ML20212H544

Proposed Tech Specs Reflecting Installation of Addl SFP Storage Racks That Will Accommodate Increase in Spent Fuel Assemblies Beyond Existing Storage Capacity of SFP as Described in Licensing Rept

ML20212H544
Person / Time
Site:	Oyster Creek
Issue date:	06/18/1999
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20137U375	List: ML20212H528 ML20212H544 ML20212H549
References
NUDOCS 9906240020
Download: ML20212H544 (6)

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Revised Technical Specification Pages i

9906240020 990618 ADOCK 05000219 PDR PDR p

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5.3 bUXILIARY EOUiPMENT 5.3.1 Fuel Storage A.

The fuel storage facilities are designed and shall be maintained with a K-effective equivalent to less than or equal to 0.95 including all calculational uncertainties.

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B.

1.

Loads greater than the weight of one fuel assembly shall not be moved over stored I

irradiated fuel in the spent fuel storage facility, except as noted in 5.3.1.B.2.

2.

The shield plug and the associated lifting hardware may be moved over irradiated fuel assemblies that are in a dry shicided canister within the transfer cask in the cask drop protection system.

C.

The spent fuel shipping cask shall not be lifled more than six inches above the top plate of the cask drop protection system. Vertical limit switches shall be operable to assure the six inch vertical limit is met when the cask is above the top plate of the cask drop protection system.

D.

The temperature of the water in the spent fuel storage pool, measured at or near the surface, shall not exceed 125"F.

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The maximum amount of spent fuel assemblics stored in the spent fuel storage pool shall l

be 3035.

BASIS The specification of a K-effective less than or equal to 0.95 in fuel storage facilities assures an ample margin from criticality. This limit applies to unirradiated fuel in both the dry storage vault and the spent fuel racks as well as irradiated fuel in the spent fuel racks. Criticality analyses were performed on the poison racks to ensure that a K-effective of 0.95 would not be exceeded. The analyses took credit for burnable poisons in the fuel and included manufacturing tolerances and uncertainties as described in Section 9.1 of the FSAR. Calculational uncertainties described in 5.3.1.A are explicitly defined in FSAR Section 9.1.2.3.9. Any fuel stored in the fuel storage facilities shall be bounded by the analyses in these reference clocuments.

i The efTects of a dropped fuel bundle onto stored fuel in the spent fuel storage facility has been analyzed.

This analysis shows that the fuel bundle drop would not cause doses resulting from ruptured fuel pins that exceed 10 CFR 100 limits (1,2,3) and that dropped waste cans will not damage the pool liner.

Administrative controls over crane movements, which include mechanical rail stops, serve to prevent travel of the crane outside the analyzed load path over the cask drop protection system. A safety factor greater than 10 with respect to ultimate strength, and redundant shield plug lift cables provide adequate margin for the shield plug lift. These features combined with operator training and required inspections, contribute to the determination that dropping the shield plug onto a loaded dry shielded canister in the spent fuel pool is not a credible event.

l OYSTER CREEK 5.3-1 Amendment No.: 23r76;-7M2h47M37

r The elevation limitation of the spent fuel shipping cask to no more than 6 inches above the top plate of the cask drop protection system prevents loss of the pool integrity resulting from postulated drop accidents.

An analysis of the effects of a 100-ton cask drop from 6 inches has been done (4) which showed that the pool structure is capable of sustaining the loads imposed during such a drop. Limit switches on the crane restrict the elevation of the cask to less than or equal to 6 inches when it is above the top plate.

Detailed structural analysis of the spent fuel pool was performed using loads resulting from the dead weight of the structural elements, the building loads, hydrostatic loads from the pool water, the weight of fuel and racks stored in the pool, seismic loads, loads due to thermal gradients in the pool floor and the walls, and dynamic load from the cask drop accident. Thermal gradients result in two loading conditions; normal operating and the accident conditions with the loss of spent fuel pool cooling. For the normal condition, the reactor building air temperature was assumed to vary between 65 F and i 10"F while the pool water temperature varied between 85'F and 125"F. The most severe loading from the normal operating thermal gradient results with reactor building air temperatures at 65 F and the water temperature at 125"F. Air temperature measurements made during all phases of plant operation in the shutdown heat exchanger room, 1

which is directly beneath part of the spent fuel pool floor slab, show that 65'F is the appropriate minimum i

air temperature. The spent fuel pool water temperature will alarm control room before the water I

temperature reaches 120*F.

i Results of the stmetural analysis show that the pool structure is structurally adequate for the loadings associated with the normal operation and the condition resulting from the postulated cask drop accident (5)

(6). The floor framing was also found to be capable of withstanding the steady state thermal gradient conditions with the pool water temperature at 150 0F without exceeding ACI Code requirements. The walls are also capable of operation at a steady state condition with the pool water temperature at 140 F (5).

Since the cooled fuel pool water returns at the bottom of the pool and the heated water is removed from the surface, the average of the surface temperature and the fuel pool cooling return water is an appropriate estimate of the average bulk temperature; alternately the pool surface temperature could be conservatively used.

References 1.

Amendment No. 78 to FDSAR (Section 7) 2.

Supplement No. I to Amendment No. 78 to the FDSAR (Question 12) 3.

Supplement No. I to Amendment 78 of the FDSAR (Question 40) 4.

Supplement No. I to Amendment 68 of the FDSAR 5.

Revision No. I to Addendum 2 to Supplement No. I to Amendment No. 78 of FDSAR (Questions 5 and 10) 6.

FDSAR Amendment No. 79 7.

Deleted 8.

Holtec Report HI-981983, Revision 4 l

1 OYSTER CREEK

.5.3-2 Amendment No. 121,179,187

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l ENCLOSURE 2 i

i Certificate of Senice t

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UNITED STATES OF AMERICA l

NUCLEAR REGULATORY COMMISSION l

IN Tile MATTER OF DOCKET NO. 50-219 GPU NUCLEAR, INC.

CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No. 261 for Oyster Creek Nuclear Generating Station Technical Specifications, filed with U.S. Nuclear Regulatory Commission on 1999, has this day of f4ue /r

,1999, been served on the Mayor of Lacey Township, Ocean County, New Jersey, and the designated official of the State of New Jersey Bureau of Nuclear Engineering, by deposit in the United States mail, addressed as follows:

i The Honorable William J. Boehm Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731 Mr. Kent Tosch, Director Bureau of Nuc! car Engineering Department of Enviromnental Protection CN 411 Trenton, NJ 08625 h

BY:

Michael B. Roche Vice President and Director Oyster Creek DATE:

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ENCLOSURE 4 Licensing Report for Storage Capacitv Expansion of OCNGS Spent Fuel Pool Holtec Report HI-981983, Revision 4 Non-Proprietary i

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