ML093370002

Issuance of Amendment Relocation of Pressure and Temperature Curves to the Pressure Temperature Limits Report

ML093370002
Person / Time
Site:	Nine Mile Point
Issue date:	01/21/2010
From:	Richard Guzman Plant Licensing Branch 1
To:	Belcher S Nine Mile Point
Guzman R, NRR/DORL, 415-1030
References
TAC ME0817
Download: ML093370002 (21)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 21, 2010 Mr. Samuel L. Belcher 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.1-ISSUANCE OF AMENDMENT REGARDING RELOCATION OF PRESSURE AND TEMPERATURE LIMIT CURVES TO THE PRESSURE AND TEMPERATURE LIMITS REPORT (TAC NO. ME0817)

Dear Mr. Belcher:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 204 to Renewed Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station (NMPNS), Unit NO.1 (NMP1), in response to your application dated March 3, 2009 (Agencywide Documents Access Management System (ADAMS) Accession No. ML090640301), as supplemented on December 17, 2009 (ADAMS Accession No. ML093570097).

This amendment modifies Technical Specification (TS) Section 3.2.1, "Reactor Vessel Heatup and Cooldown Rates," and Section 3.2.2, "Minimum Reactor Vessel Temperature for Pressurization," by replacing the existing reactor vessel heatup and cooldown rate limits and the pressure and temperature limit curves with references to the Pressure and Temperature Limits Report (PTLR). In addition, the amendment adds a new section addressing administrative requirements for the PTLR to TS Section 6.0, "Administrative Controls."

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 Ope':ator Reactor Licensing Office of Nuclear Reactor RegUlation Docket No. 50-220

Enclosures:

1. Amendment No. 204 to DPR-63

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. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 204 Renewed License No. DPR-53

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 March 3, 2009, as supplemented on December 17, 2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the aCtivities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to trl.e Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-53 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 204, is hereby incorporated into this license.

Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical Specifications.

-2

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

FOR THE NUCLEAR REGULATORY COMMISSION

~/~

Nancy L. Salgado. Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: January 21, 2010

ATTACHMENT TO LICENSE AMENDMENT NO. 204 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 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 3 3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Page Insert Page 6 6 81 81 83 83 84 84 85 85 86 87 88 89 90 91 92 93 94 358 358 358a

1.16 Dose Equivalent 1-131 Dose Equivalent 1-131 shall be that concentration of 1-131 (microcurieslgram) that alone would produce the same dose as the quantity and isotopic mixture of 1-131,1-132,1-133,1-134, and 1-135 actually present. The dose conversion factors used for this calculation shall be the Committed Effective Dose Equivalent dose conversion factors listed in Table 2.1 of Federal Guidance Report No. 11, EPA, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion." 1988.

1.17 Recently Irradiated Fuel Recently irradiated fuel is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1.18 Pressure and Temperature Limits Report (PTLRl The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates. for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 6.6.7.

1.19 (Deleted) 1.20 (Deleted) 1.21 (Deleted)

AMENDMENT NO. 142, 172, 178, 194, 204 6

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.1 REACTOR VESSEL HEATUP AND COOLDOWN RATES Applicability:

Applies to the reactor vessel heating or cooling rate.

Objective:

To assure that thermal stress resulting from reactor heatup and cooldown are within allowable code limits.

Specification:

During the startup and shutdown operations of the reactor, the reactor vessel heatup and cooldown rates shall be maintained within the limits specified in the Pressure and Temperature Limits Report (PTLR).

AMENDMENT NO.~. 204 81

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION PRESSURIZATION Applicability: Applicability:

Applies to the minimum vessel temperature required Applies to the required vessel temperature for for vessel pressurization. pressurization.

Objective: Objective:

To assure that no substantial pressure is imposed on To assure that the vessel is not subjected to any the reactor vessel unless its temperature is substantial pressure unless its temperature is greater considerably above its Nil Ductility Transition than its Nil Ductility Transition Temperature (NDTT).

