Text

Issuance of Amendments Regarding Technical Specifications Revisions Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report (TAC Nos. ME9256 And.
	ML13249A386
Person / Time
Site:	Farley
Issue date:	10/02/2013
From:	Martin R; Plant Licensing Branch II
To:	Pierce C; Southern Nuclear Operating Co
	Barillas Martha, NRR/DORL 415-2760
References
	NL-12-0868, TAC ME9256, TAC ME9257
	Download: ML13249A386 (40)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555"()001 October 2, 2013 Mr. C. R. Pierce Regulatory Affairs Director Southern Nuclear Operating Company, Inc.

P. O. Box 1295/ Bin - 038 Birmingham, AL 35201-1295

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATIONS REVISIONS ASSOCIATED WITH THE LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM AND THE PRESSURE AND TEMPERATURE LIMITS REPORT (TAC NOS. ME9256 AND ME9257) (NL-12-0868)

Dear Mr. Pierce:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 193 to Renewed Facility Operating License No. NPF-2 and Amendment No. 189 to Renewed Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant, Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) Low Temperature Overpressure Protection (LTOP) System and the Pressure and Temperature Limits Report (PTLR) in response to your application dated August 15, 2012, as supplemented by letters dated March 14, and June 14, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML12229A521, ML130740897, and ML13165A368 respectively).

These amendments implement new 54 Effective Full Power Years (EFPY) pressure and temperature limit curves as a result of adopting the NRC-approved methodology in WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 4, May 2004.

The amendments revise the Technical Specifications associated with LTOP and PTLR.

The TS revisions, as proposed in the license amendment request, include the following:

1. a change to the definition of the PTLR in TS Section 1.1 (TS 1.1);

2. changes to PTLR administrative controls specified in TS 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limit Report (PTLR)," to incorporate a new PTLR methodology;

3. a change to the LTOP system applicability temperature;

4. the relocation of the LTOP system applicability temperature from the TS to the Farley 1 and 2 PTLRs; and

5. revisions to LTOP TS 3.4.12 limiting condition for operation, action statements, and surveillance requirements to incorporate an additional limitation on the maximum number of charging pumps capable of injecting into the RCS.

C. Pierce -2 A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

~~1e~~t Plant Licensing Branch 11-1 Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosures:

1. Amendment No. 193 to NPF-2

2. Amendment No. 189 to NPF-8

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 193 Renewed License No. NPF-2

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

A. The application for amendment to the Joseph M. Farley Nuclear Plant, Unit 1, Renewed Facility Operating Licenses No. I\lPF-2, filed by Southern Nuclear Operating Company, Inc. (the licensee), dated August 15, 2012, as supplemented by letters dated March 14, and June 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 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. NPF-2, is hereby amended to read as follows:

{2} Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 193, are hereby incorporated in the renewed facility operating license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: (k;tober 2, 2013

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 189 Renewed License No. NPF-8

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

A. The application for amendment to the Joseph M. Farley Nuclear Plant, Unit 2, Renewed Facility Operating Licenses No. NPF-8, filed by Southern Nuclear Operating Company, Inc. (the licensee), dated August 15,2012, as supplemented by letters dated March 14, and June 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 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 2

-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. NPF-8 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 189, are hereby incorporated in the renewed facility operating license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION

?1I)P~-

Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: Q:tober 2, 2013

ATTACHMENT TO LICENSE AMENDMENT NO. 193 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 AND ATTACHMENT TO LICENSE AMENDMENT NO. 189 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the following pages of the Renewed Facility Operating License and Appendix "A" Technical Specifications (TSs) with the enclosed pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License Page License Page NPF-2, page 4 NPF-2, page 4 NPF-8, page 3 NPF-8, page 3 TSs 1.1-5 1.1-5 3.4.6-1 3.4.6-1 3.4.7-1 3.4.7-1 3.4.10-1 3.4.10-1 3.4.12-1 3.4.12-1 3.4.12-2 3.4.12-2 3.4.12-4 3.4.12-4 5.6-5 5.6-5

-4 (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 193 , are hereby incorporated in the renewed license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the renewed license or within the operational restrictions indicated.

The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission.

a. Southern Nuclear shall not operate the reactor in Operational Modes 1 and 2 with less than three reactor coolant pumps in operation.

b. Deleted per Amendment 13

c. Deleted per Amendment 2

d. Deleted per Amendment 2

e. Deleted per Amendment 152 Deleted per Amendment 2

f. Deleted per Amendment 158

g. Southern Nuclear shall maintain a secondary water chemistry monitoring program to inhibit steam generator tube degradation.

This program shall include:

1) Identification of a sampling schedule for the critical parameters and control points for these parameters;

2) Identification of the procedures used to quantify parameters that are critical to control points;

3) Identification of process sampling points;

4) A procedure for the recording and management of data; Farley - Unit 1 Renewed License No. NPF-2 Amendment No. 193

-3 (2) Alabama Power Company, pursuant to Section 103 of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess but not operate the facility at the designated location in Houston County, Alabama in accordance with the procedures and limitations set forth in this renewed license.

(3) Southern Nuclear, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) Southern Nuclear, 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; (5) Southern Nuclear, 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; and (6) Southern Nuclear, 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 license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I 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 Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 2775 megawatts thermal.

(2) Technical Specifications The Technical SpeCifications contained in Appendix A, as revised through Amendment No. '189, are hereby incorporated in the renewed license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

Farley - Unit 2 Renewed License No. NPF-B Amendment No. 189

Definitions 1.1 1.1 Definitions PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE liMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates and the Low Temperature Overpressure Protection System applicability temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6.

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 2775 MWt.

REACTOR TRI P The RTS RESPONSE TIME shall be that time interval from SYSTEM (RTS) RESPONSE when the monitored parameter exceeds its RTS trip setpoint TIME at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn.

With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and (continued)

Farley Units 1 and 2 1.1-5 Amendment No. 193 (Unit 1)

Amendment No. 189 (Unit 2)

RCS Loops - MODE 4 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Loops-MODE 4 LCO 3.4.6 Two loops consisting of any combination of RCS loops and residual heat removal (RHR) loops shall be OPERABLE, and one loop shall be in operation.

I\JOTES-----------------------------------_.-

1. All reactor coolant pumps (RCPs) and RHR pumps may not be in operation for $ 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause reduction of the RCS boron concentration; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

2. No RCP shall be started with any RCS cold leg temperature ~ the Low Temperature Overpressure Protection (LTOP) System applicability temperature specified in the PTLR unless:

a. The secondary side water temperature of each steam generator (SG) is < 50"F above each of the RCS cold leg temperatures; or

b. The pressurizer water volume is less than 770 cubic feet (24%

of wide range, cold, pressurizer level indication).

APPLICABILITY: MODE 4.

ACTIONS CONDITION REQUIRED ACTION I COMPLETION TIME A. One required RCS loop A.1 Initiate action to restore a Immediately inoperable. second loop to OPERABLE status.

Two RHR loops inoperable.

Farley Units 1 and 2 3.4.6-1 Amendment No. 193 (Unit 1)

Amendment No. 189 (Unit 2)

RCS Loops - MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops-MODE 5, Loops Filled LCO 3.4.7 One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either:

a. One additional RHR loop shall be OPERABLE; or

b. The secondary side water level of at least two steam generators (SGs) shall be;::: 75% (wide range).

NOTES--------------------------------------

1. The RHR pump of the loop in operation may not be in operation for 52 hours6.018519e-4 days 0.0144 hours 8.597884e-5 weeks 1.9786e-5 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause reduction of the RCS boron concentration; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

2. One required RHR loop may be inoperable for 5 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided that the other RHR loop is OPERABLE and in operation.

3. No reactor coolant pump shall be started with one or more RCS cold leg temperatures 5 the Low Temperature Overpressure Protection (LTOP) System applicability temperature specified in the PTLR unless:

a. The secondary side water temperature of each SG is < 50°F above each of the ReS cold leg temperatu res; or

b. The pressurizer water volume is less than 770 cubic feet (24%

of wide range, cold, pressurizer level indication).

4. All RHR loops may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.

5. The number of operating Reactor Coolant Pumps is limited to one at RCS temperatures < 110°F with the exception that a second pump may be started for the purpose of maintaining continuous flow while taking the operating pump out of service.

