NLS2012062, 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 Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2012062, TAC ME7324 | |
| Download: ML12258A072 (6) | |
Text
N Nebraska Public Power District Always there when you need us 50.90 NLS2012062 September 10, 2012 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
Response to Request #2 for Additional Information Re: License Amendment Request 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-46
References:
- 1. Letter from Lynnea E. Wilkins, U.S. Nuclear Regulatory Commission, to Brian 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, "License Amendment Request to Revise Technical Specification Pressure/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 a request for additional information (RAI) from the Nuclear Regulatory Commission (NRC)
(Reference 1). The RAI requested information in support of NRC's review of a license amendment request (LAR) for the Cooper Nuclear Station (CNS) facility operating license to revise Technical Specification Pressure/Temperature Limit Curves and Surveillance Requirements (Reference 2).
Responses to the specific RAI questions are provided in the Attachment. Two regulatory commitments are made in Response #2 to resubmit the curves without the analysis of the P/T nozzles by September 30, 2012 as a supplement to this LAR. Then later, after NRC approval of the generic methodology for nozzles, NPPD will submit another LAR to revise the curves considering the nozzles.
The information submitted by this response to the RAI does not change the conclusions or the basis of the no significant hazards consideration evaluation provided with Reference 2.
COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 A 001 r,.C www.nppd.com
NLS2012062 Page 2 of 2 If 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 ,4 Vice President - Nuclear and Chief Nuclear Officer
/em
Attachment:
Response to Nuclear Regulatory Commission Request for Additional Information Re: Technical Specification 3.4.9, "RCS Pressure and Temperature (P/T) Limits" (TAC NO. ME7324) cc: Regional Administrator w/ attachment USNRC - Region IV Cooper Project Manager w/ attachment USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ attachment USNRC - CNS Nebraska Health and Human Services w/ attachment Department of Regulation and Licensure NPG Distribution w/o attachment CNS Records w/ attachment
NLS2012062 Attachment Page 1 of 4 Attachment Response to Nuclear Regulatory Commission Request for Additional Information Re: Technical Specification 3.4.9, "RCS Pressure and Temperature (P/T) Limits" (TAC NO. ME7324)
Cooper Nuclear Station, Docket No. 50-298, DPR-46 NRC Question #1 The regulations in Title 10 of the Code of FederalRegulations (10 CFR) Part 50, Appendix G, "FractureToughness Requirements," state, in part, that This appendix specifies fracture toughness requirementsfor ferritic materials of pressure-retainingcomponents of the reactorcoolant pressure boundary of light water nuclear power reactorsto provide adequate margins of safety...
In addition, 10 CFR Part50, Appendix G, paragraphIV.A states, in part, that The pressure-retainingcomponents of the reactorcoolant pressure boundary that are 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 additionalrequirements set forth below [paragraphIV.A. 2, "Pressure-TemperatureLimits and Minimum Temperature Requirements"] ...
Therefore, 10 CFR Part 50, Appendix G requires that P-T limits be developed for the entire reactorcoolant pressure boundary (RCPB), consisting of ferritic RCPB materials in the reactorvessel (RV) beltline (neutron fluence > I x 1017 n/cm 2, E> 1 MeV), as well as ferritic RCPB materials not in the RV beltline (neutron fluence < I x 1017 n/cm2 , E> I MeV).
P-T limit calculations for ferritic RCPB components that are not RV beltline shell materials,may define curves that are more limiting than those calculated for the RV beltline shell materials. This may be due to the following factors:
- a. RV nozzles, penetrations,and other discontinuitieshave complex geometries that may 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 the reference temperature (RTNDT) for these components is not as high as that of RV beltline shell materials that have simpler geometries.
- b. FerriticRCPB components that are not part of the RV may have initialRTNDT values, which may define a more restrictivelowest operatingtemperature in the P-T limits than those for the RV beltline shell materials.
I NLS2012062 Attachment Page 2 of 4 Please describe how the P-T limit curves, and the methodology used to develop these curves considered all RV materials (beltline and non-beltline) and the lowest service temperatureof all ferritic RCPB materials,consistent with the requirements of 10 CFR Part 50, Appendix G.
Response #1 Nebraska Public Power District (NPPD) calculates the fluence for the reactor vessel plates and welds in accordance with the BWRVIP RAMA code for 32 effective full power years (EFPYs). Then we develop Adjusted Reference Temperature (ART) and Reference Temperature Shift (ARTNDT) values for the reactor pressure vessel plates and welds exposed to fluences greater than 1.0 x 1017 n/cm2 in accordance with Regulatory Guide 1.99, Revision 2.
