ML13032A526

Issuance of Amendment No. 245, Revise Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits, to Revise Limit Curves and Surveillance Requirements

ML13032A526
Person / Time
Site:	Cooper
Issue date:	02/22/2013
From:	Lynnea Wilkins Plant Licensing Branch IV
To:	Limpias O Nebraska Public Power District (NPPD)
Wilkins L
References
TAC ME7324
Download: ML13032A526 (24)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 22, 2013 Mr. Oscar A. Limpias Vice President-Nuclear and CNO Nebraska Public Power District 72676 648A Avenue Brownville, NE 68321

SUBJECT:

COOPER NUCLEAR STATION - ISSUANCE OF AMENDMENT RE:

REVISIONS TO TECHNICAL SPECIFICATION 3.4.9, "RCS PRESSURE AND TEMPERATURE (PIT) LIMITS," FOR 32 EFFECTIVE FULL POWER YEARS (TAC NO. ME7324)

Dear Mr. Limpias:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 245 to Renewed Facility Operating License No. DPR-46 for the Cooper Nuclear Station. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated September 22, 2011, as supplemented by letters dated March 30, September 10, and September 28, 2012, and January 3, 2013.

The amendment revises the Cooper Nuclear Station's TS 3.4.9, "RCS [Reactor Coolant System]

Pressure and Temperature (PIT) Limits," regarding the current 28 effective full power year (EFPY) pressure-temperature (P-T) limit curves with new curves that are valid for 32 EFPY.

The proposed 32 EFPY P-T limit curves were developed based on the consideration of a measurement uncertainty recapture power uprate that was approved with the issuance of Amendment No. 231 dated June 2, 2008.

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

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Lynnea E. Wilkins, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosures:

1. Amendment No. 245 to DPR-46

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 245 License No. DPR-46

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

A. The application for amendment by Nebraska Public Power District (the licensee),

dated September 22, 2011, as supplemented by letters dated March 30, September 10, September 28, 2012, and January 3, 2013, 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 license 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.

Enclosure 1

-2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-46 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 245, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-46 and Technical Specifications Date of Issuance: February 22, 2013

ATTACHMENT TO LICENSE AMENDMENT NO. 245 RENEWED FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 Replace the following pages of the Renewed Facility Operating License No. DPR-46 and Appendix A Technical Specifications with the enclosed revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License REMOVE INSERT 3 3 Technical Specifications REMOVE INSERT 3.4-22 3.4-22 3.4-23 3.4-23 3.4-24 3.4-24 3.4-25 3.4-25

-3 (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: 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 2419 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 245, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Cooper Nuclear Station Safeguards Plan," submitted by letter dated May 17, 2006.

NPPD shall fully implement and maintain in effect all provisions of the Commission approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The NPPD CSP was approved by License Amendment No. 238 as supplemented by a change approved by License Amendment No. 244.

(4) Fire Protection The licensee shall implement and maintain in effect a" provisions of the approved fire protection program as described in the Cooper Nuclear Station (CNS) Updated Safety Analysis Report and as approved in the Safety Evaluations dated November 29, 1977; May 23, 1979; November 21, 1980; April 29, 1983; April 16, 1984; June 1, 1984; January 3, 1985; August 21, 1985; April 10, 1986; September 9, 1986; November 7, 1988; February 3, 1989; August 15,1995; and July 31, 1998, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Amendment No. 245

-5 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 Proposed TS Changes By letter dated September 22, 2011, as supplemented by letter dated September 28, 2012, the licensee proposed the following revisions to TS 3.4.9 requirements for the RCS P-T limits:

• TS Figure 3.4.9-1, "PressurelTemperature Limits for Non-Nuclear Heatup or Cooldown Following Nuclear Shutdown," will be revised to include new P-T limits for heat-up and cool-down operations with the core not critical that are valid through 32 EFPY. For a markup of proposed changes to TS Figure 3.4.9-1, see Attachment 2 of the licensee's letter dated September 22, 2011, as supplemented by Attachment 2 of the licensee's letter dated September 28, 2012.

• TS Figure 3.4.9-2, "PressurelTemperature Limits for Inservice Hydrostatic and Inservice Leakage Tests," will be revised to include new P-T limits for pressure test conditions that are valid through 32 EFPY. For a markup of proposed changes to TS Figure 3.4.9-2, see Attachment 2 of the licensee's letter dated September 22, 2011, as supplemented by Attachment 2 of the licensee's letter dated September 28, 2012.