Farley Units 1 and 2 3.4.7-1 Amendment No.193 (Unit 1)

Amendment No.189 (Unit 2)

Pressurizer Safety Valves 3.4,10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves LCO 3.4,10 Three pressurizer safety valves shall be OPERABLE with lift settings 2:: 2460 psig and::; 2510 psig.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 with all RCS cold leg temperatures> the Low Temperature Overpressure Protection (LTOP) System applicability temperature specified in the PTLR,

--- -------------- ---- ----NOTE ---------- ----------- ---- --------------------

The lift settings are not required to be within the LCO limits during MODES 3 and 4 for the purpose of setting the pressurizer safety valves under ambient (hot) conditions. This exception is allowed for 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months following entry into MODE 3 provided a preliminary cold setting was made prior to heatup, ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pressurizer safety A.1 Restore valve to 115 minutes valve inoperaole, OPE RABLE status, B. Required Action and B.1 Be in MODE 3, 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 4 with any 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months RCS cold leg Two or more pressurizer temperatures::; the safety valves inoperable. LTOP System applicability temperature specified in the PTLR, Farley Units 1 and 2 3.4.10-1 Amendment No.193 (Ulit 1)

Amendment No.189 (Ulit 2)

LTOP System 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (L TOP) System LCO 3.4.12 An LTOP System shall be OPERABLE with a maximum of one charging pump capable of injecting into the RCS when one or more of the RCS cold legs is ~ 180°F and a maximum of two charging pumps capable of injecting into the RCS when all of the RCS cold legs are> 180°F and the accumulators isolated and either a or b below.

a. Two residual heat removal (RHR) suction relief valves with setpoints

~ 450 psig.

b. The RCS depressurized and an RCS vent of 2 2.85 square inches.

NO TE S---------------------------------------------

1. With one or more of the RCS cold legs::; 180°F, two charging pumps may be capable of injecting into the RCS during pump swap operations for a period of no more than 15 minutes provided that the RCS is in a non-water solid condition and both RHR relief valves are OPERABLE or the RCS is vented via an opening of no less than 5.7 square inches in area.

2. Accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by the PIT limit curves provided in the PTLR.

APPLICABILITY: MODE 4 when the temperature of one or more RCS cold legs is ~ the LTOP System applicability temperature specified in the PTLR, MODE 5, MODE 6 when the reactor vessel head is on.

Farley Units 1 and 2 3.4.12-1 Amendment NO.193 (Unit 1)

Amendment No:189 (Unit 2)

LTOP System 3.4.12 ACTIONS

NOT E -------------- ---------- ----------- ----------------

LCO 3.0Ab is not applicable when entering MODE 4.

CONDITION REQUIRED ACTION COMPLETION TIME A. More than the maximum A.1 Initiate action to verify ~ Immediately required charging the maximum required pump(s) capable of charging pump(s) injecting into the capable of injecting into RCS. the RCS.

B. An accumulator not B.1 Isolate affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolated when the accumulator.

accumulator pressure is greater than or equal to the maximum RCS pressure for existing cold leg temperature allowed in the PTLR.

C. Required Action and C.1 Increase RCS cold leg 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion temperature to > the Time of Condition B not LTOP System met. applicability temperature specified in the PTLR.

OR C.2 Depressu rize affected 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months accumulator to less than the maximum RCS pressure for existing cold leg temperature allowed in the PTLR.

Farley Units 1 and 2 3.4.12-2 Amendment NO.193 (Unit 1)

Amendment No.189 (Unit 2)

LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of one charging pump is In accordance with capable of injecting into the RCS when one or more the Surveillance RCS cold legs is s 1 SO°F. Frequency Control Program SR 3.4.12.2 Verify a maximum of two charging pumps are In accordance with capable of injecting into the RCS when all RCS cold the Surveillance legs are> 1S0°F. Frequency Control i, Program SR 3.4.12.3 Verify each accumulator is isolated. In accordance with the Surveillance Frequency Control Program SR 3.4.12.4 Verify RHR suction isolation valves are open for each In accordance with required RHR suction relief valve. the Surveillance Frequency Control I Program SR 3.4.12.5 -----------------------------NOTE-------------------------------

Only required to be performed when complying with LCD 3.4.12.b.

Verify RCS vent 2:: 2.S5 square inches open. In accordance with the Surveillance Frequency Control Program SR 3.4.12.6 Verify each required RHR suction relief valve In accordance with setpoint. the Inservice Testing Program In accordance with the Surveillance Frequency Control Program Farley Units 1 and 2 3.4.12-4 Amendment NO.193 (Unit 1)

Amendment Nc.189 (Unit 2)

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

7. WCAP-11397-P-A "Revised Thermal Design Procedure," April 1989 (Methodology for LCO 2.1.1-Reactor Core Safety Limits, LCO 3.4.1 RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)

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.

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

a. The reactor coolant system pressure and temperature limits, including heatup and cooldown rates and the LTOP System applicability temperature, shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (PrT) Limits," and LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System."

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 WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.

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

5.6.7 EDG Failure Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures shall be reported within 30 days.

Reports on EDG failures shall include a description of the failures, underlying causes, and corrective actions taken per the Emergency Diesel Generator Reliability Monitoring Program.

(continued)

Farley Units 1 and 2 5.6-5 Amendment No. 193 (Unit 1)

Amendment No. 189 (Unit 2)

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 193 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO. 189 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY. INC.

JOSEPH M. FARLEY NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364

1.0 INTRODUCTION

By letter dated AUgust 15, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12229A521), Southern Nuclear Operating Company, Inc., (the licensee), submitted a license amendment request (LAR) for Joseph M. Farley Nuclear Plant, Units 1 and 2 (FNP Units 1 and 2), pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 90. The LAR proposes several changes to FNP Units 1 and 2 Technical Specifications (TSs) associated with the Pressure Temperature Limits Report (PTLR) methodology and Low Temperature Overpressure Protection (LTOP) system, and includes new PTLRs that would be valid through 54 effective full power years (EFPY) of facility operation, corresponding to the end of the 60-year extended license term. The proposed FNP Units 1 and 2 PTLRs for 54 EFPY were developed based on the NRC-approved methodology of WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004, which is the PTLR methodology referenced in the proposed revision to TS 5.6.6.

The supplements dated March 14, (ADAMS Accession No. ML130740897), and June 14, 2013 (ADAMS Accession No. ML13165A368), provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on October 2, 2012 (77 FR 60153).

2.0 REGULATORY EVALUATION

The proposed amendment involves changes to the TS requirements for the pressure and temperature (pn-) limits, and the LTOP system. The FNP Units 1 and 2 TS Limiting Condition of Operation (LCO) 3.4.3, "Reactor Coolant System (RCS) Pressure and Temperature (pn-) Limits,"

Enclosure 3

-2 contains the TS requirements for the PfT limits. The PfT limit curves preserve the integrity of the reactor vessel. The FNP Units 1 and 2 TS LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," contains the TS requirements for the LTOP system. The LTOP system controls RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the PfT limits of 10 CFR 50, Appendix G, "Fracture Toughness Requirements." The reactor vessel material is less tough at low temperatures than at normal operating temperature. Therefore, RCS pressure is maintained low at low temperatures and is increased only as temperature is increased.

The Commission's regulatory requirements related to the content of the TS are contained in 10 CFR, Section 50.36, "Technical Specifications." The TS requirements in 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety systems settings and control settings, (2) LCOs, (3) surveillance requirements (SRs), (4) design features, and (5) administrative controls.

The TS revisions, as proposed in the Ff\lP Units 1 and 2 LAR, include the following:

1. a change to the definition of the PTLR in TS Section 1.1;

2. changes to PTLR administrative controls specified in TS 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limit Report (PTLR)," to incorporate a new PTLR methodology;

3. a change to the LTOP system applicability temperature;

4. the relocation of the LTOP system applicability temperature from the TS to the Farley 1 and 2 PTLRs; and

5. revisions to LTOP TS 3.4.12 limiting condition for operation, action statements, and surveillance requirements, to incorporate an additional limitation on the maximum number of charging pumps capable of injecting into the RCS.

2.1 Requirements for PfT Limits 10 CFR 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events," requires reference temperatures be calculated for a reactor vessel material under any conditions. For the reactor vessel materials, the reference temperature must account for the effects of neutron radiation.

10 CFR 50, Appendix A, General Design Criterion (GDC) 14, "Reactor Coolant Pressure Boundary," requires the design, fabrication, erection, and testing of the reactor coolant pressure boundary so as to have an extremely low probability of abnormal leakage, o"f rapidly propagating failure, and of gross rupture.

GDC 30, "Quality of Reactor Coolant Pressure Boundary," requires, in part, that components comprising the reactor coolant pressure boundary be designed, fabricated, erected, and tested to the highest quality standards practical.

- 3 GDC 31, "Fracture Prevention of Reactor Coolant Pressure Boundary," states 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 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.

10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," requires that the PrT limits for an operating light-water nuclear reactor be at least as conservative as those that would be generated if the methods of the ASME Code,Section XI, Appendix G were used to generate the PrT limits. This appendix specifies requirements in order to protect the integrity of the RCPS in nuclear power plants. 10 CFR Part 50, Appendix G also requires that applicable surveillance data from reactor pressure vessel (RPV) material surveillance programs be incorporated into the calculations of plant-specific PrT limits and that the PrT 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 for neutron fluence levels greater than or equal to (~) 1 x 10 17 n/cm 2 (E >

1.0 MeV).

Table 1 of 10 CFR Part 50, Appendix G provides the criteria for meeting the PrT 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 the evaluation of neutron embrittlement for PrT limit curves is found in Regulatory Guide (RG) 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Revision (Rev.) 2. Additional guidance related to the staff's review of PrT limit curve submittals is found in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Chapter 5.3.2, "Pressure-Temperature Limits and Pressurized Thermal Shock."

