ML20117E207

Forwards Results of Analyses Discussed Before ASLB on 960807 Re 960415 TS Change Request 244

ML20117E207
Person / Time
Site:	Oyster Creek
Issue date:	08/23/1996
From:	Roche M GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
6730-96-2265, NUDOCS 9608290180
Download: ML20117E207 (4)

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GPU Nuclear,Inc.

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U.S. Route #9 Scuth NUCLEAR Post Office Box 388 Forked River, NJ 087310388 Tel 609-9714000 6730-96-2265-August 23, 1996 i

U. S. Nuclear Regulatory Commission Attn.: Document Control Desk i

Washington, DC 20555

Subject:

Oyster Creek Nuclear Generating Station (OCNGS)

Docket No. 50-219 NUHOMS* Transfer Cask Shield Plug i

Reference:

Technical Specification Change Request No. 244 dated 4/15/96 from Michael B.

Roche (GPU Nuclear) to USNRC As requested by the NRC staff, the attachments to this letter provide the results of the analyses discussed before the Atomic Safety and Licensing Board on August 7,1996 regarding the reference license amendment request.

The license amendment request proposes to revise Technical Specification 5.3.1.B to allow the top shield plug for the dry shielded canister to be moved over spent fuel assemblies in the canister while it is in the plant's cask drop protection system. The basis for the request is that a drop is not a credible event. Attachment I addresses criticality potential and attachment 2 provides an evaluation of radiological consequences for a hypothetical drop of the shield plug.

j Sincerely, 7kLftsPs Michael B. Roche Vice President and Director Oyster Crcek MBR/PFC/ pip Attachments c:

Administrator, USNRC Region I g

USNRC Senior Resident Inspector 1,

0 Oyster Creek USNRC Project Manager C

9608290180 960823 PDR ADOCK 05000219

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Potential for Criticality The potential for criticality as a result of dropping the shield plug lid onto fuel assemblies in the dry shielded canister (DSC) was determined by GPU Nuclear based on guidelines provided in NUREG-0612, " Control oflleavy Loads at Nuclear Power Plants" All the fuel in the cask (52 assemblies)is assumed to crush together such that k.ais maximized. The impact of the shield i

plug drop is not considered severe enough to significantly damage the rigid structural material of the cask containing the DSC and, therefore, the borated stainless steel plates are expected to remain intact. An ENC 3e-3f fuel assembly,7x7 lattice with 2.63 weight % enriched Um was used for this analysis as it bounds the reactivity of all fuel available for dry storage.

Criticality analysis was performed using the Monte-Carlo code KENO-Va using the 27 group cross section library collapsed from ENDF-IV. Mixture cross sections were developed using the material information processor sequence CSAS25 of SCALE 4.2 which uses the BONAMI, NITAWL and ICE modules. In order to insure confidence in the cross section library and the KENO model of the DSC a comparison was made to results found in the cask safety analysis repon (CSAR)(Section 3.3) for a GE 7x7 4.0 weight % enriched bundle at beginning of core life.

Results compared well, with a 0.880 k.a in the CS AR compared with 0.882 k.a for the GPU Nuclear analysis including all biases and uncenainties.

Using the above assumptions, the maximum DSC k a, with a 95/95 confidence level, as a result of the dropped shield plug was determined to be 0.957. This includes all biases and uncenainties associated with KENO and mechanical uncertainty in the DSC design. The result is conservative because the analysis does not include fuel bumup, which will significantly lower the k a since burnup averages above 23 GWD/MT for this fael. In addition, this analysis assumes all bundles in the DSC are affected by the dropped shield plug whereas geometric considerations show that only 16 bundles would be directly impacted.

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References:

1) NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants"

2) Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System, Revision 3

1

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Radiological Consequences The radiological consequences of dropping the shield plug onto the NUHOMS*-52B dry shielded canister (DSC) is bounded by the radiological analysis summarized in Table 2.1-1 of NUREG-06 2.

The NUREG-0612 analysis for fuel that has been subcritical for 120 days indicates that 16,000 fuel assemblies must be damaged to reach % of Part 100 exposure limits, or 75 rem thyroid and 6.25 rem whole body. The DSC contains 52 fuel assemblies and is in the transfer cask located in the Oyster Creek Cask Drop Protection System (CDPS) twenty feet below the surface of the spent fuel pool. A comparison of assumptions used in the NUREG-0612 analysis to Oyster Creek data is provided in Table 1. Clearly, the NUREG-0612 analysis is bounding for Oyster Creek.

Assuming that all 52 fuel assembles are damaged by the load drop, radiological releases from the fuel (minimum ten year decay) will have minimal consequences. The fission gases (primarily Kr-85 with a half-life of 10 years) are released inside secondary containment. Due to the length of time for decay, there is no iodine to release. Even ifconserva*ive NUREG-06 2 assumptions are applied and fuel only i

120 days subcritical assumed, radiological consequences as a result of damage to all 52 fuel assemblies in the DSC could be no more than 20 millirem whole body dose at the site boundary. For the case of 16 directly impacted fuel assemblies (maximum possible due to shield plug drop), the whole body dose could be no more than 6.25 millirem. In reality, given the conservatism of NUREG-0612 assumptions for power level, X/Q, and cooling time relative to actual Oyster Creek data, radiological consequences for a shield plug drop would be expected to be essentially zero.

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Table 1 Comparison of NUREG-%12 Ileavy Load Drop Assumptions To Oyster Creek Data Parameters NUREG-0612 Oyster Creek Power Level (MWm) 3000 1930 0-2 hour X/Q, sec/M' l.0x 10-3 In 1.Ix105 isj (exclusion area boundary) 0-2 hour X/Q LPZ, 1.0x10" 14 1.1x10 l53 sec/M' l

Peaking Factor 1.2 (2i

< l.0 163 No. of Assembliesin Core 760 560 Pool Water 100 'l N/A l

Decontamination Factor l

Filter Efficiency 95 % 'l N/A Elemental Iodine Filter Efliciency 95 %

N/A Organic Iodine 4

Cooling Time (hours) 100 or greater 4.38x10 l

Notes:

1.

Based on 5% worst meteorological conditions.

l 2.

Value is 1.2 for greater than one damaged fuel assembly. For a single assemble the values are 1.65 and 1.5 for PWRs and BWRs, respectively.

3.

See Regulatory Guide 1.25.

4.

See Regulatory Guide 1.52.

5.

Oyster Creek UFS AR Table 2.3-30 for an elevated level release. Values based on Regulatory Guide 1.145 rev 1. LPZ (Low Population Zone) value is for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

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6.

Fuel loaded in cask is fuel discharged at end oflife.