ML12258A072

Cooper - Response to Request #2 for Additional Information License Amendment Request to Revise Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits.

ML12258A072
Person / Time
Site:
Issue date:	09/10/2012
From:	O'Grady B J Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2012062, TAC ME7324
Download: ML12258A072 (6)

Text

NNebraska Public Power DistrictAlways there when you need us50.90NLS2012062September 10, 2012U.S. Nuclear Regulatory CommissionAttention: Document Control DeskWashington, D.C. 20555-0001Subject: Response to Request #2 for Additional Information Re: License AmendmentRequest to Revise Technical Specification 3.4.9, "RCS Pressure and Temperature(P/T) Limits" (TAC NO. ME7324)Cooper Nuclear Station, Docket No. 50-298, DPR-46References: 1. Letter from Lynnea E. Wilkins, U.S. Nuclear Regulatory Commission, toBrian J. O'Grady, Nebraska Public Power District, dated August 10, 2012,"Cooper Nuclear Station -Request for Additional Information Re:License Amendment Request to Revise Technical Specification 3.4.9,'RCS Pressure and Temperature (P/T) Limits' (TAC No. ME7324)"2. Letter from Brian J. O'Grady, Nebraska Public Power District, to U.S.Nuclear Regulatory Commission, dated September 22, 2011, "LicenseAmendment Request to Revise Technical SpecificationPressure/Temperature Limit Curves and Surveillance Requirements"(NLS2011015)Dear Sir or Madam:The purpose of this letter is for Nebraska Public Power District (NPPD) to submit a response to arequest for additional information (RAI) from the Nuclear Regulatory Commission (NRC)(Reference 1). The RAI requested information in support of NRC's review of a licenseamendment request (LAR) for the Cooper Nuclear Station (CNS) facility operating license torevise Technical Specification Pressure/Temperature Limit Curves and SurveillanceRequirements (Reference 2).Responses to the specific RAI questions are provided in the Attachment. Two regulatorycommitments are made in Response #2 to resubmit the curves without the analysis of the P/Tnozzles by September 30, 2012 as a supplement to this LAR. Then later, after NRC approval ofthe generic methodology for nozzles, NPPD will submit another LAR to revise the curvesconsidering the nozzles.The information submitted by this response to the RAI does not change the conclusions or thebasis of the no significant hazards consideration evaluation provided with Reference 2.COOPER NUCLEAR STATIONP.O. Box 98 / Brownville, NE 68321-0098 A 001Telephone: (402) 825-3811 / Fax: (402) 825-5211 r,.Cwww.nppd.com NLS2012062Page 2 of 2If you have any questions concerning this matter, please contact David Van Der Kamp,Licensing Manager, at (402) 825-2904.I declare under penalty, of perjury that the foregoing is true and correct.Executed on __ _ _ _(date)Sincerely,Brian J. O'Grady ,4Vice President -Nuclear andChief Nuclear Officer/emAttachment: Response to Nuclear Regulatory Commission Request for Additional InformationRe: Technical Specification 3.4.9, "RCS Pressure and Temperature (P/T) Limits"(TAC NO. ME7324)cc: Regional Administrator w/ attachmentUSNRC -Region IVCooper Project Manager w/ attachmentUSNRC -NRR Project Directorate IV-1Senior Resident Inspector w/ attachmentUSNRC -CNSNebraska Health and Human Services w/ attachmentDepartment of Regulation and LicensureNPG Distribution w/o attachmentCNS Records w/ attachment NLS2012062AttachmentPage 1 of 4AttachmentResponse to Nuclear Regulatory Commission Request for Additional InformationRe: Technical Specification 3.4.9, "RCS Pressure and Temperature (P/T) Limits"(TAC NO. ME7324)Cooper Nuclear Station, Docket No. 50-298, DPR-46NRC Question #1The regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 50,Appendix G, "Fracture Toughness Requirements," state, in part, thatThis appendix specifies fracture toughness requirements for ferriticmaterials of pressure-retaining components of the reactor coolantpressure boundary of light water nuclear power reactors to provideadequate margins of safety...In addition, 10 CFR Part 50, Appendix G, paragraph IV.A states, in part, thatThe pressure-retaining components of the reactor coolant pressure boundary thatare made of ferritic materials must meet the requirements of the ASME Code[American Society of Mechanical Engineers Boiler and Pressure Vessel Code],supplemented by the additional requirements set forth below [paragraph IV.A. 2,"Pressure-Temperature Limits and Minimum Temperature Requirements"] ...Therefore, 10 CFR Part 50, Appendix G requires that P-T limits be developed for theentire reactor coolant pressure boundary (RCPB), consisting of ferritic RCPB materials inthe reactor vessel (RV) beltline (neutron fluence > I x 1017 n/cm2, E> 1 MeV), as well asferritic RCPB materials not in the RV beltline (neutron fluence < I x 1017 n/cm2, E> IMeV).P-T limit calculations for ferritic RCPB components that are not RV beltline shellmaterials, may define curves that are more limiting than those calculated for the RVbeltline shell materials. This may be due to the following factors:a. RV nozzles, penetrations, and other discontinuities have complex geometries thatmay exhibit significantly higher stresses than those for the RV beltline shell region.These higher stresses can potentially result in more restrictive P-T limits, even if thereference temperature (RTNDT) for these components is not as high as that of RVbeltline shell materials that have simpler geometries.b. Ferritic RCPB components that are not part of the RV may have initial RTNDT values,which may define a more restrictive lowest operating temperature in the P-T limitsthan those for the RV beltline shell materials.

