ML14126A378

Issuance of License Amendment Nos. 250 and 254 Regarding Change to Technical Specification 5.6.5, Reactor Coolant System Pressure and Temperature Limits Report

ML14126A378
Person / Time
Site:	Point Beach
Issue date:	06/30/2014
From:	Beltz T Plant Licensing Branch III
To:	Mccartney E Point Beach
Beltz T
References
TAC ME0532, TAC ME0533
Download: ML14126A378 (26)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 30, 2014 Mr. Eric McCartney Site Vice President NextEra Energy Point Beach, LLC 6610 Nuclear Road Two Rivers, WI 54241

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2- ISSUANCE OF AMENDMENT REGARDING CHANGE TO TECHNICAL SPECIFICATION 5.6.5, "REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)" (TAC NOS. MF0532 AND MF0533)

Dear Mr. McCartney:

The U.S. Nuclear Regulatory Commission has issued Amendment Nos. 250 and 254 to renewed Facility Operating License Nos. DPR-24 and DPR-27 for the Point Beach Nuclear Plant (Point Beach), Units 1 and 2, respectively. These amendments consist of changes to the facility technical specifications (TSs) in response to your application dated January 15, 2013, as supplemented by letters dated March 1, April 18, and September 12, 2013, and March 11, 2014.

The amendment modifies the Point Beach, Unit 1 and 2, TSs to revise TS 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," to allow the use of two new methodologies for determining RCS pressure and temperature limits. In addition, an exemption was requested from portions of Appendix G to 10 CFR Part 50, and 10 CFR 50.61.

The exemption is addressed under separate correspondence (Agencywide Documents Access and Management System Accession No. ML14126A594).

A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301

Enclosures:

1. Amendment No. 250 to DPR-24

2. Amendment No. 254 to DPR-27

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY POINT BEACH. LLC DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 250 License No. DPR-24

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

A. The application for amendment by NextEra Energy Point Beach, LLC (the licensee) dated January 15, 2013, as supplemented by letters dated March 1, 2013, April 18, 2013, September 12, 2013, and March 11, 2014, 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 4.B of the Renewed Facility Operating License No. DPR-24 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 250, are hereby incorporated in the Enclosure 1

renewed operating license. NextEra Energy Point Beach shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 180 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and the Renewed Facility Operating License Date of Issuance: June 30, 2014

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT. UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 254 License No. DPR-27

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

A. The application for amendment by NextEra Energy Point Beach, LLC (the licensee) dated January 13, 2013, as supplemented by letters dated March 1, 2013, April 18, 2013, September 12, 2013, and March 11, 2014, 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 4. B of the Renewed Facility Operating License No. DPR-27 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as Enclosure 2

revised through Amendment No. 254, are hereby incorporated in the renewed operating license. NextEra Energy Point Beach shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 180 days.

FOR THE NUCLEAR REGULATORY COMMISSION Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and the Renewed Facility Operating License Date of Issuance: June 30, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 250 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 AND LICENSE AMENDMENT NO. 254 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 DOCKET NOS. 50-266 AND 50-301 Replace the following pages of Renewed Facility Operating License Nos. DPR-24 and DPR-27, and Technical Specifications with the attached revised page. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

Renewed Facility Operating License REMOVE INSERT 3 3 Technical Specifications REMOVE INSERT 5.6-5 5.6-5

D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach 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 E. Pursuant to the Act and 10 CFR Parts 30 and 70, NextEra Energy Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.

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

A. Maximum Power Levels NextEra Energy Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1800 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 250, are hereby incorporated in the renewed operating license.

NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.

C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.

Renewed License No. DPR-24 Amendment No. 250

C. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach to receive, possess and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed source for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach to receive, possess and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30 and 70, Next Era Energy Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.

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

A. Maximum Power Levels NextEra Energy Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1800 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 254, are hereby incorporated in the renewed operating license.

NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.

C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.

Renewed License No. DPR-27 Amendment No. 254

Reporting Requirements 5.6 5.6 Reporting Requirements

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

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, LTOP enabling, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

(1) LCO 3.4.3, "RCS Pressure and Temperature (PIT) Limits" (2) LCO 3.4.6, "RCS Loops-MODE 4" (3) LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled" (4) LCO 3.4.1 0, "Pressurizer Safety Valves" (5) LCO 3.4.12, "Low Temperature Overpressure Protection (L TOP)"

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the NRC Letters dated October 6, 2000, July 23, 2001, and October 18, 2007, and June 30, 2014.

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

5.6.6 PAM Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Point Beach 5.6-5 Unit 1 - Amendment No. 250 Unit 2 -Amendment No. 254

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 250 and 254 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27 NEXTERA ENERGY POINT BEACH. LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301

1.0 INTRODUCTION

By letter dated January 15, 2013 (Reference 1), as supplemented on March 1, April 18, and September 12, 2013, and March 11, 2014 (References 2, 3, 4 and 5, respectively), NextEra Energy Point Beach, LLC (NextEra or the licensee) submitted a license amendment request (LAR) for the Point Beach Nuclear Plant (Point Beach), Units 1 and 2. NextEra requested to revise its Technical Specifications (TSs) regarding the Pressure Temperature Limit Report (PTLR), which contains the pressure-temperature (P-T) limit curves for the reactor pressure vessel (RPV) utilizing BAW-2308, Revision 1-A and 2-A, "Initial RTNor of Linde 80 Weld Materials" (References 6, 7, 8, 9), and WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (Reference 10). The LAR application included the P-T limit curves and supporting information representing operation to 50 effective full power years (EFPY). The BAW-2308 reports were incorporated to allow for an alternative estimation of the initial nil-ductility reference temperature {RT Nor) of Linde 80 weld materials.

