ML14245A151

Issuance of Amendments Regarding Technical Specification 3.1.3, Moderator Temperature Coefficient Measurement

ML14245A151
Person / Time
Site:	Beaver Valley
Issue date:	09/17/2014
From:	Jeffrey Whited Plant Licensing Branch 1
To:	Emily Larson FirstEnergy Nuclear Operating Co
Whited J
References
TAC ME9144, TAC ME9145
Download: ML14245A151 (22)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 17, 2014

SUBJECT:

BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2- ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATION 3.1.3, "MODERATOR TEMPERATURE COEFFICIENT MEASUREMENT" (TAC NOS.

ME9144 AND ME9145)

Dear Mr. Larson:

The Commission has issued the enclosed Amendment No. 291 to Renewed Facility Operating License No. DPR-66 for the Beaver Valley Power Station, Unit No. 1 (BVPS-1) and Amendment No. 178 to Renewed Facility Operating License No. NPF-73 for the Beaver Valley Power Station, Unit No. 2 (BVPS-2). These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated July 25, 2012, as supplemented by letters dated June 1, October 21, and November 14, 2013.

The amendments revise the BVPS-1 and BVPS-2 TS 3.1.3, "Moderator Temperature Coefficient Measurement (MTC)," to allow the normally required near end-of-life MTC measurement to not be performed under certain conditions.

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

frey A. Whited, Project Manager ant Licensing Branch 1-2 1 ivision of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-334 and 50-412

Enclosures:

1. Amendment No. 291 to DPR-66

2. Amendment No. 178 to NPF-73

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

FIRSTENERGY NUCLEAR OPERATING COMPANY FIRSTENERGY NUCLEAR GENERATION. LLC.

DOCKET NO. 50-334 BEAVER VALLEY POWER STATION, UNIT NO.1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 291 License No. DPR-66

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

A. The application for amendment by filed by FirstEnergy Nuclear Operating Company (FENOC}* acting on its own behalf and as agent for FirstEnergy Nuclear Generation, LLC (the licensees), dated July 25, 2012, as supplemented by letters dated June 1, October 21, and November 14, 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 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.

• FENOC is authorized to act as agent for FirstEnergy Nuclear Generation, LLC, and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

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-66 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 291, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION

~11 Robert G. Schaaf, Acting Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: September 17, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 291 RENEWED FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334 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 Insert 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 Insert 3.1.3-2 3.1.3-2 5.6-2 5.6-2 5.6-3 5.6-3

(3} FENOC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) FENOC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) FENOC, 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 the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1:

Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level FENOC is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 291, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Auxiliary River Water System (Deleted by Amendment No.8)

Amendment No. 291 Beaver Valley Unit 1 Renewed Operating License DPR-66

MTC 3.1.3 FREQUENCY SR 3.1.3.2

-NOTES-

1. Not required to be performed until 7 effective full power days (EFPD) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.

2. If the MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR, SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle.

3. SR 3.1.3.2 need not be repeated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of s 60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR.

4. SR 3.1.3.2 is not required to be performed provided that the benchmark criteria specified in WCAP-13749-P-A and the COLR requirements for the calculated revised predicted MTC are satisfied.

Verify MTC is within lower limit. Once each cycle Beaver Valley Units 1 and 2 3.1.3- 2 Amendments 291 /178

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued)

LCO 3.1.5, "Shutdown Bank Insertion Limits" LCO 3.1.6, "Control Bank Insertion Limits" LCO 3.2.1, "Heat Flux Hot Channel Factor (Fo(Z))"

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor ( F~H )"

LCO 3.2.3, "Axial Flux Difference (AFD)"

LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation"- Overtemperature and Overpower ~ T Allowable Value parameter values LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits" LCO 3.9.1, "Boron Concentration"

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:

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"

WCAP-87 45-P-A, "Design Bases for the Thermal Overtemperature ~ T and Thermal Overpower ~T Trip Functions,"

WCAP-12945-P-A, Volumes 1 through 5, "Code Qualification Document for Best Estimate LOCA Analysis,"

(For Unit 1 only) WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

WCAP-1 0216-P-A, "Relaxation of Constant Axial Offset Control/

Fa Surveillance Technical Specification,"

WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis,"

WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report,"

WCAP-15025-P-A, "Modified WRB-2 Correlation, WRB-2M, for Predicating Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids,"

WCAP-137 49-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," March 1997 (Westinghouse Proprietary),

Beaver Valley Units 1 and 2 5.6-2 Amendments 291 /178

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued)

WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON,"

WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology."

