ML14325A501

Issuance of Amendment Revision to Pressure and Temperature Limit Curves

ML14325A501
Person / Time
Site:	Browns Ferry
Issue date:	02/02/2015
From:	Farideh Saba Plant Licensing Branch II
To:	James Shea Tennessee Valley Authority
Saba F DORL/LPL2-2 301-415-1447
References
TAC MF3260
Download: ML14325A501 (20)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Joseph W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 3D-C Chattanooga, TN 37402-2801 February 2, 2015

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNIT 1 - ISSUANCE OF AMENDMENT REVISING PRESSURE AND TEMPERATURE LIMIT CURVES (TAC NO. MF3260)

Dear Mr. Shea:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 287 to Renewed Facility Operating License No. DPR-33 for the Browns Ferry Nuclear Plant, Unit 1. This amendment is in response to your application dated December 18, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13358A067), as supplemented by letter dated June 13, 2014 (ADAMS Accession No. ML14167A407). The amendment revises Technical Specification (TS) 3.4.9, "RCS [Reactor Coolant System] Pressure and Temperature (PIT) Limits," Figures 3.4.9-1 through 3.4.9-2. The PIT limits are based on proprietary topical report NEDC-33178P-A, Revision 1, "GE [General Electric] Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves." NED0-33178-A, Revision 1, is the non-proprietary version of this NRC-approved topical report (ADAMS Accession No. ML092370487).

The NRC staff has completed its review of the information provided by the licensee. The NRC staff's safety evaluation (SE) is enclosed (Enclosure 2). The NRC staff has determined that its documented SE does not contain sensitive security-related information described in Title 1 0 of the Code of Federal Regulations (10 CFR), Section 2.390, "Public inspections, exemptions, requests for withholding." However, the NRC will delay placing the enclosed SE in the public document room for a period of 1 0 working days from the date of this letter to provide the Tennessee Valley Authority with the opportunity to comment on any aspects of the SE that it considers to be sensitive. If you believe that any information in Enclosure 2 represents sensitive information, please identify such information line-by-line and define the basis for withholding it pursuant to the criteria of 10 CFR 2.390. Otherwise, after 10 working days, the enclosed SE will be made publicly available.

The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Docket No. 50-259

Enclosures:

1. Amendment No. 287 to License No. DPR-33

2. Safety Evaluation cc with enclosures:

Addressee only Sincerely, rM,~ /, ~ Sz-t?"

Farideh E. Saba, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Distribution via Listserv (1 0 days after issuance of the amendment to the licensee)

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 287 Renewed License No. DPR-33

1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated December 18, 2013, as supplemented by a letter dated June 13, 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 1 0 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 page 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-33 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 287, are hereby incorporated in the renewed operating license. The licensee 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 60 days from the date of issuance.

Attachment:

Changes to the Operating License and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Shana R. Helton, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: February 2, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 287 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Replace Page 3 of Renewed Facility Operating License DPR-33 with the attached Page 3.

Revise Appendix A, Technical Specifications, by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the areas of change.

REMOVE 3.4-29 3.4-29a 3.4-29b 3.4-29c INSERT 3.4-29 3.4-29a 3.4-29b 3.4-29c (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 equipment and instrument calibration or associated with radioactive apparatus or components; (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 operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1:

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; 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 or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.

(2)

Technical Specifications BFN-UNIT 1 The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 287

, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 234 to Facility Operating License DPR-33, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 234. For SRs that existed prior to Amendment 234, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 234.

Amendment No. 287 Renewed License No. DPR-33

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These curves include sufficient margin to provide protection against feeclwater nozzle degradation. The curves allovv for shifts in RT NDT of the Reactor vessel beltline materials, in accordance with Reg. Guide 1.99 Rev. 2 to compensate for radiation embrittlement for 38 EFPY.

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 287 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-33

1.0 INTRODUCTION

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT. UNIT 1 DOCKET NO. 50-259 By letter dated December 18, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13358A067), as supplemented by a letter dated June 13, 2014 (ADAMS Accession No. ML14167A407), the Tennessee Valley Authority (TVA, the licensee) submitted an application for amendment, requesting changes to the Browns Ferry Nuclear Plant, Unit 1 (BFN-1) Technical Specifications (TSs). TVA submitted this license amendment request to satisfy a commitment to prepare and submit revised BFN-1 Pressure and Temperatur limits prior to the start of the period of operation. The requested amendment would revise TS 3.4.9, "RCS

[Reactor Coolant System] Pressure and Temperature (PfT) Limits," Figures 3.4.9-1 through 3.4.9-2.