INLS2012062AttachmentPage 2 of 4Please describe how the P-T limit curves, and the methodology used to develop these curvesconsidered all RV materials (beltline and non-beltline) and the lowest service temperature of allferritic RCPB materials, consistent with the requirements of 10 CFR Part 50, Appendix G.Response #1Nebraska Public Power District (NPPD) calculates the fluence for the reactor vesselplates and welds in accordance with the BWRVIP RAMA code for 32 effective fullpower years (EFPYs). Then we develop Adjusted Reference Temperature (ART) andReference Temperature Shift (ARTNDT) values for the reactor pressure vessel plates andwelds exposed to fluences greater than 1.0 x 1017 n/cm2 in accordance with RegulatoryGuide 1.99, Revision 2.The analyzed Reactor Pressure Vessel (RPV) wall's local fracture toughness, at thepostulated flaw location (1/4t), is determined from considerations of initial RTNDT, localfluence, margins, and chemical composition. The ART is used to determine the fracturetoughness described in ASME Code,Section XI, Appendix G.Vessel nozzles are generally incorporated into P/T curve calculations using stressdistributions from Finite Element Analyses and applying them to geometry specificfracture mechanics models. The feedwater nozzle (upper vessel region) and coredifferential pressure (CDP) nozzle require this type of analysis due to the boundingtransients they experience and/or stress concentration effects. The core differentialpressure CDP nozzle (bottom head region) is analyzed because it is the limitingdiscontinuity in the thin portion of the bottom head.The feedwater nozzle is the bounding component in the upper vessel because it is a stressconcentrator (essentially a hole in a plate) and because it typically experiences moresevere thermal transients compared to the rest of the upper vessel region. A two-dimensional finite element model of the feedwater nozzle is created as described inSection 2.0 of the calculation. The stress distribution acting normal to the postulated 1/4thickness crack (or hoop stress distribution) due to a 1,000 psig unit pressure is obtainedalong a limiting path in the nozzle-to-RPV blend radius. Pressure stress coefficients areused to calculate the applied pressure stress intensity factor.The material property values contained in the BWRVIP ISP are incorporated in thecalculation where appropriate. The material properties documented in the calculation areconsidered to be the most recent based on the review of references and are considered tobe most appropriate values for computation of ARTNDT and ART. Since neither thefeedwater nozzle nor the CDP nozzle experience fluences greater than 1.0 x 1017 n/cm2,there is no calculation of ART for them.

NLS2012062AttachmentPage 3 of 4In addition to the above, it is also recognized that P/T limits generated for the RPV alsoare considered to cover all portions of the Reactor Coolant System (RCS) piping. Thereare at least four reasons why the RPV P/T limits are considered to adequately boundfracture toughness requirements for the RCS piping: (1) the RPV is irradiated (therebyexperiencing material degradation due to neutron embrittlement) whereas the RCS pipingis not, (2) the philosophy behind the design codes used to evaluate the design of the RPVand piping generally recognize that the RPV is more limiting than the RCS piping from astructural standpoint, (3) much of the RCS piping is austenitic stainless steel, which hasductile behavior and does not experience the fracture concerns that ferritic materialexperiences, and (4) stresses are typically higher in the thicker-walled RPV than in thethin-walled RCS piping, which is less than 2.5 inches in thickness.More detail on the calculation methodology can be found in Structural IntegrityAssociates calculation 1100445.303, Revision 0, "Revised P/T Curve Calculation", whichwas included in the submittal.NRC Question #2Linear Elastic Fracture Mechanics (LEFM) evaluation of the N16 Water Level InstrumentNozzles: The licensee's LAR submittal, which includes Structural Integrity Associates (SIA)calculation package 1100445.303, provides a reference to the generic LEFM methodology usedfor calculating the applied stress intensity factor values for the N16 instrument nozzles. ForCooper, the N16 nozzles define part of the bounding beltline region P-T curves at lowtemperatures.The generic LEFM methodology for boiling-water reactor instrument nozzles, provided in SIAReport No. 0900876.401, Revision 0, "Linear Elastic Fracture Mechanics Evaluation of GeneralElectric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature CurveEvaluation," November 2011 (ADAMS Accession No. ML 11325A074), is currently under reviewby NRC staff Please provide an alternate methodology for the stated instrument nozzles.Response #2NPPD will resubmit the curves without the analysis of the P/T nozzles by September 30,2012. Since there is no currently approved methodology for addressing the instrumentnozzles in the beltline region of a Boiling Water Reactor, NPPD will commit to providingnew P/T curves after the generic methodology is approved, but before the end of 2016(prior to exceeding 32 EFPY).

NLS2012062AttachmentPage 4 of 4LIST OF REGULATORY COMMITMENTSThe following table identifies those actions committed to by Nebraska Public PowerDistrict in this document. Any other actions discussed in this submittal are provided forinformation purposes and are not considered to be regulatory commitments.TYPECOMMITMENT/COMMITMENT NO. (Check one) SCHEDULEDCOMPLETIONONE-TIME CONTINUING DATEACTION COMPLIANCENPPD will resubmit the curves without X September 30,the analysis of the P/T nozzles. 2012[NLS2012062-01]Since there is no currently approvedmethodology for addressing the instrumentnozzles in the beltline region of a Boiling WaterReactor, NPPD will commit to providing new PITcurves after the generic methodology isapproved, but before the end of 2016 (prior toexceeding 32 EFPY. [NLS2012062-02]XDecember 31,2016