• TS Figure 3.4.9-3. "PressurelTemperature Limits for Criticality," will be revised to include new P-T limits for heat-up and cool-down operations, with the core critical that are valid through 32 EFPY. For a markup of proposed changes to TS Figure 3.4.9-3, see Attachment 2 of the licensee's letter dated September 22, 2011, as supplemented by Attachment 2 of the licensee's letter dated September 28, 2012.

• SRs 3.4.9.5, 3.4.9.6, and 3.4.9.7 will be revised to reflect the decrease in the minimum bolt-up temperature value from 80 OF to 70 of, as specified for the proposed 32 EFPY P-T limits for pressure test and core not critical conditions.

The NOTES in SR 3.4.9.6 and SR 3.4.9.7 for MODE 4 (cold shutdown) will also be changed to reduce by 10°F the points at which the surveillances are required to be performed in order to maintain the same margins from the minimum operating parameters.

Specifically, current SRs 3.4.9.5, 3.4.9.6, and 3.4.9.7 state that Verify reactor vessel flange and head flange temperatures are

> 80°F.

RCS PfT Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.5 --------------------------NOTE-----------------------

Only required to be performed lNhen tensioning the reactor vessel head bolting studs.

Verify reactor vessel flange and head flange 30 minutes temperatures are> 70°F.

SR 3.4.9.6 ----------------------------NOTE-------------------------

Not required to be performed until 30 minutes after RCS temperature ~ 80°F in MODE 4.

Verify reactor vessel flange and head flange 30 minutes temperatures are> 70°F.

SR 3.4.9.7 ---------------------------NOTE-------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature ~ 90°F in MODE 4.

Verify reactor vessel flange and head flange 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> temperatures are> 70°F.

Cooper 3.4-22 Amendment 245

RCS PIT Limits 3.4.9 Cooper Heatup/Cooldown, Core Not Critical Curve (Curve B), 32 EFPY

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Figure 3.4.9-1 PressurelTemperature Limits for Non-Nuclear Heatup or Cooldown Following Nuclear Shutdown Valid Through 32 EFPY Cooper 3.4-23 Amendment No. 245

RCS PfT limits 3.4.9 Cooper Pressure Test Curve (Curve A), 32 EFPY 1,300

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Figure 3.4.9~3 Pressureffemperature Limits for Criticality Valid Through 32 EFPY Cooper 3.4-25 Amendment No. 245

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 245 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-46 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298

1.0 INTRODUCTION

By letter dated September 22, 2011, as supplemented by letters dated March 30, September 10, and September 28,2012, and January 3,2013 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML11272A057, ML12094A118, ML12258A072, ML12278A138, and ML13008A181, respectively), Nebraska Public Power District (NPPD, the licensee) submitted a license amendment request (LAR) to revise the Technical Specifications (TSs) for Cooper Nuclear Station (Cooper). Portions of the letter dated March 30, 2012, contain sensitive unclassified non-safeguards information, and, accordingly, those portions have been withheld from public disclosure.

The LAR would revise TS 3.4.9, "RCS Pressure and Temperature (PfT) Limits," to include new reactor coolant system (RCS) pressure-temperature (P-T) limits for heat-up and cool-down operations with the core critical and core not critical, as well as for pressure test conditions. The proposed P-T limits would be valid for 32 effective full power years (EFPY) of facility operation and are based on operating conditions associated with the measurement uncertainty recapture (MUR) power uprate approved by the U.S. Nuclear Regulatory Commission (NRC) in Amendment No. 231 dated June 30,2008 (ADAMS Accession No. ML081540278). The proposed revisions to TS 3.4.9 also include revised surveillance requirements (SRs) for verifying that the reactor pressure vessel (RPV) flange and RPV head flange temperatures are greater than the revised RPV bolt-up temperature (70 degrees Fahrenheit (OF>> specified in the proposed 32 EFPY P-T limits.

The supplemental letters dated March 30, September 10, and September 28,2012, and January 3, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on March 6, 2012 (77 FR 13372).

Enclosure 2

-2

2.0 REGULATORY EVALUATION

2.1 System Description By letter dated September 22,2011, the licensee provided the following system description:

The reactor pressure vessel (RPV) is a vertical cylindrical pressure vessel with hemispherical heads of welded construction. The RPV is designed and fabricated for a useful life of 40 years based upon the specified design and operating conditions. It is designed, fabricated, inspected, tested, and stamped in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III (1965 Edition and January 1966 Addenda), its interpretations, and applicable requirements for Class A Vessels as defined therein.

RCS components are designed to withstand effects of cyclic loads due to system pressure and temperature changes introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. The prr limit curves apply to the RPV, since it is the RCS component most subject to brittle failure, and is bounding over other components that comprise the reactor coolant pressure boundary (RCPB). TS 3.4.9 establishes operating limits that provide a margin to brittle failure of the RPV and piping of the RCPB.

