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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 13, 2012 Mr. Kenneth Langdon Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093 SUBJECT: NINE MILE POINT NUCLEAR STATION, UNIT NO.2 - ISSUANCE OF AMENDMENT REGARDING A CHANGE TO THE UPDATED SAFETY ANALYSIS REPORT ALLOWING THE USE OF MODIFIED ALLOY 718 MATERIAL IN JET PUMP HOLDDOWN BEAMS (TAC NO. ME7800) Dear Mr. Langdon: The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 141 to Renewed Facility Operating License No. NPF-69 for the Nine Mile Point Nuclear Station, Unit No.2 (NMP2), in response to your application dated December 30, 2011, as supplemented by letter dated March 20, 2012. The proposed amendment authorizes changes to the NMP2 Updated Safety Analysis Report allowing the use of Modified Alloy 718 material for fabrication of the NMP2 reactor recirculation system jet pump holddown beams. A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice. Sincerely, Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-410 Enclosures: 1. Amendment No. 141 to NPF-69 2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION · The inlet-mixer contains a pin, insert, and beam made of Inconel X-7S0 to Specification ASTM A637, Grade 688, or ASTM B637, Grade UNS

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  • UNITED STATES NUCLEAR REGULATORY COMMISSION

    WASHINGTON, D.C. 20555-0001

    April 13, 2012

    Mr. Kenneth Langdon Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093

    SUBJECT: NINE MILE POINT NUCLEAR STATION, UNIT NO.2 - ISSUANCE OF AMENDMENT REGARDING A CHANGE TO THE UPDATED SAFETY ANALYSIS REPORT ALLOWING THE USE OF MODIFIED ALLOY 718 MATERIAL IN JET PUMP HOLDDOWN BEAMS (TAC NO. ME7800)

    Dear Mr. Langdon:

    The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 141 to Renewed Facility Operating License No. NPF-69 for the Nine Mile Point Nuclear Station, Unit No.2 (NMP2), in response to your application dated December 30, 2011, as supplemented by letter dated March 20, 2012. The proposed amendment authorizes changes to the NMP2 Updated Safety Analysis Report allowing the use of Modified Alloy 718 material for fabrication of the NMP2 reactor recirculation system jet pump holddown beams.

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

    Sincerely,

    Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

    Docket No. 50-410

    Enclosures: 1. Amendment No. 141 to NPF-69 2. Safety Evaluation

    cc w/encls: Distribution via Listserv

  • UNITED STATES

    NUCLEAR REGULATORY COMMISSION

    WASHINGTON, D.C. 20555·0001

    NINE MILE POINT NUCLEAR STATION, LLC (NMPNS)

    DOCKET NO. 50-410

    NINE MILE POINT NUCLEAR STATION. UNIT NO.2

    AMENDMENT TO FACILITY OPERATING LICENSE

    Amendment No. 141 Renewed License No. NPF-69

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

    A. The application for amendment by Nine Mile Point Nuclear Station, LLC (the licensee) dated December 30,2011, as supplemented on March 20,2012, 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;

    S. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

    C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations;

    D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

    E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

    2. Accordingly. the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-69 is hereby amended to read as follows:

    (2) Technical Specifications

    The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix S, both of which are attached hereto, as revised through Amendment No. 141 are hereby incorporated into this license. Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

  • - 2

    3. Further, Renewed Facility Operating License No. DPR-69 is hereby amended to authorize changes to the NMP2 Updated Safety Analysis Report (USAR) as set forth in the license amendment application dated December 30, 2011, as supplemented by letter dated March 20, 2012, and evaluated in the safety evaluation dated Apri113 ,2012. The amendment provides authorization to use Modified Alloy 718 material for fabrication of the NMP2 reactor recirculation system jet pump holddown beams.