As described in reference documents listed above, when an initial assumed power level of 102% of RATED THERMAL POWER is specified in a previously approved method, 100.6% of RATED THERMAL POWER may be used when input for reactor thermal power measurement of feedwater flow is by the leading edge flow meter (LEFM).

Caldon, Inc. Engineering Report-SOP, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM ..J TM System" Caldon, Inc. Engineering Report-160P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM ..J TM System"

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

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

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

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, Overpressure Protection System (OPPS) enable temperature, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (PIT) Limits," and LCO 3.4.12, "Overpressure Protection System (OPPS)"

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 documents:

NRC Letter, "Beaver Valley Power Station, Units 1 and 2- Acceptance of Methodology for Referencing Pressure and Temperature Limits Report (TAC Nos. MB3319 and MB3320)," dated October 8, 2002.

Beaver Valley Units 1 and 2 5.6-3 Amendments 291 I 178

FIRSTENERGY NUCLEAR OPERATING COMPANY FIRSTENERGY NUCLEAR GENERATION, LLC.

OHIO EDISON COMPANY THE TOLEDO EDISON COMPANY DOCKET NO. 50-412 BEAVER VALLEY POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 178 License No. NPF-73

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

A. The application for amendment by FirstEnergy Nuclear Operating Company (FENOC}*-acting on its own behalf and as agent for FirstEnergy Nuclear Generation, LLC, Ohio Edison Company, and The Toledo Edison Company (the licensees),dated July 25, 2012, as supplemented by letters dated June 1, October 21, and November 14, 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 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.

• FENOC is authorized to act as agent for FirstEnergy Nuclear Generation, LLC., Ohio Edison Company, and The Toledo Edison Company and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

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 Facility Operating License No. NPF-73 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 178, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

FOR THE NUCLEAR REGULATORY COMMISSION

?U).Jg Robert G. Schaaf, Acting Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: September 17, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 178 RENEWED FACILITY OPERATING LICENSE NO. NPF-73 DOCKET NO. 50-412 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 Insert 4 4 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 Insert 3.1.3-2 3.1.3-2 5.6-2 5.6-2 5.6-3 5.6-3

(b) Further, the licensees are also required to notify the NRC in writing prior to any change in: (i) the term or conditions of any lease agreements executed as part of these transactions; (ii) the BVPS Operating Agreement, (iii) the existing property insurance coverage for BVPS Unit 2, and (iv) any action by a lessor or others that may have adverse effect on the safe operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations set forth in 10 CFR Chapter 1 and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level FENOC is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 178, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 178 Beaver Valley Unit 2 Renewed Operating License NPF-73

MTC 3.1.3 FREQUENCY SR 3.1.3.2

-NOTES-

1. Not required to be performed until 7 effective full power days (EFPD) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.

2. If the MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR, SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle.

3. SR 3.1.3.2 need not be repeated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of : : ; 60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR.

4. SR 3.1.3.2 is not required to be performed provided that the benchmark criteria specified in WCAP-13749-P-A and the COLR requirements for the calculated revised predicted MTC are satisfied.

Verify MTC is within lower limit. Once each cycle Beaver Valley Units 1 and 2 3.1.3- 2 Amendments 291 I 178

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT {COLR) (continued)

LCO 3.1.5, "Shutdown Bank Insertion Limits" LCO 3.1.6, "Control Bank Insertion Limits" LCO 3.2.1, "Heat Flux Hot Channel Factor (Fo(Z))"

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor ( F~H )"

LCO 3.2.3, "Axial Flux Difference (AFD)"

LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation"- Overtemperature and Overpower AT Allowable Value parameter values LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits" LCO 3.9.1, "Boron Concentration"

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:

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"

WCAP-87 45-P-A, "Design Bases for the Thermal Overtemperature AT and Thermal Overpower AT Trip Functions,"

WCAP-12945-P-A, Volumes 1 through 5, "Code Qualification Document for Best Estimate LOCA Analysis,"

(For Unit 1 only) WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

WCAP-1 0216-P-A, "Relaxation of Constant Axial Offset Control/

Fa Surveillance Technical Specification,"

WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis,"

WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report,"

WCAP-15025-P-A, "Modified WRB-2 Correlation, WRB-2M, for Predicating Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids,"

WCAP-137 49-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," March 1997 (Westinghouse Proprietary),

Beaver Valley Units 1 and 2 5.6-2 Amendments 291 /178

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued)

WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON,"

WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology."

As described in reference documents listed above, when an initial assumed power level of 102% of RATED THERMAL POWER is specified in a previously approved method, 100.6% of RATED THERMAL POWER may be used when input for reactor thermal power measurement of feedwater flow is by the leading edge flow meter (LEFM).