The use of the BAW-2308 reports requires exemption from certain requirements in Section 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events," of Part 50 of Title 10 of the Code of Federal Regulations (1 0 CFR Part 50) and Appendix G, "Fracture toughness requirements," to 10 CFR Part 50. An exemption has been granted separately from this safety evaluation (SE) (Agencywide Documents Access and Management System Accession No. ML14126A594).

In addition, the NextEra provided revised RT Prs values, which is a material's RT Nor based on the RPV end-of-license (EOL) inside diameter (I D) neutron fluence, for the purpose of protecting the RPV against pressurized thermal shock {PTS) events.

The licensee's supplemental letters dated March 1, April 18, and September 12, 2013, and March 11, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original Enclosure 3

proposed no significant hazards consideration determination as published in the Federal Register on June 11, 2013 (78 FR 35062).

2.0 REGULATORY EVALUATION

The regulations in 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," allow a licensee to apply to amend the license or permit. The regulations in 10 CFR 50.92, "Issuance of Amendment," specify that the NRC staff will be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate in determining whether an amendment will be issued to the applicant.

The U.S. Nuclear Regulatory Commission (NRC) has established requirements in10 CFR Part 50 to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The NRC staff evaluates the acceptability of a facility's proposed P-T limits based on the following NRC regulations and guidance:

• Appendix G of 10 CFR Part 50, "Fracture Toughness Requirements" Appendix G of 10 CFR Part 50 requires that facility P-T limits for the RPV be at least as conservative as those obtained by following the methods of analysis and the margins of safety in Appendix G of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).

• Appendix H of 10 CFR Part 50, "Reactor Vessel Material Surveillance Program Requirements" Appendix H of 10 CFR Part 50 establishes requirements related to facility RPV material surveillance programs.

• Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials" RG 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels.

• Generic Letter (GL) 92-01, Revision 1, "Reactor Vessel Structural Integrity" GL 92-01, Revision 1, requested that licensees submit the RPV data for their plants to the NRC staff for review given concerns regarding neutron embrittlement of the Yankee Nuclear Power Station reactor vessel.

• GL 92-01, Revision 1, Supplement 1, "Reactor Vessel Structural Integrity" and

GL 92-01, Revision 1, Supplement 1, requested that licensees provide and assess data from other licensees that could affect their RPV integrity evaluations.

• NUREG-0800, NRC Standard Review Plan (SRP) Section 5.3.2.

SRP Section 5.3.2 provides an acceptable method for determining the P-T limits for ferritic materials in the beltline of the RPV based on the ASME Code Appendix G methodology.

The most recent version of Appendix G to Section XI of the ASME Code which has been endorsed in 10 CFR 50.55a, and therefore by reference in 10 CFR Part 50, Appendix G, is the 2010 Edition of the ASME Code. This edition of Appendix G to Section XI of the ASME Code incorporates the provisions of ASME Code Case N-588, "Attenuation to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels," and ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves."

Additionally, Appendix G to 10 CFR Part 50 imposes minimum head flange temperatures when system pressure is at or above 20 percent of the preservice hydrostatic test pressure.

GL 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," identifies information that a licensee can use to establish an acceptable PTLR methodology and an acceptable PTLR. The PTLR also needs to comply with TSTF-419, which documents revised guidance for a plant's PTLR. GL 96-03 states that subsequent changes in the methodology must be approved by a license amendment, and that 10 CFR 50.59 does not apply.

For protection against PTS events, 10 CFR 50.61 specifies screening criteria for the calculated RPV beltline material RT PTS* For materials above the screening criteria, licensees may submit plant-specific analysis described in 10 CFR 50.61, or alternate criteria provided in 10 CFR 50.61a.

The NRC staff also considered the following regulatory requirements in 10 CFR 50.36, "Technical specifications," in which the Commission established its regulatory requirements related to the contents of the TS. Specifically, 10 CFR 50.36(a)(1) states that "Each applicant for a license authorizing operation of a production or utilization facility shall include in its application proposed technical specifications in accordance with the requirements of this section."

3.0 TECHNICAL EVALUATION

3.1 Licensee's Evaluation The revised P-T limits are based on application of the WCAP-14040-A, Revision 4, methodology (henceforth, the Westinghouse methodology) to Point Beach, Units 1 and 2. The Westinghouse methodology is an approved generic methodology for generating P-T limits and Cold Overpressure Mitigation System (COMS) setpoints based on the plant-specific adjusted reference temperatures (ARTs) consistent with NRC PTLR development guidance. The Westinghouse methodology was implemented via analysis contained in WCAP-16669-NP, "Point Beach Unit 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," dated

December 2006, which was submitted as part of this application via the March 1, 2013 (Reference 2).