Temperature (NDTT).

Specification: Specification:

a. During reactor vessel heatup and cooldown when a. Reactor vessel temperature and pressure shall be the reactor is not critical, the reactor vessel monitored and controlled to assure that the temperature and pressure shall be maintained pressure and temperature limits specified in the within the limits specified in the Pressure and PTLR are met.

Temperature Limits Report (PTLR).

b. During reactor vessel heatup and cooldown when the reactor is critical, the reactor vessel temperature and pressure shall be maintained within the limits specified in the PTLR.

AMENDMENT NO. 142.184, 204 83

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. During leakage and hydrostatic testing, the reactor vessel temperature and pressure shall be maintained within the limits specified in the PTLR.

d. The reactor vessel head bolting studs shall not be under tension unless the temperature of the vessel head flange and the head are equal to or greater than the values specified in the PTLR.

AMENDMENT NO. 142. 164. 183. 184, 204 84

Pages 85 Through 94 Deleted AMENDMENT NO. 142, 164, 183, 204 85

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

1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel," U.S. Supplement, (NRC approved version specified in the COLR).

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

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

6.6.6 Special Reports Special reports shall be submitted within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a. - f. (Deleted)

g. Sealed Source Leakage In Excess Of Limits, Specification 3.6.5.2 (Three months).

h. Accident Monitoring Instrumentation Report. Specification 3.6.11.a (Table 3.6.11-2, Action 3 or 4) (Within 14 days following the event).

6.6.7 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

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

1. Limiting Condition for Operation Section 3.2.1. "Reactor Vessel Heatup and Cooldown Rates."

2. Limiting Condition for Operation Section 3.2.2, "Minimum Reactor Vessel Temperature for Pressurization."

3. Surveillance Requirement Section 4.2.2, "Minimum Reactor Vessel Temperature for Pressurization."

AMENDMENT NO. 142, 157, 162, 181, 184, 204 358

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

1. SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors,"

Revision 0, April 2007.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel f1uence period and for any revision or .

supplement thereto.

AMENDMENT NO. 204 358a

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 204 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 NINE MILE POINT NUCLEAR STATION, LLC NINE MILE POINT NUCLEAR STATION, UNIT NO.1 DOCKET NO. 50-220

1.0 INTRODUCTION

By letter dated March 3, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090640301), as supplemented by letter dated December 17, 2009 (ADAMS Accession No. ML093570097), Nine Mile Nuclear Station, LLC (NMPNS or the licensee) submitted a request for changes to the Nine Mile Point Unit No.1 (NMP1) Technical Specifications (TSs). The supplement dated December 17,2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination noticed in the Federal Register on May 19, 2009 (74 FR 23447).

The proposed amendment would revise the NMP1 TSs as necessary to relocate the pressure and temperature (P-T) limit curves and associated references to a Pressure and Temperature Limits Report (PTLR) (Reference 1). The request is submitted consistent with the guidance provided in NRC Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," (Reference 2), as supplemented by TS Task Force (TSTF) traveler TSTF-419-A, "Revise PTLR Definition and References in ISTS [Improved Standard Technical Specifications] 5.6.6, RCS PTLR,"

(Reference 3). Specifically, the request would modify TS Section 3.2.1, "Reactor Vessel Heatup and Cooldown Rates," and Section 3.2.2, "Minimum Reactor Vessel Temperature for Pressurization," by replacing the existing reactor vessel heatup and cooldown rate limits and the P-T limit curves with references to the PTLR. The request would also add a new definition for the PTLR to TS Section 1.0, "Definitions," and a new section addressing administrative requirements for the PTLR would be added to TS Section 6.0, "Administrative Controls."

2.0 REGULATORY EVALUATION

In Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR), the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plant's TSs.

