

~ 138 oOlL 128 8

188 288 388 488 TIME AFTER REACTOR SHUTDOWN, HRS FIGURE 2.6 SFP BULK WATER TEMPERATURE PROFILE

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HOLTEC INTERNATIONAL DONALD C.

COOK SPENT FUEL POOL BACK-TO-BACK FULL CORE OFFLOAD, 2388 GPl1 SFP FLOWiCOOLER TWO COOLING TRAINS, CASE 3 REACTOR SHUTDOWN 168 REACTOR SHUTDOWN I- ~

oO~ 138 128 118 8

488 888 TINE AFTER REACTOR SHUTDOWN, HRS FIGURE 2.6 SFP BULK WATER TEMPERATURE PROFILE 1288

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HOLTEC INTERNATIONAL DONALD C.

COOK SPENT FUEL POOL BACK-TO-BACK FULL CORE OFFLOAD. 2388 GPM SFP FLOWrCOOLER ONE COOLING TRAIN, CASE 0 FOR REFERENCE ONLY REACTOR SHUTDOWN 288 REACTOR SHUTDOWN 188

~ 168 OO 128 8

488 888 TIME AFTER REACTOR SHUTDOWN.

HRS FIGURE 2.7 SFP BULK WATER TEMPERATURE PROFILE 1288

&1

HOLTEC INTERNATIONAL DONALD C.

COOK SPENT FUEL POOL NORMAL DISCHARGE (

88 ASSEMBLIES )

2388 GPM SFP FLOW r COOLER ONE COOLING TRAIN, CASE )A REACTOR SHUTDOWN

4. 88E+7

, NET HEAT LOAD

3. 88E+7 o~ 2.88E+7

1. 88E+7 EVAPORATION HEAT LOSSES 8

$ 88 288 388 488 688 TIME AFTER REACTOR SHUTDOWN, HRS FIGURE 2.8 SFP NET DECAY HEAT LOAD AND HEAT LOSSES FOR CASE iA

HOLTEC INTERNATIONAL DONALD C.

COOK SPENT FUEL POOL NORMAL DISCHARGE ( 88 ASSEMBLIES )

2888 GPM SFP FLOW r COOLER ONE COOLING TRAIN. CASE 18 REACTOR SHUTDOWN

d. 88E+7 NET HEAT LOAD

3. 88E+7

~~ 2. 88E+7

1. 88E+7 EVAPORATION HEAT LOSSES 8

188 288 388 488.

688 TIME AFTER REACTOR SHUTDOWN.

HRS

-'IGURE 2.9 SFP NET DECAY HEAT LOAD AND HEAT LOSSES FOR CASE 1B

HOLTEC INTERNATIONAL DONALD C.

COOK SPENT FUEL POOL NORMAL DISCHARGE (

88 ASSEMBLIES ), 2388 GPM SFP FLOW z COOLER TWO COOLING TRAINS, CASE 2 REACTOR SHUTDOWN

d. 88E+7 NET HEAT LOAD

3. 88E+7

~o 2'88E+7

1. 88E+7 EVAPORATION HEAT LOSSES 8

188 288 388 488 688 TIME AFTER REACTOR SHUTDOWN, HRS FIGuRE 2.18 SFP NET DECAY HEAT LOAD AND HEAT LOSSES FOR CASE 2

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HOLTEC INTERNATIONAL

. DONALD C.

COOK SPENT FUEL POOL BACK-TO-BACK FULL CORE OFFLOAD, 2388 GPM SFP FLOW r COOLER TWO COOLING TRAINS, CASE 3 REACTOR SHUTDOWN REACTOR SHUTDOWN

6. 88E+7

d. 88E+7 NET HEAT LOAD

2. 88E+7 EVAPOR TION I HEAT LOSSES 8.88E+8 8

488 888 1288 TIME AFTER REACTOR SHUTDOWN, HRS FIGURE 2.11 SFP NET DECAY HEAT LOAD AND HEAT LOSSES FOR CASE 3

HOLTEC INTERNATIONAL DONALD C.

