ML20214V186

Application for Amend to License DPR-66,revising Tech Specs to Permit Design of Reactor Coolant Pump & Steam Generator Supports

ML20214V186
Person / Time
Site:	Beaver Valley
Issue date:	06/01/1987
From:	Sieber J DUQUESNE LIGHT CO.
To:
Shared Package
ML19292H353	List: ML20214V178 ML20214V186 ML20214V192 ML20214V195 ML20214V202 ML20214V237
References
NUDOCS 8706120011
Download: ML20214V186 (8)

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UNITED STATES OF AMERICA 1 NUCLEAR REGULATORY COMMISSION In the Matter of )

) Docket No. 50-334 Duquesne Light Company )

(Beaver Valley Power Station, Unit No. 1) )

APPLICATION FOR AMENDMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the regulations of the U. S. Nuclear Regulatory Commission (the " Commission"), Duquesne Light Company

("DLC"), holder of Operating License No. DPR-66, hereby requests that the Operating License be revised to permit the design of the reactor coolant pump and steam generator supports to be revised in accordance with this submittal.

The proposed license change is set forth in Attachment A to this Application. A No Significant Hazard Evaluation is set forth in Attachment B. Based on the guidance provided by the Commission, the proposed changes have been determined to involve no significant hazards consideration in accordance with 10 CFR 50.92.

WHEREFORE, Applicant respectfully requests that Operating License No. DPR-66 be amended in the form attached hereto as Attachment A.

Duquesne Light Company By A -

gf D. Sieber Vice President, Nuclear l

Subscribed and sworn to before me i

on this /M day of b G M /Ib

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/ us , . .: +-r Lf ELVA G. LESONDAK, NOTARY PUBUC I

SHlJrlNGPORT, GEAVER COUNTY MY COMMISSION EXPlRES OCTOBER 20,1990

$fS63S8N$$0h$7 P

B3Svar Vollsy Powar Station, Unit No. 1

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Docket No. 50-334, License No. DPR-66 Proposed Operating License Change 1A-140 Page 3 cc: Mr. S. M. Pindale, Resident Inspector U. S. Nuclear Regulatory Commission Beaver Valley Power Station Shippingport, PA 15077 i U. S. Nuclear Regulatory Commission Regional Administrator Region 1 631 Park Avenue King of Prussia, PA 19406 Mr. Peter S. Tam U. S. Nuclear Regulatory Commission Project Directorate No. 2 Division of PWR Licensing - A Washington, DC 20555

- Mail Stop 316 Addressee only Director, Safety Evaluation & Control Virginia Electric & Power Company P.O. Box 26666 One James River Plaza Richmond, VA 23261 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection Pennsylvania Dept. of Environmental Resources P.O. Box 2063 Harrisburg, PA 17120 i

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A'TTACHMENT

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. Revise the Operating License as follows:

2.G The design of the reactor . coolant pump and steam generator supports may be revised in accordance with the licensee's submittal dated June 1, 1987.

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ATTACHMENT B No Significant Hazard Determinations Proposed Change Request No. 140 amends the Beaver Valley Power Station, Unit No. 1 operating license to allow the design of the reactor coolant pump and steam generator supports to be revised in accordance with the technical justification provided within this submittal (WCAPs 11317 and 11318, and Attachments C and D)

Description of knendment Request: Based on the recent revision to GDC-4 (Reference 1), Duquesne Light Company requests approval for a redesign of the reactor coolant pump and steam generator supports at Beaver Valley Unit 1. The new GDC-4 eliminates the need for consideration of postulated breaks in the RCS primary loop piping and its associated dynamic and other effects such as pipe whip, jet impingement, asymmetric pressure loading, and primary component sub-compartment pressurization. Approval of this request will also allow us to eliminate certain snubbers which are now required solely to mitigate a pipe rupture event and to replace with rigid restraints certain snubbers which have minimal thermal movement.

Specifically, approval of this request will allow us to redesign the primary reactor coolant loop equipment supports:

1) Eliminate two snubbers per loop acting in a direction perpendicular to the reactor coolant hot leg at the steam generator upper support ring and replace them with rigid struts.

2) Eliminate two snubbers per loop which are parallel to the reactor coolant cold leg at the reactor coolant pump support.

3) Eliminate two snubbers per loop which are parallel to the reactor coolant hot leg at the steam generator lower support.

4) Eliminate four snubbers per loop which are cross over restraints between the reactor coolant pump support and the steam generator lower support.

Granting this request would not affect:

ECCS design basis

- Reactor containment building and compartment design basis Equipment qualification basis Engineered safety feature systems response.

It is our intent to implement these changes during the Beaver Valley Unit 1 sixth refueling outage scheduled to begin in November, 1987.

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, AttachmSnt B (Continued)

Page 2 This request is based upon the use of advanced fracture mechanics technology as applied to primary system piping in Westinghouse Electric Corporation topical reports (References 3, 4 and 5 with approval by the USNRC in Reference 2), and by WCAP 11317 and 11318 (Reference 6) prepared specifically for Beaver Valley Unit 1.

Generic Letter 84-04 provided the NRC staff Safety Evaluation Report for analysis of materials submitted for a group of utilities operating PWRs to resolve unresolved safety issue A-2. The staff evaluation concluded that provided certain conditions were met, an acceptable technical basis exists so that asymmetric blowdown loads resulting from large breaks in main coolant loop piping need not be considered as a design basis. Beaver Valley Unit 1 was not included with the group of plants for which the Unresolved safety Issue A-2 was addressed. Therefore, to supplement the fracture mechanics studies performed for the A-2 owner's group, a plant specific fracture mechanics study was undertaken for Beaver Valley Unit 1.