    4. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

    FOR THE NUCLEAR REGULATORY COMMISSION

    George A. Wilson, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

    Attachment: Changes to Page 4 of Renewed Facility

    Operating License No. NPF-69

    Date of Issuance: Apr; 1 13, 2012

  • ATTACHMENT TO LICENSE AMENDMENT NO. 141

    TO RENEWED FACILITY OPERATING LICENSE NO. NPF-69

    DOCKET NO. 50-410

    Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

    Remove Page Insert Page

    4 4

  • -4

    (1) Maximum Power Level

    Nine Mile Point Nuclear Station, LLC is authorized to operate the facility at reactor core power levels not in excess of 3988 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

    (2) Technical Specifications and Environmental Protection Plan

    The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 141 are hereby incorporated into this license. Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

    (3) Fuel Storage and Handling (Section 9.1, SSER 4)*

    a. Fuel assemblies, when stored in their shipping containers, shall be stacked no more than three containers high.

    b. When not in the reactor vessel, no more than three fuel assemblies shall be allowed outside of their shipping containers or storage racks in the New Fuel Vault or Spent Fuel Storage Facility.

    c. The above three fuel assemblies shall maintain a minimum edgeto-edge spacing of twelve (12) inches from the shipping container array and approved storage rack locations.

    d. The New Fuel Storage Vault shall have no more than ten fresh fuel assemblies uncovered at anyone time.

    (4) Turbine System Maintenance Program (Section 3.5.1.3.10, SER)

    The operating licensee shall submit for NRC approval by October 31, 1989, a turbine system maintenance program based on the manufacturer's calculations of missile generation probabilities. (Submitted by NMPC letter dated October 3D, 1989 from C.D. Terry and approved by NRC letter dated March 15, 1990 from Robert Martin to Mr. Lawrence Burkhardt, III).

    * The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report (SER) and/or its supplements wherein the license condition is discussed.

    Amendment 141

    http:3.5.1.3.10

  • UNITED STATES

    NUCLEAR REGULATORY COMMISSION

    WASHINGTON, D.C. 20555-0001

    SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

    RELATED TO AMENDMENT NO. 141

    TO RENEWED FACILITY OPERATING LICENSE NO. NPF-69

    NINE MILE POINT NUCLEAR STATION, LLC

    NINE MILE POINT NUCLEAR STATION, UNIT NO.2

    DOCKET NO. 50-410

    1.0 INTRODUCTION

    By letter dated December 30, 2011, (Agencywide Documents Access Management System {ADAMS} Accession No. ML 12009A118), as supplemented on March 20, 2012 (ADAMS Accession No. ML 120820407), Nine Mile Point Nuclear Station. LLC (NMPNS. the licensee) submitted a License Amendment Request (LAR) to amend Renewed Facility Operating License NPR-69 for Nine Mile Point, Unit No.2 (NMP2). The licensee requests Nuclear Regulatory Commission (NRC) approval to use Alloy 718 with a modified heat treatment for fabrication of the jet pump holddown beams (JP beams) at NMP2. NRC approval of this proposed change to the NMP2 current licensing basis would be reflected in a revision to Section 4.5 of the NMP2 Updated Safety Analysis Report (USAR), which provides a description of the materials used in the reactor internals. including the jet pump assemblies. The Technical Specifications would not be affected by this LAR.

    The supplemental letter dated March 20, 2012, provided additional information that clarified the application and did not expand the scope of the application as originally noticed, and did not change the NRC staff's initial proposed no significant hazards consideration determination.

    2.0 REGULATORY EVALUATION

    Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50). Appendix A. includes the general design criterion for nuclear power plants. General Design Criterion (GDC) 1, "Quality standards and records," requires that structures, systems and components (SSC) important for safety be designed. fabricated, and tested to quality standards commensurate with the importance of the safety functions to be performed, and that, where generally recognized codes and standards are used, they be identified and evaluated to determine their applicability. adequacy, and sufficiency, and are supplemented or modified as necessary to assure a quality product in keeping with the required safety function.

