L-13-359, End-of-Life Moderator Temperature Coefficient Testing Revision

End-of-Life Moderator Temperature Coefficient Testing Revision

ML13319A882
Person / Time
Site:	Beaver Valley
Issue date:	11/14/2013
From:	Emily Larson FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-13-359, TAC ME9144, TAC ME9145
Download: ML13319A882 (9)

Text

FENOC' Beaver Valley Power Station P.O. Box 4 Shippingport, PA 15077 FirstEnergy Nuclear Operating Company Eric A. Larson 724-682-5234 Site Vice President Fax: 724-643-8069 November 14, 2013 L-13-359 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit Nos. 1 and 2 Docket No. 50-334, License No. DPR-66 Docket No. 50-412, License No. NPF-73 End-of-Life Moderator Temperature Coefficient Testing Revision (TAC Nos. ME9144 and ME9145)

By correspondence dated July 25, 2012 (Accession No. ML12208A309), as supplemented by correspondence dated June 1, 2013 (Accession No. ML13155A021),

FirstEnergy Nuclear Operating Company (FENOC) submitted to the Nuclear Regulatory Commission (NRC) a proposed amendment to the Beaver Valley Power Station, Unit Nos. 1 and 2, Technical Specifications (TSs). The proposed amendment would modify TS 3.1.3, "Moderator Temperature Coefficient (MTC)," to allow the normally required near-end-of-life MTC measurement to not be performed under certain conditions.

By correspondence dated October 21, 2013 (Accession No. ML13295A106), FENOC expressed its intention to submit a revision to the Technical Specification 5.6.3, "CORE OPERATING LIMITS REPORT (COLR)," list of analytical methods used to determine core operating limits that was provided in the amendment application. The proposed license amendment is hereby supplemented by replacing proposed changes to TS pages 5.6-2 and 5.6-3 provided with the July 25, 2012 letter with the changes shown in the attachments to this letter. Attachment 1 provides TS pages marked to show the proposed changes. Attachment 2 provides typed TS pages incorporating the proposed changes.

The information provided by this submittal does not invalidate the significant hazards consideration analysis submitted by FENOC in the July 25, 2012 correspondence.

There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - FENOC Fleet Licensing, at (330) 315-6810.

Beaver Valley Power Station, Unit Nos. 1 and 2 L-13-359 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on November .li. 2013.

Sincerely, Eric A Larson Attachments:

1 Technical Specification Pages Marked to Show Proposed Changes 2 Typed Technical Specification Pages Incorporating Proposed Changes cc: NRG Region I Administrator NRG Resident Inspector NRG Project Manager Director BRP/DEP Site BRP/DEP Representative

Attachment 1 L-13-359 Technical Specification Pages Marked to Show Proposed Changes The following pages are included in Attachment 1:

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 (F 0 (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-10216-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.,.J."

WCAP-13749-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 2-00TBD I +64TBD

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 --./TM System" Caldon, Inc. Engineering Report-160P, "Supplement to Topical Report ER-SOP: Basis for a Power Uprate with the LEFM --./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 SOM, 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 (P/T) 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 S, 2002.

Beaver Valley Units 1 and 2 5.6 - 3 Amendments ~TBD I 4-e+TBD

Attachment 2 L-13-359 Typed Technical Specification Pages Incorporating Proposed Changes The following pages are included in Attachment 2:

5.6-2 5.6-3 5.6-4

Reporting Requirements 5.6

!Revised Typed Pagesj 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 (Fa(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-8745-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-10216-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-13749-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 TBD I TBD

Reporting Requirements

!Revised Typed Pagesj 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 -'1 ' System" Caldon, Inc. Engineering Report-160P, "Supplement to Topical Report ER-BOP: Basis for a Power Uprate with the LEFM -'1 ' 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 B, 2002.

Beaver Valley Units 1 and 2 5.6- 3 Amendme~s TBD/TBD

Reporting Requirements No Change Proposed - Provided 5.6 for Information Only 5.6 Reporting Requirements 5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves. 11 The methodology listed in WCAP-14040-NP-A was used with two exceptions:

• ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1."

• ASME,Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1996 version.

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

5.6.5 Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.6 Steam Generator Tube Inspection Report 5.6.6.1 Unit 1 SG Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.1, Unit 1 Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service-induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h. The effective plugging percentage for all plugging in each SG.

Beaver Valley Units 1 and 2 5.6-4 Amendments TBD I TBD