The ASME Code,Section XI, Appendix G methodology for generating PrT limit curves is based upon the principles of linear elastic fracture mechanics (LEFM). The basic parameter of this methodology is the stress intensity factor, K" which is a function of the stress state in the component and flaw configuration. The ASME Code,Section XI, Appendix G requires a safety factor of 2.0 on stress intensities resulting from reactor pressure during normal operating conditions and a safety factor of 1.5 on these stress intensities for hydrostatic and pressure testing limits. The ASME Code,Section XI, Appendix G specifies that the PrT limits shall be generated by postulating a flaw with a depth that is equal to 1/4 of the RPV section thickness (1/4T) and a length equal to 1.5 times the RPV section thickness. The critical locations in the RPV section thickness for calculating heat-up and cool-down PrT limit curves are the 1/4T and 3/4T locations, which correspond to the maximum depth of the postulated inside surface and outside surface flaws, respectively.

The ASME Code,Section XI, Appendix G specifies PrT limit curve calculations are based, in part, on the nil-ductility reference temperature (RT NOT) for the material. The RT NOT is the critical parameter for determining the plane strain stress intensity factor (Kd for the material. 10 CFR Part 50, Appendix G requires that RT NOT values for materials in the RPV beltline region be adjusted to account for the effects of neutron irradiation. RG 1.99, Rev. 2 contains methodologies for calculating the adjusted RT NOT (ART) due to neutron irradiation. The ART is defined as the

-4 sum of the initial (unirradiated) reference temperature (initial RT NDT), the mean value of the shift in reference temperature caused by irradiation (llRT NDT), and a margin term. The llRT NDT is a product of a chemistry factor (CF) and a fluence factor. The CF is dependent upon the amount of copper (Cu) and nickel (Ni) in the material and may be determined from tables in RG 1.99, Rev. 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence at the postulated flaw depths described above. 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, Rev. 2, or surveillance data. The margin term is used to account for uncertainties in the values of the initial RT NDT, the Cu and Ni contents, the neutron fluence and the calculational procedures. RG 1.99, Rev. 2, describes the methodology to be used in calculating the margin term.

To satisfy the requirements of 10 CFR Part 50, Appendix G, methods for determining fast neutron fluence are necessary to estimate the fracture toughness of the RPV materials. 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," requires the installation of surveillance capsules, including material test specimens and flux dOSimeters, to provide data for material damage correlations as a function of neutron fluence. RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."

describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence with respect to meeting the regulatory requirements discussed above.

The licensee refers in Section 3.0, "Reactor Vessel Material Surveillance Program," of enclosure 4 to the licensee's LAR, to neutron 'l'luence calculations performed in support of withdrawal and analysis of Surveillance Capsules Z and V. The results are discussed in WCAP-16964-NP, "Analysis of Capsule Z from the Southern Nuclear Operating Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program" (ML082890113) and WCAP-16918-NP, "Analysis of Capsule V from Southern Nuclear Operating Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program" (ML081130652). Chapters 6 of WCAP-16964-NP and WCAP-16918-NP, respectively, describe discrete ordinates transport analyses performed to determine the neutron fluence for the RPV and surveillance specimens.

2.2 Criteria for PTLRs On January 31, 1996, the staff issued Generic Letter (GL) 96-03 to inform licensees that they may request a license amendment to relocate the PIT limit curves and LTOP system limits from the TS LCOs to a PTLR or other licensee-controlled document that would be governed by the TS administrative controls. In order to permit relocation of the PIT limits and LTOP system limits, GL 96-03 indicated that licensees would need to generate their PIT limits and LTOP system limits in accordance with an NRC-approved methodology and that the methodology used to generate the PIT limits and LTOP system limits would need to comply with the requirements of 10 CFR Part 50, Appendices G and H. Furthermore, the methodology used to generate the PIT limits and LTOP system limits would need to be incorporated by reference in the administrative controls section of the TS. The GL 96-03 also indicated that the TS administrative controls section for the PTLR would need to reference the PTLR methodology approved by the NRC and that the PTLR be defined in Section 1.0 of the TS. Attachment 1 to GL 96-03 provided a list of the criteria that the proposed PTLR methodology and plant-specific PTLR license amendments would be required to meet.

- 5 TSTF-233-A amended the Standard Technical Specification (STS) to provide the option for relocating the LTOP system arming temperature (or LTOP system applicability temperature) from the TS to the PTLR. The arming temperature is the temperature below which STS LCO 3.4.12, LTOP is applicable in MODE 4, "Hot Shutdown." This temperature is a plant-specific requirement that is established based on the limiting ART for the RPV beltline region, and it is periodically revised and updated as required based on the extent of RPV neutron ernbrittlement. The relocation of the arming temperature from the TS to the PTLR is consistent with the relocation of the prr limits to the PTLR since both are dependent on neutron fluence. The methodology for determining the LTOP arming temperature would continue to be controlled by the TS.

TSTF-233-A was approved by the NRC in a letter dated July 16, 1998, and it has since been incorporated into NU REG-1431.

TSTF-419-A amended the STS to: (1) delete references to the TS LCO for the prr limits and LTOP system limits in the TS definition of the PTLR, and (2) revised the administrative controls for the PTLR in STS Section 5.6.6 (STS 5.6.6) to allow NRC-approved topical reports for PTLR methodologies to be identified by number and title. TSTF-419-A did not change the requirements associated with the NRC review and approval of the PTLR methodology or the requirement to operate within the limits specified in the PTLR. Any changes to a PTLR methodology that had not been approved by the NRC staff would continue to require staff review and approval pursuant to the LAR provisions of 10 CFR 50.90. TSTF-419-A was approved by the !\IRC in a letter dated March 21,2002, and it has since been incorporated into the STS for Westinghouse plants, NUREG-1431, "Standard Technical Specifications - Westinghouse Plants," Rev. 4, April 2012.

3.0 TECHNICAL EVALUATION

The guidance provided in RG 1.190 indicates that the following comprises an acceptable fluence calculation: 1) a fluence calculation performed using an acceptable methodology, 2} analytic uncertainty analysis identifying possible sources of uncertainty, 3} benchmark comparison to approved results of a test facility, and 4} plant-specific qualification by comparison to measured fluence values.

The fast neutron exposure parameters were determined for Southern Nuclear Operating Company by Westinghouse, using the methods discussed in Rev. 4 of WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating Systems Setpoints and RCS Heatup and Cooldown Limit Curves" (ML050120209). As noted by the safety evaluation, "Final Safety Evaluation for Topical Report WCAP-14040, Revision 3, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cool down Limit Curves,"

dated February 27,2004 (ML040620297), the staff determined that this methodology is generically acceptable for the reference in licensing actions.

As described in WCAP-16964-NP and WCAP-16918-NP, the licensee is using the two-dimensional discrete ordinates code, "Two-Dimensional Discrete Ordinates Transport Code System (DORT)," dated August 1993, with the BUGLE-96 cross section library in ENDP/B-VI:

"Evaluated Nuclear Data Library for Nuclear Science and Technology," dated December 1996, which was derived from the Evaluated Nuclear Data File (ENDF/B-VI) in "Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," dated March 1996.

- 6 Approximations include a P5 Legendre expansion for anisotropic scattering and a S16 order of angular quadrature. These approximations are of a higher order than the P3 expansion and S8 quadrature suggested in RG 1.190. Space and energy dependent core power (neutron source) distributions and associated core parameters are treated on a fuel cycle specific basis. Three dimensional flux solutions are constructed using a synthesis of azimuthal, axial, and radial flux.

Source distributions include cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions, which are used to develop spatial and energy dependent core source distributions that are averaged over each fuel cycle. This method accounts for source energy spectral effects by using an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of each fuel assembly. The neutron transport calculations, as described above, are performed in a manner consistent with the guidance set forth in RG 1.190.

Also described in WCAP-16964-NP and WCAP-16918-NP, the licensee performed an analytiC uncertainty analysis by combining the uncertainties associated with the individual components of the transport calculations in square-root-of-the-sum-of-the-squares. The calculations were compared with the benchmark measurements from the Poolside Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL), and with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment. As noted in Regulatory Position 1.4.2.2 of RG 1.190, these constitute acceptable test facilities.

The licensee stated that the proposed revision to TS 5.6.6 would change the NRC-approved PTLR methodology currently cited in TS 5.6.6b to the latest NRC-approved methodology in WCAP-14040-A, Rev. 4. The licensee stated that any future changes to the Farley 1 and 2 PfT limits contained in the PTLR will be determined in accordance with this NRC-approved methodology, consistent with the proposed revision to TS 5.6.6b:

PTLRs for Farley 1 and 2 were provided in an enclosure to the licensee's LAR submittal, as well as Westinghouse technical reports documenting the methods and calculations employed for developing the PTLRs. The proposed PTLRs contain PfT limit curves that would be valid through 54 EFPY, corresponding to the end of the 60-year extended license term. The PTLRs specify that they were prepared in accordance with the requirements of TS 5.6.6, and that revisions to the PTLRs shall be provided to the NRC upon issuance. In accordance with the TS change described in the LAR, the PTLRs now include the revised LTOP system applicability temperature (s 275 OF for both units).