2.2 Regulatory Requirements Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The TSs ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The NRC's regulatory requirements related to the content of the TSs are contained in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, "Technical specifications,"

which requires that the TSs include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) SRs; (4) design features; and (5) administrative controls. However, the regulation does not specify the particular requirements to be included in TSs.

The regulations in 10 CFR 50.36(c)(2), "Limiting conditions for operation," state, in part, that Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

-3 The regulations in 10 CFR 50.36(c)(2)(ii) state that:

A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(8) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The regulations in 10 CFR 50.36(c)(3), ".Surveillance requirements," state that Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The NRC has established requirements in Appendix G, "Fracture Toughness Requirements," to 10 CFR Part 50, in order to protect the integrity of the RCP8 in nuclear power plants. The regulations in 10 CFR Part 50, Appendix G, require that the P-T limits for an operating light water nuclear reactor be at least as conservative as those that would be generated if the methods of Appendix G, "Fracture Toughness Criteria for Protection Against Failure," to Section XI of the ASME Code were used to generate the P-T limits. The regulations in 10 CFR Part 50, Appendix G, also require that applicable surveillance data from RPV material surveillance programs be incorporated into the calculations of plant-specific P-T limits, and that the P-T limits for operating reactors be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials.

Table 1 of 10 CFR Part 50, Appendix G, provides the NRC staff's criteria for meeting the P-T limit requirements of the ASME Code,Section XI, Appendix G, as well as the minimum temperature requirements of the rule during normal and pressure testing operations.

In addition, the NRC staff's regulatory guidance related to P-T limit curves is found in Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988 (ADAMS Accession No. ML003740284), and NUREG-0800, "Standard Review Plan for the

- 4 Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Edition" (SRP), Section 5.3.2, "Pressure-Temperature Limits and Pressurized Thermal Shock" (ADAMS Accession No. ML070380185).

Pressure-temperature limit curve calculations are based, in part, on the reference nil-ductility temperature (RT NDT) for the material, as specified in the ASME Code,Section XI, Appendix G.

The regulations in 10 CFR Part 50, Appendix G, require that RT NDT values for materials in the RPV beltline region be adjusted to account for the effects of neutron radiation. RG 1.99, Revision 2, contains methodologies for calculating the adjusted RT NDT (ART) due to neutron irradiation. The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RT NDT), the mean value of the adjustment in reference temperature caused by irradiation

(~RT NDT), and a margin term.

The ~RT NDT is a product of a chemistry factor (CF) and a fluence factor. The CF is dependent upon the amount of copper and nickel in the material and may be determined from tables in RG 1.99, Revision 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the initial RT NDT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99, Revision 2, or surveillance data. The margin term is used to account for uncertainties in the values of the initial RT NDT, the copper and nickel contents, the neutron fluence, and the calculational procedures. RG 1.99, Revision 2, describes the methodology to be used in calculating the margin term.

Appendix H, "Reactor Vessel Material Surveillance Program Requirements," to 10 CFR Part 50 provides the NRC staff's criteria for the design and implementation of RPV material surveillance programs for operating light-water reactors.

In March 2001, the NRC staff issued RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (ADAMS Accession No. ML010890301).

Fluence calculations for use in ART and P-T limit curve analyses are acceptable if they are performed with approved methodologies or with methods which are shown to conform to the guidance in RG 1.190.

RG 1.190 describes methods and assumptions acceptable to the NRC staff for determining the pressure vessel neutron 'fluence with respect to the General Design Criteria (GDC) contained in Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50. In consideration of the guidance set forth in RG 1.190, GDC 14, "Reactor coolant pressure boundary," GDC 30, "Quality of reactor coolant pressure boundary," and GDC 31, "Fracture prevention of reactor coolant pressure boundary," are applicable. GDC 14 requires the design, fabrication, erection, and testing of the reactor coolant pressure boundary so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. GDC 30 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 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

-6 Revised SRs 3.4.9.5, 3.4.9.6, and 3.4.9.7 would state that Verify reactor vessel flange and head flange temperatures are

> 70°F.

Current SR 3.4.9.6 NOTE states that Not required to be performed until 30 minutes after RCS temperature ~ 90°F in MODE 4.

Revised SR 3.4.9.6 NOTE would state that Not required to be performed until 30 minutes after RCS temperature ~ 80°F in MODE 4.

Current SR 3.4.9.7 NOTE states that Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature ~ 100°F in MODE 4.

Revised SR 3.4.9.7 NOTE would state that Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature ~ 90°F in MODE 4.

