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 NUCLEAR FUEL POOL THERMAI HYDRAULICANALYSIS REPORT for DONALD C. COOK NUCLFAR PLANT INDIANAMICHIGANPOWZR COMPANY by HOLTEC INI'ERNATIONAL HOLTEC PROJECT 40224 HOLTEC REPORT HI-941183-REPORT CATEGORY: I AUGUST, 1994 9411220367 941116 PDR P

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SUMMARY

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SUMMARY

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Title Page Review and Cetti6cation Log Summaty of Revisions Log Section 1 Section 2 Section 3 Section 4

~ E555 H 0 LTEC REVE@ AND CERTIHCATION LOG DOCUMENT NAME: SPENT NUCLEAR HJEL POOL TIIERMAL-HYDRAULICANALYSIS REPORT for DONALD C. COOK NUCLEAR PLANT HOLTEC DOCUMENT LD. NUMBER: HI-941183 HOLTEC PROJECT NUMBER: ~ 44'jlf/~wg~iw .40224 CUSTOMER/CLIENT: INDIANAMICHIGANPOWER COMPANY REVISION BLOCK ISSUE AUTHOR 8r, REVIEWER 8c QA APPROVED NUMBER DATE DATE MANAGER 8c DATE 8c DATE ga.Af CO+

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REVISION 2 8/~ $ '0 ~~ a~A REVISION 3 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.

KC

'I

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 with Holtec 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 flexibility in 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 of fuel 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 of the analyses carried out to demonstrate the acceptability of reduction of incore 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, it is 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.

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

e 3

2.0 THERMA HYDRAULICEVALUATION The thermal-hydraulic considerations documented in Section 5.0 of Ref, 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 of the 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 if all 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 of the 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 All discharge scenaxios considered herein are intended to be predicated on the maximum residual heat load fxom previously discharged fuel. Accordingly, all four discharge scenarios (Case 1 through 4 below) are considered during a refueling outage close to the end of the 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 will be 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 ~

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. All the 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 flow of 2800 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. A Rll core of 193 assemblies are discharged to the pool after the Unit 2 normal discharge. The Rll core 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 of 4 assemblies per hour. The average burnup of the 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 of operation. Two spent fuel pool cooling trains are running to cool the pool. Figure 22 schematically shows the discharge. A maximum 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). It is presented for reference purposes only.

2-3

The calculated maximum accumulating long term decay heat during the outages close to the end of the 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 of maximum 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. It is 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 of the spent fuel pool cooling system have sufEcient heat removal capacity to maintain the spent fuel pool bulk water temperature below 151'F (Case 3) during an assumed back-to-back full core 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 with the previous rerack submittal analysis bulk pool 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 of all forced cooling is conservatively assumed to occur at the instant of peak pool temperature. Table 2.3 summarizes the results of the 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 will not occur even

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

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

&el pool reinforced concrete structure was based on a temperature differential, AT, of 85'P between the inside and outside faces of the pool structure. A thermal heat Qow path analysis across the reinforced concrete sections for the highest peak pool bulk temperature case shows 6T to be 69'F. Therefore, the margins of safety 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 bulk spent 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 Cumulative Total Assemblies Discharge ~

Pool Into Pool Inventory from Unit 1 1A 18-Jan-75 23-Dec-76 463 65 65 65 64 .

2A 20-Feb-77 06-Apr-78 129 18-Jun-78 06-Apr-79 193 193 4A 08-Jul-79 30-May-80 268 65 338 5A '. 04-Aug-80 29-May-81 217 64 322 494 6A 01-Aug-81 '04-Jul-82 64 386 558 7A 16-Sept-82 ~ 17-Jul-83 265 80 466 710 SA 21-Oct-83 06-Apr-85 410 80 546 882 9A 17-Nov-85 22-Jun-87. 80 626 1050

,10A 05-Oct-87 19-Mar-89 428.5 80 1210 11A 30-Jun-89 11-Oct-90 437 )80 786 1367 23-Jan-91 22-Jun-92 459 80 866 1523 2-7

Table 2.1 (continued)

FUEL CYCLE AND SPBNT FUEL DISCHARGE

SUMMARY

UNlT 1 Cyclo BOC Date EOC Date Cycle EFPD Discharge Cumulative Total Pool Assemblies Discharge Inventory Into Pool from Unit 1 13A 28-OctM 12-Feb-94 445 80 1603 14A 11-May-94 05-Jul-95 420 80 1026 1759 15A 02-Nov-95 .26-Dec-96 420 80 1106 1915 16A 21-Mar-97 15-May-98 420 80 1186 2071 17A 08-Aug-98 02-Oct-99 420 80 1266 1&A 26-Dec-99 18-Feb%1 420 80 1346 2383 19A 14-May1 08-Ju142 420 80 1426 2539 20A 0]-OcWQ 25-Nov43 420 80 1506 2695 21A 24-Mar44 18-May45 420 80 1586 2851 11-Aug45 05~t46 420 80 1666 29-Dec46 '2-Feb48 420 80 1746 3163 17-May4& . 11-Ju149 420 t 80 1826 3319 28-Nov-10 420 80 1906 3475