Two sets of input values have changed since the generation ofWCAP-16669-NP, namely fluence and RPV beltline material properties values. The results of WCAP-16669-NP were originally drafted for 53 EFPY based on the assumption of a 10 percent power up rate beginning in 2008, and utilizing the material properties found in BAW-2308, Revision 1-A. The licensee's response to request for additional information (RAI) 1 regarding inconsistent neutron fluence values in the submittal and the supplements indicated that the results of WCAP-16669-NP were compared to fluence projections supporting the 2009 Point Beach extended power uprate (EPU) application (Reference 11 ), and the comparison established that the results of WCAP-16669-NP would be bounding until 50 Effective Full-Power Years (EFPY) under the current fluence estimates.

For the limiting RPV beltline materials, the licensee identified the intermediate-to-lower shell circumferential weld and the lower shell axial weld for Point Beach, Unit 1. For Point Beach, Unit 2, the limiting RPV beltline materials were cited as the intermediate to lower shell circumferential weld and the intermediate shell forging. ART values were calculated for 50 EFPY. The parameters used to determine the licensee's ART values for the limiting materials at one-quarter of the RPV wall thickness (1/4T) and three-quarter (3/4T) location for 50 EFPY are shown in Table 4-10 of WCAP-16669-NP. In the case of Linde 80 materials, the properties in WCAP-16669-NP have been updated using BAW-2308, Revision 2-A.

The WCAP-16669-NP report contained P-T limits for several scenarios, varying EFPY and whether hafnium rods were removed. The applicant selected the most conservative P-T limits from WCAP-16669-NP on the basis that both the fluence and material property values used in generating the selected curves bound current estimates for plant conditions up to 50 EFPY.

The applicant also submitted updated RT PTS values for the purpose of confirming continued PTS rule compliance using the new fluence estimates and results from BAW-2308, Revision 1-A.

3.2 NRC Staff Evaluation 3.2.1 PTLR Implementation The licensee utilized the Westinghouse methodology to generate their P-T limits. The Westinghouse methodology was approved for use in generating PTLRs by the NRC staff.

As noted in Section 2.0 of this SE, GL 96-03 provides that a licensee may use seven technical criteria to demonstrate the acceptability of its PTLR. The NRC staff examined the proposed PTLR and determined that it was developed appropriately based on the Westinghouse methodology and meets the seven technical criteria:

(1) The PTLR methodology describes the transport calculation methods including computer codes and formula used to calculate neutron fluence.

The Westinghouse methodology specified how the fluence was determined. The NRC staff approved the fluence prediction method within the methodology. The applicant further specified

neutron fluence values derived through this method within the context of the application. Two important sets of values were presented. First, the licensee cited 50 EFPY values from the Point Beach, Units 1 and 2, EPU application. These values were then compared to the values used in WCAP-16669-NP and shown to be bounding. The NRC staff reviewed these values and finds them acceptable.

(2) The PTLR methodology describes the surveillance program.

The relevant changes to the Reactor Vessel Material Surveillance Program (RVMSP) were documented and approved through the Point Beach, Units 1 and 2, EPU application, and the current RVMSP schedule was provided by the applicant within the updated PTLR.

(3) The PTLR methodology describes how the low temperature overpressure protection (LTOP) system limits are calculated applying system/thermal hydraulics and fracture mechanics.

The Westinghouse methodology used in the application is approved for the purpose of generating LTOP limits relevant to the criteria. The applicant provided this limit in Reference 2.

(4) The PTLR methodology describes the method for calculating the ART values using RG 1.99, Revision 2.

The submittal indicated that RG 1.99, Revision 2, provides the methods for determining the ARTs for the beltline materials. The initial RT NDT values for Linde 80 materials were updated via BAW-2308, Revision 1-A, as approved in response to an exemption request submitted by the licensee on January 15, 2013. The NRC staff reviewed the information provided, independently verified the calculations, and found the information acceptable.

(5) The PTLR methodology describes the application of fracture mechanics in the construction of P-T limits based on ASME Code,Section XI, Appendix G, and the SRP.

Page 1 of the PTLR states that the P-T limits were calculated in accordance with the Westinghouse methodology. The Westinghouse methodology applies fracture mechanics in construction of P-T limits based on ASME Code,Section XI, Appendix G, and the SRP. The NRC staff reviewed the Westinghouse methodology, and found that it met the fifth criteria.

(6) The PTLR methodology describes how the minimum temperature requirements in Appendix G to 10 CFR Part 50 are applied to P-T limits for boltup temperature and hydrotest temperature.

Again, the licensee referenced the Westinghouse methodology and asserted that it is sufficient because the report contains detailed information regarding the minimum temperature requirements for boltup temperature and hydrotest temperature. The NRC staff reviewed the Westinghouse methodology, and found that it met the sixth criteria.

(7) The PTLR methodology describes how the data from multiple surveillance capsules are used in the ART calculation.

Section 2.4 of the Westinghouse report contains detailed information regarding how data from the surveillance capsules are used in the ART calculation. The NRC staff reviewed the report and approved the Westinghouse methodology under the seventh criteria.

In summary, the implementation of the Point Beach, Units 1 and 2, PTLRs are acceptable. The NRC staff notes that these PTLRs do not extend to the end of the period of extended operation.

Rather, the submitted PTLRs will remain adequate approximately through the end of 2029 as stated in the Reference 2.