By letter dated September 28, 2012, the licensee stated that the revised neutron fluence calculation used for the proposed 32 EFPY P-T limit curves includes the effects of the previously-approved 1.62% MUR power uprate. According to the licensee, the neutron fluence calculations were based on the NRC-approved Radiation Analysis Modeling Application (RAMA) neutron fluence methodology. The licensee stated that the RAMA neutron fluence methodology has been benchmarked using RG 1.190. The RAMA fluence calculations were applied to the RPV beltline region, defined as the region near the core where the neutron fluence exceeds 1.0 x 10 17 n/cm 2 (E > 1.0 MeV).

The licensee stated that the proposed 32 EFPY TS P-T limit curves were determined based on calculations documented in Calculation Package 1100445.303, "Cooper P-T Curve Revision,"

dated August 5, 2011, which was provided in an enclosure to the LAR. According to the licensee, the P-T limit curves were developed by evaluating three regions of the RPV: (1) the beltline region, (2) the non-beltline upper RPV region, and (3) the non-beltline bottom head region.

The licensee's evaluation of the beltline region included analyses of the limiting beltline shell material and the water level instrument nozzles, Nozzles N16A and N16B. For the bottom head region, P-T limits were calculated for both the bottom head penetration discontinuities and the core differential pressure (CDP) nozzle. The P-T limits for the upper RPV region were determined based on the analysis of the feedwater (FW) nozzles, as well as the determination of minimum temperature requirements based on the limiting RT NDT value for the closure flange region.

-7 The licensee's submittal included information on the methods employed for determining the fracture toughness, Klc . and the applied stress intensity factors due to pressure and thermal stresses, KIP and KIT, which were used for the P-T limit calculations. For all locations, Klc was established based on the RT NOT for the material. For beltline components, RTNOT were adjusted to account for the effects of neutron embrittlement through 32 EFPY using the procedures in RG 1.99, Revision 2. For the RPV shell, KIP and KIT, values were calculated using the formulations specified in the ASME Code,Section XI, Appendix G, and G-2214. For the nozzles, KIP and KIT, values were determined using finite element analysis techniques applied to component-specific fracture mechanics models. The licensee stated that all regions of the RPV were evaluated by considering a postulated flaw that extends to one-quarter of the RPV wall (1/4T location), accounting for the specific RPV section thickness (shell thickness, nozzle forging thickness, etc.).

The proposed P-T limits incorporated minimum temperature criteria based the limiting RTNDT for the closure flange region, as required by Table 1 of 10 CFR Part 50, Appendix G. P-T limits for core critical conditions were calculated by applying a 40 OF shift to the core not critical P-T limits, as required by Table 1 of 10 CFR Part 50, Appendix G. The final TS P-T limit curves were established by applying pressure and temperature adjustment factors to the calculated P-T limits to account for pressure and temperature instrument uncertainties and the static pressure head from the weight of the water in the RPV.

3.2 NRC Staff Evaluation The NRC staff reviewed the licensee's submittals, including Calculation Package 1100445.303, to determine whether the licensee's proposed 32 EFPY P-T limit curves are in compliance with the requirements of 10 CFR Part 50, Appendix G. The staff verified that the proposed 32 EFPY P-T limits were developed by taking into consideration three regions of the RPV: the beltline region, the non-beltline upper RPV region, and the non-beltline bottom head region. The staff noted that, based on the evaluation of these three regions of the RPV, the bounding P-T limits for 32 EFPY are controlled by the RPV beltline region, which includes the RPV beltline shell and the water level instrument nozzles, as well as the minimum temperature criteria based on Table 1 of 10 CFR Part 50, Appendix G.

3.2.1 Evaluation of RPV 8eltline Region P-T Limits and ART Values The NRC staff verified that the licensee's proposed 32 EFPY P-T limits were calculated based on an evaluation of the RPV beltline region, accounting for neutron embrittlement through 32 EFPY. The licensee's evaluation of the RPV beltline region included P-T limit calculations

. for two beltline components: the limiting RPV beltline shell material and the water level instrument nozzles. The staff noted discrepancies in the submittal regarding the identification of the limiting beltline shell material and limiting ART value. Specifically, Section 3.3 of the LAR states that the most limiting beltline material is the lower longitudinal weld with a 1/4T ART value of 103.5 OF. However, page 10 of Calculation Package 1100445.303 identifies the limiting beltline material as the Lower Intermediate Shell Plate (Heat No. C2307-2), with a 1/4T ART value of 105.8 OF. The staff also noted that the licensee did not document any of the input parameters necessary for calculating the limiting ART value, By letter dated February 29,2012 (ADAMS Accession No. ML120590085), the NRC staff issued a request for additional information (RAI), requesting, in RAI-1, that the licensee (a) identify the

-8 limiting beltline shell material and 32 EFPY limiting ART value, and (b) provide the input parameters necessary for calculating this 32 EFPY ART value, based on RG 1.99, Revision 2 procedures.

