ML083530806

R. E. Ginna Nuclear Power Plant - Amendment Re Revised Methodology for Determining Reactor Coolant System Pressure and Temperature and Low Temperature Over-Pressure Limits

ML083530806
Person / Time
Site:	Ginna
Issue date:	02/23/2009
From:	Pickett D Plant Licensing Branch 1
To:	John Carlin Ginna
Pickett D
References
TAC MD8069
Download: ML083530806 (19)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Fehnmy 23, 7M Mr. John T. Carlin Vice President RE. Ginna Nuclear Power Plant RE. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519

SUBJECT:

RE. GII'JI'JA NUCLEAR POWER PLANT - AMENDMENT RE: REVISED METHODOLOGY FOR DETERMINING REACTOR COOLANT SYSTEM PRESSURE AND TEMPERATURE AND LOW TEMPERATURE OVER-PRESSURE LIMITS (TAC NO. MD8069)

Dear Mr. Carlin:

The Commission has issued the enclosed Amendment NO.1 06 to Renewed Facility Operating License No. DPR-18 for the RE. Ginna Nuclear Power Plant. This amendment is in response to your application dated February 8, 2008, as supplemented by letter dated April 25, 2008, and email dated January 7, 2009.

The amendment revises Technical Specification 5.6.6, "Reactor Coolant System (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," to change the method of determining reactor coolant system pressure and temperature and low temperature over pressure limits. The new PTLR methodology is documented in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," dated May 2004.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244

Enclosures:

1. Amendment NO.106 to Renewed License No. DPR-18

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 R.E. GINNA NUCLEAR POWER PLANT. LLC DOCKET NO. 50-244 R.E. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment NO.1 06 Renewed License No. DPR-18

1. The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A. The application for amendment filed by the R.E. Ginna Nuclear Power Plant, LLC (the licensee) dated February 8,2008, as supplemented by letter dated April 25, 2008, and email dated January 7,2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-18 is hereby amended to read as follows:

-2 (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment NO.106, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION ark G. Kowal, Chief lant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: Fehruary 23, 2ffR

ATTACHMENT TO LICENSE AMENDMENT NO. 106 RENEWED FACILITY OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Replace the following page of the Facility Operating License with the attached revised page.

The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove 3 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 5.6-1 5.6-1 5.6-2 5.6-2 5.6-3 5.6-3 5.6-4 5.6-4 5.6-5 5.6-5

-3 (b) Pursuant to the Act and 10 CFR Part 70, to possess and use four (4) mixed oxide fuel assemblies in accordance with the RG&E's application dated December 14, 1979 (transmitted by letter dated December 20, 1979), as supplemented February 20, 1980, and March 5, 1980; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

(1) Maximum Power Level Ginna LLC is authorized to operate the facility at steady-state power levels up to a maximum of 1775 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment NO.106, are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Fire Protection (a) The licensee shall implement and maintain in effect all fire protection features described in the licensee's submittals referenced in and as approved or modified by the NRC's Fire Protection Safety Evaluation (SE) dated February 14, 1979, and Amendment NO.106

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Deleted 5.6.2 Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring activities for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the plant shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.6.4 Deleted R.E. Ginna Nuclear Power Plant 5.6-1 Amendment 106

Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

The following administrative requirements apply to the COLR:

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

2.1, "Safety Limits (SLs)";

LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

LCO 3.1.3, "MODERATOR TEMPERATURE COEFFICIENT (MTC)";

LCO 3.1.5, "Shutdown Bank Insertion Limit";

LCO 3.1.6, "Control Bank Insertion Limits";

LCO 3.2.1, "Heat Flux Hot Channel Factor (Fa(Z))";

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FNt>H)";

LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

LCO 3.3.1, "Reactor Protection System (RPS) Instrumentation";

LC03.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and LCO 3.9.1, "Boron Concentration."

R.E. Ginna Nuclear Power Plant 5.6-2 Amendment 106

Reporting Requirements 5.6

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

(Methodology for 2.1, LCO 3.1.1, LCO 3.1.3, LCO 3.1.5, LCO 3.1.6, LCO 3.2.1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)

2. WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty (ASTRUM)," January 2005.

3. WCAP-10216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control / FQ Surveillance Technical Specification,"

February 1994.