-2 In general, there are two classes of changes to TSs: (1) changes needed to reflect modifications to the design basis, as TSs are derived from the design basis, and (2) changes to take advantage of the evolution in policy and guidance as to the required content and preferred format of TSs over time. In determining the acceptability of such changes, the NRC staff interprets the requirements in 10 CFR 50.36 using as a model the accumulation of generically approved guidance in the ISTS. For this review, the NRC staff used NUREG-1433, Revision 3, "Standard Technical Specifications, General Electric Plants BWR/4."

The NRC has established requirements in 10 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 PTLR based on the following NRC regulations and guidance: Appendix G to 10 CFR Part 50; Appendix H to 10 CFR Part 50; Regulatory Guide (RG) 1.99, Revision 2 (Rev. 2); GL 92-01, Rev. 1; GL 92-01, Rev. 1, Supplement 1; Standard Review Plan (SRP)

Section 5.3.2; and GL 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits." Appendix G to 10 CFR Part 50 requires that facility P-T limits for the 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). Appendix H to 10 CFR Part 50 establishes requirements related to facility RPV material surveillance programs. RG 1.99, Rev. 2 contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation.

GL 92-01, Rev. 1 requested that licensees submit the RPV data for their plants to the NRC staff for review, and GL 92-01, Rev. 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 provides an acceptable method for determining the P-T limits for ferritic materials in the beltline of the RPV based on the ASME Code,Section XI, Appendix G methodology.

The most recent version of Appendix G to Section XI of the ASME Code which has been endorsed in 10 CFR 50.55a, and therefore by reference in 10 CFR Part 50, Appendix G, is the 2004 Edition. Additionally, Appendix G to 10 CFR Part 50 imposes minimum head flange temperatures when system pressure is at or above 20% of the preservice hydrostatic test pressure.

GL 96-03 provides for the relocation of P-T limit curves from the plant TS to a PTLR, which remains under administrative control, and is incorporated by reference into the plant TS. To implement GL 96-03, a licensee is required to use, and reference in the TS, NRC-approved methodologies to develop the P-T limits. In this case, the licensee is using the methodologies described in (1) Technical Report MPM-402781, "Benchmarking of Nine Mile Point Unit 1 and Unit 2 Neutron Transport Calculations" (Reference 4), and (2) Structural Integrity Associates Topical Report SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors" (Reference 5).

TSTF-419-A provides additional guidance which provides an alternative format for documenting the implementation of a PTLR in the "Administrative Controls" section of a facility's TS. Since this license amendment request sought the initial implementation of a PTLR for NMP1, the staff's review focused on both the implementation of the NMP 1 PTLR and the appropriate application of the SIR-05-044-A methodology to generate the proposed NMP1 P-T limits.

-3 The NRC staff reviewed the information supplied by NMPNS regarding its fluence calculations to establish that the calculations were performed using the methodology that adheres to RG 1.190, "Calculational and Dosimetry Techniques for Determining Pressure Vessel Neutron Fluence,"

(Reference 7).

3.0 TECHNICAL EVALUATION

3.1 Licensee's Evaluation 3.1.1 PTLR Implementation of GL 96-03 contains seven technical criteria that the contents of a proposed methodology should conform to if license amendments requesting PTLRs are to be approved by the NRC staff. The NRC staffs evaluations of the contents of the Boiling Water Reactor Owners Group (BWROG) methodology against the seven criteria in Attachment 1 of GL 96-03 are provided in Section 3.1 of the safety evaluation (SE) for the SIR-05-044-A report dated February 6, 2007 (ADAMS Accession No. ML070180483). The licensee stated in its March 3, 2009, application, that a PTLR has been prepared for NMP1 based on the methodology and template provided in SIR-05-044-A and that the TS changes are consistent with the guidance provided in GL 96-03.