The licensee stated that revised LTOP applicability temperature for Farley 1 and 2 was calculated in accordance with ASME Code Case N-641. The licensee noted that Section 3.4 of WCAP-14040-A, Rev. 4 also incorporates the provisions of ASME Code Case N-641 for calculating the LTOP applicability temperature. Thus, the licensee proposes to revise LTOP applicability temperature to be consistent with Section 3.4 of the WCAP for a 3-loop plant.

Section 3.4 of the LAR describes the licensee's proposal to revise TS 3.4.12 to incorporate a new limitation specifying that when all RCS cold leg temperatures> 180 OF, but still below the MODE 4 LTOP applicability temperature, a maximum of two charging pumps shall be capable of injecting into the RCS. The licensee stated that the current TS 3.4.12 LCO only specifies that a maximum of one charging pump shall be capable of injecting into the RCS when one or more RCS cold leg temperatures is s 180 OF; however, the current LCO does not specify any limit on the maximum number of charging pumps capable of RCS injection when all RCS cold leg temperatures are>

- 7 180 of, but still below the MODE 4 LTOP applicability temperature (s 325 of in the current TS, and s 275 of, as revised per the LAR) for the LCO.

The licensee noted that the LTOP system design bases, as described in the Updated Final Safety Analysis Report (UFSAR), Section 5.2.2.4.3, assume that the worst-case mass input event assumed for the design of the LTOP pressure relief capability is the inadvertent operation of three high-head safety injection pumps, which are the same as the RCS charging pumps. Accordingly, the worst-case mass injection is conservatively limited by TS 3.4.12 to the start of a single charging pump at RCS temperatures s 180 of. Therefore, according to the licensee, the proposed change provides an additional TS requirement for ensuring that LTOP design bases are satisfied when all RCS temperatures> 180 of. TS 3.4.12 action statements and surveillance requirements were similarly revised to re'flect the new limitation on RCS charging pump capability.

3.2 Staff Evaluation 3.2.1 Evaluation of the Proposed TS Revisions (1) Revision to PTLR Definition in TS 1.1 Both the TS 1.1 definition of the PTLR and the TS 5.6.6 administrative controls currently identify LCO 3.4.3 as the applicable requirement for plant operation within the PIT limit curves in the PTLR. This change deletes the citation of LCO 3.4.3 in the PTLR definition, while leaving the citation of LCO 3.4.3 in TS 5.6.6. The licensee noted that the proposed change is administrative in nature and serves to eliminate duplication of information in the TS. As discussed in Section 2.2 of this safety evaluation (SE), the subject change to the PTLR definition is identified in item (1) of TSTF-419-A, which is now incorporated into NUREG-1431. Since this change is reflected in NUREG-1431 and will not affect the technical aspects of the PIT limit curves and LTOP system limits, the staff finds the licensee's proposed revision to the PTLR definition acceptable.

In addition, the first sentence of the PTLR definition is supplemented to state that the PTLR provides the LTOP system applicability temperature (in addition to the PIT limits and heat-up/cool-down rates previously included in the definition). This change is necessary to appropriately reflect the relocation of the LTOP system applicability temperature from the TS to the PTLR. Therefore it is acceptable.

(2) Revision to PTLR methodology referenced in TS 5.6.6b of GL 96-03 identifies seven technical criteria that must be satisfied for PTLR methodologies in order for the methodologies to be approved for incorporation by reference in the administrative controls sections of licensee's TS. The NRC staff's approval letter and SE for WCAP-14040-A, Rev. 4 are included as an attachment to the topical report. The SE describes how the seven technical criteria for PTLR methodologies are satisfied by WCAP-14040-A, Rev. 4.

The revision to Farley 1 and 2 TS 5.6.6b to reference WCAP-14040-A, Rev. 4 is consistent with the acceptance criteria for plant TS identified in Attachment 1 of GL 96-03, which state that TS administrative controls must reference an NRC-approved PTLR methodology, including the staff approval document. Therefore, the TS 5.6.6b revision is acceptable.

-8 (3) Revision to L TOP System Applicability Temperature TS 3.4.12 currently specifies that the LTOP system shall be operable in MODE 4 when the temperature of one or more RCS cold legs is s 325 of. This applicability temperature for the LTOP system is revised to s 275 of, based on the licensee's use of the LTOP system enable temperature methodology identified in Section 3.4 of WCAP-14040-A, Rev. 4 and ASME Code Case N-641. The staff confirmed that the revised LTOP system applicability temperature meets the criteria for these temperatures in Section 3.4 of WCAP-14040-A and ASME Code Case 1\J-641 ,

paragraph 2215.2(a) for postulated 1/4T axial flaws. ASME Code Case N-641 is listed as acceptable without conditions in RG 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Rev. 16. Therefore, the licensee's proposed revision to the LTOP system applicability temperature is acceptable.

(4) Relocation of L TOP System Applicability Temperature from the TS to the PTLRs The LTOP system applicability temperature is currently incorporated in several TS sections, as identified above. As such, this temperature is used to identify the applicability of LCO 3.4.12 for the LTOP system and LCO 3.4.10 for the pressurizer safety valves, as well as the low temperature limit for starting an RCS pump in LCO 3.4.6 and LCO 3.4.7. As discussed in Section 2.2 of this SE, the licensee's proposal to relocate the LTOP system applicability temperature from these TS sections to the PTLRs is consistent with TSTF-233-A, which is now incorporated into NUREG-1431. Since this change is reflected in NUREG-1431 the staff finds the relocation of the LTOP system applicability temperature acceptable.

(5) Revision to TS 3.4. 12 to Address a Maximum of Two Charging Pumps Capable of Injecting Into the RCS As discussed above, the current TS 3.4.12 LCO limits the number of charging pumps capable of injecting into the RCS for the condition when one or more RCS cold leg temperatures is s 180 of.

The LCO is silent regarding the maximum number of charging pumps capable of RCS injection when all RCS cold leg temperatures are> 180 OF, but still below the MODE 4 LTOP applicability temperature (s 325 OF in the current TS, and s 275 OF, as revised per the LAR) for the LCO. Even though the LTOP system design bases assume that the worst-case mass input event is the inadvertent operation of three high pressure safety injection pumps, the current TS LCO does not limit the number of RCS charging pumps capable of injection when all cold leg temperature are>

180 OF but below the LTOP applicability temperature. Therefore, the proposed change provides the necessary operability requirement for ensuring that the LTOP system does not exceed its design basis when all RCS cold leg temperatures are> 180 OF, but still below the LTOP applicability temperature. The proposed change serves exclusively to provide additional operating restrictions for the RCS to ensure LTOP system functionality over all temperature ranges below the LTOP applicability temperature, and does not affect the actual LTOP system pressure relief capability. Therefore the staff finds the subject revisions to TS 3.4.12 acceptable.

3.2.2 Evaluation of the Farley 1 and 2 PTLRs for 54 EFPY of G L 96-03 identifies seven technical criteria that must be satisfied for plant-specific PTLRs. The proposed Farley 1 and 2 PTLRs satisfy the GL 96-03 criteria as follows:

-9 (1) The PTLR should provide the values of neutron fluence that are used in the ART calculation for the RPV beltline materials. The 54 EFPY neutron fluence values for the RPV beltline materials are provided in Section 5.0, Table 5-4 of the Farley 1 and 2 PTLRs. Therefore, PTLR criterion (1) is satisfied. The staff's evaluation of the acceptability of the neutron fluence calculations for generating the PfT limit curves is discussed below in Section 3.2.2.1 of this SE.

(2) The PTLR should provide the surveillance capsule withdrawal schedule, or reference by title and number the documents in which the schedule is located. The PTLR must also reference the surveillance capsule reports by title and number if ARTs are calculated using surveillance data.

The surveillance capsule withdrawal schedules are provided in Section 3.0, Table 3-1 of the PTLRs. References for the surveillance capsule reports are provided in Section 6.0 of the PTLRs.

The staff noted that all surveillance capsules have been withdrawn from the Farley 1 and 2 RPVs, all capsule analysis reports have been provided to the NRC, and the capsule withdrawals satisfy the requirements of 10 CFR Part 50, Appendix H for up two 20-year renewal terms (80 years of facility operation). Therefore, PTLR criterion (2) is satisfied.

(3) If LTOP system limits are relocated to the PTLR, the PTLR should provide the setpoint curves or parameters. In accordance with the TS revisions evaluated above, the licensee has elected to relocate only the LTOP system applicability temperature, as it is a time-dependent parameter that is calculated based on the limiting ART for the RPV beltline region. The LTOP system applicability temperature is now specified in Section 2.3 of the PTLRs. Since GL 96-03 provides the option for licensees to relocate LTOP system limits, and the relocation of only the arming temperature (equivalent to the applicability temperature at Farley) is consistent with NUREG-1431, PTLR criterion (3) is satisfied.