Caldon, Inc. Engineering Report-BOP, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM -.,J TM System" Caldon, Inc. Engineering Report-160P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM -.,J TM System"

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

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

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

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, Overpressure Protection System (OPPS) enable temperature, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (PfT) Limits," and LCO 3.4.12, "Overpressure Protection System (OPPS)"

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 documents:

NRC Letter, "Beaver Valley Power Station, Units 1 and 2- Acceptance of Methodology for Referencing Pressure and Temperature Limits Report (TAG Nos. MB3319 and MB3320)," dated October 8, 2002.

Beaver Valley Units 1 and 2 5.6-3 Amendments 291 I 178

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 291 AND 178 TO FACILITY OPERATING LICENSE NOS. DPR-66 AND NPF-73 FIRSTENERGY NUCLEAR OPERATING COMPANY FIRSTENERGY NUCLEAR GENERATION. LLC.

OHIO EDISON COMPANY THE TOLEDO EDISON COMPANY BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-334 AND 50-412

1.0 INTRODUCTION

By application dated July 25, 2012, 1 as supplemented by letters June 1, October 21, and November 14, 2013, 2 FirstEnergy Nuclear Operating Company (the licensee), requested changes to the Technical Specifications {TSs) for Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2). The proposed changes would revise the BVPS-1 and 2 TS 3.1.3, "Moderator Temperature Coefficient Measurement (MTC)," to allow the normally required near end-of-life (EOL) MTC measurement to not be performed under certain conditions. If these conditions are met, the MTC measurement would be replaced by a calculated value.

The supplements provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on September 4, 2012 (77 FR 53929).

2.0 BACKGROUND

On October 9, 1996, the NRC approved Westinghouse topical report WCAP-13749-P-A, 3 "Safety Evaluation Supporting the Conditional Exemption of the Most Negative [End-of-Life]

EOL Moderator Temperature Coefficient Measurement," for referencing in license applications.

The NRC's safety evaluation (SE) conclusion stated that" ... the analysis for the proposed TS change is acceptable provided {1) only PHOENIX/ANC calculation methods are used for the 1 Agencywide Documents Access and Management System (ADAMS) Accession No. ML12208A309.

2 ADAMS Accession Nos. ML13155A021, ML13295A106, and ML13319A882, respectively.

3 A non-proprietary version of WCAP-137 49-P-A, WCAP-14851-A, with the same title as WCAP-137 49-P-A, is available at ADAMS Legacy Accession No. 9704230055. ADAMS Legacy documents are available through the NRC's Public Document Room.

individual plant analyses relevant to the determinations for the EOL MTC methodology, and (2) the predictive correlation is reexamined if changes in core fuel designs or continued MTC calculation/measurement data show significant effect on the predictive correction."

On March 18, 2004, the NRC approved Westinghouse topical report WCAP-16045, "Qualification of the Two-Dimensional Transport Code PARAGON."4 The conclusion of theSE states, in part that, "[i]n addition, the staff considers the new PARAGON code to be well qualified as a stand-alone code replacement for the PHOENIX-P lattice code, wherever the PHOENIX-P code is used in NRC-approved methodologies. The staff considers it acceptable for licensing applications."

By email dated August 16, 2012, 5 the NRC staff accepted the license amendment request for review.

By letter dated December 28, 2012, 6 the NRC staff issued a request for additional information (RAI) with three questions regarding FENOC's letter dated July 25, 2012. On February 4, 2013, the NRC staff issued a public meeting notice for a public meeting to be held on February 20, 2013/ between the NRC and representatives of FENOC. The purpose of the meeting was to discuss an RAI sent by the NRC staff to FENOC by letter dated December 28, 2012. The NRC staff provided a meeting summary on March 11, 2013. 8 By letter dated May 9, 2013, 9 the NRC staff stated that a response to RAI question number 3 on the letter dated December 28, 2012, is not required.

By letter dated September 18, 2013, 10 the NRC staff determined that the information contained in the following document should be withheld from public disclosure pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) Section 2.390:

FirstEnergy Nuclear Operating Company Letter L-13-07 4, Attachment 1, "Response to December 28, 2012 Request for Additional Information" (Proprietary Version).

By letter dated September 21,2013, 11 the NRC staff requested additional information.

4 ADAMS Accession No. ML040780402.

5 ADAMS Accession No. ML12229A465.

6 ADAMS Accession No. ML12340A256.

7 ADAMS Accession No. ML13023A175.

8 ADAMS Accession No. ML13057A678.

9 ADAMS Accession No. ML13084A346.

10 ADAMS Accession No. ML13241A275. A nonproprietary version has been placed in the NRC's Public Document Room under ADAMS Accession No. ML13155A021.