3.1.2 P-T limits The adjusted reference temperature (ART) values and P-T limits valid for 46 effective full-power years (EFPY) of facility operation using the SIR-05-044-A methodology were documented in the proposed NMP1 PTLR. The licensee identified the limiting beltline material as the P2112 lower shell plate. The key parameters in determining the licensee's ART value for the limiting material at the one-quarter of the RPV wall thickness (1/4T) location are shown in Table 11 and Table 12 of the PTLR for 36 EFPY and 46 EFPY, respectively. Corresponding parameters at the three quarter of the RPV wall thickness (3/4T) are not provided in the PTLR because the PTLR indicated in Section 5, "Discussion," that the P-T limit curves based on the cooldown transient (where the relevant critical location is at the 1/4T depth) are more conservative than the P-T limit curves based on the heatup transient (where the relevant critical location is at the 3/4T depth).

The licensee will use the developed cooldown curves for both heatup and cooldown of the NMP1 RPV.

3.1.3 Fluence Methodology In March 2001, the NRC staff issued RG 1.190 which provides methods for determining RPV fluence. Fluence calculations are acceptable if they are performed with approved methodologies or with methods which are shown to conform with the guidance in RG 1.190.

The NRC staff approved the fluence methodology described in Technical Report MPM-40278 for plant specific use at NMP1 in its SE dated October 27,2003 (Reference 6). Reference 6 discusses the NRC staffs review of the methodology and concludes that the code is acceptable based on its adherence to the guidance contained in RG 1.190. The fluence values used for the determination of the NMP1 P-T limits curves used this methodology.

-4 3.2 Staff Evaluation 3.2.1 PTLR Implementation As mentioned in Section 3.1.1 of this SE, Attachment 1 of GL 96-03 requires the licensee to evaluate and document seven technical criteria to demonstrate the acceptability of its PTLR methodology. The NRC staff examined the proposed PTLR and determined that it was developed from the template PTLR of the SIR-05-044-A report and meets the seven technical criteria:

(1) The PTLR methodology describes the transport calculation methods including computer codes and formula used to calculate neutron fluences (Section 3.0, "Methodology," page 2 of the NMP1 PTLR).

(2) The PTLR methodology describes the surveillance program (Appendix A, "NMP1 Reactor Vessel Materials Surveillance Program," page 41 of the NMP1 PTLR).

(3) The PTLR methodology describes how the low temperature overpressure protection system limits are calculated applying system/thermal hydraulics and fracture mechanics (not applicable to BWRs).

(4) The PTLR methodology describes the method for calculating the ART values using RG 1.99, Rev. 2 (Section 3.0, Methodology," page 2 of the NMP1 PTLR).

(5) The PTLR methodology describes the application of fracture mechanics in the construction of P-T limits based on ASME Code,Section XI, Appendix G and the guidance in the NRC's SRP. The NMP1 PTLR provided information regarding the 'finite element analyses performed to generate part of the P-T limits. The submittal stated that the equations and values were calculated in accordance with the SIR-05-044-A report.

(6) The PTLR methodology describes how the minimum temperature requirements in Appendix G to 10 CFR Part 50 are applied to P-T limits for boltup temperature and hydrotest temperature. (Page 1 of the NMP1 PTLR stated that the P-T limits were calculated in accordance with the SIR-05-044-A Report. This description is sufficient because the SIR-05-044-A report contained detailed information regarding the minimum temperature requirements for boltup temperature and hydrotest temperature.)

(7) The PTLR methodology describes how the data from multiple surveillance capsules are used in the ART calculation. Appendix A of the NMP1 PTLR was developed using the PTLR template in the SIR-05-044-A Report which does not describe how the data from multiple surveillance capsules are used in the ART calculation. However, the !\IRC staff accepts the NMP1 PTLR because the generic information is available in Appendix A of the SIR-05-044-A Report.

Given that the licensee has adequately addressed the criteria from GL 96-03, the NRC staff finds that the implementation of the NMP1 PTLR is acceptable.

