ML20247F176

Requests Mod to Existing Plant Licensing Basis,Per 10CFR50.59.Mod Identified While Addressing Concerns Re NRC Bulletin 96-003, Potential Plugging of ECC Suction Strainer by Debris in Bwrs. Calculation Details & Evaluation,Encl

ML20247F176
Person / Time
Site:	Oyster Creek
Issue date:	05/05/1998
From:	Roche M GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20247F180	List: ML20247F176 ML20247F183 ML20247F189
References
1940-98-20124, IEB-96-003, IEB-96-3, NUDOCS 9805190186
Download: ML20247F176 (5)

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{ CPU Nuclear. inc.

A U.S. Route #9 South NUCLEAR Post 0tfice Box 388 i Forked River. NJ 087310388 Tel 609 971-4000 May 5,1998 1940-98-20124 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington DC 20555

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Request for Change to the Licensing Basis While addressing the concerns identified in NRC Bulletin 96-03 " Potential Plugging of l Emergency Core Cooling Suction Strainers by Debris in Boiling Water Reactors", the '

calculation determined that a modification would have to be made to the existing plant. An evaluation was performed as required by 10 CFR 50.59. The evaluation determined that the i

proposed modification involved an unreviewed safety question.

Specifically, it was necessary to utilize a small amount of containment overpressure to ensure sufficient Net Positive Suction Head for the Emergency Core Cooling System pumps under post Loss of Cooling Accident conditions. Pursuant to 10 CFR 50.90, this letter requests that change to the Oyster Creek Nuclear Generating Station licensing basis.

Attachment I provides a brief description of the modification and the determination of No Significant Hazards. Attachment Il provides the details of the calculation which was performed. Attachment III provides the Risk Evaluation for this request.

The requested change is in support of a major modification scheduled to be installed during the upcoming 17R refueling outage (September 1998). As planning, scheduling, and procurement ,

have already commenced, approval of this change is requested by August 1,1998. Significant delays past that time could necessitate a deferral of the modification to 18R.

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8 1940-98-20124 Page 2 If any additional information or assistance is required, please contact Mr. John Rogers of my staff at 609.971.4893.  !

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Very truly yours,

/b Michael B. Roche Vice President and Director Oyster Creek l

MBR/JJR Attachments

-cc: Administrator, Region I NRC Project Manager Senior Resident Inspector i i

Sworn to and Subscribed before me this day of N1Au- ,1998.

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) sD M - lJ1 0 $LD A Notary Public of NJ My commission expires O(ta /3, JOo/ .

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Attachment I Description of Modification and Statement of No Significant Hazards In response to the concerns identified in NRC Bulletin 96-03, " Potential Plugging of Emergency Core Cooling (ECCS) Suction Strainers by Debris in Boiling Water Reactors", 1 GPU Nuclear, Inc. has determined that a modification to the Oyster Creek Nuclear Generating i Station will be required. A vendor has been selected and a new suction strainer has been chosen.

The existing ECCS strainers at the Oyster Creek Nuclear Generating Station consists of three strainers which are attached to sixteen inch pipes and feed a common ring header. The ring header feeds all four main core spray pumps and all four containment spray pumps. The  !

proposed modification will replace the existing strainers with a stacked disk design by General Electric.

The strainers will be procured as Nuclear Safety Related. They are designed to strain out particulate debris larger than 3/16 inch nominal diameter. The strainers are of a slip on flange design.

The new design replaces all three of the existing strainers. However, in order to ensure proper {

operation of the core spray system under the accident conditions, a positive wetwell pressure i of 1.25 psig is required for the first hour of the accident. The basis for this number is contained in Attachment II. Therefore, a change to the licensing basis of the Oyster Creek Nuclear Generating Station from the existing zero to 1.25 psig overpressure in the wetwell is requested. The consequences of the offsite release due to a postulated l_oss of Coolant i Accident (LOCA) have already assumed a containment overpressure of approximately 40 psig, which rapidly decays to approximately 20 psig. This new analysis assumes that operators take manual actions to start containment spray pumps when required by procedure, and subsequently secure them when required by procedure. The pumps also presently trip automatically at a drywell pressure of 0.6 psig. A change to the switch setpoint will ensure that the pumps automatically trip above 1.25 psig wetwell pressure to provide defense in depth for the operator actions. Therefore, containment overpressure is maintained. These switches are not Nuclear Safety Related. However, they are routinely surveilled and provide an excellent defense in depth for the operator actions. The risk evaluation contained in Attachment III models both the proceduralized operator action to trip the containment spray pumps as well as the automatic pump trip function.

GPU Nuclear, Inc. has evaluated this change as required by 10 CFR 50.90 using the specifics detailed in 10 CFR 50.92, and has determined that No Significant Hazard exists.

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.< 1940-98-20124 Attachment I Page 2

' The proposed change to the licensing basis does not " Involve a significant increase in the probability or consequences of an accident previously evaluated...". As the strainers have no function until after the design basis LOCA occurs, the design of the strainer cannot affect the probability of a Large Break LOCA.

The requested change to raise the assumed containment over pressure for suction strainer design to 1.25 psig is less than that which is already used in LOCA analyses for offsite releases. Therefore, this change will not increase the offsite consequences of any previously analyzed accident. The frequency of a design basis LOCA occurrence at the Oyster Creek d

Nuclear Generating Station is conservatively estimated at 5.67 x 10 per year. The frequency of a design basis LOCA with a loss of containment overpressure is conservatively estimated at 4

2.46 x 10 per year. Since the frequency of the design basis LOCA coincident with a loss of containment overpressure is insignificant (2.46 x 104), the requested increase does not significantly impact the probability of exceeding the existing design bases. The core damage frequency increase due to the request for overpressure is mitigated, in part, by the current procedural requirement to flood containment following the design basis LOCA, thereby obviating the need for overpressure in the long term. The risk evaluation, performed in )

support of the request for overpressure, indicated a non-risk significant change in the core i damage frequency.

The proposed change to the licensing bases does not " Create the possibility of a new or I different kind of accident from any accident previously evaluated...". Both the new and  !

existing strainers are passive. They function solely to prevent debris from entering the suction of the core and containment spray pumps. The only significant difference is that the new strainers can remove more debris without clogging. The slight amount of containment l

overpressure does not affect the operation of the strainers, and improves the ability of the core i spray and containment spray systems to continue operation. Therefore, no new or different kind of accident is created or possible. i i

The proposed change to the licensing bases does not " Involve a significant reduction in a l margin of safety...." The modification increases the amount of debris that can be removed l while maintaining core spray system operation. The requested change takes credit for 1.25 psig of wetwell overpressure. However, as the requested change is bounded by existing

calculations for offsite release, no significant reduction in the margin of safety can occur.

. Additionally, as demonstrated in Attachment III, the probability of a LOCA with a loss of l containment overpressure is not significant.

Guidance has been provided in " Final Procedures and Standards on No Significant Hazards Considerations," Final Rule,51 FR 7744, for the application of standards to license change l requests for determination of the existence of significant hazards considerations. This -

document provided examples of amendments which are and are not considered likely to involve significant hazards considerations.

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4 1940-98-20124 Attachment I Page 3 Based on the above evaluation and the review of 51 FR 7744, this proposed change to the licensing basis of the Oyster Creek Nuclear Generating Station does not involve irreversible changes, a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings, or a significant relaxation of the bases for the limiting conditions for operations. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed change does not constitute a significant hazard.

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