    In 10 CFR 50.55a(b), "Standards approved for incorporation by reference," the requirements that must be met for the systems and components of boiling water reactors (BWR) are addressed. Part of those requirements is the NRC Regulatory Guide (RG) 1.84, Revision 35 {July 2010}.

  • - 2

    Code Case N-60-S, "Material for Core Support Structures, Section III, Division 1" (Feb. 20, 2004) is the current code case that covers the jet pump assembly.

    The JP beam is specifically included in the NMP2 USAR under Section 4.S.2.1 that states:

    The inlet-mixer contains a pin, insert, and beam made of Inconel X-7S0 to Specification ASTM A637, Grade 688, or ASTM B637, Grade UNS N077S0, Type 3.

    The USAR section goes on to say that non-American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code) components, like the JP beams, are required to be fabricated from American Society for Testing and Materials (ASTM) or ASME specification materials, which is consistent with the acceptance criteria in NUREG-0800, Standard Review Plan (SRP), Section 4.S.2. In addition, the SRP states that additional permitted materials are identified in ASME Code Cases approved for use by an NRC RG 1.84.

    3.0 TECHNICAL EVALUATION

    The information provided by the licensee in support of the LAR to evaluate the use of the Modified Alloy 718 to fabricate the JP beams at NMP2 has been reviewed and the basis for disposition is documented below.

    3.1 Licensee's Request

    Background

    The licensee is requesting NRC approval to use a Modified Alloy 718 material for fabrication of the JP beams in the reactor recirculation system at NMP2. The existing material is Alloy X-7S0. The Modified Alloy 718 material, specified as SB-637 and identified as Grade 718 Type 2 in ASME Code Case N-60-6, has the same chemical composition as the conventional Grade 718 Type 1 material, but the heat treatment conditions have been modified to improve alloy's resistance to stress-corrosion cracking (SCC).

    The jet pump assemblies are part of the reactor vessel internals, but are not core support structures. As noted in the NMP2 USAR, such reactor internals are not ASME Code components, but are fabricated from ASTM or ASME specification materials which are consistent with the acceptance criteria in NUREG-0800, SRP Section 4.S.2. In addition, the Modified Alloy 718 material has been included in ASME Code Case N-60-6. This code case revision was approved by ASME on December 6,2011; however, it has not yet been approved for use by the NRC in RG 1.84. Therefore, the licensee has submitted this LAR for NRC approval.

    Licensee's Basis for Request

    Each jet pump assembly contains one jet pump beam, which is designed to hold the transition piece in contact with the riser pipe and the inlet mixer by maintaining sufficient bolt preload on the transition piece. The preload must be large enough to overcome the hydraulic force applied to the transition piece by the flow inside the jet pump assembly. To accommodate the high preload on the beams, the currently installed beams are fabricated from Alloy X-7S0 material

  • -3

    with a minimum yield strength (YS) of 85 kilopounds per square inch (ksi). As an alternative to the Alloy X-7S0 material, the licensee is proposing to use the Modified Alloy 718 with a minimum YS of 100 ksi, which has been developed based on the conventional Alloy 718 material. The Modified Alloy 718 has similar or improved material properties and improved resistance to SCC initiation and propagation as compared to Alloy X-7S0 material.

    Information that supported the addition of Modified Alloy 718 material to Code Case N-60-S has been provided in Enclosure 3 to the December 30, 2011, submittal, including data on YS, tensile strength, stress intensity, fatigue, and other physical and material properties. The data and evaluations demonstrate that Modified Alloy 718 is comparable to Alloy X-7S0. Details of the modified heat treatment process and the SCC resistance of Modified Alloy 718 are also included. Based on the direct comparison with Alloy X-7S0, the licensee concludes that the proposed change of JP beam material from Alloy X-7S0 to Modified Alloy 718 will not affect the design function of the ~IP beams and will improve the beam's resistance to SCC.