By letter dated March 30, 2012, the licensee stated that the limiting beltline shell material is the Lower Intermediate Shell Plate (Heat No. C2307-2), which has a 32 EFPY ART of 105.8 of at the 1/4T location. The licensee clarified that this information was incorrectly stated in Section 3.3 of the LAR. The licensee also provided the input parameters (initial RT NDT , CF, 32 EFPY fluence, and margin term) used for determining the 32 EFPY ART value for this material. Based on this information, the NRC staff verified that the 32 EFPY ART for the limiting beltline shell material (Lower Intermediate Shell Plate) was correctly determined using the procedures in RG 1.99, Revision 2, Position 2.1.

The NRC staff verified that the initial RTNDT, CF, and margin term for the limiting beltline shell material are consistent with those identified in Table 4.2-2 of the Cooper License Renewal Application (LRA, ADAMS Accession No. ML083030239), and approved by the staff in the Cooper LRA final safety evaluation report, NUREG-1944, "Safety Evaluation Report Related to the License Renewal of Cooper Nuclear Station," October 2010 (ADAMS Accession No. ML103070009). The staff noted that the CF value for the limiting material was determined based on the application of the credible surveillance data from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP), as documented in the proprietary BWRVIP-135 report, "BWR Vessel and Internal Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations," Revision 2, using the procedures in RG 1.99, Revision 2, Position 2.1. As documented in NUREG-1944, the staff determined that the licensee's participation in the BWRVIP ISP satisfies the requirements of 10 CFR Part 50, Appendix H. Therefore, based on its review of the licensee's RAI response, including the licensee's acceptable ART calculation for the limiting material as discussed above, the staff concludes that RAI-1 is resolved.

Regarding the 32 EFPY neutron fluence used for the limiting ART calculation, the NRC staff verified that the neutron fluence was calculated using the NRC-approved RAMA neutron fluence methodology, as described in BWRVIP-114, "BWR Vessel and Internals Project, RAMA Fluence Methodology Manual," May 2003 (ADAMS Accession No. ML031640195), and BWRVIP-115, "BWR Vessel and Internals Project, RAMA Fluence Methodology Benchmark Manual Evaluation of Regulatory Guide 1.190 Benchmark Problems," June 2003 (ADAMS Accession No. ML031820254). In its safety evaluation related to Amendment No. 219, dated April 27, 2006 (ADAMS Accession No. ML060810339), which authorized the implementation of current TS P-T limit curves through 30 EFPY, the NRC approved the RAMA neutron fluence methodology for use in P-T limit curves analyses at Cooper. The primary staff consideration for acceptability is the fact that RAMA has been found adherent to RG 1.190, and in particular for calculating vessel fluence values for BWRl4 vessel geometries such as Cooper's. The staff determined that the BWRVIP-114 and BWRVIP-115 documents provide the technical basis for approval of the RAMA methodology on a generic basis, and the safety evaluation dated April 27,2006, supports plant-specific implementation of RAMA for P-T limit curve analyses. The staff also noted that Section 3.2 of the LAR states that the RAMA neutron fluence calculation supporting the proposed 32 EFPY P-T limit curves accounts for the effects of the MUR power uprate approved in Amendment No. 231 dated June 2,2008. This information indicates that the neutron fluence calculation takes into account current core operating conditions. Based on the

- 9 above, the NRC staff concludes that the neutron fluence calculation supporting the proposed 32 EFPY P-T limits is acceptable.

By letter dated September 22, 2011, Section 3.2, the licensee also stated that the RAMA calculations account for a MUR uprate that has been implemented at Cooper. This information indicates that the fluence calculations take into account current core operating conditions, which the NRC staff also concludes is acceptable.

For the limiting beltline shell material, the NRC staff performed a set of confirmatory calculations for verifying that the licensee's 32 EFPY P-T limits for heat-up and cool-down operations, and for pressure test conditions, are consistent with the ASME Code,Section XI, Appendix G.

Based on its confirmatory calculations, the staff determined that the P-T limits for the limiting beltline shell material, Lower Shell Plate (Heat No. C2307-2), meet the criteria of the ASME Code,Section XI, Appendix G, G-2215 for heat-up and cool-down operations, and G-2400 for pressure test conditions, as required by 10 CFR Part 50, Appendix G, and, therefore, the P-T limits are acceptable.

