ML17292A376

Forwards Response to RAI Re NRC Bulletin 96-002, Movement of Heavy Loads Over Spent Fuel,Over Fuel in Reactor Core,Or Over Safety-Related Equipment.

ML17292A376
Person / Time
Site:	Columbia
Issue date:	07/19/1996
From:	Parrish J WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GO2-96-141, IEB-96-002, IEB-96-2, NUDOCS 9607290226
Download: ML17292A376 (7)

Text

CATEGORY j.

REGULA%1'Y INFORMATION DZSTRZBUT101%YSTEN (RIDE)

ACCESSION NBR:9607290226 DOC.DATE: 96/07/19 NOTARIZED: YES 'OCKET 6 FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION PARRISH,J.V. Washin'gton Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Forwards response to RAI re NRC Bulletin 96-002, "Movement of Heavy Loads Over Spent Fuel, Over Fuel in Reactor Core, Or Over Safety-Related Equipment."

DISTRIBUTION CODE: ZE11D COPIES RECEIVED:LTR I ENCL  ! SIZE:

TITLE: Bulletin Response (50 DKT)

NOTES: E RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD4-2 PD 1 1 COLBURNFT 1 1

'NTERNA FILE CENT 1 1 NRR/DE/EMEB 1 1 PECB 1 1 NRR/DSSA 1 1 NRR/DSSA/SCSB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SPLB/A 1 1 NRR/DSSA/SRXB 1 1 NRR/PD3-2 1 1 NUDOCS-ABSTRACT 1 1 RES/DSIR/EIB 1 1 RGN4 FILE Ol 1 1 EXTERNAL: NOAC 1 1 NRC PDR 1 1 D

0 E

N NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-5(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 16 ENCL 16

WASHINGTON PUBLIC POWER SUPPLY SYSTEM PO. Box 968 ~ 3000 George Washington Way ~ Richland, Washington 99352-0968 ~ (509) 372-5000 July 19, 1996 G02-96-141 Docket No. 50-397 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Gentlemen:

Subject:

WNP-2, OPERATING LICENSE NPF-21, RESPONSE TO NRC BULLETIN 96-02, "MOVEMENT OF HEAVY LOADS OVER SPENT FUEL, OVER FUEL IN THE REACTOR CORE, OR OVER SAFETY-RELATED EQUIPMENT" - ADDITIONALINFORMATION

References:

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

2) Letter, G02-96-100, dated May 10, 1996, JV Parrish (SS) to NRC, "Response to NRC Bulletin 96-02, Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or Over Safety-Related Equipment" Per the above references, the Supply System has conducted an extensive review of our plans and capabilities for handling heavy loads while the reactor is at power. The review included a walkdown to assess general plant condition in the crane and hoist areas and assess handling system load paths. The review concluded that present activities are in accordance with existing guidelines and our licensing basis, with the exceptions listed in the enclosed attachment.

The review identified procedure enhancements that will be made to ensure consistency and provide commitment clarification. These will be tracked by the Plant Tracking Log (PTL).

In addition, the Supply System will be pursuing future use of a spent fuel dry cask storage facility. As requested by Reference 1, a license amendment request will be submitted 6 to 9 months in advance for planned movement of loads involving the handling of any new 9607290226 9607i9 PDR ADQCK 05000397 8 PDR

Page 2 RESPONSE TO NRC BULLETIN96-02 heavy loads, including dry storage casks, that create a potential load drop accident not previously evaluated in the FSAR, or that require changes to Technical Specifications. If necessary, the capability of performing actions required for safe shutdown in the presence of a source term resulting from a dropped storage cask will also be addressed in the license amendment request.

Should you have any questions or desire additional information regarding this matter, please call me or Ms. L. C. Fernandez (509) 377-4147.

Respectfully, V. Parrish Chief Executive Officer Mail Drop 1023 Attachment cc: LJ Callan - NRC RIV TG Colburn - NRR .

KE Perkins, Jr. -.NRC RIV, WCFO DL Williams - BPA/399 NS Reynolds - Winston & Strawn NRC Sr. Resident Inspector - 927N

STATE OF WASHINGTON )

Subject:

Response to NRC Bulletin 96-02 "Move-

) ment of Heavy Loads Over Spent Fuel, COUNTY OF BENTON ) Over Fuel in the Reactor Core, or Over Safety-Related Equipment" - Additional Information I, J. V. PARRISH, being duly sworn, subscribe to and say that I am the Chief Executive Officer for the WASHINGTON PUBLIC POWER SUPPLY SYSTEM, the applicant herein; that I have the full authority to execute this oath; that I have reviewed the foregoing; and that to the best of my knowledge, information, and belief the statements made in it are true.

DATE ~'P J , 1996 J. V.P sh Chief Executive Officer On this date personally appeared before me J. V. PARRISH, to me known to be the individual who executed the foregoing instrument, and acknowledged that he signed the same as his free act and deed for the uses and purposes herein mentioned.

GIVEN under my hand and seal this ~day of 0 1996.

o Public in 'he STATE OF WASHINGTON Residing a My Commission Expires

RESPONSE TO NRC BULLETIN96-02 Attachment A Page 1 of 2 Plant activities that deviate from descriptions provided in the WNP-2 FSAR:

The FSAR states that new fuel is loaded into the spent fuel pool using the auxiliary hook on the Reactor Building crane, whereas pool side jib cranes have been used since initial plant startup. The jib cnnes are specifically designed for handling new fuel, and for related refueling activities.

The FSAR states that crane hook periodic inspection will be done using the Dye Penetrant (PT) method, whereas piesent practice employs the equivalent Magnetic Particle Testing method.

FSAR changes will be processed and will be tracked by the plant corrective action program and PTL.

2) The impact testing recommendation of ANSI N14.6-1978, as referenced in NUREG-0612, to prevent brittle fracture has not been performed for special lifting devices.

A review of the design adequacy of existing special lifting devices was performed. The review considered device material, the stresses due to the specified load limits, member thickness, and temperatures to which the devices could be exposed. This review concluded that impact testing was not required for one of two reasons:

the stress levels were lower than the threshold stress for performing impact testing; or, the threshold temperatures for performing impact testing are less than iniiurnum plant design temperatures.

Two enhancements are planned as precautionary measures. Administrative controls will be incorporated, where appropriate, to ensure that devices are not used below their threshold temperatures. In addition plant procedures will be revised accordingly to ensure the ANSI N14.6-1978 impact testing recommendation is addressed in the future design, fabrication, and procurement of special lifting devices.

One exception was noted for the lifting device used for the Reactor Recirculation Pump motors which was supplied by General Electric. The thickness of members used is such that impact testing may be required. It willbe necessary to determine the device material to ascertain whether impact testing is required prior to release for use.

RESPONSE TO NRC BULLETIN96-02 Attachment A Page 2 of 2 This device would be used only in the shutdown refueling or defueled mode to resolve a significant equipment problem. Therefore, there are no plans to use this device in the near future. Administrative controls will be imposed to prevent use of this device until this evaluation is complete.

The identified administrative controls and procedure enhancements willbe tracked by the plant corrective action program and PTL.