The PfT limits are based on proprietary topical report NEDC-33178P-A, Revision 1, "GE [General Electric] Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves." NED0-33178-A, Revision 1, is the non-proprietary version of the topic.al report (ADAMS Accession No. ML092370487). Proposed changes to the TSs, TS Bases, Updated Final Safety Analysis Report (UFSAR), and supporting information representing operation to 25 and 38 effective full-power years (EFPYs) were included in the license amendment request.

The supplement dated June 13, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on May 6, 2014 (79 FR 25902).

2.0 REGULATORY EVALUATION

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has established requirements in Title 10 of the Code of Federal Regulations (1 0 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 PfT limits based on the following NRC regulations and guidance:

Appendix G, "Fracture Toughness Requirements," to 10 CFR Part 50; Appendix H, "Reactor Vessel Material Surveillance Program Requirements," to 1 0 CFR Part 50; Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials";

Generic Letter (GL) 92-01, Revision 1, "Reactor Vessel Structural Integrity"; GL 92-01, Revision 1, Supplement 1, "Reactor Vessel Structural Integrity"; and NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants-LWR [light-water reactor] Edition," (SRP) Section 5.3.2, "Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock."

Section 50.60 of 10 CFR imposes fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary, which are set forth in 10 CFR Part 50, Appendices G and H.

Appendix G to 10 CFR Part 50 establishes fracture toughness requirements. Appendix G.ll, "Definitions," paragraph D.(ii) states, "For the reactor vessel beltline materials, RT NoT [reference nil-ductility transition temperature] must account for the effects of neutron radiation." Paragraph A under Appendix G.IV, "Fracture Toughness Requirements," states, in part, "... the values of RT NoT and Charpy upper-shelf energy must account for the effects of neutron radiation, including the results of the surveillance program of appendix H of this part." The effects of neutron radiation are determined, in part, by estimating the neutron fluence on the reactor vessel.

Paragraph G.IV.A.2.b requires that the PIT limits for the facility's reactor pressure vessel (RPV) be at least as conservative as those obtained by applying the linear elastic fracture mechanics methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). The most recent version of Appendix G to Section XI of the ASME Code that has been endorsed by the NRC 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, "Alternative 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 PIT Limit Curves." Additionally, Table 1 of 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.

Appendix H to 1 0 CFR Part 50 establishes requirements for each facility related to its RPV material surveillance program. RG 1.99, Revision 2, contains guidance on methodologies that the NRC staff considers acceptable for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation. Appendix H compliance was addressed in detail in the topical report NEDC-33178P-A referenced by the licensee for this proposed amendment (see Section 3.0 below); accordingly, the NRC staff did not specifically review the proposed amendment against Appendix H.

GL 92-01, Revision 1, requested that licensees submit the RPV data for their plants to the NRC staff for review, 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. SRP Section 5.3.2 describes a method acceptable to the NRC for determining the Pff limits for ferritic materials in the beltline of the RPV based on the ASME Code Appendix G methodology.

RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence with respect to the General Design Criteria (GDC) 14, 30, and 31 contained in Appendix A to 10 CFR Part 50. GDC 14, "Reactor coolant pressure boundary," requires the reactor coolant pressure boundary be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

GDC 30, "Quality of reactor coolant pressure boundary," requires, among other things, that components comprising the reactor coolant pressure boundary be designed, fabricated, erected, and tested to the highest quality standards practical. GDC 31, "Fracture prevention of reactor coolant pressure boundary," pertains to the design of the reactor coolant pressure boundary, stating:

The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

3.0 TECHNICAL EVALUATION

3.1 The Licensee's Evaluation The operating limits for pressure and temperature are required for three categories of operation:

Hydrostatic pressure tests and leak tests, referred to as Curve A; Non-nuclear heatup/cooldown (core not critical), referred to as Curve B; and Core critical operation, referred to as Curve C.

These Pff operating limits are currently set forth as Technical Specification Figures 3.4.9-1 and 3.4.9-2, valid for up to 12 and 16 EFPYs. The licensee proposed to replace these existing Figures 3.4.9-1 and 3.4.9-2 with corresponding new ones for up to 25 and 38 EFPYs. The licensee's proposed Pff limits are based on application of the GE methodology to BFN-1. NEDC-33178P-A provides the NRC-approved generic GE methodology for generating Prr limits based on the plant-specific adjusted reference temperature, described in RG 1.99, Revision 2. The GE methodology provides generic upper vessel and bottom head Pff limit curves along with beltline curves that are shifted by the plant-specific adjusted reference temperature, as well as guidance on the application of the 1998 edition, 2000 addenda of the ASME Code, Appendix G and 10 CFR Part 50, Appendix G.