At low temperatures (70 of < T < 110 of, for heat-up and cool-down operations with the core critical and core not critical), the bounding P-T limits for the RPV beltline region are defined by the water level instrument nozzles, rather than the limiting beltline shell material. The NRC staff noted that since the instrument nozzle material is not ferritic, the nozzle material properties do not require specific evaluation. However, the stress-concentrating effects of the instrument nozzles in the surrounding ferritic plate material must be considered in the calculation of the applied stress intensity factors for determining the P-T limits at this location.

The licensee's LAR included calculations of the applied KIP and KIT for the water level instrument nozzles. The NRC staff determined that the KIP and KIT calculations for the water level instrument nozzles are acceptable because they are based on the appropriate 1/4T postulated flaw, plant-specific instrument nozzle configuration, and bounding pressure and thermal loading conditions at Cooper. Therefore, the staff concludes that the KIP and KIT calculations for the water level instrument nozzles are consistent with the ASME Code,Section XI, Appendix G, and 10 CFR Part 50, Appendix G.

Since the water level instrument nozzles are located in the beltline region, the NRC staff determined that additional information was needed to confirm the validity of the ART value used for analyzing these components. Therefore, by letter dated February 29.2012. the NRC staff requested in RAI-2 that the licensee (a) identify the RPV beltline plates that contain the water level instrument nozzles, and (b) provide the ART calculation for the nozzle/plate configuration, including the material property inputs used for performing this calculation. By letter dated March 30. 2012, the licensee identified the RPV beltline plates and provided the requested ART calculation. The staff determined that the material properties for this plate (initial RT NDT. copper content, nickel content, and margin term) are consistent with those documented in the Cooper LRA and are, therefore, acceptable. The staff noted that, considering the water level instrument nozzle configuration and 1/4T postulated corner flaw, the applicant's ART calculation for the plate is appropriate for determining the P-T limits for these nozzles. The staff verified that the 32 EFPY ART value for this location was correctly determined using the procedures in RG 1.99, Revision 2. Position 1.1. Based on the above, the NRC staff concludes that RAI-2 is resolved.

- 10 Based on the licensee's calculations for the applied stress intensity factor values and 32 EFPY ART, the NRC staff concludes that the P-T limits for the water level instrument nozzles meet the acceptance criteria of the ASME Code,Section XI, Appendix G, as required by 10 CFR Part 50, Appendix G, and, therefore, the P-T limits are acceptable.

Based on its review of the licensee's analysis of the limiting beltline shell material and the water level instrument nozzles, the NRC staff determined that the licensee's evaluation of the RPV beltline region is acceptable for determining the proposed 32 EFPY p.T limit curves.

Regarding ferritic RCPB components that are not part of the RPV beltline region, 10 CFR Part 50, Appendix G, Paragraph IV.A states, in part, that The pressure-retaining components of the reactor coolant pressure boundary that are made of ferritic materials must meet the requirements of the ASME CodeL Section III], 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 by considering beltline and non-beltline ferritic RCPB components. Further, 10 CFR Part 50. Appendix G, requires that all ferritic RCPB components must meet the applicable ASME Code,Section III requirements. For RCPB piping, pumps, and valves greater than 2.5 inches in thickness, the relevant ASME Code,Section III requirement that will affect the P-T limits is the lowest service temperature (LST) requirement specified in Section III, NB-2332(b).

The NRC staff notes 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. This may be due to the following factors:

1. RPV nozzles, penetrations, and other discontinuities exhibit higher stresses than the RPV beltline shell region, which could result in more restrictive P-T limits, even if the RT NOT for these components is not as high as that of RV beltline shell materials.

2. Ferritic RCPB components that are not part of the RPV may have initial RT NOT values, which may define a more restrictive LST in the P-T limits than the minimum temperature requirements for the RPV.

By letter dated August 10, 2012 (ADAMS No. ML12205A216), the NRC staff requested, in RAI-3, that the licensee describe how the proposed 32 EFPY P-T limit curves, and the methodology used to develop these curves, considered all ferritic RCPB components, consistent with the requirements of 10 CFR Part 50, Appendix G.

By letter dated September 10, 2012, the licensee stated, in part, that:

Vessel nozzles are generally incorporated into PIT curve calculations using stress distributions from Finite Element Analyses and applying them to geometry specific fracture mechanics models. The feedwater nozzle (upper vessel region) and core differential pressure (COP) nozzle require this type of analysis due to

- 11 the bounding transients they experience and/or stress concentration effects. The core differential pressure COP nozzle (bottom head region) is analyzed because it is the limiting discontinuity in the thin portion of the bottom head.