Westinghouse reports, WCAP-ll317 and 11318 (Reference 6) document the results of the plant specific study.

These Westinghouse reports, WCAP-ll317 and 11318 in association with the other references provide a substantial and adequate basis for eliminating postulated breaks in the Beaver Valley Unit 1 stainless steel reactor coolant system piping. The analyses demonstrate that the probability of rupturing such piping is extremely low under design basis conditions.

The bases for the request for Beaver Valley Unit 1 are as follows:

1. Extensive operating experience has demonstrated the integrity of the PWR reactor coolant system krimary loop including the fact that there has never been a leakage crack.

2. Pre-service and in-service inspections performed on the RCS piping minimize the possibility of flaws existing in such piping. The application of advanced fracture mechanics has demonstrated in other applications that if such flaws exist they will not grow to a leakage crack when subjected to the worst case loading condition over the life of the plant.

3. If a large through-wall flaw is postulated, large margins against unstable crack extension exist for the stainless steel primary coolant piping even if subjected to the safe shutdown earthquake in combination with the loads associated with normal operation.

4. Beaver Valley Unit 1 has adequate leakage detection systems such that a postulated reference flaw would yield detectable leakage with margin when subjected to normal operating load condition.

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Attechm3nt B (Continued) l Page 3 In addition to WCAP 11317 and 11318, which document the plant specific facture mechanics study, further information is submitted addressing loading evaluation and leakage detection systems evaluation (Attachments C and D).

I Basis for no significant hazards consideration determination:  !

Based on the criteria for defining no significant hazards consideration set forth in 10 CFR 50.92(c), operation of Beaver Valley Unit 1 in accordance with this proposed change would not:

1) Involve a significant increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated. We have determined that the probability or consequences of an

] accident previously evaluated are not increased by our application of leak before break technology. Through proper application, we have demonstrated that: a) advanced fracture mechanics analysis is an acceptable alternative to maintaining systems or features solely to mitigate the consequences of postulated pipe ruptures; b) that high safety margins are maintained for the remaining snubbers and rigid restraints as shown in our loading evaluation to mitigate the effects of all postulated events (except the now eliminated pipe rupture); c) that our leakage detection systems are capable of detecting leakage from postulated through-walls flaws and allow operating personnel to take appropriate responses in a timely fashion, and d) the ECCS design basis, reactor containment and compartment design basis, equipment qualification basis, and engineered safety feature system response are unaffected by this change.

2) Create the possibility of a new or different kind of accident from any accident previously evaluated. It has

been demonstrated that one type of accident - a postulated RCS pipe rupture - has been eliminated from the design basis through application of leak before break technology. More specifically, our submittals have established that certain snubbers we propose to eliminate are required only to mitigate the effects of pipe rupture. Our analyses have demonstrated that with the redesigned support system, a) the loading of the primary loop is enveloped by the Beaver Valley Unit 1 specific analysis which supplements and extends the generic analyses submitted by Westinghouse on behalf of the USI A-2 Owners Group, and approved by the NRC staff in Generic Letter 84-04; and b) the reactor coolant system equipment, piping, and supports continue to have acceptable margins of safety under licensed loading conditions other than the now-eliminated RCS main loop rupture. The twenty four snubbers which we propose to eliminate have been demonstrated to serve only as pipe whip restraints. The six snubbers which we propose to replace with rigid struts exhibit minimal thermal movement.

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3) Involve a significant reduction in a margin of safety, our loading evaluation with the revised support load path configuration (i.e., snubbers removed) establishes that the piping components and supports are stressed within UFSAR acceptable limits. Adequate safety margins exist in a seismic event in excess of code allowables and the forces and maximum moment in the reactor coolant loop piping is within the forces and moment used in a plant specific fracture mechanics analysis, which is an extension to an earlier fracture mechanics study approved by the NRC on February 1, 1984, for a group of Westinghouse plants.

Conclusion This proposed change to allow the reactor coolant pump and steam generator supports to be redesigned in accordance with this submittal does not involve a significant increase in the probability or consequence of an accident previously evaluated, create the possibility of a new or different kind of accident than previously evaluated, or involve a significant reduction in the margin of safety. Therefore, it is proposed to characterize this change as involving a no significant hazards consideration.

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Page 5 References

1. Modification of General Design Criterion 4, requirements for protection against Dynamic Effects of Postulated Pipe Ruptures; 51 FR 12502.

2. Generic Letter. 84-04, February 1, 1984, " Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops."

3. WCAP-9558, Revision 2 (May 1981) " Mechanistic Fracture Evaluation of. Reactor Coolant Pipe Containing a Postulated Circumferential Through-wall Crack."

4. WCAP-9787 (May 1981) " Tensile and Toughness Properties of Primary Piping Weld Metal for use in Mechanistic Fracture Evaluation."

5. Letter Report NS-EPR-2519, E. P. Rahe to D. G. Eisenhut (November 10, 1981) Westinghouse Response to Questions and Comments Raised l by Members of ACRS Subcommittee on Metal Components During the Westinghouse Presentation on September 25, 1981.

6. WCAP-11317 Westinghouse Proprietary Class 2 and WCAP-11318 Westinghouse Proprietary Class 3; Technical basis for eliminating large primary loop pipe rupture as a structural design basis for Beaver Valley Unit 1.