COOK SPENT FUEL POOL BACK-TO-BACK FULL CORE OFFLOAD, 2388 GPN SFP FLOW r COOLER ONE COOLING TRAIN, CASE 4 FOR REFERENCE ONLY REACTOR SHUTDOWN REACTOR SHUTDOWN

6. 88E+7 4.SHE+7 3

NET HEAT'OAD

2. 88E+7 EVAPOR TION HEAT LOSSES E+8

~ 8 488 888 1288 TINE AFTER REACTOR SHUTDOWN, HRS FIGURE 2.12 SFP NET DECAY HEAT LOAD AND HEAT LOSSES FOR CASE 4

3.0 Referring to Holtec Report HI-90488, submitted as an attachment to the 1992 licensing submittal (Amendment 169 for Unit 1 and 152 for Unit 2), the following two editorial changes are documented herein.

ao Number of different cell types: Figure 4.1 ofthe Licensing Report provided the storage pattern for Regions 1, 2 and 3 cells.

While the storage cell designations in that Ggure are correct, the total cell counts next to.the legend are not. The correct counts aie as follows:

Region 1: 503 cells Region 2: 1440 cells Region 3: 1670 cells Figure 4.1 (revised) is attached herein.

b.

Poison Surveillauce Program:

The Boral surveillance program presented in

, Section 10 of the rerack licensing report [1] is somewhat unclear with respect to coupon pre-characterization and post-irradiation tests.

The following paragraph is intended to clarify this item.

All 12 coupons presently installed in the Cook Nuclear Plant fuel pool have been pre-characterized by measuring their length, width, and their thickness

't discrete need locations.

In addition, their neutron transmission.'haracteristics at discrete marked points have also been quantiGed using standard Holtec quality procedures for coupon testing.

This pre-characterization data will serve as benchmark for future post-inadiation evaluations.

The coupon tree willbe placed in a storage cell, normally used for storing spent nuclear fuel, such that the coupons are exposed to as high a gamma Geld as practicable. At the time of the second discharge into the pool, number one coupon from the tree willbe removed and the tree reinstalled in a storage cell, such that the coupons will, once again, continue to receive as much gamma dose as is practicable (this is evidently realized by placing the tree in a storage location which is surrounded by &eshly discharged fuel).

3-1

As a aunimum, the coupon removed &om the tree willbe measured to determine its variation in length, width, and thickness (at the pre-calibrated locations). Ifthese physical dimensions exhibit less than 1%

variation, then no further testing will be done.

However, if the measured variation in any ofthe physical dimensions exceeds 1%, then the neutron transmission ability of the coupon (at the pre-calibrated locations) willbe measured. Ifthe post-irradiation neutron attenuation is not less than 95% of the benctunark (pre-characterized value), then no Rrther action willbe necessary.

However, ifthe coupon oils to muster neutron attenuation acceptance capability, then it will be destructively tested to obtain a direct measure ofits areal boron density by using the wet chemistry method.

Should the measured boron density be found to be less than the stipulated licensing basis minimum

(.030 gm/sq.cm. B-10), then the condition would w:mant immediate reappraisal ofcriticality compliance ofthe storage system.

The Plant's standard reporting procedures for such discrepant situations will be followed.

It should be added that no plant has experienced this situation after over 200 pool years of experience with Boxal.

The schedule of coupon surveillance is provided in Table 3.1.

3~2

Table 3.1 SCHEDULE OP COUPON SURVEILLANCE COUPON PERIOD Fall of 1994 1 to 2 years...<'> o) 3 to 5 years 6 to 8 years..."> o) 9 to 11'ears

...after removal of Coupon No. 1.

The coupon shall be removed one or two months preceding a reactor refueling (either Unit One or Two).

Coupon tree willbe moved to a region of high gamma Qux during the reactor refueling outage (i.e., surrounded by &eshly discharged fuel) when a coupon has been removed Rom the pool.

Repeat the test every Gve years for the remaining duration of wet. storage in the Donald C. Cook spent fuel pool.

3-3

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4.0 REFERENCES

f1]

Letter from E.E. Fitzpatrick to T.E Mulrey, VSNRC, AEP: NRQ 1146, dated July 26, 1991 and attachments (includes Holtec Licensing Report HI-90488 as one of the attachments).