(4) The PTLR should identify both the limiting ART values and limiting RPV beltline materials at the 1/4T and 3/4T locations. The criterion also states that pressurized water reactors (PWRs) should identify the reference temperature for pressurized thermal shock (RTPTS) value in accordance with 10 CFR 50.61. The 54 EFPY ART values for all RPV beltline materials are provided in Section 5.0, Table 5-5 of the PTLRs. The ART calculations for the limiting beltline shell materials are provided in Section 5.0, Table 5-6 of the PTLRs. 54 EFPY RT PTS values for all RPV beltline materials are provided in Section 5.0, Table 5-7 of the PTLRs. Therefore, PTLR criterion (4) is satisfied.

(5) The PTLR should provide the PfT limit curves for heatup, cooldown, criticality, and pressure testing conditions. The PfT limit curves and tabulated PfT limit data points for these conditions are provided in Section 2.0, Figures 2-1 and 2-2, and Tables 2-1 and 2-2 of the PTLRs. Therefore, PTLR criterion (5) is satisfied.

(6) The PTLR should identify the minimum temperatures on the PfT limit curves, such as the minimum boltup temperature and the hydrotest temperature. The minimum temperature criteria, including the minimum bolt up temperature and the minimum temperature for criticality are identified in the PfT limit curves provided in Section 2.0, Figures 2-1 and 2-2 of the PTLRs.

Therefore, PTLR criterion (6) is satisfied.

- 10 (7) The PTLR should evaluate RPV surveillance data to determine if they meet the credibility criteria of RG 1.99, Rev. 2 and provide the results of the credibility assessment. The results of the surveillance data credibility assessment are provided in Section 4.0 of the PTLRs. Therefore, PTLR criterion (7) is satisfied.

Based on its evaluation of the seven PTLR technical criteria, the staff finds that the proposed Farley 1 and 2 PTLRs are consistent with GL 96-03. To determine whether the actual 54 EFPY PIT limit curves meet the requirements of 10 CFR Part 50, Appendix G, the staff performed an independent evaluation of the curves, which is discussed below.

3.2.2.1 Evaluation of the 54 EFPY PIT Limit Curves Limiting ART Values and PIT Limit Curves-RPV Beltline Shell Region The 54 EFPY PIT limit curves provided in the Farley 1 and 2 PTLRs were generated based on the calculations documented in WCAP-17122-NP, Rev. 0, "J. M. Farley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," October 2009 for Unit 1, and WCAP-17123-NP, Rev. 1, J. M.

Farley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," July 2011 for Unit 2.

These reports were provided in enclosures to the August 15, 2012 LAR. The PIT limit curves were generated based on the analysis of the limiting beltline shell material (Lower Shell Plate B6919-1 for Unit 1 and Intermediate Shell Plate B7212-1 for Unit 2). The staff independently verified that the ART values for these materials at the 1/4T and 3/4T locations were correctly calculated using the procedures in RG 1.99, Rev. 2, Position 2.1 (procedures for ART calculations when surveillance data is available) using valid input parameters, as summarized in Table 1 below:

- 11 l

T a bl e 1 L'Imllng Be ItI'me SheII ART Cacua r for 54 EFPY I I f Ions at the 1/4T an d 3/4T Localons Material Location Initial I Fluence Fluence CF(l) ilRTNDT I Margin(') ART RT NDT i at RPV at (OF) (OF) (OF) (OF)

(OF) Inside Location Surface (n/cm2)

(n/cm2) I I Farley, Unit 1

Lower Shell !

I

Plate 1/4T 15 5.810 x 3.622 x 106.7 142.4 34 191 I 86919-1 10 19 1019 (surveillance 1

data)(l) I i Lower Shell

Plate 3/4T 15 5.810 x 1.408 x 106.7 116.8 34 166 19 1019 (surveillance

86919-1 10 I data)(l)

Farley, Unit 2 Intermediate Shell Plate 1/4T -10 5.760 x 3.591 x 144.6 192.7 17 200 87212-1 19 19 (surveillance 10 10 data)(l)

Intermediate Shell Plate 3/4T -10 5.760 x 1.396 x 144.6 158.1 17 165 19 87212-1 1019 10 (surveillance dEita)(1)

(1) The CF was determined based on non-credible surveillance data (I.e., data that does not meet the five credibility criteria in Section 8 of RG 1.99, Rev. 2) for Unit 1 and credible surveillance data for Unit 2, in accordance with RG 1.99, Rev. 2, Position 2.1. The Unit 1 ART based on Position 2.1 non-credible surveillance data is more conservative than the ART based on Position 1.1 and is therefore acceptable, as discussed below.

(2) The margin term for each ART calculation was based on the establishment of an initial material property uncertainty (0'1) and shift in material property uncertainty (aLl), consistent with the guidance in RG 1.99, Rev. 2.

The staff verified that the initial RT NDT values and the Cu and Ni contents for the RPV beltline shell materials are consistent with those tabulated in the UFSAR and in previous docketed correspondence. It should be noted that for Unit 1, the licensee elected to use non-credible surveillance data for its limiting material ART calculation. The licensee PTLR for Unit 1 shows that the use of the CF values based on Position 1.1 from RG 1.99, Rev. 2, when surveillance data is not available results in lower (Le. less conservative) ARTs than when surveillance data is available and values are calculated using Position 2.1 of RG 1.99, Rev. 2. The staff finds the licensee's approach acceptable since the licensee's use of non-credible surveillance data per RG 1.99, Rev. 2, Position 2.1 results in higher ARTs than Position 1.1 and therefore, more bounding PIT limits for this material.

- 12 The staff confirmed that the ARTs are based on 54 EFPY neutron fluence projections that were calculated using the methods described in WCAP-14040-A, Rev. 4. As documented in the NRC SE approving WCAP-14040-A, Rev. 4, these neutron fluence calculational methods adhere to the guidance contained in RG 1.190. Based on this consideration, the NRC staff determined that the 54 EFPY neutron fluence values are acceptable.

Using the above ARTs for the limiting beltline shell materials, the staff performed a set of confirmatory calculations for verifying that the licensee's 54 EFPY PIT limits for heat-up and cool-down operations, and for pressure test conditions, are at least as conservative as those that would be generated using the methods and acceptance criteria of the ASME Code,Section XI, Appendix G, as required by 10 CFR Part 50, Appendix G. Based on our confirmatory calculations, the staff determined that the PIT limits for the limiting beltline shell materials, Lower Shell Plate B6919-1 for Unit 1 and Intermediate Shell Plate B7212-1 for Unit 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. Therefore, the staff finds them acceptable.

The staff confirmed that the PIT limit curves for both units incorporate minimum temperature criteria that meet the requirements of Table 1 of 10 CFR Part 50, Appendix G. For core not critical conditions and pressure test conditions, minimum temperature criteria were correctly determined, based on the limiting RTNDT for the closure flange region, as required by Table 1 of 10 CFR Part 50, Appendix G. For core critical conditions, the minimum temperature is set equal to the minimum hydrostatic test temperature, which is more bounding than the closure flange limit for core critical conditions, as required by Table 1 of 10 CFR Part 50, Appendix G.

Based on the above evaluation, the staff finds that the licensee's PIT limit curves for the limiting beltline shell materials (Lower Shell Plate B6919-1 for Unit 1 and Intermediate Shell Plate B7212-1 for Unit 2) are acceptable for 54 EFPY.

Consideration of Ferritic RCPB Components Outside of the RPV Beltline She" Region Regarding ferritic RCPB components that are not part of the RPV beltline shell region, 10 CFR Part 50, Appendix G, and Paragraph IV.A states the following:

The pressure-retaining components of the reactor coolant pressure boundary that are made of ferritic materials must meet the requirements of the [ASME Code, 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 PIT limits be developed for the ferritic materials in the RPV beltline, as we" as ferritic materials not in the RPV beltline. Further, 10 CFR Part 50, Appendix G requires that all ferritic RCPB components must meet the applicable ASME Code,Section III requirements. For replacement components, the relevant ASME Code,Section III requirements that will affect the PIT limits are the lowest service ternperature requirement of NB-2332(b) for piping, pumps, and valves, and the fracture toughness requirements of NB-3211 (d) for vessels.

- 13 The staff noted that prr limit calculations for ferritic RCPB components that are not RPV beltline shell materials may define prr curves that are more limiting than those calculated for the RPV beltline shell materials. This may be due to the following factors:

Factor 1: RPV nozzles, penetrations, and other discontinuities have complex geometries that may exhibit significantly higher stresses than those for the RPV beltline shell region. These higher stresses can potentially result in more restrictive prr limits, even if the RTNDT for these components is not as high as that of RPV beltline shell materials that have simpler geometries.

Factor 2: Ferritic components of replacement RCPB vessels, such as the steam generators (SGs), may potentially define more restrictive prr limits than those for the RPV.

Factor 3: Ferritic components of replacement RCPB piping, pumps, and valves may have initial RT NDT values that may define a more restrictive lowest service temperature in the prr limits than those for the RPV materials.