-5 3.2.2 P-T limits To evaluate the proposed P-T limits for the NMP1 RPV, the NRC staff confirmed the licensee's selection of the lower shell plate P2112 as the limiting beltline material and performed an independent calculation of the ART values for this material using the RG 1.99, Rev. 2, methodology. The NRC staff's ART values for the limiting beltline material at the 1/4T location are 162.2 of and 171.5 of for 36 EFPY and 46 EFPY, respectively, which were calculated using materials information for NMP1 in the NRC Reactor Vessel Integrity Database (RVID) and the RPV inner diameter (10) fluence in the NMP1 PTLR. The licensee's ART values of 159.0 of and 167.4 of at the 1/4T location for 36 EFPY and 46 EFPY, respectively, for the limiting beltline material are close to the NRC staff's values based on identical initial reference temperature (RT NDT) and copper (Cu) and nickel (Ni) values. The licensee did not calculate the ART value at the 3/4T location, which is relevant to the heatup P-T limit calculation, because the SIR-05-044-A Report concluded that P-T limits for the cooldown transient are bounding.

The proposed NMP1 P-T limits are composite curves, representing the most limiting P-T limits for the RPV beltline, the bottom head, and the upper vessel. The NRC staff has independently verified that, for a cooldown of 100 of per hour, the bottom head P-T limits are the least limiting.

Consequently, only the RPV beltline P-T limits (upper segment) and the upper vessel P-T limits (lower segment) are represented in the proposed normal operation core not critical heatup/cooldown curve (Curve B). This is also true for the proposed NMP1 P-T limits for pressure test (Curve A). The NRC staff performed independent calculations for both segments of the proposed P-T limits valid for 36 EFPY and 46 EFPY.

For the RPV beltline P-T limit segment, the NRC staff utilized the ASME Code,Section XI, Appendix G methodology in its independent evaluation, using the K1c curve as resistance and the pressure-dependent Kim formula and the cooldown rate dependent Kit formula as driving forces. The NRC staff used plant-specific information submitted by the licensee, which included:

the temperature measurement instrument uncertainty, the pressure measurement instrument uncertainty, and the pressure head for accounting the column of water in the RPV provided in the licensee's calculation package. The NRC staff produced almost identical beltline P-T limits for the two ends of the upper segment. The NRC staff's calculation indicated that the licensee's "Minimum Reactor Vessel Metal Temperature (degrees F)" in the proposed P-T limits is actually the metal temperature at 1/4T. This represents a deviation from the SIR-05-044-A approach which recommended use of the coolant temperature, instead of the metal temperature at 1/4T, to evaluate Klc . The SIR-05-044-A report further stated that "[t]he use of the coolant temperature is considered to be a necessary conservatism in P-T curve development to ensure that all design margins and safety factors are maintained." However, since this "necessary conservatism" is not required by the ASME Code,Section XI, Appendix G or by 10 CFR Part 50, Appendix G, the NRC staff considers the licensee's deviation acceptable. The NRC staff expects that the licensee will, however, impose the necessary controls on the NMP1 reactor coolant temperature to ensure that the P-T limits, based on RPV metal temperature, will be met.

For the upper vessel P-T limit segment, the NRC staff utilized the Kit and Kim formulas in the SIR-05-044-A report to calculate driving forces and the ASME Code,Section XI, Appendix G Klc curve to calculate resistance. The input nozzle corner pressure and thermal hoop stresses were based on plant-specific finite element model (FEM) results for the NMP1 feedwater nozzle under the limiting turbine roll event. This transient represents the most severe event for the feedwater

-6 nozzle. The NRC staff, therefore, agrees with the licensee that this event is equivalent to the limiting normal/upset design transient for a BWR feedwater nozzle. The NRC staff's calculation produced consistent P-T values for a randomly selected point along the lower segment of the proposed P-T limits.