(Methodology for LCO 3.2.1 and LCO 3.2.3.)

4. WCAP-12610-P-A, "VANTAGE + Fuel Assembly Reference Core Report," April 1995.

(Methodology for LCO 3,2.1.)

5. WCAP 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.

(Methodology for LCO 3.4.1 when using RTDP.)

6. WCAP-1 0054-P-A and WCAP-1 0081-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.

(Methodology for LCO 3.2.1.)

7. WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and COSI Condensation Model," July 1997.

(Methodology for LCO 3.2.1)

8. WCAP-11145-P-A, "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study With the NOTRUMP Code,"

October 1986.

(Methodology for LCO 3.2.1)

9. WCAP-10079-P-A, "NOTRUMP - A Nodal Transient Small Break and General Network Code," August 1985.

(Methodology for LCO 3.2.1 )

10. WCAP-8745, "Design Basis for the Thermal Overpower Delta T and Thermal Overtemperature Delta T Trip Functions,"

March 1977.

(Methodology for LCO 3,3.1.)

R.E. Ginna Nuclear Power Plant 5.6-3 Amendment 106

Reporting Requirements 5.6

11. WCAP-14710-P-A, "1-0 Heat Conduction Model for Annular Fuel Pellets," May, 1998.

(Methodology for LCO 3.2.1)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

The following administrative requirements apply to the PTLR:

a. RCS pressure and temperature limits for heatup, cooldown, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (PIT) Limits"

b. The power operated relief valve lift settings required to support the Low Temperature Overpressure Protection (LTOP) System, and the LTOP enable temperature shall be established and documented in the PTLR for the following:

LCO 3.4,6, "RCS Loops - MODE 4";

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled";

LCO 3.4,10, "Pressurizer Safety Valves"; and LCO 3.4,12, "LTOP System,"

c. The analytical methods used to determine the RCS pressure and temperature and LTOP limits shall be those previously reviewed and approved by the NRC. Specifically, the methodology is described in the following documents:

1. WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 4.

R.E. Ginna Nuclear Power Plant 5.6-4 Amendment 106

Reporting Requirements 5.6

2. As an alternative to the use of WCAP-14040-A Section 3.2 methodology, the existing Ginna specific LTOP Setpoint Methodology submitted to the NRC in the letter to Guy S.

Vissing (NRC) from Robert C. Mecredy (RG&E), "Application for Amendment to Facility Operating License, Revision to Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) Administrative Controls Requirements,"

Attachment VI, Section 3.2, dated September 29, 1997 and approved in letter to Robert C. Mecredy (RG&E) from S.

Singh Bajwa (NRC), "R.E. Ginna - Acceptance for Referencing of Pressure Temperature Limits Report,"

Revision 2 (TAC No. M96529), dated November 28, 1997, may be utilized.

d. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for revisions or supplement thereto.

5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged to date, and

g. The results of condition monitoring, including the results of tube pulls and in-situ testing.

R.E. Ginna Nuclear Power Plant 5.6-5 Amendment 106

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 106 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 RE. GINNA NUCLEAR POWER PLANT. LLC RE. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244

1.0 INTRODUCTION

By letter dated February 8, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML080460504), as supplemented on April 25, 2008 (ADAMS Accession No. ML081280095), and email dated January 7,2009 (ADAMS Accession No. ML090070445), RE. Ginna Nuclear Power Plant, LLC, the licensee for the RE. Ginna Nuclear Power Plant, submitted a license amendment request to revise Technical Specification (TS) 5.6.6, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," to change the method of determining reactor coolant system (RCS) pressure and temperature and low temperature over pressure (LTOP) limits.

The licensee's application of February 8, 2008, requested an implementation period of 60 days following NRC staff approval. The licensee's email of January 7,2009, requested that the time frame to implement the new license amendment be revised from 60 to 90 days.

The supplemental letter of April 25, 2008, and the email of January 7, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register.

