JAFP-97-0083, Forwards Response to NRC RAI Re Bulletin 96-002, Movement of Dry Storage Casks Over Spent Fuel,Fuel in Reactor Core or Safety-Related Equipment

Forwards Response to NRC RAI Re Bulletin 96-002, Movement of Dry Storage Casks Over Spent Fuel,Fuel in Reactor Core or Safety-Related Equipment

ML20136F529
Person / Time
Site:	FitzPatrick
Issue date:	03/05/1997
From:	Michael Colomb POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
IEB-96-002, IEB-96-2, JAFP-97-0083, JAFP-97-83, TAC-M97448, NUDOCS 9703140166
Download: ML20136F529 (4)

Text

. _ . ._. - .

, . . JimJs A. Fit 2 Patrick

. Nucittr Pow 2r Plint P.O. Box 41

- Lycoming. New York 13093

, 315-342-3840 i I

I Michael J. Colomb 4# Authon.ty sne aecuse Omce, March 5,1997 JAFP-97-0083 U.S. Nuclear Regulatory Commission ,

Attn: Document Control Desk l Mail Station P1-137 l l Washington, D.C. 20555 l

SUBJECT:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Response to Reauest for Additional Information Renardina NRC Bjgetin 96-02

References:

1. NRC Letter to Mr. William J. Cahill, Jr., " Request for Additional Information Related to Bulletin 96-02, ' Movement of Dry Storage Casks Over Spent Fuel, Fuel in the Reactor Core, or Safety-Related Equipment' (TAC NO. M97448)," dated January 9,1997.

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2. NRC Bulletin 96-02, " Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or Over Safety-Related Equipment," dated April 11,1996.

3. NYPA Letter, Michael J. Colomb to the NRC (JAFP-96-0200),

" Response to NRC Bulletin 96-02," dated May 10,1996.

Dear Sir:

This letter is the Authority's 60-day response to an NRC Request for Additional Information (RAl) (Refererue 1) regarding NRC Bulletin 96-02 (Reference 2) for the James A. FitzPatrick Nuclear Power Plant. Reference 1 requests the Authority to evaluate a potential spent fuel cask drop scenario while the reactor is at power (i.e., in all modes other than cold shutdown, refueling, and defueled) and forward the results of this evaluation to the NRC staff for their review. The NRC staff requests that this evaluation include the FitzPatrick crane design, load path, and spent fuel cask loading and unloading processes that supports a determination that the potential spent fuel cask drop scenario is not credible at FitzPatrick. If the evaluation results show that the potential spent fuel cask drop scenario is credible at FitzPatrick, further evaluation results are requested by the NRC.

The Authority stated in Reference 3 that FitzPatrick did not plan to move spent fuel storage casks over spent fuel, fuel in the reactor core, or safety-related equipment at power for a two year time period. In addition, the Authority stated in Reference 3 that an amendment request would be submitted to the NRC 6-9 months prior to moving spent fuel storage casks at power. Since no moves will occur, the potential spent fuel cask drop /

scenario is not credible at this time. ,i .[

Q l1 9703140166 970305  !

PDR ADOCK 05000333  %,E%%%,E%,E%E PDR ,

, U.S. Nuclear Regul: tory Commission Attn: Document Control Desk

Subject:

Response to RAI Regarding NRC Bulletin 96-02 Page ' 2-4 The Reactor Building Crane at FitzPatrick does not meet the requirements of a single failure proof crane. The Authority is currently reviewire the crane design to determine whether an upgrade to a single failure proof crane design is necessary for any future moves of spent fuel storage casks at power. The Authority will provide the NRC with the information requested in Reference 1, which includes crane design, in any future submittal regarding the movement of spent fuel storage casks at power. This commitment is identified in Attachment 1 (List of Commitments) of this letter and enhances Commitment Number 2 in Reference 3.

If you have any questions, please contact Mr. A. Zaremba at (315) 349 6365.

Very Truly Yours, l

( -

i MICHAEL J. COLOMB Site Executive Officer l MJC:JJC:las Att.: As stated cc: U.S. Nuclear Regulatory Commission Regional Administrator 475 Allendale Road King of Prussia, PA 19406 U.S. Nuclear Regulatory Commission Office of the Resident inspector P.O. Box 136 Lycoming, New York 13093 U.S. Nuclear Regulatory Commission Ms. K. Cotton, Acting Project Manager Project Directorate 1-1 Division of Reactor Projects 1/11 Mail Stop 14 B2 Washington, DC 20555 l

i ATTACHMENT 1 TO JAFP-97 0083 LIST OF COMMITMENTS Commitment No. Description Due Date JAFP 97 0083-01 Provide the NRC staff with the following 6-9 months prior information regarding the movement of spent fuel to moving spent storage casks at power: fuel storage casks at power.

1. An evaluation of the FitzPatrick crane design, load path, and cask loading and unloading processes that supports a determination that the scenario described in the RAI dated 1/9/97 is not credible at FitzPatrick.

OR

2. If the Authority determines that the event is credible, provide the following information to the NRC staff:

(a) An analysis of a possible drop of a spent fuel storage or transportation cask involving a drop that results in the tipping over of the spent fuel cask, loss of the cask lid, or loss of the cask lid and ejection of the spent fuel from the cask into the spent fuel pool or areas adjacent to the pool.

This load drop / consequence analysis should include a dose analysis to personnelinvolved in the cask movement for the time immediately following the accident. Also, the analysis should address personnel exposure resulting from required entry into plant areas affected by the event and the impact of elevated dose fields on the ability to reach safe shutdown or continue normal plant operation.

.- ATTACHMENT 1 TO JAFP-97-0083 LIST OF COMMITMENTS Commitment No. Description Due Date JAFP 97 0083-01 (b) An evaluation addressing the potential for 6-9 months prior (cont'd) criticality resulting from the postulated to moving spent i cask drop accident scenario described fuel storage casks l above. I t power.

(c) An evaluation that addresses possible 1

means of recovering from the postulated caak drop accident scenario described above.

(d) An evaluation that addresses whether the 4

potential impact of the scenario described

above on other parts of the facility (e.g.,

the spent fuel pool) is bounded by previous load drop analyses.

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