11 ADAMS Accession No. ML13252A258.

3.0 REGULATORY EVALUATION

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:

Criterion 2 of 10 CFR 50.36(c)(2)(ii) states that:

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.

The limitations on MTC contained in TS 3.1.3 and verified by Surveillance Requirement (SR) 3.1.3.2 provide assurance that the value of the coefficient remains within the limiting condition assumed in the final safety analysis report accident and transient analyses. This, in turn, provides assurance that the reactor will be operated in a safe manner.

Given that the Topical Report WCAP-13749-P-A has been approved by the NRC staff, the focus of this review will be verification that the conditions and limitations of the generic approval are satisfied for the Beaver Valley specific application.

4.0 TECHNICAL EVALUATION

4.1 Description of Changes The proposed changes are:

1. SR 3.1.3.2 would be revised to exempt the requirement for a near-EOL MTC measurement, if the specified benchmark criteria and Core Operating Limits Report (COLR) requirements for near-EOL MTC are satisfied.

2. WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative [End-of-Life] EOL Moderator Temperature Coefficient Measurement," March 1997, would be added to the list of references for the COLR in TS 5.6.3.b.

3. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON,"

would be added to the list of references for the COLR in TS 5.6.3.b.

4. WCAP-16045-P-A. Addendum 1-A. "Qualification of the NEXUS Nuclear Data Methodology,"

would be added to the list of references for the COLR in TS 5.6.3.b.

4.2 Reason for Changes The TS 3.1.3 places limits on the MTC based on the accident analysis assumptions for the moderator density coefficient. A positive moderator density coefficient corresponds to a negative MTC. The most negative MTC limiting condition for operation (LCO) limit requires that the MTC be less negative than the specified limit for the all rods withdrawn, EOL, rated thermal power condition. To demonstrate compliance with the most negative MTC LCO, the surveillance requires verification of the MTC after 300 parts per million (ppm) equilibrium boron

concentration is reached. From the time that 300 ppm is reached until EOL, the hot full power (HFP) MTC will gradually become more negative due to additional core burnup and boron concentration reduction. To account for this effect, the 300 ppm MTC surveillance limit is sufficiently less negative than the EOL LCO limit to ensure that the LCO limit will be met, as long as the 300 ppm MTC surveillance criterion is met.

Currently, TS 3.1.3 requires measurements of MTC at beginning-of-life (BOL) to verify the most positive MTC limit and at near-EOL to verify the most negative MTC limit. At BOL, the measurement of the isothermal temperature coefficient is relatively simple to perform since it is performed at hot zero power (HZP) isothermal conditions and is not complicated by changes in the reactor coolant enthalpy rise or the presence of xenon. The measurement made near-EOL is performed at or near HFP conditions. MTC measurements at HFP are more difficult to perform than at HZP due to small variations in soluble boron concentration, changes in xenon concentration and distribution, changes in fuel temperature, and changes in reactor coolant enthalpy rise created by small changes in the core average power during the measurement.

Unless changes in each of these parameters are accounted for when reducing the measurement data, additional measurement uncertainties would be introduced. Although these additional uncertainties and the total reactivity change associated with the swing in moderator temperature would be minimal, the resulting MTC measurement uncertainty created by even a small change in power level could become significant and, if improperly accounted for, could yield inaccurate measurement results.

The MTC measurement typically includes time at reduced power as a result of the MTC determination measurement procedures. Additional manpower is also required to perform the test. This measurement disrupts normal reactor operation and increases the potential for a reactivity event due to a human performance error or unanticipated equipment issues.

Therefore, in order to improve availability and minimize disruptions to normal reactor operation, FENOC proposes an alternative to the EOL MTC measurement. If predefined conditions are met, the SR 3.1.3.2 EOL MTC measurement would be replaced by a calculated verification of an acceptable margin to the surveillance required MTC limit.

The proposed change would modify the EOL MTC SR by placing a set of conditions on core operations. If these conditions are met, that is, the specified revised prediction of the MTC and several core parameters measured during the cycle are within specified bounds, performing the surveillance measurement would not be required.

4.3 Evaluation FENOC has proposed to compare an alternate predicted MTC to the EOL (300 ppm) MTC. If the predicted value is less negative than the EOL MTC, then an EOL MTC measurement would not be required. The proposed calculational method is consistent with the method described and approved in WCAP-13749-P-A. It should be noted that the NRC staff approved the calculational method with the following two conditions:

1. Only PHOENIX/ANC calculation methods should be used for the plant-specific analyses for the determination of the EOL MTC, and

2. The predictive correction will be reexamined if changes in core fuel designs or continued MTC calculation/measurement data show significant effect on the predictive correction.