2.0 REGULATORY EVALUATION

The NRC has established requirements in Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50) to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The staff evaluates the acceptability of a facility's proposed PTLR based on the following NRC regulations and guidance:

• Appendix G to 10 CFR Part 50, "Fracture toughness requirements," requires that facility poT limit curves for the RPV be at least as conservative as those obtained by applying the linear elastic fracture mechanics (LEFM) methodology of Appendix G to Section XI of the ASME Code (ASME Code, Appendix G methodology),

-2

• Regulatory Guide 1.99, Revision 2 (RG 1.99, Rev. 2), "Radiation Embrittlement of Reactor Vessel Materials," contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy (USE) resulting from neutron radiation,

• Standard Review Plan (SRP) Section 5.3.2, "Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock," provides an acceptable method for determining the P-T limit curves for territic materials in the beltline of the RPV based on the ASME Code,Section XI, Appendix G methodology,

• RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," provides methodologies for determining reactor vessel neutron fluences, and

• GL 96-03, "Relocation of Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," addresses the technical information necessary for a licensee's implementation of a PTLR.

3.0 TECHNICAL EVALUATION

3.1 Background The Ginna RCS pressure and temperature limits for heatup, cooldown, criticality, and hydrostatic testing as well as heatup and cool down rates are established and documented in the PTLR.

The power operated relief valve lift settings required to support the LTOP system, and the LTOP enable temperature are also established and documented in the PTLR. The PTLR is a Iicensee controlled document and is maintained separate from the TSs. Changes to the PTLR are made under the controls of 10 CFR 50.59.

The methodology used to develop the RCS heatup and cooldown curves and LTOP setpoints must receive prior review and approval by the NRC staff. The approved methodology is included in Ginna TS 5.6.6. The methodology to generate Ginna's current RCS heatup and cooldown curves is identified in TS 5.6.6, Section c.2, which references Westinghouse topical report WCAP-14040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Sections 1 and 2, January 1996. The methodology to generate Ginna's current LTOP setpoints is identified in TS 5.6.6, Section c.1 , which references site specific analyses using the RELAP5 computer program. It should be noted that while LTOP setpoints could be generated using WCAP-14040-NP-A, Revision 2, the site specific analyses referenced above is an acceptable alternative.

As the result of license renewal and the 16.8 percent extended power uprate license amendments for Ginna, it is necessary to employ current up-to-date methodology in support of continued plant operation. This is primarily due to the additional neutron fluence which will occur in the reactor vessel due to both license renewal and the extended power uprate. Thus, the NRC review will include the methodology used to generate the assumed neutron fluence over the duration of the operating license.

The licensee's application requested that TS 5.6.6 be revised such that the current reference to WCAP-14040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Sections 1 and 2, January 1996, be replaced with WCAP-14040-A, Revision 4, "Methodology used to Develop Cold

-3 Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004. The NRC has previously reviewed and approved WCAP-14040-A, Revision 4, and found it acceptable for referencing in licensing applications (ADAMS Accession No. ML050120209).

The licensee further requested that the site specific analyses currently referenced in TS 5.6.6 and currently used to generate the LTOP setpoints be retained as an approved, alternative method.

In support of its application, the licensee explained that the proposed poT limit methodology of WCAP-14040-A, Revision 4 incorporates the provisions of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Cases N-588, "Attenuation to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels,"

N-640, "Alternative Reference Fracture Toughness for Development of poT Limit Curves" and N-641 , "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements."

WCAP-14040-A, Revision 4, provides a generic methodology for generating poT limits and licensees referencing this topical report must generate a plant-specific version. WCAP-15885, Revision 0, "R.E. Ginna Heatup and Cooldown Limit Curves for Normal Operation," July 2002, was generated by the licensee for the specific application to the Ginna facility. The licensee stated that data generated from WCAP-15885 was input into the proposed PTLR that developed the RCS heatup and cooldown curves. Since WCAP-15885 (July 2002) predates WCAP 14040-A, Revision 4 (May 2004), both Westinghouse and the licensee reviewed WCAP-15885 and determined that it is consistent with WCAP-14040-A, Revision 4 methodology. The NRC staff requested that the proposed PTLR and detailed P-T limit information be submitted in order to ensure appropriate application of the new methodology and to allow comparison with the staff's confirmatory evaluation. Therefore, by letter dated April 25, 2008, the licensee provided the proposed PTLR and WCAP-15885.