For the RPV beltline materials, adjusted reference temperature values were calculated for 25 and 38 EFPYs using plant-specific properties in most cases. The licensee identified the axial electroslag welds (ESWs) with no heat number as the limiting beltline weld material and C2886-2 as the limiting beltline plate for BFN-1. The licensee noted that the N 16 water level instrument nozzles were fabricated from Alloy 600, a material that does not require evaluation for fracture toughness; therefore, the water level instrument nozzles were evaluated using the limiting material properties (chemistry and initial RTNoT) established in NB-2330 of Section XI from the ASME Code of the adjoining plates.

The parameters used to determine the licensee's adjusted reference temperature values for the limiting materials at the one-quarter of the RPV wall thickness (1/4T) location for 25 and 38 EFPYs are shown in Enclosure 2 of the submittal. The estimated adjusted reference temperature of the limiting weld was significantly above that of the limiting beltline plate. Corresponding parameters at the three-quarter of the RPV wall thickness (3/4T) were not provided in the attachments. Instead, the licensee applied the maximum tensile stress for both heatup and cooldown at the 1/4T location.

The licensee stated that this approach is conservative as the 1/4T material toughness is lower than that in the 3/4T locations.

The PIT curves for the non-beltline region were developed for a Boiling-Water Reactor (BWR)

Product Line 6 (BWR/6) with nominal inside diameter of 251 inches. The licensee considers this analysis appropriate for BFN-1, which is a BWR/4 with nominal inside diameter of 251 inches, because the plant-specific geometric values are bounded by the generic analysis for the BWR/6.

The generic value was adapted to the conditions at BFN-1 using plant-specific RT NDT values for the RPV.

For Curve A, the maximum coolant heatup and cooldown temperature rate is 15 degrees Fahrenheit (°F)/hour (per hr); for Curves B and C, the PIT curves specify a maximum coolant heatup and cooldown temperature rate of 100 °F/hr. However, these curves were also developed to bound transients defined on the RPV thermal cycle diagram and the nozzle thermal cycle diagrams. The PIT limits and corresponding heatup/cooldown rates of either Curve A orB may be applied while achieving or recovering from test conditions. Curve A applies during pressure testing and when the limits of Curve B cannot be maintained.

The licensee stated that there are two thickness discontinuities in the vessel; each is discussed and the plant-specific RT NDT values used in the development of the PIT curves bound the RT NDT values associated with the thickness discontinuities in most cases. The licensee also stated that the PIT limit curves were limited by beltline materials for portions of the curves, detailing precisely which portions these were.

The licensee provided data from the Integrated Surveillance Program BWRVIP-135, "BWR Vessel and Internals Project Integrated Surveillance Program (BWRVIP ISP) Data Source Book and Plant Evaluations," in compliance with a requirement in the GE methodology, but since the target plate material of the BFN-1 RPV did not match the representative material, the data from BWRVIP-135 was not used. The BWRVIP-135 source book is used by the industry in compliance with BWRVIP-86, Revision 1, "BWR Vessel and Internals Project Updated BWR Integrated Surveillance Program Implementation Plan" (ADAMS Accession No. ML090300556). Information was also included detailing the determination process for evaluating non-beltline but potentially limiting components.

3.2 The NRC Staff's Evaluation The licensee proposed to replace the current TS Figures 3.4.9-1 and 3.4.9-2, valid for up to 12 and 16 EFPYs, with corresponding new figures for up to 25 and 38 EFPYs. The NRC staff's evaluation of these new figures and their supporting technical information is set forth below.

The NRC guidance provided in RG 1.190 indicates that the following attributes comprise an acceptable fluence calculation:

A fluence calculation performed using an acceptable methodology Analytic uncertainty analysis identifying possible sources of uncertainty Benchmark comparison to approved results of a test facility Plant-specific qualification by comparison to measured fluence values TVA determined the fast neutron exposure parameters using the methods discussed in topical report NED0-32983-A, Revision 2, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations" (ADAMS Accession No. ML072480121). As noted by the November 17, 2005, safety evaluation enclosed with the approved topical report, the NRC staff determined that this methodology is generically acceptable for referencing in licensing actions for GE-designed BWRs.