10 CFR Part 50, Appendix G contains additional requirements for the minimum metal temperature of the closure head flange and vessel flange regions. These considerations were reflected in the "notch" of the upper vessel P-T limits. The NRC staff verified that when the pressure is greater than 20% of the hydro test pressure, the temperature for the pressure test P-T limits is greater than the RT NDT of the limiting flange material plus 90 of (130 OF), and the temperature for the core not critical P-T limits is greater than the RT NDT of the limiting flange material plus 120 of (160 OF). NMP1 pressure test P-T limits also show a 70 of straight line on the low pressure end, meeting the 10 CFR Part 50, Appendix G minimum temperature requirement for pressure tests which limits the operating temperature to the highest RT NDT of the closure flange that is highly stressed by the bolt preload. Since this value for the NMP1 RPV limiting flange material is 40 of, the licensee's approach is conservative.

Based on the above evaluation, the NRC staff determined that the licensee's proposed P-T limits are in accordance with the SIR-05-044-A Report and satisfy the requirements of Appendix G to Section XI of the ASME Code and Appendix G to 10 CFR Part 50. Hence, the licensee's proposed P-T limit curves are acceptable for operation of the NMP1 RPV and are valid for 46 EFPY.

3.2.3 Fluence Methodology The licensee stated that the methodology used to calculate RPV neutron fluence values utilized in the development of the NMP1 P-T limit curves is in accordance with the recommendations of RG 1.190. The NRC documented its determination that the NMP neutron fluence calculations were acceptable in Reference 6.

The NRC staff reviewed the licensee's references that pertain to its reactor vessel neutron fluence calculations and determined the following:

The methods rely on S8 angular quadrature and P-3 Legendre expansion. The calculations use the BUGLE-96 cross section library, which is based on ENDF/B-VI nuclear data. The methodology calculates a three-dimensional neutron flux distribution via two-dimensional flux synthesis.

Both generic and plant-specific benchmarking have demonstrated that average deviation between calculated and measured dosimetry results is well within RG 1.190 limits of +20 percent. These considerations reflect adherence to the guidance in RG 1.190.

Because the methodology meets the guidance set forth in RG 1.190 and has been adequately benchmarked, the NRC staff finds that it is acceptable for use with the SIR-05-044-A methodology. Therefore, the NRC staff finds that the fluence criterion of GL 96-03 is adequately addressed, and on that basis, the licensee's requested implementation of the PTLR is acceptable.

-7 3.2.4 Generic Resolution of TSTF-419 By letter dated November 23,2009 (ADAMS Accession No. IVIL093270101), the NRC staff identified an issue concerning the inclusion of revision numbers and dates in the TSs for the PTLR References list. The issue identified by the NRC staff is discussed in a separate letter dated November 2,2009 (ADAMS Accession No. ML092151016), issued by the NRC to the TSTF, regarding the use of TSTF-419-A to remove revision numbers and dates from the TS References list.

In the November 2, 2009 letter, the NRC staff requested NMPNS to add the revision number and date to the reference requested for inclusion in the new TS 6.6.7. The licensee responded by providing a revised, proposed TS page including the revision number and date to SIR-05-044A.

The NRC staff reviewed this supplement and determined that it was consistent with the guidance contained in GL 96-03, and is more restrictive than the format specified in TSTF-419 A. Based on these considerations, the NRC staff determined that the modification to TS 6.6.7 is acceptable.

3.2.4 Conclusion Based on the NRC staff's review of the information provided in the licensee's March 3, 2009 submittal and the topics assessed in this SE, the NRC staff concludes:

The proposed NMP1 PTLR meets GL 96-03 requirements for implementation and, therefore, is approved as part of the NMP1 licensing basis.

The I\JI\IIP1 RPV P-T limits are based on an acceptable methodology documented in the SIR-05 044-A Report. The staff performed independent evaluations and verified that the P-T limits were developed appropriately using the SIR-05-044-A methodology, and the proposed P-T limits for 46 EFPY satisfy the requirements of Appendix G to Section XI of the ASIVIE Code and Appendix G to 10 CFR Part 50.