3.2.2 BAW-2308-1A, BAW-2308-2A Implementation Material properties for unirradiated Linde-80 weld materials for Point Beach originate from BAW-2308, Revisions 1-A. The applicant conditions set forth for use of BAW-2308, Revision 1-A, are as follows:

(1) The heat specific value of IRTTO (initial RTNoT) must be used if it is higher than the "all heats" value.

(2) When the NRC staff accepted values of IRTTO are used, the methodology of RG 1.99, Revision 2, may be used to assess the shift. Additionally, a minimum chemistry factor of 167 degrees Fahrenheit CF) must be applied in this case.

(3) When the NRC staff accepted values of IRTTO are used, a value of at.= 28°F must be applied.

(4) Any licensee using BAW-2308, Revision 1 must request an exemption, per 10 CFR 50.12, from the requirements of Appendix G to 10 CFR Part 50 and 10 CFR 50.61.

The NRC staff confirmed that the applicant fulfilled the conditions for BAW-2308, Revision 1-A.

Since the facility does not have welds with the specific heat numbers referenced in BAW-2308, Revision 2-A, this specific revision has no current applicability. Based on the above, the NRC staff concludes that the applicant has adequately implemented BAW-2308, Revisions 1-A and 2-A. ,

3.2.3 Fluence Comparisons Pages 2-3 of Enclosure 1 to Reference 2 states:

WCAP-16669-NP ... PT curves were generated based on the reactor vessel information and calculated fluencies based on the plant operating conditions listed in Table 1-1 of WCAP-16669-NP.

The fluence projections were based on the FERRET Code best-estimate values verified by Westinghouse in letter WEP-06-13, dated February 14, 2006, as approved for use by the NRC in WCAP-16083-NP-A, Revision 0.

To establish the applicability of the P-T limits generated for WCAP-16669-NP to this LAR, the licensee compared the individual material fluence estimates to those generated in support of the Point Beach, Units 1 and 2, EPU submittal and found the fluence values for 50

EFPY under EPU conditions to be bounded by those used in the most limiting set of P-T limits of WCAP-16669-NP. Consequently, the projections of WCAP-16669-NP can be considered sufficient for Point Beach, Units 1 and 2, at 50 EFPY.

3.2.4 P-T Limits The proposed P-T limits were generated using the Westinghouse methodology and ASME Code,Section XI, Appendix G. The P-T limits based on WCAP-16669-NP were presented in the PTLR attached to Reference 2.

3.2.4.1 RPV Beltline Shell Region To evaluate the proposed Point Beach, Units 1 and 2, RPV beltline P-T limits, the NRC staff first confirmed the licensee's selection of limiting materials. For the Point Beach beltline materials, the NRC staff found that the initial RT NoT. copper (Cu), and nickel (Ni) values are in agreement with the information in the NRC's Reactor Vessel Integrity Database (RVID) and BAW-2308, Revisions 1-A and 2-A for Linde 80 materials (Revision 2-A being inapplicable). The licensee calculated ART values for the RPV 1/4T and 3/4T locations. The NRC staff independently calculated ART values and confirmed the accuracy of the calculated ART values. The NRC staff also confirmed through independent calculation that the updated material chemistry and fluence result in nearly identical ART values to those in WCAP-16669-NP, hence the results presented in WCAP-16669-NP, with a revised EPFY value of 50, are acceptable for this application. The NRC staff then generated P-T Limit curves to compare with the applicant generated curves. The staff found that both sets of curves were in good agreement and consistent with the requirements of Appendix G to 10 CFR Part 50, and ASME Code,Section XI, Appendix G.

The licensee made use of the Westinghouse methodology in generating P-T limits. The NRC staff verified that the P-T limits were consistent with the Appendix G to 10 CFR Part 50 requirements for the minimum metal temperature of the closure head flange and vessel flange regions. In the case of Point Beach, this resolved as "notches" in the P-T limits consisting of vertical lines at constant temperature above approximately 621 psig in the licensee's proposed P-T limits. For all Point Beach, Unit 1 and 2, curves, the lowest temperature is defined by the minimum boltup temperature of 60°F. The ASME Code,Section XI, Appendix G requires that when the flange and adjacent shell regions are stressed by the full intended bolt preload under 20 percent of the preservice hydrostatic test pressure, the temperature at these regions must be at least their RTNoT temperatures. WCAP-16669-NP showed that the highest RT NOT value for these regions for the Point Beach units is 60°F. Hence, the NRC staff confirmed that the lowest temperature segment of the P-T limits also meet the ASME Code,Section XI, Appendix G requirements.

3.2.4.2 Ferritic Reactor Coolant Pressure Boundary (RCPB) Components Outside of the RPV Beltline Shell Region The regulations in 10 CFR Part 50, Appendix G, and Paragraph IV.A state the following:

The pressure-retaining components of the reactor coolant pressure boundary that are made of ferritic materials must meet the requirements of the [ASME Code, Section Ill], supplemented by

the additional requirements set forth in [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 the ferritic materials in the RPV beltline, as well as ferritic materials not in the RPV beltline. Further, 10 CFR Part 50, Appendix G, requires that all ferritic RCPB components must meet the applicable ASME Code, Section Ill, requirements. The relevant ASME Code, Section Ill, requirements that will affect the P-T limits are the lowest service temperature requirement of NB-2332(b) for piping, pumps, and valves, and the fracture toughness requirements of NB-3211(d) for vessels.