    The inspection intervals for the Modified Alloy 718 beams are not currently addressed in any BOiling-Water Reactor Vessel and Internals Project (BWRVIP) documents; however, the licensee has committed to performing evaluations to establish inspection intervals for the Modified Alloy 718 beams, and expects that inspection intervals similar to those for Alloy X-7S0 Group 3 beams in BWRVIP-41 and BWRVIP-138 can be demonstrated. Specifically, in its December 30, 2011, submittal, NMPNS makes the following regulatory commitment regarding the inspection intervals for the Modified Alloy 718 jet pump holddown beams:

    In accordance with BWRVIP guidelines, NMPNS will submit to the NRC an evaluation to support establishment of inspection intervals for the new holddown beams in accordance with criteria contained in the latest revision of BWRVIP-41, "BWR Vessel and Internals Project, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines." This NMPNS evaluation will be submitted approximately 1 year prior to the next scheduled NMP2 refueling outage following installation of the Modified Alloy 718 jet pump holddown beams.

    The licensee requests NRC approval of their proposed license amendment request in time to support jet pump modifications scheduled for implementation during the planned NMP2 refueling outage, scheduled to begin in April 2012.

    3.2 NRC Staff Evaluation

    The NRC staff has reviewed the submittal consistent with the SRP, Section 4.5.2, Revision 3, March 2007. The review has considered the adequacy and suitability of the Modified Alloy 718 material proposed for the JP beams at NMP2 in terms of the alloy's fabricability, strength, ductility, fracture toughness, stress corrosion resistance, and other physical and mechanical properties.

    Fabricability

    The NRC staff considers the main fabricability issue to consider for this LAR is the heat treatment given to the Modified Alloy 718 component, which is the last major step in the

  • - 4

    fabrication process. Enclosure 4, Section 2.6 of the licensee's December 30, 2011, submittal describes the heat treatment as follows:

    Solution heat treatment (SHT) for modified alloy 718 shall be 1850°F-1922°F (1010°C-1050°C), target temperature of 1886°F (1030'C), for 1 to 2 hours, followed by rapid quenching in oil or water. SHT for modified alloy 718 shall be followed by precipitation hardening heat treatment at 1300°F+15°F (704°C+8°C) for 6 hours, +1 1-0 hours, followed by air cooling. Solution and precipitation hardening heat treatment conditions were determined from the T-T-P (time-temperature-precipitation) curve (Reference 3 and Figure 2-3) and isothermal aging curves for yield strength and elongation of alloy 718 (Figure 2-4). In order to improve SCC resistance and ductility, SHT temperature 1010-1 050'C was selected for complete solution and the precipitation hardening heat treatment condition 704°C/6 hours was selected to avoid the precipitation of 8-phase ...

    Microstructure has been shown to playa determinant role in the SCC behavior of high-strength Ni-based alloys [Reference 1]. The two most important microstructural characteristics of alloy Modified Alloy 718 are grain structure (size and distribution) and is-phase structure. After the recommended solution heat treatment for this alloy (1850 °F-1922 OF for 1 to 2 hours), the grain structure should be relatively uniform (an average grain size of ASTM No.2 to NO.6 is considered optimum for Alloy X-750 and would be similar for the Modified Alloy 718) while the is phase is completely dissolved in the matrix. The grain boundaries should be free of precipitates. The final, low-temperature, aging step (1300 °F+15 OF for 6 hours) is sufficient to increase the YS to the desired range through intragranular precipitation of y" and avoid any grain-boundary precipitation.

    The March 30, 2012, submittal for Modified Alloy 718 and other related sources of information [Reference 1 and 2] include extensive microstrucural characterization for the different versions of Alloy 718. With the microstructural characterization available for the Modified Alloy 718 components to be installed at NMP-2, the staff is satisfied that the resistance to SCC will be similar to that which has been demonstrated in the December 30, 2011 submittal.

    Tensile Properties

    Enclosure 3 of the licensee's December 30, 2011, submittal describes the tensile properties of the Modified Alloy 718. The Alloy 718 material with the modified heat treatment exhibits consistent tensile properties that meet the specification requirement.