As a separate issue, the staff determined that the licensee has met its commitment, made as part of the NRC approval of the NMP1 license renewal application, to submit revised P-T limits for NRC review and approval as part of a plan for managing P-T limits prior to the period of extended operation. As long as the P-T limit methodology remains the same, this implementation of a PTLR allows the licensee to revise P-T limits for any licensed period of operation under the 10 CFR 50.59 process.

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.

-8

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

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 (May 19, 2009 (74 FR 23447)). 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. Polson, Keith J., Constellation Energy, letter to U.S. Nuclear Regulatory Commission, "License Amendment Request Pursuant to 10 CFR 50.90: Relocation of Pressure and Temperature Limits to the Pressure and Temperature Limits Report," Docket No. 50-220, ADAMS Accession No. ML090640301, March 3, 2009.

2. U.S. Nuclear Regulatory Commission, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," Generic Letter 96-03, ADAMS Accession No. ML031110004, January 31, 1996.

3. Pietrangelo, A. R., Nuclear Energy Institute, letter to W. D. Beckner, U.S. Nuclear Regulatory Commission, "Forwarding of TSTFs," Enclosure 2, "Revise PTLR Definition and References in the ISTS 5.6.6., RCS PTLR," ADAMS Accession No. ML012690199, September 19, 2001.

4. Manahan, M. P., "Benchmarking of Nine Mile Point Unit 1 and 2 Neutron Transport Calculations," MPM-402781, ADAMS Accession No. ML032681023, September 2003.

5. Stevens, G. L., Structural Integrity Associates, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," Report Number SIR-05-044-A, ADAMS Accession No. ML072340283, Centennial, Colorado, April 30, 2007.

6. Tam, P., U.S. NRC, letter to Katz, P. E., Constellation Energy, "Nine Mile Point Nuclear Station, Unit NO.1-Issuance of Amendment Re: Pressure-Temperature Limit Curves and Tables," ADAMS Accession No. ML032760696, October 27,2003.

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7. U.S. Nuclear Regulatory Commission, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," RG 1.190, ADAMS Accession No. ML010890301, No. March 31, 2001.

Principal Contributors: C. Fairbanks B. Parks Date: January 21, 2010

ML093570097).

This amendment modifies Technical Specification (TS) Section 3.2.1, "Reactor Vessel Heatup and Cooldown Rates," and Section 3.2.2, "Minimum Reactor Vessel Temperature for Pressurization," by replacing the existing reactor vessel heatup and cooldown rate limits and the pressure and temperature limit curves with references to the Pressure and Temperature Limits Report (PTLR). In addition, the amendment adds a new section addressing administrative requirements for the PTLR to TS Section 6.0, "Administrative Controls."

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, IRA!

Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosures:

1. Amendment No. 204 to DPR-63

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

PUBLIC RidsOgcRp RidsOGCMailCenter RidsNrrDssSrxb LPL1-1 R/F RidsNrrDorlLPL 1-1 RidsNrrLASUttie RidsRgn1 MailCenter RidsAcrsAcnw_MailCenter RidsNrrDirsltsb RidsNrrDorlDpr RidsNrrPMNineMilePoint BParks, NRR CFairbanks, NRR CSchulten, NRR RidsNrrDeCvib A DA MSAccesslon. N0.: ML093370002 *SE provi'dedb)y memo, No su bstanrla I chanQes rnade. ** Concurrence via e-mal'I NRR 106 OFFICE LPL1-1/PM LPL 1-1/LA DSS/SRXB/BC DE/CVIB/BC 01 RS/ITSB/BC OGC/NLO LPL 1-1/BC NAME RGuzman SUttle GCranston* MMitchell* RElliott CBoote NSalgado (wI comments)

DATE 01/04/10 01/04/10** 09/09/09 SE DTD 12/10109 SE DTD 01/12/10 01/06/10 01/21/10