ML20198N341

Submits Interim Rept on Plant Efforts to Address NRC Bulletin 96-003, Potential Plugging of ECC Suction Strainers by Debris in BWRs & to Demonsrate Compliance with Corollary Requirements of Reg Guide 1.82,rev 2

ML20198N341
Person / Time
Site:	Pilgrim
Issue date:	12/30/1998
From:	Alexander J BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-REGGD-01.082, RTR-REGGD-1.082 BECO-LTR-2.98.1, IEB-96-003, IEB-96-3, NUDOCS 9901060101
Download: ML20198N341 (5)

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. 0: NRC Bullstin 96 '

Pilgrim Nuclear Power Station

. Rocky Hill Road Plymouth, Massachusetts 02360-5599 i

December 30,1998  !

BECo Ltr.-2.98.167 U.S. Nuclear' Regulatory Commission Attn.: Document Control Desk Washington, D.C. 20555 Docket No. 50-293 License No. DPR-35 Summary Final Response to NRCB 96-03

1

~ This letter is an interim report on Pilgrim efforts to address NRC Bulletin (NRCB) 96-03,

" Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-

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. Water Reactors,";and to demonstrate compliance with the corollary requirements of -

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. Regulatory Guide .1.82, Revision 2, i Water Sources For Long-Terrn Recirculation Cooling Following A Loss-Of-Coolant Accident." : A more detailed submittal will be made after completion of additional analysis that was identified as necessary. in the

' final stages of preparing the final report required by NRCB 96-03.

Due to concerr5 identified in Bulletin 96-03, Pilgrim installed two large passive

- emergency core cooling system (ECCS) suctionistrainers, during refueling outage

-(RFO) 11, which was completed.in the Spring of 1997.. The new strainers and their ff  ;

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effect con Pilgrim's Mark 1 containment design 'were -evaluated using methods  !

consistent with the Mark 1 containment program contained in NUREG-0661, " Safety Evaluation Report Mark l Containment Long-Term Program. Resolution of Generic Technical Activity A-7," and the " Pilgrim Plant Unique-Analysis Reports of the Torus

. Suppression Chamber and Torus Attached Piping." Design, fabrication, examination d and installation of the strainers conform to the requirements of " ASME lil B&PV Code, [.

. Subsection NF,1977. Edition through Summer 1978 Addenda." The strainers and suction piping were fabricated from 304/304L stainless steel and the support brackets were fabricated from carbon steel. Attachment 1 to this letter is a 3-dimensional

. cutaway depiction of a single torus bay showing +he strainer with the attached suction piping.nThis is typical of the strainer installat5n for two bays. Attachment 2 depicts the

' new strainer locations in a plan view of the torus.

9901060101 981230 3 PDR- ADOCK'05000293 PDb L. .g L

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Subs qu nt to tha str:inir install tion, a d:bris g:n: ration / transport end strain:r head loss gn: lysis w:s p:rform:d in compliance with tha r:quirem:nts of R:gulttory Guida 1.82, Revision 2. The debris generation and transport analysis was performed using the Regulatory Guide 1.82, Revision 2, guidance as implemented by the NRC approved BWR Owners' Group (BWROG) Utility Resolution Guidelines (URG).

Each of these debris types was addressed in the calculations:

. Fibrous debris from NUKON@ insulation installed on piping and the reactor vessel;

. Metallic foil debris from insulation installed on the reactor vessel; )

. Suppression pool sludge; I

. Particulate debris (e.g., paint chips, dirt, rust), and; e Other miscellaneous debris.

The general methods used for the estimation of the head loss across the strainers at the suction of the ECCS of Pilgrim are based on the NUREG/CR-6224, " Parametric 1 Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA [ loss-of-coolant I accident] Generated Debris." Such use is consistent with NRC practice because the l NRC implemented NUREG/CR-6224 fundamental models in the BLOCKAGE 2.5R computer code, which was used to calculate Pilgrim's estimated head losses.

Results of the analysis indicated that G. pired pumi NPSH was greater than the ,

maximum currently licensed containment pressure of 21/2 pounds / square inch gauge (psig). A maximum of 5 psig of overpressure is indicated to be required for the worst-case post-LOCA debris loading. The required overpressure for pump NPSH requirements is less than the calculated available containment pressure from the beginning of a design-basis accident (DBA) LOCA until the tcrus pressure decays to )

atmospheric pressure. The NRC was notified of this by a supplement to Licensing l Event Report (LER) 97-017-01, " Past Operation With Service Water Temperatures j Greater Than Design Due To Licensing Basis Ambiguity." dated December 18,1998.  !

Operability was demonstrated in accordance with the requirements of Generic Letter C 91-18, Revision 1. A license amendment requesting credit for additional overpressure (approximately 5 psig) is being developed and submittal is planned for January 15, 1999. Further details on our NRCB 96-03 efforts supplementing this summary will be submitted with the license amendment request i Pilgrim submits this summary in response to the NRCB 96-03 requirement to report on this issue by December 31,1998. Should the NRC require further information on this .j issue, please contact P.M. Kahler at (508) 830-7939.  ;

J. exand r  ;

uclear Assessment  !

l Group Manager Attachments JFA/PMK/cis 2.98.167 i

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-cc: - Mr. Alin B. Wing, Project M:nagar. 1 i:

Project Directorrts 1 l- Mail Stop: 14B20 L Office of Nuclear Reactor Regulation

,, . U. S. Nuclear Regulatory Commission

.1 White Flint North .

' 11555 Rockville Pike I Rockville, MD 20852 '

U.S. Nuclear Regulatory Commission .

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475 Allendale Road {

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' Senior Resident inspector Pilgrim Nuclear Power Station 4 1 i .

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