In its letter dated July 25, 2012, FENOC stated that BVPS would meet both of these above conditions. In the letter dated November 14, 2013, FENOC stated that the PARAGON/NEXUS code system will be used in place of the PHOENIX/ANC system. The PARAGON/NEXUS code system is an NRC approved code system and is an acceptable substitute for the PHOENIX/ANC code system when used with WCAP-13749-P-A and WCAP-9272-P-A methodologies. Therefore, the NRC staff finds that FENOC will meet the first of the two conditions.

The second condition is for re-evaluation of the predictive correction if the measured and predictive (M-P) data shows a "significant effect" on the predictive correction or if core fuel design changes are implemented that could have such an effect. In the absence of future surveillances there will not be M-P data for this evaluation at the EOL condition and, if there were data, the threshold of "significant effect" is not defined in the Technical Report.

The NRC staff understands the second condition to mean that the EOL MTC surveillance will be performed in the cycle following implementation of a core fuel design change that is reasonably expected to alter standard deviation of the M-P data or MTC behavior of the reactor. If the predictive correction is not altered by the addition of new M-P data, then future surveillances may be exempted using the WCAP-13749-P-A methodology. The NRC staff defines "significant effect" for this application as a change in the standard deviation such that the stated uncertainty of 3 percent milli-rho per degrees Fahrenheit (pcm/°F) is no longer bounding.

The core fuel design changes that are expected to alter the MTC behavior of the reactor are the following:

1. An increase in the allowable core thermal power ( > 2%), or

2. A change in designed operating cycle length from the current strategy of 18 month operating cycles. More precisely, a change to annual or 24 months operating cycles, or

3. Introduction of a new reload batch fuel product line (excludes Lead Test Assembly program).

The NRC staff finds that FENOC, by Note 4 of SR 3.1.3.2, will meet the second of the two conditions making the proposed TS revision acceptable.

The NRC staff finds that the application of WCAP-13749-P-A methodology is not compatible with:

1. Mixed Oxide (MOX) Fuel, or

2. Burnable Poison materials not previously approve and operated in the BVPS-1 and BVPS-2 reactors.

5.0 STATE CONSULTATION

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

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change 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 amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding (September 4, 2012, 77 FR 53929). Accordingly, the amendments meet 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 amendments.

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 Contributor: J. Dean and J. Lamb Date: September 17, 2014

September 17, 2014 Mr. Eric A. Larson, Site Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mail Stop A-BV-SEB1 P.O. Box 4, Route 168 Shippingport, PA 15077

SUBJECT:

BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2- ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATION 3.1.3, "MODERATOR TEMPERATURE COEFFICIENT MEASUREMENT" (TAC NOS.

ME9144 AND ME9145)

Dear Mr. Larson:

The Commission has issued the enclosed Amendment No. 291 to Renewed Facility Operating License No. DPR-66 for the Beaver Valley Power Station, Unit No. 1 (BVPS-1) and Amendment No. 178 to Renewed Facility Operating License No. NPF-73 for the Beaver Valley Power Station, Unit No. 2 (BVPS-2). These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated July 25, 2012, as supplemented by letters dated June 1, October 21, and November 14, 2013.

The amendments revise the BVPS-1 and BVPS-2 TS 3.1.3, "Moderator Temperature Coefficient Measurement (MTC)," to allow the normally required near end-of-life MTC measurement to not be performed under certain conditions.

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

Sincerely, IRA/

Jeffrey A. Whited, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-334 and 50-412

Enclosures:

1. Amendment No. 291 to DPR-66

2. Amendment No. 178 to NPF-73

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

PUBLIC LPLI1-1 R/F RidsNrrDorllpll-2 RidsNrrPMBeaverValley RidsNrrDirsltsb RidsNrrLAABaxter RidsAcrsAcnwMaiiCenter RidsNrr DoriDpr RidsRegioniMaiiCenter ADAMS Accession* ML14245A151 *via e-mail OFFICE LPLI-2/PM LPLI-2/LA SNPB/BC OGC LPLI-2/BC (A) LPLI-2/PM NAME JWhited A Baxter JDean* BMizuno RSchaaf JWhited (Jlamb for) w/comments (Jlamb/for)

DATE 09/02/14 09/03/14 08/29/14 9/16/14 9/17/14 9/17/14 OFFICIAL RECORD COPY