Therefore, in RAI-1 , the staff requested that the licensee describe how the proposed 54 EFPY prr limit curves, and the methodology used to develop these curves, considered all RPV materials (beltline and non-beltline) and the replacement ferritic RCPB materials, consistent with the requirements of 10 CFR Part 50, Appendix G.

In its March 14,2013, non-proprietary response (ADAMS Accession No.(ML13074A801) to RAI-1, the licensee stated that the 54 EFPY prr limit curves were developed using the generic methodology described in WCAP-14040-A, Rev. 4. The licensee noted that WCAP-14040-A, Rev.

4 did not consider the RPV inlet and outlet nozzles, which are the most highly stressed ferritic components in the RPV. Therefore, to demonstrate that the 54 EFPY prr limit curves are bounding for the entire RPV, the licensee developed component-specific prr limit curves for the RPV inlet and outlet nozzles.

Evaluation of Response to RAI-1, Factor 1- RPV Inlet/Outlet Nozzle PIT Limit Curves Tables 1 and 2 of the RAI response include 54 EFPY ART calculations for the Farley 1 and 2 inlet and outlet nozzle forgings. There are three inlet nozzles and three outlet nozzles, based on the Westinghouse three-loop design. Several of the critical input parameters used for the nozzle ART calculations - notably, all of the initial RTNDT values, the generic Cu content, and the neutron fluence value for one of the nozzles - differ from those provided in the PTLRs. The staff's evaluation of the changes to these parameters is discussed below.

Table 5-3 of the 54 EFPY PTLRs, and Tables 5.2-24 and 5.2-25 of the Farley 1 and 2 FSAR, list initial RTNDT values that were determined based on NUREG-0800, Branch Technical Position MTEB 5-2, "Fracture Toughness Requirements." However, Tables 1 and 2 of the licensee's RAI response list initial RT NDT values for the Farley 1 and 2 inlet and outlet nozzles that are lower (Le.

less conservative) than those listed in the PTLR submittals and the FSAR. The revised initial RT NDT values were used as material property inputs for generating the inlet and outlet nozzle prr limit curves for Farley 1 and 2. The RAI response states that the initial RTNDT values for the inlet and outlet nozzles have been updated using the proprietary Boiling Water Reactor Vessel and Internals Project (BWRVIP) technical report, BWRVIP-173-A, "BWRVIP-173-A: BWR Vessel and Internals Project, Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials," July

- 14 2011 (non-proprietary version of BWRVIP-173-A available under ADAMS Accession No. ML12083A268).

To ensure that the BWRVIP-173-A methodology was correctly implemented, in a follow-up RAI (RAI-2), the staff requested that the licensee provide the calculations employed for deriving the new initial RT NOT values for the Farley 1 and 2 inlet and outlet nozzles.

In its June 14,2013 response (ADAMS Accession No. ML13165A368) to RAI-2, the licensee provided the details of the initial RT NOT calculation for each of the twelve nozzle forging materials for Farley Units 1 and 2. Individual plots of Charpy impact energy (ft-Ibs) versus temperature (OF),

along with hyperbolic tangent curve fits, were included for every nozzle material.

The staff reviewed these calculations and finds that they are consistent with the methods described in BWRVIP-173-A. Therefore, the staff finds that the updated initial RT NOT values listed in Table 1 and 2 of the RAI-1 response are acceptable for implementation in the PrT limit curve analysis of the inlet and outlet nozzles. Thus, RAI-2 is resolved.

The neutron fluence values listed in Tables 1 and 2 of the RAI-1 response are the same as those listed in Table 5-4 of the PTLRs, with the exception of Unit 1 Inlet Nozzle B6917-2. The licensee stated that for the PTLRs, the nozzle neutron fluence values were conservatively calculated at the lowest extent of the nozzle to upper shell welds, which is located closer to the active core region than the nozzle forgings. However, for its RAI response, the licensee elected to recalculate the neutron fluence used for determining the Unit 1 Inlet Nozzle B6917-2 PrT limits by conSidering the more realistic elevation of the postulated 1/4T nozzle corner flaw in the nozzle forging. The licensee's decision to recalculate this particular nozzle neutron fluence was based on its consideration of the impact of the revised initial RT NOT values discussed above. Specifically, the licensee noted that the revised initial RT NOT values listed in Tables 1 and 2 are less than zero for all inlet and outlet nozzles, with the exception of Unit 1 Inlet Nozzle B6917-2, which has a revised initial RT NOT value of 29 OF. The licensee stated that the Significantly higher initial RT NOT value for this nozzle warranted a more realistic treatment of the nozzle with regard to the neutron fluence calculation based on the actual elevation of the postulated flaw in the nozzle forging. The licensee stated that the elevation of the postulated flaw was determined based on the RPV design drawings. Furthermore, the licensee stated that the 54 EFPY nozzle ARTs listed in Table 1 and 2 of the RAI-1 response were conservatively calculated at the RPV clad/base metal interface rather than the 1/4T location.

Since the elevation of the postulated flaw for the Unit 1 Inlet Nozzle B6917-2 revised neutron fluence was determined based on RPV design drawings, and the fluence used for all nozzle ART calculations was at the RPV clad/base metal interface, the staff finds the recalculated neutron fluence for this nozzle to be acceptable. As discussed previously, the 54 EFPY neutron fluence values for all RPV beltline materials, including the inlet/outlet nozzles, were calculated using the NRC-approved methods in WCAP-14040-A, Rev. 4. Therefore, they are acceptable.

As shown in Tables 1 and 2 of the RAI-1 response, the licensee updated the generic Cu content for all of the nozzle forgings from that found in the plant-specific PTLRs. The licensee noted that the certified material test reports (CMTRs) for these SA-508, Class 2 nozzle forgings do not contain measured values of Cu content because, at the time of fabrication, measured values for the specific heat (I.e., production run) of material were not required for SA-508, Class 2 forging

- 15 material. The licensee stated that the PTLR values were taken from ORNL Report, ORNUTM-2006/530, "A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels," November 2007 (ADAMS Accession No. MLOB1000630).

The licensee stated that the updated Cu content value was obtained from Section 4 of BWRVIP-173-A. The staff verified that the updated Cu content value in Tables 1 and 2 of the RAI-1 response corresponds to the "best-estimate" generic value recommended by BWRVIP-173-A when no measured values for the particular heat of material are available. The staff finds the licensee's application of the BWRVIP-173-A methods acceptable for determining the updated generic Cu content because the recommendations of the report are based on analyses of an industry-wide database of SA-SOB, Class 2 forging material chemistries from both BWRs and PWRs.

The licensee noted in its RAI-1 response that the Ni content values are unchanged from those listed in the PTLRs. The licensee also stated that these Ni content values were obtained directly from material heat-specific CMTRs. The staff confirmed that these values are the same as those listed in Table 5.2-24 and 5.2-25 of the FSAR. The staff also verified that the margin term values were correctly determined based on the RG 1.99, Rev. 2 procedures. Therefore, the Ni content values and margin terms are acceptable.

The staff verified that the 54 EFPY ART values for the nozzles listed in Tables 1 and 2 of the RAI-1 response were correctly calculated using the procedures in Position 1.1 of RG 1.99, Rev. 2.

This is the appropriate method given that the nozzle materials are not represented in the RPV surveillance program. Therefore, these 54 EFPY ART values are acceptable for use in generating PIT limit curves for the nozzles.

The licensee selected the most limiting ART value from each set of three inlet and outlet nozzles for each unit to generate a bounding set of nozzle PIT limit curves for the 100 OF per hour cool-down transient. The bounding 54 EFPY ART values for these nozzles at the RPV clad/base metal interface, along with the valid input parameters, are summarized in Table 2 below:

- 16 Ta bl e 2 - Ine Material I tlO utl et Initial RTNDT (OF) 1\1ozze Fluence at RPV Clad/Base CF(~)

(OF) liRTN DT (OF)

Margin (OF)

ART (OF) l I Boun d"mg 54 EFPY ARTs at the RPV Clad/Base Metallnterface(1)

Metal

~t Interface(1)

(n/cITJ 2 )

I Nozzle ,29. /7.78 x 10 16 i

I' 141.0 I 13.0 i

113.0 i 55.1

~ ---+-----+'----+-- ----+-----+-------\

iO~~

Nozzle -23 2.31 x 10 17 1

140.3 26 26.4 29.8 B6916-3 1 .4 i

i i Inlet Nozzle -55 ' 4.49 x 10 17 137.9 138.1 ' 34 i

16.6

, B7218-1 I

Outlet I

i Nozzle -43 3.04 x 1017 138.2 30 6 30 6 18.3 B7217-3 1 . 1 . i

~

(1) As discussed above, the nozzle 54 EFPY ARTs were conservatively determined at the RPV clad/base metal interface.

(2) The CF was determined using Table 2 in RG 1.99, Rev. 2, in accordance with Position 1.1 of the RG.