The analyzed Reactor Pressure Vessel (RPV) wall's local fracture toughness, at the postulated flaw location (1/4t), is determined from considerations of initial RTNDT, local fluence, margins, and chemical composition. The ART is used to determine the fracture toughness described in ASME Code,Section XI, Appendix G.
Vessel nozzles are generally incorporated into P/T curve calculations using stress distributions from Finite Element Analyses and applying them to geometry specific fracture mechanics models. The feedwater nozzle (upper vessel region) and core differential pressure (CDP) nozzle require this type of analysis due to the bounding transients they experience and/or stress concentration effects. The core differential pressure CDP nozzle (bottom head region) is analyzed because it is the limiting discontinuity in the thin portion of the bottom head.
The feedwater nozzle is the bounding component in the upper vessel because it is a stress concentrator (essentially a hole in a plate) and because it typically experiences more severe 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 in Section 2.0 of the calculation. The stress distribution acting normal to the postulated 1/4 thickness crack (or hoop stress distribution) due to a 1,000 psig unit pressure is obtained along a limiting path in the nozzle-to-RPV blend radius. Pressure stress coefficients are used to calculate the applied pressure stress intensity factor.
The material property values contained in the BWRVIP ISP are incorporated in the calculation where appropriate. The material properties documented in the calculation are considered to be the most recent based on the review of references and are considered to be most appropriate values for computation of ARTNDT and ART. Since neither the feedwater nozzle nor the CDP nozzle experience fluences greater than 1.0 x 1017 n/cm 2, there is no calculation of ART for them.
NLS2012062 Attachment Page 3 of 4 In addition to the above, it is also recognized that P/T limits generated for the RPV also are considered to cover all portions of the Reactor Coolant System (RCS) piping. There are at least four reasons why the RPV P/T limits are considered to adequately bound fracture toughness requirements for the RCS piping: (1) the RPV is irradiated (thereby experiencing material degradation due to neutron embrittlement) whereas the RCS piping is not, (2) the philosophy behind the design codes used to evaluate the design of the RPV and piping generally recognize that the RPV is more limiting than the RCS piping from a structural standpoint, (3) much of the RCS piping is austenitic stainless steel, which has ductile behavior and does not experience the fracture concerns that ferritic material experiences, and (4) stresses are typically higher in the thicker-walled RPV than in the thin-walled RCS piping, which is less than 2.5 inches in thickness.
More detail on the calculation methodology can be found in Structural Integrity Associates calculation 1100445.303, Revision 0, "Revised P/T Curve Calculation", which was included in the submittal.
NRC Question #2 Linear Elastic FractureMechanics (LEFM) evaluation of the N16 Water Level Instrument Nozzles: The licensee's LAR submittal, which includes StructuralIntegrityAssociates (SIA) calculation package 1100445.303, provides a reference to the generic LEFM methodology used for calculatingthe applied stress intensity factor values for the N16 instrument nozzles. For Cooper, the N16 nozzles define part of the bounding beltline region P-T curves at low temperatures.
The generic LEFM methodology for boiling-waterreactorinstrument nozzles, provided in SIA Report No. 0900876.401, Revision 0, "LinearElastic FractureMechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-TemperatureCurve Evaluation," November 2011 (ADAMS Accession No. ML11325A074), is currently underreview by NRC staff Pleaseprovide an alternate methodology for the stated instrument nozzles.
Response #2 NPPD 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 instrument nozzles in the beltline region of a Boiling Water Reactor, NPPD will commit to providing new P/T curves after the generic methodology is approved, but before the end of 2016 (prior to exceeding 32 EFPY).
NLS2012062 Attachment Page 4 of 4 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Nebraska Public Power District in this document. Any other actions discussed in this submittal are provided for information purposes and are not considered to be regulatory commitments.
TYPE COMMITMENT/COMMITMENT NO. (Check one) SCHEDULED ONE-TIME CONTINUING COMPLETION DATE ACTION COMPLIANCE NPPD will resubmit the curves without X September 30, the analysis of the P/T nozzles. 2012
[NLS2012062-01]
Since there is no currently approved X December 31, methodology for addressingthe instrument 2016 nozzles in the beltline region of a Boiling Water Reactor, NPPD will commit to providing new PIT curves after the generic methodology is approved, but before the end of 2016 (priorto exceeding 32 EFPY. [NLS2012062-02]