The NRC staff noted that P-T limit calculations for ferritic RCPB components that are not RPV beltline shell materials may define P-T curves that are more limiting than those calculated for the RPV beltline shell materials because:

1. RPV nozzles, penetrations, and other discontinuities may exhibit significantly higher stresses than those for the RPV beltline shell region. These higher stresses can potentially result in more restrictive P-T limits, even if the RT NDT for these components is not as high as that of RPV beltline shell materials that have simpler geometries.

2. Ferritic components that are not part of the RPV may have initial RT NDT values, which may define a more restrictive lowest operating temperature in the P-T limits than those for the RPV beltline shell materials.

Therefore, in RAI 4, the NRC staff requested that the licensee describe how the proposed P-T limit curves and the methodology used to develop these curves considered all RPV materials (beltline and non-beltline) and the replacement ferritic RCPB materials, consistent with the requirements of 10 CFR Part 50, Appendix G.

In its response to RAI 4 in Reference 4, the licensee stated that WCAP-14040-A, Revision 4, did not consider the RPV inlet and outlet nozzles. Therefore, to demonstrate that the 50 EFPY P-T limit curves are bounding for the entire RPV, the licensee provided component-specific P-T limit curves at 60 EFPY for the RPV inlet and outlet nozzles.

Evaluation of Response to RAI 4 - RPV Inlet/Outlet Nozzle P- T Limit Curves Tables 2 and 3 of the RAI response include 60 EFPY ART calculations for the Point Beach, Units 1 and 2, inlet and outlet nozzle forgings. There are two inlet nozzles and two outlet nozzles, based on the Westinghouse two-loop design. The critical input parameters used for the nozzle ART calculations- the initial RT NoT value, the Cu and Ni contents, and the neutron fluence value for each nozzle -were not provided in the PTLRs. The NRC staff's evaluation of these parameters is discussed below.

The initial RT NDT values, the Cu and Ni contents, and the neutron fluence values for the Point Beach inlet and outlet nozzles are provided in Tables 2 and 3 of the licensee's RAI response.

The RAI response states that the initial RT NDT values for the inlet and outlet nozzles were obtained using the proprietary Boiling Water Reactor Vessel and Internals Project (BWRVIP) technical report, BWRVIP-173-A (Reference 12).

Considering that (1) the Charpy V-notch impact energy curves for all eight of the Point Beach, Units 1 and 2, inlet and outlet nozzle materials were from the certified material test reports (CMTRs) data, (2) the 50 ft-lb transition temperatures were conservatively increased by 30oF to account for lack of information regarding specimen orientation, and (3) the approved methodology in BWRVIP-173-A was used for estimating the initial RT NOT values, the NRC staff determined that the proposed initial RT NOT values for the subject nozzles are acceptable.

Regarding the Cu and Ni contents for the inlet and outlet nozzle materials, the licensee stated that the nozzle Cu and Ni contents for Point Beach, Unit 1, were based on CMTRs. For Point Beach, Unit 2, although the Ni content is based on CMTRs, the Cu content is based on the "best-estimate" generic value recommended by BWRVIP-173-A when no measured values for the particular heat of material are available. The NRC staff finds the licensee's determination of the Ni contents acceptable because they are based on CMTRS. The NRC staff also finds the licensee's application of the BWRVIP-173-A methods acceptable for determining the updated generic Cu content because the recommendations of the report are based on analyses of an industry-wide database of SA-508, Class 2 forging material chemistries from both BWRs and PWRs. Further, the NRC staff accepts the margin term values because they were determined based on RG 1.99, Revision 2, procedures.

Regarding the neutron fluence values listed in Table 1 of the RAI 4 response, the licensee stated that the nozzle neutron fluence values were calculated in accordance with the provisions of RG 1.190. Furthermore, the NRC staff noted two conservatisms in the nozzle neutron fluence calculations: firstly, the calculations were performed at the lowest extent of the nozzles, at an elevation lower than the actual elevation of the postulated flaw at the inside corner of the nozzle; secondly, the calculations were for 60 EFPY, even though the proposed P-T limits are for 50 EFPY. Therefore, the NRC staff considers the Table 1 neutron fluence values acceptable for the subsequent ART calculations for the inlet and outlet nozzle materials.

Regarding the ART calculations, the licensee revealed that the 60 EFPY nozzle ARTs listed in Table 2 and 3 of the RAI 4 response were calculated at the RPV clad/base metal interface rather than the 1/4T location, and they were calculated using Position 1.1 of RG 1.99, Revision 2. The NRC staff noted that using the neutron fluence at the RPV clad/base metal interface for the 1/4T location is very conservative and using Position 1.1 is appropriate when surveillance data is not available. Therefore, the staff finds that these 60 EFPY ART values are acceptable for generating P-T limit curves for the inlet and outlet nozzles.

The licensee selected the most limiting nozzle ART value for each unit to generate a bounding set of nozzle P-T limit curves for the 100°F per hour cooldown transient. The limiting 60 EFPY ART value is 13.1 oF for the outlet nozzle BT2305 for Unit 1 and 2.8°F for the inlet nozzle 9-5414 for Unit 2.