3.2 Vessel Fluence Methodology The Ginna license renewal and extended power uprate license amendments will have a significant impact on the overall neutron fluence of the reactor vessel. Long term embrittlement of the reactor vessel and P-T heatup and cooldown curves are highly impacted by the assumed neutron fluence. Therefore, the NRC staff review included a detailed study of the methodology used to calculate long term neutron fluence.

The current poT limit curves and LTOP setpoints are valid to 32 EFPYs of operation, Le., to the end of the original 40-year operating license. WCAP-15885 states that the fluence methodology adheres to the guidance of RG 1.190, the cross section data is derived from EI\lDF/B-VI (BUGLE-96 cross section library), and the calculational methodology was approved in the context of WCAP-14040-A, Revision 4. The calculation uses the synthesis method and two, two-dimensional solutions (r, 6) and (r, z) to reproduce the (r, 6, z) flux (fluence) distribution. The neutron source was derived from fuel cycle design reports (for cycles 1 through 29) for each fuel cycle average assembly power, burnup, and axial distribution. Future cycles were represented as the average of cycles 26 to 29. In this manner, the clad metal interface azimuthal values at 0, 15, 30 and 45 degrees were calculated for 28, 32, 36, 40, 44, 48, 52, and 54 EFPYs of operation.

-4 WCAP-15885 includes a comparison of the measured to calculated results for the surveillance capsules that have been withdrawn (V, R, S, T) and the calculated lead factors for the remaining capsules P and N. The report's comparison of the calculated dosimeter response versus the corresponding measured value indicates excellent agreement.

The proposed PTLR includes P-T limit curves and LTOP setpoint limits for 47.3 EFPYs. This value accounts for the Ginna power uprate and license renewal. The NRC staff finds the fluence values in Table PTLR-3 of the proposed PTLR acceptable because they are an interpolation of the computed values. A linear interpolation is acceptable because, given (1) a relatively constant operating strategy and (2) after exposure beyond a marginal period of initial operation, reactor vessel neutron fluence increases linearly with exposure. Based on these considerations, the staff concludes that the RV fluence calculations are acceptable because (1) the fluence calculations are consistent with WCAP-14040-A, Revision 4, and the recommendations of RG 1.190, and (2) the NRC has previously approved the use of WCAP-14040-A, Revision 4, for referencing in licensing applications.

3.3 P-T Limits Licensee's Evaluation The licensee's adjusted reference temperature (ART) values and P-T limit curves, valid for up to 47.3 effective full-power years (EFPYs) of facility operation, are documented in the proposed PTLR. The licensee identified the limiting materials for the Ginna RPV as the intermediate-to lower shell forging circumferential weld and the intermediate shell forging. Both heatup and cooldown curves are composite curves with one part limited by the intermediate-to-Iower shell forging circumferential weld and the other by the intermediate shell forging. The key parameters in determining the licensee's ART values for the limiting materials for both the one-quarter RPV wall thickness (1/4t) and three-quarter RPV wall thickness (3/4t) locations are shown in Table 22 of WCAP-15885.

WCAP-15885 documents the detailed thermal analysis and fracture mechanics evaluations used to establish the proposed Ginna P-T limits. The RPV temperature distribution at any specified time during heatup or cooldown is generated analytically. Based on the subsequent thermal stress results, the applied thermal stress intensity factors (Kit) at the tip of the postulated flaw at the desired locations are generated for heatup and cooldown transients using the alternative thermal stress-based formula provided in the ASME Code,Section XI, Appendix G methodology. In the final step, the applied Kit values and the plane-strain fracture toughness (K 1c ) values at the crack tip were used to calculate the corresponding applied pressure stress intensity factors (Kip) at the tip of the postulated flaw at the 1/4t and 3/4t locations and then the pressure itself.

