ML20117H847

Responds to 960814 Memo Re Petitioners Related Communications to Board,Lewis & Hodgdon 960819 Joint Conference W/Chairman Bollwerk & 960823 Submittal Re NUHOMS Transfer Cask Shield Plug

ML20117H847
Person / Time
Site:	Oyster Creek
Issue date:	08/27/1996
From:	Blake E GENERAL PUBLIC UTILITIES CORP., SHAW, PITTMAN, POTTS & TROWBRIDGE
To:	Bollwerk G, Kelber C, Lam P Atomic Safety and Licensing Board Panel
References
CON-#396-17893 OLA, NUDOCS 9609100050
Download: ML20117H847 (6)

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- . Administrative Judge G. Paul Bollwerk, Chairman Administrative Judge Dr. Peter S. Lam l Administrative Judge Dr. Charles N. Kelber Atomic Safety & Licensing Board l U.S. Nuclear Regulatory Commission

Washington, D.C. 20555 i

In the Matter of GPU Nuclear Corporation (Oyster Creek Nuclear Generating Station)

Docket No. 50-219-OLA i

Chairman Bollwerk and Judges Lam and Kelber

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Upon my retum yesterday from a trip outside the U.S., I learned for the first time of the Board's 1 August 14 Memorandum, Petitioners' related communications to the Board (telecon on August

15 and letter mailed on August 16 and received by Licensee counsel on August 20), Mr. Lewis'

' j and Ms. Hodgdon'sjoint conference with Chairman Bollwerk by telephone on August 19, and an 1 August 23 submittal to NRC Staff by GPU Nuclear concerning the NUHOMS Transfer Cask 3 Shield Plug. i The latter document is enclosed for the information of the Board and participants because it re-

! lates directly to the subject matter of this proceeding. Additionally, I feel constrained to clarify the' substance of settlement discussions between Licensee and Petitioners. l 4

j It is only with reluctance that I discuss the substance of settlement discussions at all. It is my ex-

perience that for well understood and sensible reasons this topic is regarded as confidential be-tween parties. Mr. Gunter and I have discussed the matter of confidentiality of settlement - l l discussions quite explicitly, and a voice mail I had from him on August 15 confirmed this when  !

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he related his intention to call Chairman Bollwerk and also send a short letter saying " simply that 4 no agreement was reached. It doesn't detail anything." To the contrary, however, the Petitioners'

{ ' letter of August 15 reports on a settlement discussion Mr. Gunter and I had immediately follow-

ing the bench conference with the Board at the end of the August 8 prehearing conference.' Peti-j tioners' summary is "Mr. Blake informed Mr. Gunter that the GPUN safety analysis could not be made available to the Petitioners."

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I SHAW, PITTMAN, POTTs & TROWBRIDGE

& PARTNE RSMer thCLUDING PROFESSION AL CORPOmatIONS Administrative Judge G. Paul Bollwerk, Chairman Administrative Judge Dr. Peter S. Lam Administrative Judge Dr. Charles N. Kelber August 27,1996 l

What actually occurred was that I told Mr. Gunter that I believed I had no obligation to provide i him the dose assessment analysis that I had referred to during the prehearing conference, but if he really wanted it to use in subsequent broader hearings we both anticipated may occur on actual transfer cask movements of spent fuel (as opposed to this very particularized DSC shield plug amendment proceeding), I would make it available to him in return for Petitioners' withdrawal from the instant case. He said he would consider the offer and several days later called me to say no. No other settlement discussions have occurred.

Respectfully, Gw.y Ernest L. Blake Counsel for Licensee Enclosure cc: Sersice List h789941/ DOCSDCI

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( GPU Nuclear. Inc U.S. Route #9 Soutn NUCLEAR Post Office Box 388 Forked River, NJ 08731-0388 Tet 609-971-4000 6730-96-2265 August 23, 1996 l

U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555 i 1

Subject:

Oyster Creek Nuclear Generating Station (OCNGS)

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

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 discussed before the Atomic Safety and Licensing Board on August 7,1996 regarding th

, reference license amendment request.