The licensee generated the P-T limit curves for the 1OOoF per hour cool down transient and the steady state condition for the limiting inlet nozzle in Figure 1 and for the limiting outlet nozzle in Figure 2 of the RAI 4 response. Also plotted in each figure for comparison are the cooldown curves of 20°F to 1OOoF per hour, with an increment of 20°F, for the limiting beltline shell materials good for both Units. The licensee stated that the method used for determining the applied stress intensity factor (K1) due to the pressure and thermal gradient is based on the ORNL/TM-201 0/246 report (Reference 13).

The licensee stated that the through-wall stress distributions in the nozzle inside corner region were fitted based on a third-order polynomial, and the coefficients from the polynomial stress distribution were used to compute K1 values for a postulated 1/4T inside corner flaw. The NRC staff confirmed that both the stress distribution and the K1 formulas in the RAI 4 response were from Reference 13 for a rounded nozzle corner. Further, this K1 formula has been approved by the NRC for implementation in RPV nozzle K1 calculations, as documented in the SE for BWR Owners Group Licensing Topical Report SIR-05-044-A (Reference 14). Hence, the NRC staff finds the licensee's method for calculating the K1 values for the nozzles acceptable.

The licensee also indicated that nozzle P-T limit curves were not provided for heat-up conditions, as they would be less limiting than the nozzle P-T limit curves for cool down conditions at the 1/4T location.

The NRC staff noted in Figures 1 and 2 that the beltline cooldown P-T limits are far more limiting than either the inlet nozzle cooldown P-T limits or the outlet nozzle cooldown P-T limits (the difference exceeds 150°F at any pressure). This phenomenon is primarily caused by the much smaller 1/4T ART values of the limiting inlet and outlet nozzles as compared to the limiting beltline material. Since the similar magnitude of difference in 3/4T ART values also exists between the limiting nozzle material and the limiting beltline material, the NRC staff concluded that the beltline heatup P-T limits will be more limiting than either the inlet nozzle heatup P-T limits or the outlet nozzle heatup P-T limits. Consequently, the NRC staff finds the licensee's focus on cooldown P-T limit calculations justifiable and the licensee's conclusion that the TS P-T limit curves are controlling for the entire RPV acceptable. The first concern of RAI 4 is resolved.

In response to the second concern of RAI 4, the licensee stated that the Point Beach, Units 1 and 2, reactor coolant systems do not have ferritic materials in the Class 1 piping, pumps, or valves. Therefore, the LST requirements of the ASME Code, Section Ill, NB-2332(b) are not applicable to the Point Beach units. The NRC staff agrees with this determination because the requirements of NB-2332(b) and 10 CFR Part 50, Appendix G are only applicable to ferritic components of the RCPB, and not to stainless steel components. Accordingly, the licensee focused its evaluation on the replacement ferritic components of the RCPB.

Evaluation of Response to RAJ 4 - Replacement Ferritic Components of the RCPB The licensee stated in its RAI 4 response that both units have replaced their steam generators (SGs) since original construction. The licensee indicated that the Point Beach, Unit 2, replacement SGs (RSG) meet the ASME Code, Section Ill, Appendix G requirements for protection against non-ductile fracture analysis, and no further consideration is necessary for these components with regards to P-T limits. However, since a similar evaluation was not performed for the Point Beach, Unit 1, RSGs, the licensee presented the P-T limits for the RSGs in the RAI 4 response to demonstrate that the beltline P-T limits remain limiting.

In this demonstration, two RSG locations were reviewed based on the ASME Code,Section XI, Appendix G fracture mechanics analysis: the RSG channel head to tube sheet region lower junction and the primary nozzle knuckle region. Both were selected because they are highly stressed and contain discontinuities. The licensee stated that its fracture mechanics analysis of the RSG tube sheet to the channel head junction was performed based on postulated inside surface axial and circumferential flaws with aspect ratios of 6:1, per the ASME Code,Section XI, Appendix G. The licensee analyzed the case for a 1OOoF per hour cooldown transient, as it will

produce high tensile stresses on the inside surface for pressure and thermal transients. The licensee stated that the limiting tensile stress components are chosen to determine the stress intensity factors, with the appropriate safety factors, consistent with ASME Code,Section XI, Appendix G.

The licensee stated that the initial RT NDT value of 60°F for this location is based on the design specification for the Point Beach, Unit 1 RSG ferritic materials (base metals and welds). Based on these analyses, the licensee generated a P-T limit curve for the Unit 1 RSG tube sheet to channel head junction, and compared it to the 1OOoF per hour cooldown curve for the limiting beltline shell materials for Point Beach, Units 1 and 2. These curves, which are provided in Figure 3 of the RAI 4 response, demonstrate that the P-T limit curve for the RSG tube sheet to channel head junction is bounded by the P-T limit curves for the limiting beltline shell materials.