As described in NED0-32983-A, Revision 2, the neutron fluence was calculated using the following methods. A solution to the Boltzmann transport equation was approximated using the two-dimensional discrete ordinates transport code. The licensee used a cross-section library that the NRC staff had found generically acceptable based on the guidance contained in RG 1.190 (Section 3.1 of the September 14, 2001, safety evaluation approving NED0-32983-A, Revision 1, enclosed in ADAMS Accession No. ML072480121). Approximations included a P3 1 Legrendre expansion for anisotropic scattering and an Sa order of angular quadrature (page 4 of NRC staff safety evaluation in NED0-32983-A, Revision 2, Section 3.1, "Pressure Vessel Fluence Calculation Methodology"). These approximations are consistent with the minimum P3 expansion and Sa quadrature suggested in RG 1.190. The three-dimensional flux field was approximated using synthesis of lower-dimension calculations. Therefore, the NRC staff finds that these neutron calculations, as described above, were performed in a manner consistent with the guidance set forth in RG 1.190.

Also described in NED0-32983-A, Revision 2, an analytic uncertainty analysis was performed by combining the uncertainties associated with the individual components of the transport 1 P3 and S8 are discussed in Regulatory Positions 1.1.2.2 and 1.3.1 of RG 1.190, respectively.

calculations in square-root-of-the-sum-of-the squares. The calculations were found to be comparable to the benchmark measurements from the vessel fluence benchmark problems provided in NUREG/CR-6115, "PWR [Pressurized-Water Reactor] and BWR Pressure Vessel Fluence Calculation Benchmark Problems and Solutions," by Brookhaven National Laboratory.

As such, the NED0-32983-A, Revision 2, method is considered suitably qualified for use in solving BWR fluence problems.

Finally, NED0-32983-A, Revision 2, is acceptably bench marked for BFN-1, as it contains a database of BWR dosimetry benchmarking, and BFN-1 's unit geometry (BWR/4 Reactor Vessel) is well represented within the database. Source term uncertainty for fast neutrons was approximated at 13 percent. In addition, benchmarking demonstrated that calculated reaction rates were within 20 percent of measured values from approved test facilities, as suggested in RG 1.190; therefore, the NRC staff finds that these uncertainties are acceptable.

In summary, TVA has provided fluence calculations performed using acceptable methodology, supported by analytic uncertainty analysis and comparison to approved test facilities, along with a plant-specific comparison to measured fluence values from surveillance capsules. Based on these considerations, the NRC staff concludes that the TVA's fluence calculations adhere to the guidance in RG 1.190, and that the neutron exposures reported in the licensee's submittals are, therefore, acceptable.

To evaluate the licensee's input material property values for calculating the PIT limit curves, the NRC staff first examined the licensee's selection of limiting materials. Specifically, for beltline materials, the NRC staff found that the initial RT Nor, copper, and nickel values had not changed since the NRC approval for the TS change associated with the updated PIT limits, dated January 15, 1999 (ADAMS Accession No. ML020100017).

With regard to the suitability of the values for ESWs to the PIT limits at BFN-1, the NRC staff noted that the NRC approval for the unirradiated RT Nor of the axial welds (summarized in the staff's safety evaluation dated August 31, 2005, for license renewal (ADAMS Accession No. ML052210484)) was based on a small database of similar welds. By a Request for Additional Information (RAI) dated April 23, 2014 (ADAMS Accession No. ML14114A475), the NRC staff asked the licensee to update the approved values with any new information that may have been identified regarding fracture toughness of ESWs and justify why the previously approved values can be used in the current submittal.

By letter dated June 13, 2014 (ADAMS Accession No. ML14167A407), the licensee responded to the NRC staffs RAI by providing a detailed summary of past communications between the NRC and the licensee on this subject. No drop weight specimens were tested, but three sets of Charpy V-Notch impact toughness data were available from the surveillance welds made at the same time as the RPV axial welds. Based on the three Charpy V-Notch results in BAW-1845, the licensee originally used +0 oF as the estimated unirradiated RT Nor for the axial ESWs at BFN-1, which was approved by the NRC in License Amendment No. 121 on September 16, 1985. Later, in response to GL 92-01, the licensee proposed a value of +1 0 oF as a conservative value for the unirradiated RT NDT* As a result of the subsequent approval of the unirradiated RT Nor in BAW-2258 and 2259 for other plants with axial ESWs (approved by the NRC in 1997 (ADAMS Accession No. ML021150615)), the NRC staff asked the licensee to explain why the +1 0 oF value is appropriate for the BFN ESWs. In response, the licensee decided to adopt the more conservative values of 23. 1 oF for the unirradiated RT NOT and 13 oF for the standard deviation for the measurement of the initial unirradiated value determined from BAW-2258 and 2259 for all three units at Browns Ferry Nuclear Plant.