Staff Evaluation To evaluate the proposed P-T limits for Ginna, the NRC staff performed an independent calculation of the ART values for the limiting materials for the Ginna RPV using the methodology of RG 1.99, Rev. 2. The staff's ART values for the limiting materials at the 1/4t and 3/4t locations were calculated using materials information for Ginna in the NRC Reactor Vessel Integrity Database (RVID). Based on these calculations, the staff verified that the licensee's limiting materials are the intermediate-to-Iower shell forging circumferential weld and the

-5 intermediate shell forging. The staff determined that the licensee's ART values for the limiting materials of the Ginna RPV are consistent with the staff's values because the licensee's initial reference temperature (RT NDT) and copper (Cu) and nickel (Ni) values for the limiting materials based on the Certified Material Test Report are almost the same as those in the RVID.

The NRC staff, however, noted that a significant difference exists between the licensee's and the RVID's information on the Cu content of the nozzle forging material. The RVID indicated that the default Cu value of 0.35 percent is used for the nozzle forging because the Certified Material Test Report lacks information on this material. WCAP-15885 reported a Cu content of 0.07 percent for the nozzle forging with the justification that the nozzle forging was made from the same material as the intermediate and lower shell forgings at the same time period and, therefore, it is a safe assumption to assign the highest Cu value of the known Ginna RPV forgings to the nozzle forging material.

The NRC staff reviewed the Cu information for RPV forgings from all plants in the RVID and determined that using the RG default Cu value of 0.35 percent for the Ginna RPV nozzle forging was unnecessarily conservative. The staff also found that the WCAP's justification for using a Cu value of 0.07 percent for this material was not appropriate either. The staff found that there are multiple plants which have very low copper content RPV shell forgings (around 0.07 percent) but have moderate Cu content RPV nozzle forgings (around 0.16 percent). Consequently, the staff determined that an appropriate Cu value for Ginna RPV nozzle forging should be close to 0.17 percent, the highest Cu content for the RPV shell and nozzle forgings of the entire domestic fleet based on the RVID. The staff concluded that using this Cu value for the nozzle forging will not change the selection of the RPV limiting materials and, therefore, will have no impact on the proposed Ginna P-T limits.

The NRC staff then evaluated the licensee's poT limit curves for acceptability by performing independent calculations using the ASME Code,Section XI, Appendix G methodology (as indicated by SRP 5.3.2) based on information submitted by the licensee. The proposed poT limits were based on the use of the K lc curve in accordance with Section XI of the ASME Code (the 2004 Edition of the ASME Code, endorsed in 10 CFR 50.55a).

ASME Code,Section XI, Appendix G permits two approaches to calculate Kit. They include (1) use of the bounding Kit formulas based on heatup and cooldown rates, or (2) use of the Kit formulas based on the thermal stress distribution from a thermal model (e.g., a finite element model, FEM) for heatup and cooldown. WCAP-15885 used the latter approach but, unlike some recent WCAPs on P-T limits prepared for other dockets, WCAP-15885 did not provide relevant RPV coolant temperature, metal temperature, and the applied Kit values at the tip of the postulated flaw at the 1/4t location during cooldown and at the 1/4t and 3/4t locations during heatup. In order to verify the Ginna licensee's P-T limit curves, the NRC staff applied previous review experience on typical poT limit curves prepared by Westinghouse, such as WCAP-16346 NP prepared for Comanche Peak, Units 1 and 2 (NRC safety evaluation in ADAMS Accession No. ML070320825). This NRC's experience indicates that, Kit values based on the FEM results are about 75% of those based on the ASME Code,Section XI, Appendix G bounding Kit formula, while the temperature difference between water and RPV metal at the 1/4t or 3/4t from these two approaches is small. Applying this experience, the staff verified that the proposed poT limits are valid and satisfy the requirements of Appendix G to 10 CFR Part 50. Hence, the licensee's proposed poT limit curves are acceptable for operation of the Ginna RPV for 47.3 EFPY.