The licensee stated that the RSG primary nozzle knuckle region is also considered as potentially limiting, per the ASME Code,Section XI, Appendix G criteria, due to the discontinuity at the nozzle corner. To demonstrate that the P-T limits for the RSG primary nozzle knuckle regions are less limiting than those for the RPV beltline, the licensee performed a comparison of the component stresses and RT NDT values. Since it was established earlier in this SE that the P-T limits for the RPV outlet nozzle inside corner regions are less limiting than the limiting RPV beltline shell material P-T limits, the licensee needs only to establish that the RSG primary nozzle knuckle region P-T limits are less limiting than those for the RPV outlet nozzle for a 1/4T inside surface corner flaw, under the cooldown transient.

The NRC staff's evaluation of the factors supporting this qualitative comparison is:

• Higher K1c for the RSG primary nozzle: The licensee stated that the maximum RT NDT of 10°F, per the CMTR for the RSG primary nozzles, was considered for the evaluation.

The licensee noted that this value can be considered as the ART due to negligible neutron irradiation on the RSG nozzles. For the RPV outlet nozzle, however, the 60 EFPY ART is 13.1 °F, making them more limiting. The NRC staff finds this determination acceptable because the lower the ART value, the greater the K1c.

• Lower K1p for the RSG primary nozzle: The licensee stated that the stresses due to 1000 psi were compared for the RSG primary nozzle and the Point Beach, Unit 1, RPV outlet nozzle corners. This comparison revealed that the pressure stresses at the RSG primary nozzle corner region were less than the RPV outlet nozzle corner regions from the inside surface up to 90 percent of the nozzle thickness. Since for a 1/4T postulated corner flaw at the RSG primary nozzle corner, a pressure stress profile, which is lower from the inside surface up to 90 percent of the nozzle thickness than that of the RPV outlet nozzle, will produce a lower K1p value, the NRC staff finds that the licensee's determination regarding the lower K1p for the RSG primary nozzle relative to the RPV outlet nozzles is acceptable.

• Lower K1T for the RSG primary nozzle: The licensee stated that the RSG primary nozzle wall thickness at the knuckle region is 10.2 inches while the RPV outlet nozzle corner through-wall thickness is 18.1 inches. Based on the lower wall thickness for the RSG primary nozzle, the licensee determined that the cooldown transient thermal stresses and associated K1T values would be lower for the RSG primary nozzle knuckle region than those for the RPV outlet nozzles. The NRC staff accepts this determination

because K1T generally increases as a function of the section thickness, as shown by the thermal stress intensity correlations from G-2214.3 in Appendix G of the ASME Code,Section XI.

Since the RSG nozzle has lower K1p and K1T values and a higher K1c, the NRC staff finds that the licensee adequately demonstrated that the P-T limits for the RSG primary nozzle are bounded by those for the RPV outlet nozzles; therefore, the RSG primary nozzle is also bounded by the P-T limits for the RPV beltline shell region.

Based on its review of the licensee's P-T limit curve for the RSG tube sheet to channel head junction and its analysis of the RSG primary nozzle knuckle region, as documented above, the NRC staff determined that the licensee has adequately demonstrated that the ferritic materials of the RSGs are bounded by the 50 EFPY P-T limits in the proposed PTLRs.

Based on the resolution of all concerns raised in RAI 4, the NRC staff determined that the licensee adequately demonstrated that the 50 EFPY P-T limit curves for the limiting beltline shell materials in the proposed PTLRs are bounding for all ferritic RPV materials and the replacement ferritic RCPB materials, consistent with the requirements of 10 CFR Part 50, Appendix G. Thus, the proposed 50 EFPY P-T limit curves and PTLRs are acceptable for implementation.

3.2.5 Pressurized Thermal Shock (PTS) Update The licensee provided updated RT PTs values as part of the April 18, 2013, supplement within the marked up TS changes. For Unit 1, the licensee identified the Intermediate to Lower Shell Circumferential Weld at an RTPTs of 236°F, and the Lower Shell Axial Weld at 234.4°F as the limiting materials. For Unit 2, the licensee identified the Intermediate to Lower Shell Circumferential Weld at an RT PTs of 280.6°F, and the Intermediate Shell Forging at 155.4°F as the limiting materials. Since the chemistry, initial RTNDT. and Margin of the beltline materials are the same as those used for the ART calculations supporting the proposed P-T limits, as evaluated in Section 3.2.4 of the SE, independent verification of the proposed RT PTs is straightforward. The NRC staff verified the calculation of these RT PTs values, focusing on the

~RT PTs calculation based on the much higher neutron fluence at clad-base metal interface (as opposed to the 1/4T or 3/4T location for the P-T limits), and noted that the weld RT PTS values were reduced from the previous docketed values primarily due to the incorporation of BAW-2308, Revision 1-A, information. The NRC staff verified that these RT PTS values meet the requirements of 10 CFR 50.61, and therefore determined that the licensee has appropriately updated their PTS analysis for Point Beach, Units 1 and 2.

3.3 Conclusion Based on its review of the information provided by the licensee, the NRC staff concludes that the proposed Point Beach, Units 1 and 2, PTLRs are consistent with the GL 96-03 guidance for implementation and, therefore, should be approved as part of the Point Beach, Units 1 and 2, licensing bases. Regarding the proposed P-T limits, the Point Beach, Units 1 and 2, RPV P-T limits are based on an acceptable methodology documented in WCAP-14040-A, Revision 4, and BAW-2308, Revision 1-A. The NRC staff performed independent evaluations and verified that the P-T limits were developed appropriately using the WCAP-14040-A, Revision 4, methodology, as supplemented by the BAW-2308, Revision 1-A and Revision 2-A methodology.