The NRC staff reviewed the historical documents related to the unirradiated RT NoT values and notes that the preservice Charpy impact test results for the surveillance specimens at BFN-1 are the best plant-specific data available. The surveillance specimens at BFN-1 were fabricated from the same base and weld metal at the same time as the RV. Therefore, based on the preservice Charpy impact test results for the surveillance specimens at BFN-1, the NRC staff finds that the values from BAW-2258 and 2259 of RT NOT and the standard deviation for the measurement of the unirradiated value are conservative for the axial welds at BFN-1. Accordingly, the NRC staff considers the issue identified in the April23, 2014, RAI to have been satisfactorily addressed by the licensee.

The licensee reported that there are no best estimate chemistries for the BFN-1 beltline materials described in BWRVIP-135 and, therefore, that the information from BWRVIP-135 does not change the limiting beltline material previously identified by the NRC staff. The licensee only calculated adjusted reference temperature values for the RPV 1/4T location, following NEDC-33178P-A, Revision 1. The NRC staff finds that this is reasonable because using the maximum tensile stress for either heatup or cooldown and applying it at the 1/4T location is equivalent to using the maximum thermal stress intensity factor and the minimum fracture toughness in the heatup and cooldown analysis, making the proposed PIT limits bound both the heatup and cooldown curves. The NRC staff notes that this approach is part of the NRC-approved GE methodology in NEDC-33178P-A.

The NRC staff reviewed how the nozzles are accounted for in the analysis and confirmed that only the N16 water level instrument nozzles were in the extended beltline region where the fluence is estimated to exceed 1 x 1017 n/cm2 (E greater than1 MeV). The licensee used Appendix J of the GE methodology to calculate the PIT limits for the water level instrument nozzle. The NRC staff had previously determined that the thermal stress value from Appendix J of the methodology is conservative compared to that which would result from the normal and upset transients required to be addressed by the ASME Code,Section XI, Appendix G, as it is derived from an emergency transient (ADAMS Accession No. ML13183A017). Therefore, the resulting applied stress intensity factor determined for the water level instrument nozzle in the licensee's PIT curves should still be very conservative, and, accordingly, the staff finds the analysis acceptable.

The NRC staff also evaluated the licensee's analysis of non-beltline components and materials by a systematic review of the licensee's application, as supplemented, and the BFN-1 UFSAR. The NRC staff confirmed that the material properties conveyed in the UFSAR are consistent with the material properties stated in the application. to Enclosure 1 of the application provides proposed changes to Section 4.2.4 of the UFSAR to reflect the implementation of the new PIT limit curves. The NRC staff reviewed the proposed changes to the UFSAR and determined that they are consistent with the new PIT limit curves and supporting technical information.

Based on the evaluation described above, the NRC staff has determined that the licensee's proposed PIT limits covering the two time periods follow the same approved methodology, but with different limiting inputs that are valid for the two time periods. The NRC staff finds the two new sets of PIT limit figures for 25 and 38 EFPYs acceptable; either set may be used by the licensee up until 25 EFPY while the set valid up to 38 EFPY is more restrictive and valid for the longer period. Therefore, the proposed amendment is in accordance with the NRC-approved GE methodology and satisfies the requirements of Appendix G to Section XI of the ASME Code, and Appendix G of 10 CFR Part 50. Accordingly, the licensee's proposed PIT limit curves are acceptable.

3.3 Technical Evaluation Conclusion

Based on the NRC staff's review of the information provided in the licensee's December 18, 2013, application, as supplemented by a letter dated June 13, 2014, the NRC staff concludes that the proposed BFN-1 RPV PIT limits are consistent with the NRC-approved methodology documented in topical report NEDC-33178P-A, Revision 2. The NRC staff performed independent evaluations and verified that the PIT limits were developed appropriately using the NEDC-33178P-A, Revision 2, methodology, and that the proposed PIT limits, valid for 25 and 38 EFPYs, satisfy the requirements of Appendix G to Section XI of the ASME Code, and Appendix G of 10 CFR Part 50.

The NRC staff also finds the proposed changes to the UFSAR to reflect the use of the new PIT limit curves acceptable. The licensee should incorporate the UFSAR changes according to the UFSAR update schedule as required by 10 CFR 50.71(e).

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Alabama 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 the use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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 (79 FR 25902; May 6, 2014). 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), 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Patrick Purtscher Mathew Hardgrove Date: February 2, 2015

The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRA/

Farideh E. Saba, Senior Project Manager Plant Licensing Branch 11-2 Docket No. 50-259

Enclosures:

1. Amendment No. 287 to License No. DPR-33

2. Safety Evaluation cc with enclosures:

Addressee only Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Distribution via Listserv (1 0 days after issuance of the amendment to the licensee)

DISTRIBUTION:

NON-PUBLIC (No PDC/ListServ for 10 working days)

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