The licensee generated PIT limit curves for the inlet and outlet nozzle forgings using the 54 EFPY ART values listed in Table 2, based on a 100 OF per hour cool-down rate. The inlet and outlet nozzle PIT limit curves, along with the corresponding 100 of per hour cool-down curves for the limiting beltline shell materials, are provided in Figures 1 through 4 of the RAI-1 response. The licensee stated that the methods used for determining the applied stress intensity factors due to pressure loading (KIP) and thermal gradients (KIT) are consistent with those published in the ORNL study, ORNLlTM-2010/246, "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles - Revision 1," June 2012 (ADAMS Accession No. ML12181A162).

- 17 The licensee stated that the through-wall stress distributions in the nozzle inside corner region were fitted based on a third-order polynomial of the form given in Equation (1) of ORNLlTM-2010/246. The licensee used the coefficients from the polynomial stress distribution to compute KIP and KIT values for a postulated 1/4T inside corner flaw using the nozzle stress intensity solution provided in Equation (2) of ORNLlTM-201 0/246. The licensee's RAt-1 response cited these specific ORNLlTM-201 0/246 formulas that were used for analyzing the nozzles' inside corner region.

The staff verified that the stress intensity formulation described in the licensee's RAt response has been approved by the NRC for implementation in RPV nozzle KIP and KIT calculations, as documented in the SE for BWR Owners Group Licensing Topical Report SIR-05-044-A, "Pressure-Temperature Limits Report [PTLR] Methodology for Boiling Water Reactors," Rev. 0, April 2007 (ADAMS Accession No. ML070180483). The PTLR methodology described in SIR-05-044-A was developed and approved for BWRs seeking to relocate PIT limit curves from the TS to PTLRs. The nozzle KI formulation described in these reports is based on a LEFM model that is generally considered to be applicable to postulated corner flaws in rounded corner nozzle forgings, irrespective of plant design. Based on the review of these analyses, the staff finds that the licensee's methods for calculating the KIP and KIT values for the nozzles are acceptable.

The licensee noted that an outside surface (3/4T) flaw was not considered in the development of the nozzles' PIT limit curves. The licensee explained that the pressure stress is significantly lower at the outside surface than the inside surface for the nozzles. The licensee also indicated that nozzle PIT limit curves were not provided for heat-up conditions, as they would be less limiting than the nozzle PIT limit curves for cool-down conditions at the 1/4T location.

The staff determined that the licensee's basis for excluding the 3/4T postulated flaw in the development of the PIT limit curves for the nozzles is acceptable. Specifically, the staff noted that the pressure stress decreases as a function of distance from the inside corner along the through-wall nozzle corner path, as shown in Figure 24 of ORNLlTM-201 0/246. Therefore, the KIP value for a 3/4T postulated flaw at the outside corner of the nozzle would be lower than that for the 1/4T flaw postulated for the inside corner region. The ORNL Report ORNUTM-201 0/246 LEFM analyses do not address 3/4T postulated flaws for this reason. Additionally, based on the analysis of the 1/4T location, the nozzle PIT limits for heat-up conditions would be less restrictive than those calculated for cool-down conditions, due to the fact that the 1/4T thermal stresses are compressive for heat-up. Therefore, the analysis of the 1/4T location during the cool-down transient generates the most bounding PIT limits for the nozzles. Thus, the staff finds that the licensee's consideration of the nozzle PIT limits for cool-down conditions, based on the analysis of the 1/4T location, is acceptable.

Based on the PIT limit curves shown in Figures 1 through 4 of the RAI-1 response, the licensee determined that the nozzle PIT limits are bounded by the traditional curves for the limiting beltline shell materials. The licensee thus concluded that the 54 EFPY PIT limits provided in the Farley 1 and 2 PTLRs are valid for the entire RPV.

The staff performed an independent confirmatory calculation to verify that the PIT limits for the inlet and outlet nozzle forgings are bounded by the TS PIT limit curves. The staff's nozzle calculation employed a simplified, but conservative approach by applying a stress concentration factor to the RPV shell membrane stress to account for the elevated stress levels in the nozzle

- 18 corner region. The staff's confirmatory calculation verified that the inlet and outlet nozzles' PIT limits are less restrictive and therefore bounded by those for the limiting beltline she" materials for each unit. Given that the inside corner regions of the inlet and outlet nozzles are the most highly stressed ferritic components in the RPV, the staff determined that the licensee adequately demonstrated that the TS PIT limit curves are controlling for the entire RPV. Therefore, the staff finds that RAI-1 regarding Factor 1 is resolved.

Evaluation of Response to RAI-1, Factors 2 and 3 - Replacement Ferritic Components of the RCPB Regarding replacement ferritic components of the RCPB, the licensee stated in its RAI-1 response that both units have replaced their SGs and RPV closure heads since original construction. The licensee indicated that the Farley 1 and 2 replacement SGs (RSGs) and replacement RPV closure head components were evaluated and designed for protection against non-ductile failure. The licensee stated that the RSGs were designed to the fracture mechanics requirements of the 1989 Edition of the ASME Code,Section III, Appendix G, and the replacement RPV closure heads were designed to the requirements of the 1998 Edition, with 2000 Addenda of the ASME Code,Section III, Appendix G. The licensee noted that the ASME Code,Section III, Appendix G, presents a method for obtaining a"owable loadings for protection against non-ductile failure for ferritic Class 1 components. The licensee stated that these replacement components do not have to be addressed for the 54 EFPY PIT limits since they have been designed to the requirements of the ASME Code, Section '" and have not undergone neutron embrittlement that would affect their PIT limits. The licensee did not indicate whether the 54 EFPY PIT limits incorporate the lowest service temperature (LST) requirement of Section III, NB-2332(b) for replacement ferritic RCS piping system components.

For the RSGs, the 1989 Edition of the ASME Code,Section III, NB-3211 (d) specifies that an acceptable procedure for non-ductile failure prevention is provided in Appendix G of that section.

The ASME Code,Section III, Appendix G provides methods for calculating PIT limit for these vessel components that are the same as the ASME Code,Section XI, Appendix G methods, although neutron embrittlement does not need to be considered for RCPB components outside the RPV. Based on these requirements, the staff determined that it requires demonstration that the proposed 54 EFPY PIT limits for the RPV are bounding with respect to the RSGs.

Therefore, in RAI-3, the staff requested that the licensee (a) provide information that demonstrates that the ferritic components of the RSGs are bounded by the proposed 54 EFPY PIT limits for the RPV, and (b) identify whether there are any replacement ferritic RCPB piping, pump, or valve components at either Farley unit and, if so, address whether the proposed 54 EFPY PIT limits for the RPV incorporate the LST requirement of the ASME Code,Section III, NB-2332(b).

In response to part (a) of RAI-3, the licensee stated that, in order to demonstrate that the 54 EFPY PIT limit curves bound the RSG ferritic components, two locations in the RSGs were evaluated.

The first location is the RSG tube sheet to channel head junction, and the second location is the RSG primary nozzle knuckle region, which is the inside corner of this nozzle. The licensee indicated that these are highly stressed discontinuity locations that should be considered for a non-ductile failure evaluation.

- 19 The licensee stated that, to determine if the RSG tube sheet to channel head junction location is more limiting for the PIT limits than the RPV beltline shell region, a PIT limit curve is calculated for this RSG location. The licensee stated that its fracture mechanics analysis of the RSG tube sheet to the channel head junction was performed based on a postulated inside surface axial flaw, per the ASME Code,Section III, Appendix G criteria. The licensee analyzed the case for a 100 of per hour cool-down transient, as it will produce the highest thermal stresses at the 1/4T location. The licensee stated that the limiting tensile stress components are chosen to determine the stress intensity factors, with the appropriate safety factors, consistent with the ASME Code,Section III, Appendix G.

The licensee stated that the initial RTNDT value for this location is conservatively set to the maximum allowable value of 10°F, based on the design specification for the Farley 1 and 2 RSG ferritic materials (base metals and welds). The licensee indicated that the RTNDT from the actual CMTRs is lower than 10 OF.

Based on these analyses, the licensee generated a PIT limit curve for the RSG tube sheet to channel head junction, and compared it to the 100 OF per hour cool-down curve for the limiting beltline shell materials for each unit. These curves, which are provided in Figure 1 of the RAI-3 response, demonstrate that the PIT limit curve for the RSG tube sheet to channel head junction is bounded by the PIT limit curves for the limiting beltline shell materials.

The licensee stated that the RSG primary nozzle corner region should also be considered as potentially limiting, per the ASME Code,Section III, Appendix G criteria, due to the nozzle discontinuity. In order to demonstrate that the PIT limits for the RSG primary nozzle knuckle regions are less limiting than those for the RPV, the licensee performed a comparison of the component stresses and RTNDT values. The licensee established in its response to RAI-1 that the PIT limits for the RPV inlet and outlet nozzle inside corner regions are less limiting than the limiting RPV beltline shell material PIT limits. In its response to RAI-3(a), the licensee sought to demonstrate that the RSG primary nozzle knuckle region PIT limits are less limiting than those for the RPV inlet and outlet nozzles for a 1/4T inside surface corner flaw, based on the 100 OF per hour cool-down transient.

The licensee established that the PIT limits for the RSG primary nozzle knuckle region would be bounded by those for the RPV inlet and outlet nozzles based on the following factors.