The proposed P-T limits, valid for 50 EFPY, satisfy the requirements of Appendix G to Section XI of the ASME Code and Appendix G to 10 CFR Part 50. Finally, the NRC staff considers the TS revision to reflect the use of these acceptable methodologies and, therefore, is appropriate.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Wisconsin State official was notified of the proposed issuance of the amendment. The State officials provided no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or 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 NRC has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published June 11, 2013 (78 FR 35062). The amendment also makes minor editorial and corrective changes. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (25). Pursuant to 10 CFR 51.22(b ), 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) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. Letter from Larry Meyer, Site Vice President, NextEra Energy Point Beach, LLC, to NRC Document Control Desk, "License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR),"

dated January 15, 2013 (ADAMS Accession No. ML13016A028).

2. Letter from Larry Meyer, Site Vice President, NextEra Energy Point Beach, LLC, to NRC Document Control Desk, "Supplement to License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," dated March 1, 2013 (ADAMS Accession No. ML13063A292).

3. Letter from Larry Meyer, Site Vice President, NextEra Energy Point Beach, LLC, to NRC Document Control Desk, "Supplement 2 to License Amendment Request 252 Technical

Specifications 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," dated April 18, 2013 (ADAMS Accession No. ML13113A008).

4. Letter from Larry Meyer, Site Vice President, NextEra Energy Point Beach, LLC, to NRC Document Control Desk, "License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

Response to Request for Additional Information," dated September 12, 2013 (ADAMS Accession No. ML13256A064).

5. Letter from Eric McCartney, Site Vice President, NextEra Energy Point Beach, LLC, to NRC Document Control Desk, "License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

Response to Request for Additional Information," dated March 11, 2014 (ADAMS Accession No. ML14071A405).

6. B&W Owners Group (BWOG) Reactor Vessel Working Group, "Initial RTNDT of Linde 80 Weld Materials, Revision 1," August 2003 (ADAMS Accession No. ML032380455).

7. Letter from Herbert N. Berkow, NRC, to Jerald S. Holm, Framatome ANP, "Final Safety Evaluation for Topical Report BAW-2308, Revision 1, 'Initial RT NDT of Linde 80 Weld Materials' (TAC NO. MB6636)," dated August 4, 2005 (ADAMS Accession No.-

ML052070408).

8. BWOG Reactor Vessel Working Group, "Initial RT NDT of Linde 80 Weld Materials, Revision 2," January 2007 (ADAMS Accession No. ML070430445).

9. Letter from Herbert N. Berkow, NRC, to Jerald S. Holm, Framatome ANP, "Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR)

BAW-2308, Revision 2, 'Initial RTNoT of Linde 80 Weld Materials' (TAC No. MD4241),"

dated March 24, 2008 (ADAMS Accession No. ML080770349).

10. Westinghouse Owners Group (WOG) WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", dated May 2004 (ADAMS Accession No. ML050120209).

11. Letter from Larry Meyer, FPL Energy, to NRC, "License Amendment Request 261, Extended Power Up rate," dated April 7, 2009 (ADAMS Accession No. ML110750120).

12. Boiling Water Reactor Vessel and Internals Project Report, BWRVIP-173-A, "BWRVIP-173-A: BWR Vessel and Internals Project, Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials," July 2011 (non-proprietary version of BWRVIP-173-A available under ADAMS Accession No. ML12083A268).

13. Oak Ridge National Laboratory Report, ORNL/TM-201 0/246, "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles- Revision 1,"June 2012 (ADAMS Accession No. ML12181A162).

14. Letter from Ho K. Nieh, NRC, to Randy C. Bunt, BWR Owners Group, "Final Safety Evaluation for the Boiling Water Reactor Owners Group (BWROG) Structural Integrity Associates Topical Report (TR) SIR-05-044, "Pressure Temperature Report Methodology for Boiling Water Reactors (TAC No. MC9694)," Revision 0, dated April 2007 (ADAMS Accession No. ML070180483).

Principal Contributor: S. Sheng Date: June 30, 2014

A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RAJ Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301

Enclosures:

1. Amendment No. 250 to DPR-24

2. Amendment No. 254 to DPR-27

3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC LPL3-1 R/F RidsAcrsAcnw_MaiiCTR Resource RidsNrrLAMHenderson Resource RidsNrrPMPointBeach Resource RidsRgn3MaiiCenter Resource RidsNrrDorllpl3-1 Resource RidsNrrDeEvib Resource RidsNrrDssSrxb Resource Rids NrrDssStsbResou rce RidsNrrDoriDpr Resource KHemphill, NRR SSheng, NRR ADAMS Accession No.: ML14126A378

• via memo dated April 30, 2014 OFFICE DORLILPL3-1/PM DORLILPL3-1 /LA EVIB/BC* STSB/BC NAME TBeltz MHenderson SRosenberg REIIiott DATE 5/7/14 5/19/14 4/30/14 5/28/14 OFFICE OGC (NLO subject DORLILPL3-1 /BC DORLILPL3-1/PM to changes)

NAME BMizuno RCa rison TBeltz DATE 6/6/14 6/16/14 6/30/14 OFFICIAL RECORD COPY