The license amendment request proposes to revise Technical Specification 5.3.1.B to allow l

top shield plug for the dry shielded canister to be moved over spent fuel assemblies in the ca '

while it is in the plant's cask drop protection system. The basis for the request is that not a credible event. Attachment I addresses criticality potential and attachment 2 provides an i

evaluation of radiological consequences for a hypothetical drop of the shield plug.

Sincerely, N MM i Michael B. Roche Vice President and Director Oyster Creek MBR/PFC/ pip Attachments c: Administrator, USNRC Region i USNPC Senior Resident inspector Oyster Creek USNRC Project Manager

r Attachment 1

' Potential for Criticality l

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The potential for criticality as a result of dropping the shield plug tid onto fuel assemblies in t l

dry shielded canister (DSC) was determined by GPU Nuclear based on guidelines pros '

i NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants" All the fuel in the ca i assemblies)is assumed to crush together such that k.a is maxinuzed The impact of the shield plug drop is not considered severe enough to significantly damage the rigid structural material !

j the cask containing the DSC and, therefore, the borated stainless steel plates are expecteI remain intact. An ENC 3e-3f fuel assembly, 7x7 lattice with 2.63 weight % enriched Um was !

used for this analysis 1ts it bounds the reactivity of all fuel available for dry storage.

l Criticality analysis was performed using the Monte-Carlo code KENO-Va using the 27 g!

cross section library collapsed from ENDF-IV. Mixture cross sections were developed usin 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 analy report (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 CSAR compared with 0.882 k.a for the GPU Nuclear analysis including all biases and uncertainties.

Using the above assumptions, the maximum DSC k.e , 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 uncertainties associated with KENO and mechanical uncertainty in the DSC design. The result is conservative because the analysis does not include fuel burnup, which will significantly lower the k.a since burnup averages above 23 GWD/MT for this fuel. In addition, this analysis assumes all bundles in the DSC are affected by the dropped shield plug whereas geo:ne:ric considerations sh 16 bundles would be directly impacted.

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 l

t 8

Attachment 2

Radiological Consequences The radiological consequences of dropping the shield plug onto the NUHOMS*-52B dry shielded j canister (DSC) is bounded by the radiological analysis summarued in Table 2.1-1 of NUREG-0612.

l The NUREG-0612 analysis for fuel that has been subcritical for 120 days indicates that 16,000 fuel

assemblies must be damaged to reach % ofPart 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 1

comparison of assumptions used in the NUREG-0612 analysis to Oyster Creek data is provided in i Table 1. Clearly, the NUREG-0612 analysis is boundmg for Oyster Creek.

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Assuming that all 52 fuel assembles are damaged by the load drop, radiologpcal releases from the fuel

! (muumum ten year decay) will have minimal consequences The fission gases (primanly Kr-85 with a j half-life of 10 years) are released inside secondary contamment. Due to the length of time for decay,

there is no iodine to release. Even ifconservative NUREG-0612 assumptions are applied and fuel only 120 days subcritical assumed, radiological consequences as a result of damage to all 52 fuel assemblies j 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, give n the conservatism ofNUREG-0612 assumptions

! for power level. X/Q, and cooling time relative to acual Oyster Creek data, radiological consequences

{ for a shield plug drop would be expected to be essentiMiy zero.

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Table 1 Comparison of NUREG-0612 Heavy land Drop Assumptions To Oyster Creek Data 4

Parameters NUREG-0612 i Oyster Creek l Power Level (MWm) 3000 t 1930 0-2 hour X/Q, sec/M' l.0x104 34 1.1x10-5151

) (exclusion area boundary) 0-2 hour X/Q LPZ, 1.oxlod 14

{ 1.1x104 151 i sec/M' -

Peaking Factor 1.2 i21 <l .0 l'1 No. of Assemblies in Core 760 560 l Pool Water 1001 'l N/A

( Decontamination Factor Filter Efficiency 95 % l'1

} N/A 3 ElementalIodine 4

Filter Efficiency 95 % N/A Organic iodine

}

Cooling Time (hours) 100 or greater 4.38x10'

Notes.

1. Based cn 5% worst meteorological conditions.

i 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.

i 3. See Regulatory Guide 1.25.

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4. See Regulatory Guide 1.52.

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5. Oyster Creek UFSAR 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 />.

4 j 6. Fuel loaded in cask is fuel discharged at end oflife.

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