-6 P-T Limits Summary Based on the NRC staff's evaluation and its review of the information provided by the licensee, the staff concludes that the proposed PTLR for Ginna continues to meet the recommendations of GL 96-03 and is, therefore, acceptable.

The NRC staff finds that the licensee has defined an acceptable PTLR methodology which is consistent with the regulatory requirements given in Section 2.0 of this safety evaluation. This methodology is documented in WCAP-14040-A, Revision 4. WCAP-15885 is a plant-specific report for the Ginna facility and has been found to be consistent with the methodology of WCAP 14040-A, Revision 4. The TS revision to reflect the use of this methodology is appropriate.

Furthermore, based on the above evaluations, the staff verified that the PTLR methodology has been implemented appropriately, so the proposed poT limits, which are valid for 47.3 EFPY, satisfy the requirements in Appendix G to 10 CFR Part 50 and Appendix G to Section XI of the ASME Code.

LTOP Setpoints The submittal requests that the site-specific analyses and the current limits for the LTOP setpoints using the RELAP5 computer program be maintained as an alternate method for the calculation of the LTOP setpoints. The licensee's rationale is that the new poT curves (which also determine the setpoints) are bounded by the original setpoints, therefore, the old setpoints can be maintained without the need to recalculate them using the WCAP-14040-A methodology.

The generic method provided in WCAP 14040-A relies on LOFTRAN analyses, which are, as approved by the staff, acceptable for evaluating the LTOP transients that are used to determine or confirm the LTOP setpoints. The licensee's plant-specific analysis uses RELAP5. The NRC staff is familiar with RELAP5 and finds that the proposed plant-specific LTOP setpoint confirmation is acceptable because RELAP5 is a thermal-hydraulic system code that is capable of modeling the LTOP transients. The I\IRC staff concludes that either method, WCAP-14040-A or RELAP5, provides a reasonable prediction of plant behavior during the LTOP transient.

Therefore, the licensee's proposed use of RELAP5 is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (73 FR 19111). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b)

-7 no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Simon Sheng Lambros Lois Date: Felmmy 23, zr>>

February 23, 2009 Mr. John 1. Carlin Vice President R E. Ginna Nuclear Power Plant RE. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519

SUBJECT:

RE. GINNA NUCLEAR POWER PLANT - AMENDMENT RE: REVISED METHODOLOGY FOR DETERMINING REACTOR COOLANT SYSTEM PRESSURE AND TEMPERATURE AND LOW TEMPERATURE OVER-PRESSURE LIMITS (TAC NO. MD8069)

Dear Mr. Carlin:

The Commission has issued the enclosed Amendment No.1 06 to Renewed Facility Operating License No. DPR-18 for the RE. Ginna Nuclear Power Plant. This amendment is in response to your application dated February 8, 2008, as supplemented by letter dated April 25, 2008, and email dated January 7,2009.

The amendment revises Technical Specification 5.6.6, "Reactor Coolant System (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," to change the method of determining reactor coolant system pressure and temperature and low temperature over pressure limits. The new PTLR methodology is documented in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," dated May 2004.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRA!

Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244

Enclosures:

1. Amendment No.1 06 to Renewed License No. DPR-18

2. Safety Evaluation cc w/encls: Distribution via Listserv ADAMS Accession No ML08350806 OFFICE LPLI-1\PM LPLI-1\LA OGC CVIB\BC SRXB\BC LPLI-1\BC NAME DPickett SLittie BMizuno MMitchell as GCranston as JBoska for signed on signed on MKowal DATE 01 /10/09 01/09/09 02/18/09 12/09/08 06/04/08 02/19/09 Official Record Copy

DATED: February 23,2009 AMENDMENT NO. 106 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 R.E. GINNA NUCLEAR POWER PLANT PUBLIC LPLI-1 R/F M. Kowal RidsNrrDorlLpla S. Little D. Pickett RidsNrrPMDPickett OGC RidsOgcMailCenter ACRS RidsAcrsAcnwMailCenter M. Mitchell RidsNrrDciCvib G. Cranston RidsNrrDssSrxb S. Sheng, CVIB G. Dentel, RI cc: Listserv