The feedwater nozzle is the bounding component in the upper vessel because it is a stress concentrator (essentially a hole in a plate) and because it typically experiences more severe thermal transients compared to the rest of the upper vessel region.

The licensee also indicated that applied stress intensity factors for these nozzles are calculated by determining the pressure and thermal stress distributions acting normal to the postulated 1/4T flaw along the limiting path in the nozzle inside corner radius.

Calculation Package 1100445.303 describes the non-beltline RPV component analyses. With the acceptance of these analyses, as indicated in the NRC staff's evaluation of the licensee's submittal relative to the non-beltline RPV components provided below, this portion of RAI-3 is resolved.

By letter dated January 3, 2013, the licensee stated, in part, that:

The Cooper Nuclear Station (CNS) Class 1 ferritic piping systems were designed to [American Nuclear Standards Institute (ANSI)] B31.1-1967. At the time of construction ASME Section III did not apply to piping, pumps or valves. Thus NB-2332(b) does not specifically apply to CNS. Furthermore, the nominal wall thicknesses of the Class 1 ferritic piping systems are less than 2.5 inches.

However brittle failure was considered in the design of the CNS piping system.

The NRC staff concludes that this response adequately described the licensee's evaluation of the non-beltline RCPB components because the ASME Code,Section III LST requirement is not applicable to these RCPB components since the components were designed prior to the incorporation of ASME Code,Section III, NB-2332(b). Furthermore, since the nominal wall thicknesses of all ferritic RCPB piping components are less than 2.5 inches, the LST requirement of NB-2332(b) is not applicable to the ferritic RCPB piping components at Cooper.

Based on the above, the NRC staff concludes that this portion of RAI-3 is resolved.

3.2.2 Evaluation of RPV Non-Beltline Region P-T Limits and Minimum Temperature Requirements For developing the 32 EFPY P-T limits, the licensee analyzed the non-beltline regions of the RPV: the upper RPV region and the bottom head region. The licensee's analysis of the upper RPV region included P-T calculations for the FW nozzles and the establishment of minimum temperature criteria based on the highest RT NDT value for the closure flange region, in accordance with 10 CFR Part 50, Appendix G, Table 1, minimum temperature requirements.

The licensee's analysis of the bottom head region included P-T limit calculations for the bottom head control rod drive stub tube penetrations and the COP nozzle.

P-T limits were calculated for the FW nozzles based on a limiting RT NDT value for these nozzles and applied stress intensity factors for pressure and thermal loading. The FW nozzles were determined to be the most limiting component for the upper RPV region, with respect to the

- 12 applied stresses due to thermal and pressure loads. The NRC staff confirmed that the RT NDT value used for analyzing the FW nozzles is consistent with previously approved P-T limit applications. The staff verified that the licensee's analysis of the FW nozzles was based on an NRC-approved methodology, Structural Integrity Associates, Inc. (SIA) topical report SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors,"

Revision 0 (ADAMS Accession No. ML072340283). SIR-05-044-A indicates that the FW nozzles should be analyzed for determining the highest stress intensities for the upper RPV region. The staff verified that the applied pressure and thermal stress intensity factors for the FW nozzle with the postulated 1/4T inside corner flaw were appropriately calculated based on a stress distribution output from a finite element analysis of the nozzle, using the fracture mechanics model prescribed in SIR-05-044-A For the closure flange region, the NRC staff verified that the licensee correctly applied the limiting RT NDT for the closure flange material to calculate minimum temperature criteria for normal operating conditions (with the core critical and core not critical) and pressure test conditions that are least as conservative as those required by Table 1 of 10 CFR Part 50, Appendix G. Minimum temperature criteria were calculated for pressures less than or equal to 20 percent (%) of the pre-service hydrostatic test pressure (20% hydro) and for pressures greater than 20% hydro, as required by 10 CFR Part 50, Appendix G. The staff determined that the minimum temperature criteria were appropriately incorporated into the 32 EFPY P-T limit curves for core critical, core not critical, and pressure test conditions. A composite P-T limit curve for the upper RPV region was determined based on the application of the minimum temperature criteria and the analysis of the FW nozzles.

The P-T limits generated for the RPV bottom head region are based on the licensee's analysis of the bottom head control rod drive stub tube penetrations and the COP nozzle. The methods for calculating P-T limits for the bottom head stub tube penetrations are similar to the methods for the shell region, except that a specified stress concentration factor is applied to the bottom head membrane stress to account for the stress concentration resulting from these penetrations. The NRC staff verified that the stress concentration factor used for the bottom head penetrations is consistent with SIR-05-044-A and, therefore, is acceptable. The COP nozzle was determined to be the limiting nozzle for the bottom head region, with respect to its stress concentrating effects. For the COP nozzle, the licensee applied finite element analysis based fracture mechanics methods similar to those used for the FW nozzles, in accordance with the methodology in SIR-05-044-A Based on the above, the NRC staff concludes that the licensee's analysis of the COP nozzle is acceptable.

