ML20217J580

Revised Pages to CAA-96-146, Criticality Analysis of Point Beach Nuclear Plant Spent Fuel Storage Racks Considering Boraflex Gaps & Shrinkage W/Credit for Integral Fuel Burnable Absorbers

ML20217J580
Person / Time
Site:	Point Beach
Issue date:	08/05/1997
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20217J575	List: ML20217J580 NPL-97-0416, Forwards Revised Pages to W Criticality Analysis Related to Misload Accident & Required Boron to Offset Accident Re Rev to TS Change Request 193, Nuclear Fuel Storage Enrichments
References
CAA-96-146, CAA-96-146-ERR, NUDOCS 9708140336
Download: ML20217J580 (4)

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.c o-Attachment to NPl. 97-0416 e Revised pages of the Criticality Analysis of the Point Beach Nuclear Plant Spent Fuel Storage Racks

.s 9708140336 DR 970805

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w umni 4.0 Discussion of Postulated Accidents Most accident conditions will not result in an increase in Kg of the rack. Examples are:

Fuel assembly drop The rack structure pertinent for criticality is not excessively deformed on ton of rack and the dropped assembly which comes to rest horizontally on top of the rack has sufficient water te parating it from the active fuel height of stored assemblies to preclude neutronic interaction.

Fuel assembly drop Design of the spent fuel racks is such that it precludes the insertion of a between rack fuel assembly between rack modules.

modules Loss of cooling Reactivity decreases since loss of cooling causes an increase in systems temperature, which causes a decrease in water density, which results in decreased reactivity.

However, two accidents can be postulated which would increase reactivity beyond the analyzed condition. One such postulated accident would be placement of a fresh fuel assembly of the highest permissible enrichment outside and adjacent to a storage rack module. This abnormal location of a fresh fuel assembly could result in an increased reactivity. To very conservatively estimate the reactivity impacts of such an occurrence in the spent fuel racks, the impact of loading a fresh assembly at 4.6 w/o 235 U enrichment adjacent to an outside face, which has no boraaex. of l a 9x9 array of the Point Beach spent fuel rack cells loaded with fresh assemblies at 4.6 w/o 235 U enrichment was determined. The reactivity increase associated with this abnormal placement of l the fresh assembly is less than 0.06985 AK A second accident which could result in increased reactivity would be a "cooldown" event dcing which the pool temperature would drop below 50*F. Calculations show that if the Point Beach spent fuel pool water temperature was to decrease from 50*F to 32*F, reactivity could increase by about 0.00081 AK.

For occurrences of any of the above postulated accidents, the double contingency principle of ANSI /ANS 8.1-1983 can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.

The worth of soluble boron in the Point Beach spent fuel pool has been calculated with PHOENIX-P and is shown in Figure 3 on page 20. As the curves show, the presence of soluble boron in the pool water reduces rack reactivity significantly and is more than sufficient to offset the positive reactivity impacts of any of the postulated accidents. To bound the 0.06985 AK reactivity increase from the most limiting accident in the : pent fuel racks, it is estimated that 700 ppm of soluble boron is required.

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. . Therefore should a postulated accident occur which causes a reactivity increase in the Point Beach spent fuel racks, Ke rr w ill be maintained less than or equal to 0,95 due to the presence of at l least 700 ppm of soluble boron in the spent fuel pool water.

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i.0' Summary of Criticality Results For the storage of fuel assemblies in the spent fuel storage racks, the acceptance criteria for criticality requires the effective neutron multiplication factor, K g,e to be less than or equal to 0.95.

l including uncertainties, under all conditions. Maintaining at least 700 ppm of soluble boron in the spent fuel pool water is required to mitigate the consequences of a postulated accidents as described in Section 4.0 of this report.

This report shows that the acceptance criteria for criticality is met for the Point Beach spent fuel storage racks for the storage of Westinghouse 14X14 OFA and 14x14 STD fuel assemblies with the following configurations and enrichment limits:

Westinghouse Storage of Westinghouse 14x14 OFA and 14x14 STD fuel assemblies 14x14 OFA and with nominal enrichments up to and including 4.6 w/o 23sU utilizing 14x14 STD Fuel all available storage cells is allowed. Fresh fuel assemblies with higher Assemblies initial nominal enrichments up to and including 5.0 w/o 235 U can also be stored in these racks provided a minimum number of IFBAs are present in each fuel assembly. IFBAs consist of neutron absorbing material applied as a thin ZrB2 coating on the outside of the UO2 fuel pellet. As a result, the neutron absorbing materialis a non removable or integral part of the fuel assembly once it is manufactured.

The analytical methods employed herein conform with ANSI N18.21973, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7 Fuel Handling System: ANSI 57.21983," Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations," Section 6.4.2 ANSI N16.9-1975, " Validation of Calculational Methods for Nuclear Criticality Safety"; and the NRC Standard Review Plan, Section 9.1.2, " Spent Fuel Storage".

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