    Fracture Toughness Properties

    The licensee's December 30, 2011, submittal does not include any information related to the fracture toughness properties of the Modified Alloy 718. A reference from the open literature for the fracture toughness properties of Alloy 718, Type 1 (higher YS than the Modified Alloy 718) in the unirradiated condition demonstrates that the typical fracture toughness properties are relatively high [Reference 3]. A second reference indicates that the fracture toughness properties are degraded after neutron irradiation, but the testing showed that fracture toughness is still acceptable up to 8 displacements per atom (dpa) [Reference 4]. The NRC staff considers this to be an adequate demonstration of acceptable fracture toughness for the Modified Alloy

  • -S

    this to be an adequate demonstration of acceptable fracture toughness for the Modified Alloy 718 JP beams at NMP2

    Stress Corrosion Resistance

    Enclosure 4 of the licensee's December 30, 2011, submittal describes the resistance to SCC in the Modified Alloy 718. For the initiation of SCC in high temperature (SSO OF) water, two different industry-standard tests were used. In a uniaxial-constant load (UCL) test, the Modified Alloy 718 displayed a higher stress threshold for cracking than the X-7S0 samples; no failures in 10,000 hours at 900 MPa (130 ksi) for two UCL samples of Modified Alloy 718 while 1 of the 2 UCL samples of Alloy X-7S0 failed before 10,000 hours. In the creviced bent beam (CBB) test, the Alloy X7S0 samples displayed more numerous cracking over the whole range of test conditions (chloride contents) tested.

    The low temperature (200 OF) susceptibility to SCC initiation was characterized in a rising load test (RL T) included in BWRVIP-84, "BWR Vessel and Internals Project: Guidelines for Selection and Use of Materials for Repairs to BWR Internals." The test is really a rising displacement test where a constant displacement rate is imposed on a fatigue precracked 3-point bend specimen. For a given test environment of interest, the results are defined in terms of the peak load (elastic stress intensity factor, K) and the time for the load to drop from the peak to SO% of the peak load. When comparing the performance of the Modified Alloy 718 with that of Alloy X-7S0, both sets of samples met the acceptance criteria in BWRVIP-84, but the Modified Alloy 718 exhibited better results, i.e., Modified Alloy 718 always displayed higher peak loads and longer times for the load to drop to SO% of the maximum load than for the Alloy X-750 samples.

    The resistance to SCC crack growth was characterized in a fracture mechanics test using the direct current (DC) potential drop method to monitor crack growth under a constant applied K, (described in Reference 2 and 5) that was varied from about 27 to 45 ksi-in 112. The SCC growth rate for Modified Alloy 718 was always below the proposed curves in BWRVIP-138 for Alloy X-750 and was at least an order of magnitude lower than for the Alloy X-750 samples at each applied K level.

    In summary, the NRC staff has reviewed the test results discussed in the December 30,2011, submittal, and finds that the licensee has demonstrated that the resistance of the Modified Alloy 718 to SCC is superior to that for Alloy X-750.

    Other Properties

    Enclosure 3 of the licensee's December 30, 2011, SUbmittal, describes several other mechanical and physical properties of the Modified Alloy 718, including fatigue testing (both load- and straincontrolled), radiation relaxation rate, spring property, thermal expansion, thermal conductivity, and elastic properties. In each case, the Modified Alloy 718 samples performed as well or better than the Alloy X-7S0 samples.

    Summary

    The NRC staff has reviewed the licensee's December 30, 2011, and March 20, 2012, submittals, and other pertinent references. The NRC staff agrees that the properties of the Modified Alloy

  • - 6

    718 material were as good as or better than the same properties of the Alloy X-750 material; therefore, the NRC staff finds that the use of Modified Alloy 718 for the JP beams at NMP2 is acceptable.