. 2-8

0 Table 2.1 (continued)

PUBL CYCLE AND SPENT FUEL DISCHARGE

SUMMARY

UNIT 2 Cycle BOC Date BOC Date Cyclo BFPD Discharge Cumulative Total Assemblies Discharge Pool Into Pool Inventory

&om Unit 2 1B 10-Mar-78 20-Oct-79 396 80 80 273 2B 18-Jan-80 15-Mar-81 335 172 430 3B 19-May-81 22-Nov-82 453 72 630 4B 21-Jan-83 10-Mar-84 .337 92 336 5B 07-Jul-S4 28-Feb-86 88 424 970 6B 11-Jul-S6 01-May-SS 428 80 1130 7B 17-Mar-89 30-Jun-90 407 77 581 1287 SB 10-Nov-90 20-Feb-92 . 420 76 657 1443 9B 17-Dec-92 02-Sep-94 428 76 733 1679 10B 26-Nov-94 20-Jan-96 420 76 1835 11B 14-Apr-96 0&-Jun-97 420 76 885 1991 12B 01-Sap-97 26-Oct-98 420 76 961 2147 13/ 19-Jan89 14-Mar40 420 76 1037 2303 14B 12-Jul40 05-Sep1 420 76 1113 2459 15B ~ 29-Nov41 23-Jan43 420 76 1189 2615 16B 18-Apr43 11-Jun44 420 76 1265 2771 2-9

Table 2.1 (continued)

FUEL CYCLE AND SPENT FUEL DISCHARGE

SUMMARY

UNIT 2 Cycle BOC Date EOC Date Cycle EFPD Discharge Cumulative Total Assemblies Discharge Pool Into Pool Inventory from Unit 2 17B 29-Oct45 420 76 1341 2927 18B 22-Jan46 18-Mar47 420 76 1417 3083 19B 11-Jun47 04-Aug 48 420 76 1493 20B 28-Oct48 22-Dec49 420 76 1569 3395 21B 21-Apr-10 15-Jun-11 420 76 1645 3551 2-10

Table 2.2 MMGMUMSFP BULKPOOL TBMPBRATURB AND COINCIDENT TIME .

Maximum Pool Temp,, 'F Present Present Present Case Number and Coincident Coincident. Coincident Number of

.Description Time After Heat Load to Evaporation Cooling Present Previous Reactor SFP HXs Heat Losses Trains Submittal Value Shutdown, 10'tu/hr 10'tu/hr hrs.

1A 160.48 . 15954 136 30.84 3.14 (normal discharge, Design Basis Flow) 1B 157.25 15631 136 31.28 2.70 (normal discharge, actual S.F. water Qow) 2 132.26 13157 129 33.62 0.72 (normal discharge, Design Basis Qow) 3 15057 143.84 155 58.66 1.96 (Back-to-back full core ofQoad) 4 185.07 ~

176.91 156 49.87 10.65 (same as 3, reference case only) 2-11

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

Case Time Required for Operator action (hours) New Maximum

.Number Evaporation Rate New Computed Value Existing Submittal (GPM) 1A 7.28 7.82 72.22 1B 7.72 8,27 .72,27 10.58 11.52 72.56 4,51 5.74 129.23 1.98 3.02 129.55 2-12

Table 2.4 hGQDMUM LOCAL POOL WATER AND FUEL CLADDINGTEMPERATURE FOR THE LIMITINGCASE

~

(CASE 1A)

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

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

HOLTEC INTERNATIONAL NORMAL REFUELING DISCHARGE SPENT FUEL INVENTORYBEFORE CYCLE 208 OUTAGE 100 HOURS 0

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

BATCH DISCHARGE FULL CORE OFFLOAD FROM THE OTHER UNIT 100 HOURS 100 HOURS FULL CORE OFFLOAD AT 4 ASSEMBUES/HR OFFLOAD AT 4 ASSEMBUES/HR I

30 DAYS REACTOR SHUTDOWN 'EACTOR SHUTDOWN FIGURE 2.2 DONALD C. 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 O

O 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 O

O 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 oO lL 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 REACTOR SHUTDOWN 168 I- ~

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~ 138 128 118 8 488 888 1288 TINE AFTER REACTOR SHUTDOWN, HRS FIGURE 2.6 SFP BULK WATER TEMPERATURE PROFILE

I A

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 REACTOR SHUTDOWN 288 188

~ 168 O

O 128 8 488 888 1288 TIME AFTER REACTOR SHUTDOWN. HRS FIGURE 2.7 SFP BULK WATER TEMPERATURE PROFILE

&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 )

ONE COOLING TRAIN. CASE 18 2888 GPM SFP FLOW r COOLER 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 2.9 SFP NET DECAY HEAT LOAD AND HEAT LOSSES FOR CASE 1B

-'IGURE

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

'r A~

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 NET HEAT'OAD 3

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 of the 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 at discrete marked points have also been quantiGed using transmission.'haracteristics standard Holtec quality procedures for coupon testing. This pre-characterization data will serve as benchmark for future post-inadiation evaluations.

The coupon tree will be 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 of the 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 will be necessary. However, if the coupon oils to muster neutron attenuation acceptance capability, then it will be destructively tested to obtain a direct measure of its 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 of criticality compliance of the 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 will be 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).