Based on the above, the NRC staff concludes that the licensee's P-T limit calculations for the non-beltline RPV components are consistent with the NRC-approved methodologies documented in SIA topical report SIR-05-044-A, Revision O. Therefore, the staff concludes that the non-beltline region P-T limit calculations meet the criteria of the ASME Code,Section XI, Appendix G, as required by 10 CFR Part 50, Appendix G, and are, therefore, acceptable. The staff also concludes that the minimum temperature criteria for the proposed 32 EFPY P-T limit curves satisfy the requirements of Table 1 of 10 CFR Part 50, Appendix G.

The NRC staff concludes that the final TS P-T limit curves were correctly established based on the evaluations of the three RPV regions, as discussed above. The bounding P-T limits for 32 EFPY are controlled by the RPV beltline region, as well as the minimum temperature criteria discussed above. The staff verified that the TS P-T limits for core critical conditions were

- 13 correctly established by applying a 40 of shift to the P-T limits for core not critical conditions, as required by Table 1 of 10 CFR Part 50, Appendix G. The staff also verified that the final TS P-T limits were determined by applying appropriate correction factors to the calculated P-T limits, to account for RCS pressure and temperature instrument uncertainties, as well as an additional static pressure adjustment factor to account for the pressure head of the water in the RPV. The staff noted that the static pressure head for all locations was conservatively calculated based on the full weight of the water in the RPV. Based on the above, the NRC staff concludes that the correction factors are acceptable.

The NRC staff concludes that the licensee's proposed revisions to TS SRs 3.4.9.5, 3.4.9.6, and 3.4.9.7 are acceptable because they reflect the decrease in the minimum bolt-up temperature from 80 of to 70 of, as specified for the proposed 32 EFPY P-T limits for pressure test and core not critical conditions. In addition, the staff determined that the NOTES in SRs 3.4.9.6 and 3.4.9.7 for MODE 4 were also correctly revised to reflect the 10°F decrease in the minimum bolt-up temperature. The NRC staff concludes that the TS revisions to SRS 3.4.9.5, 3.4.9.6 and 3.4.9.7 are acceptable since they meet the requirements of 10 CFR 50.36(c)(3), in that the SRs will ensure that the necessary quality of systems are maintained, that the facility will be maintained within safety limits, and the LCOs will continue to be met.

3.3 NRC Staff Conclusion

Based on the evaluation in Section 3.2 of this safety evaluation, the NRC staff concludes that the licensee's calculated P-T limits for all regions of the RPV meet the criteria of the ASME Code,Section XI, Appendix G, and are in compliance with the requirements of 10 CFR Part 50, Appendix G, and 10 CFR 50.60. Therefore, the staff concludes that the proposed 32 EFPY P-T limit curves are acceptable for incorporation into the Cooper TSs.

4.0 REGULATORY COMMITMENTS By letter dated September 10, 2012, the licensee made the following regulatory commitments:

Commitment Due Date

1. NPPD will resubmit the curves without X the analysis of September 30,2012 the prr nozzles.

2. Since there is no currently approved methodology for

• December 31,2016 addressing the instrument nozzles in the beltline region of a Boiling Water Reactor, NPPD will commit to providing new prr curves after the generic methodology is approved, but before the end of 2016 (prior to exceeding 32 EFPY.

By letter dated September 30, the licensee fulfilled Commitment #1. By letter dated January 3, 2013, the licensee rescinded Commitment #2. In addition, based on Section 3.0 of this safety evaluation, the NRC staff approves the use of the methodology on a plant specific basis, explicit to Cooper Nuclear Station.

- 14 5,0 STATE CONSULTATION In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the amendment. The State official had no comments.

6.0 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 published in the Federal Register on March 6, 2012 (77 FR 13372). 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.

7.0 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: B. Parks C. Sydnor Date: February 22, 2013

ML13032A526 ========;===~~;;';;;~;;';;;';~r==========="9l OFFICE NRRlDORULPL4/PM RRIDORULPL4/LA NRRlDE/EVIB/BC NAME LWilkins JBurkhardt SRosenberg*

DATE 13 2/11/13 1/16/13 OFFICE NRRlDSS/STSB/BC OGC NLO NRRlDORULPL4/BC NRR/DORULPL4/PM NAME RElliott DCylkowski MMarkley LWilkins DATE 2/12/13 2/20/13 2/22/13 2/22/13