    The NRC staff has reviewed the licensee's submittal and concludes that the licensee's evaluation regarding the mechanical and physical properties of Modified Alloy 718 is adequate to support the approval of the LAR for the use of Modified Alloy 718 material for the JP beams at NMP2. The NRC staff concludes that the change of material for the NMP2 JP beams meets the 10 CFR 50.90 process and the ASME Code, Section III requirements. Therefore, the NRC staff concludes that the licensee's change in material for the JP beams to provide improved resistance to SCC is acceptable.

    4.0 STATE CONSULTATION

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

    5.0 ENVIRONMENTAL CONSIDERATION

    The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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 (February 8,2012 (77 FR 6601 ». 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) 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.

    7.0 REFERENCES

    1. T. Yonezawa, N. Yamaguchi, Y. Okada, and M. Igarashi, "Effects of Chemical Compositions and Heat Treatment on the Stress Corrosion Cracking Resistance of Precipitation Hardened Nickel Base Alloys in High Temperature Water," J. Japan Inst. Metals, Vol. 51, No.4, 1987, pp. 309-318.

    2. Y. Katayama, M. Tsubota, Y. Saito, N. Tanaka, and S. Tanaka, "SCC Properties of Modified Alloy 718 in BWR Plant." 15th International Conference on Environmental

  • - 7

    Degradation, Edited by J.T. Busby, G. Hevbare, and P.L. Andresen, TMS (The Minerals, Metals, and materials Society) 2011.

    3. W. J. Mills and L. D. Blackburn, "Variations in Fracture Toughness for Alloy 718 Given a Modified Heat Treatment," J. Eng. Mater. Technol., January 1990, Volume 112, Issue 1, pp. 116-124.

    4. W. J. Mills, "Effect of Irradiation on the Fracture Toughness for Alloy 718 Plate and Weld", J. Nuclear Materials. December 1992, Volume 199, Issue 12, pp. 68-78.

    5. P.L. Andresen, J. Flores-Preciado, M.M. Morra, and R. Carter, "Microstructure and SCC of Alloy X-750," 15th International Conference on Environmental Degradation, Edited by J.T. Busby, G. Hevbare, and P.L. Andresen, TMS (The Minerals, Metals, and materials Society) 2011.

    Principal Contributor: P. Purtscher

    Date: April 13, 2012

  • April 13, 2012

    Mr. Kenneth Langdon Vice President Nine Mile Point Nine lVIile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093

    SUBJECT: NINE MILE POINT NUCLEAR STATION, UNIT NO.2 - ISSUANCE OF AMENDIVIENT REGARDING A CHANGE TO THE UPDATED SAFETY ANALYSIS REPORT ALLOWING THE USE OF MODIFIED ALLOY 718 MATERIAL IN JET PUMP HOLDDOWN BEAMS (TAC NO. ME7800)

    Dear Mr. Langdon:

    The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 141 to Renewed Facility Operating License No. NPF-69 for the Nine Mile Point Nuclear Station, Unit No.2 (NMP2), in response to your application dated December 30,2011, as supplemented by letter dated March 20, 2012. The proposed amendment authorizes changes to the NMP2 Updated Safety Analysis Report allowing the use of Modified Alloy 718 material for fabrication of the NMP2 reactor recirculation system jet pump holddown beams.

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

    Sincerely,

    !raJ

    Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

    Docket No. 50-410

    Enclosures: 1. Amendment No. 141 to NPF-69 2. Safety Evaluation

    cc w/encls: Distribution via Listserv

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    ADAMS Accession No.: ML120940373 * Concurrence via e-mail, **SE provided by memo. No substantial chan~es made. NRR-106 OFFICE LPL 1-1/PM LPL1-1/LA* EVIB/BC·· OGC LPL1-1/BC NAME RGuzman SUttle SRosenberg DRoth GWilson

    DATE 4/4/12 4/4/12 3/26/12 4/7112 4/12112

    OFFICIAL RECORD COPY