ML20076K645

Application for Amend to Licenses NPF-39 & NPF-85 to TS 93-18 Re Change Request That Proposed to Increase Allowable Leak Rate for MSIV & to Delete MSIV LCS

ML20076K645
Person / Time
Site:	Limerick
Issue date:	10/25/1994
From:	Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20076K647	List: ML20076K645 ML20076K658
References
NUDOCS 9411010322
Download: ML20076K645 (5)

Text

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nn PECO ENERGY reco tee,2< ce-ne v Nuclear Group Headquarters 965 Chesterbrook eoulevard Wayne, PA 19087-5691 October 25,1994 Docket Nos. 50-352 50-353 License Nos. NPF-39 ,

NPF-85 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk ,

Washington, DC 20555

SUBJECT:

Limerick Generating Station, Units 1 and 2 Technical Specifications Change Request No. 93-18-0 Request For Additional Information Gentlemen:

By letter dated January 14,1994, PECO Energy Company submitted a Limerick Generating Station (LGS), Unit 1 and Unit 2, Technical Specifications (TS) Change Request (i.e.,93-18-0) that proposed to increase the allowable leak rate for the main steam isolation valves (MSIVs) and to delete the MSIV Leakage Control System (LCS).

During a telephone conversation between PECO Energy and the NRC on October 12,1994, the NRC technical reviewer identified that the Limerick Technical Specifications Change Request 93-18-0 and a response to an earlier Request for Additional Information, dated August 1,1994, did not provide sufficient information. Therefore, additionalinformation is contained in Attachment 1 to this letter.

If you have any questions, please do not hesitate to contact us.

Very truly yours, A? f/=o t G. A. Hunger, Jr.,

Director-Licensing Attachments Enclosure cc: T. T. Martin, Administrator, Region I, USNRC (w/ attachments and enclosure)

N. S. Perry, USNRC Senior Resident inspector, LGS (w/ attachments and enclosure)

R. R. Janati, PA Bureau of Radiological Protection (w/ attachments and enclosure) 0 0\

9411010322 941025 PDR ADOCK 05000352 l \

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. COMMONWEALTH OF PENNSYLVANIA ss.

COUNTY OF CHESTER .

W. H. Smith, Ill, being first duly sworn, deposes and says: That he is Vice President of PECO Energy Company, the Applicant herein; that he has read the enclosed additional information supporting Technical Specifications Change Request No. 93-18-0 " Increase the Allowable Leak Rate for the Main Steam Isolation Valves and Delete the MSIV Leakage Control System," for Limerick Generating Station, Unit 1 and Unit 2, Facility Operating License Nos. NPF-39 and NPF-85, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.

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Vice President Subscribed and swor o before me this - day of 1 bl994.

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Notary Public NotanalSeal Enm A Santon.tbtaryPubic Tr . .*r* 'wg Crester County n '-- . .eion Exps July 10.1995 l

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s ATTACHMENT 1 LIMERICK GENERATING STATION UNIT 1 AND UNIT 2 DOCKET NOs.

50-352 50-353 LICENSE NOs.

NPF-39 NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 93-18-0

" Increase the Allowable Leak Rate for the Main Steam isolation Valves and Delete the MSIV Leakage Control System" i 4

Additional Seismic Information - 3 PAGES 1

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October 25,1994 i Page 1 j

On October 12, 1994, during a telephone conversation between PECO Energy and the NRC, the NRC )

technical reviewer identified that the Limerick Technical Specifications Change Request No. 93-18-0 for Limerick Units 1 and 2 and the response to the Request for Additional Information, dated August 1,1994, did not provide sufficient information. Specifically, the NRC Reviewer stated that the piping supports and an identified block wall evaluation intending to demonstrate that the structures would not fail following an ,

earthquake, required an evaluation using an approach prev lously approved by the NRC. This would be )

either an A46(SOUG) approach or a 1.E. Bulletin 79-02 evaluation for the piping supports, and 1.E. Bulletin '

80-11 evaluation for the block wall. In response to the reviewer's request, PECO Energy has provided the following description of the evaluations already performed.

CONCRETE BLOCK WALL EVALUATION METHODOLOGIES The following discussion summarizes the methodology used to evaluate concrete block walls for out-of-plane seismic loads recommended in Appendix R of EPRI NP-6041-SL These methodologies were utilized to determine the seismic demand and the capacity of the subject LGS block wall. The calculation determined that the capacity of the wall exceeded the seismic demand.

SEISMIC DEMAND An estimate of median centered seismic response was developed based on the procedure outlined in EPRI NP-6041-SL. Scaling factors are developed using the direct scaling procedure from EPRI ,

NP-6041-SL and are applied to appropriate floor response developed from the LGS Turbine Building j Design Basis Spectra. The LGS Turbine Building is rock founded. The scaling factors are based  ;

on a comparison of 7% damped NUREG CR-0098 spectra anchored at 0.15g and 5% damped Design Basis Earthquake ground spectra at the dominant frequencies of the structure.

LOAD COMBINATIONS The general procedure of EPRI NP-6041-SL recommends that seismic loads be combined with other loads occurring under normal operating conditions, with load factors of 1.0 typically assigned.

SEISMIC CAPACITY The seismic capacity of concrete block walls in commercial nuclear power plants is typically controlled by out-of-plane flexure. EPRI NP-6041 SL recommends that the out-of-plane flexural strength of lightly reinforced masonry walls be determined following ultimate strength provisions for reinforced concrete. A strength reduction factor of 0.90 is recommended, based on bench marking the strength evaluation method against test data.

ELASTIC SEISMIC RESPONSE The elastic seismic response analysis procedures of EPRI NP-6041-SL pcount for appropriate wall boundary conditions and crack section properties. The wall is analyzed based on the applicable floor response spectrum. An efft clive damping of 6% of critical damping incorporates nonlinear behavior discussed below.

NONLINEAR RESPONSE EPRI NP-6041-SL provides procedures to account for frequency shifting and inelastic energy absorption associated with ductile, nonlinear response. These procedures were benchmarked against nonlinear time-history analysis results.

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1 October 25,1994 Page 2 PIPING ANCHOR EVALUATION METHODOLOGIES Provided in the attached Table is a listing of the loads for the three piping supports which have been determined to be bounding and restrictive of the LGS analysis. The table provides a value for both the total load on these bounding supports and the capacity of the supports. Median centered demand was developed as described in the block wall analysis procedure. Best estimate piping loads were developed using peak spectral accelerations of 5% damped spectra from Turbine Building responses at appropriate elevations.

The piping anchorage capacity is determined by utilizing the plant design specifications and utilizing the criteria for capacities established for A-46 in EPRI NP-5228, Seismic Verification of Nuclear Plant Equipment Anchorage (Revision 1), June 1991 and EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1) August 1991, The above methodologies were utilized to determine the seismic demand and the capacity of the piping supports. As can be seen in the attached table, the capacity of the supports exceeds the seismic demand.

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