Higher K1c for the RSG primary nozzle: The licensee stated that the maximum allowable RTNDT of 10°F, per the design specification for the RSG primary nozzles, was considered for the evaluation. The licensee noted that the 54 EFPY ART values used for the development of the RPV inlet and outlet nozzles PIT limit curves are significantly greater.

The staff finds this determination acceptable because the allowable RT NDT of 10°F for the RSG primary nozzle is less than the 54 EFPY ARTs for RPV inlet and outlet nozzles listed in Table 2 of this SE.

Lower KIP for the RSG primary nozzle: The licensee stated that the stresses due to a unit pressure of 1000 psi were compared for the RSG primary nozzle corner and the Farley Units 1 and 2 RPV inlet and outlet nozzle corners. Based on this comparison, the licensee determined that the pressure stresses and associated KIP values at the RSG nozzle corner region are less than those at the RPV inlet and outlet nozzle corner regions for 80 percent

- 20 of the nozzle thickness. The staff finds since the 1/4T postulated corner flaw extends only 25 percent into the nozzle thickness from the inside corner, the pressure stress and KIP value for the RSG primary nozzle corner at the 1/4T location would therefore be lower than those for the RPV inlet and outlet nozzle corners. Thus, the staff finds that the licensee's determination regarding the lower KIP for the RSG primary nozzle (relative to the RPV inlet and outlet nozzles) is acceptable.

Lower KIT for the RSG primary nozzle: The licensee stated that the RSG primary nozzle wall thickness at the knuckle region is 10.24 inches, whereas the RPV nozzle corner through-wall thicknesses are approximately 15 inches and 18 inches for the inlet and outlet nozzles, respectively. Based on the lower wall thickness for the RSG primary nozzle relative to the RPV inlet and outlet nozzles, the licensee determined that the cool-down transient thermal stresses and associated KIT values would be lower for the RSG primary nozzle knuckle region than those for the RPV inlet and outlet nozzles. The staff finds this determination acceptable because KIT generally increases as a function of the section thickness, as shown by the thermal stress intenSity correlations from G-2214.3 in Appendix G of the ASME Code, Sections III and XI.

Since the RSG nozzle has lower KIP and KIT values and a higher K,c, the staff finds that the licensee adequately demonstrated that the PIT limits for the RSG primary nozzle are bounded by those for the RPV inlet and outlet nozzles; therefore, the RSG primary nozzle is also bounded by the PIT limits for the RPV beltline shell region.

Based on its review of the licensee's PIT limit curve for the SG tube sheet to channel head junction, and its analysis of the RSG primary nozzle knuckle region, as documented above, the staff finds that the licensee has adequately demonstrated that the ferritic materials of the RSGs are bounded by the 54 EFPY PIT limits in the proposed PTLRs. Therefore, the staff finds that Factor 2 of RAI-1 and part (a) of RAI-3 are resolved.

In response to part (b) of RAI-3, the licensee stated that the RCPB piping, pump, and valve components for Farley 1 and 2 are fabricated from stainless steel, and no replacements have been performed that altered this aspect of the original fabrication and construction of the RCPB.

Therefore, the licensee determined that the LST requirement of the ASME Code,Section III, I\IB-2332(b) is not applicable to the Farley 1 and 2 RCPS piping, pump, and valve components.

The staff agrees with this determination because the requirements of NB-2332(b) and 10 CFR Part 50, Appendix G are only applicable to ferritic components of the RCPB, and stainless steel is not ferritic. Therefore, the staff finds that Factor 3 of RAI-1 and part (b) of RAI-3 are resolved.

Based on the resolution of all three factors of RAI-1 , as docuroented above, the staff finds that the licensee adequately demonstrated that the 54 EFPY PIT limit curves for the limiting beltline shell materials, as established in the proposed PTLRs, are bounding for all territic RPV materials and the replacement territic RCPB materials, consistent with the requirements ot 10 CFR Part 50, Appendix G. Therefore, the proposed 54 EFPY PIT limit curves and PTLRs are acceptable for implementation in accordance with TS 5.6.6, as revised per the LAR.

In WCAP-16964-NP and WCAP-16918-NP, the licensee provided, and the staff reviewed, a direct comparison against the measured sensor reaction rates from Capsules Z and V. For all reactions for both units, the measured-to-calculated (M/C) ratios were very close to unity; the average ratio for Unit 1 was 0.86 with an associated standard deviation of 7.6-percent and the average ratio for

- 21 Unit 2 was 0.91 with an associated standard deviation of 6.5-percent. The distribution of (M/C) ratios for Unit 1 ranged from 0.80 to 0.97 and for Unit 2 ranged from 0.87 to 1.01. Therefore, all reaction rates were calculated within 20-percent of measured values, as suggested in RG1.190.

The staff finds the licensee has provided fluence calculations performed using an acceptable methodology, supported by analytic uncertainty analysis and comparison to approved test facilities, along with a plant-specific comparison of measured fluence values from Surveillance Capsules Z and V. Based on these considerations, the staff concludes that the licensee fluence calculations adhere to the guidance in RG 1.190, and the neutron exposures reported in the licensee's submittal are, therefore, acceptable.

As described in Section 2.0, Regulatory Evaluation of this SE, FNP proposed the following TS changes:

1. Revise the definition of the PTLR in TS 1.1, "Definitions," in accordance with item (1) of TSTF-419-A.

2. Revise TS Section 5.6.6 (TS 5.6.6), "Reactor Coolant System (RCS) Pressure and Temperature Limit Report (PTLR)," paragraph b (TS 5.6.6b) to incorporate the most recent NRC-approved version of the Westinghouse PTLR methodology, WCAP-14040-A, Rev. 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004 (ADAMS Accession No. ML050120209).

3. Revise the LTOP system applicability temperature, which is currently incorporated in several TS sections, from less than or equal to 325 of (s 325 OF) to s 275 oF.

4. Relocate the LTOP system applicability temperature, as revised above, is to be relocated from the TS to the PTLRs. Accordingly, the following TS sections, which currently incorporate the LTOP system applicability temperature, will be impacted by this change:

TS 3.4.6, "RCS Loops - MODE 4,"

TS 3.4.7, "RCS Loops - MODE 5, Loops Filled,"

TS 3.4.10, "Pressurizer Safety Valves," and

TS 3.4.12, "Low Temperature Overpressure Protection (LTOP) System,"

5. Revise the LTOP TS 3.4.12 LCO, action statements, and surveillance requirements to incorporate an additional limitation specifyill9 that when all RCS cold leg temperatures are

> 180 OF, but still below the MODE 4 LTOP applicability temperature, a maximum of two charging pumps shall be capable of injecting into the RCS.

Based on its evaluation, as documented in Section 3.0 of this SE, the NRC staff has determined the following:

The proposed TS changes identified in Section 3.1.1 are consistent with the guidance in NRC GL 96-03, TSTF-233-A, and TSTF-419-A.

- 22

The proposed TS changes associated with the LTOP system meet the criteria specified in ASME Code Case N-641.

The proposed PTLRs are in compliance with the requirements of 10 CFR Part 50, Appendix G and are consistent with GL 96-03.

The staff concludes that the proposed TS revisions are acceptable for implementation at Farley 1 and 2, and the proposed PTLRs are acceptable for implementation through 54 EFPY under the revised TS 5.6.6 administrative controls.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements and surveillance requirements 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 (77 FR 60153).

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.

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) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: C. Sydnor, NRR M. Hardgrove, NRR B. Parks, NRR Date: cttober 2, 2013

C. Pierce -2

5. revisions to LTOP TS 3.4.12 limiting condition for operation, action statements, and surveillance requirements to incorporate an additional limitation on the maximum number of charging pumps capable of injecting into the RCS.

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

Sincerely, Robert Martin, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosures:

1. Amendment No. 193 to NPF-2

2. Amendment No. 189 to NPF-8

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

PUBLIC LPL2-1 RIF RidsNrrPMFarley Resource RidsAcrsAcnw_MailCTR Resource RidsNrrLASFigueroa Resource RidsNrrDssstsb Resource RidsNrrDorlLpl2-1 Resource RidsNrrDsssrxb Resource RidsNrrDorlDpr Resource RidsNrrDeEvib Resource RidsRgn2MailCenter Resource ADAMS Accession No.: ML13249A386

via email OFFICE LPL2-1/PM LPL2-1/PM LPL2-1/LA DSS/SRXB/BC*

i

. NAME MBarillas RMartin SFigueroa CJackson I

DATE 9/11/13 9/12/13~oJo'l,../i3 9/11/13 11/27/12 OFFICE DElEVIB* DSS/STSB OGC LPL2-1/BC NAME SRosenberg REliiot MYoung RPascarelli roATE 9/4/13 9/12/13 9/27/13 10101/13 OFFICIAL RECORD COpy

v t e Letter Sequence Approval
TAC:ME9256, (Approved, Closed) TAC:ME9257, (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions 1 February 2013 Answers 14 March 2013 Results Approval Other: ML13217A309
MONTHYEAR2013-03-14 [Table View]

ML13249A386

Text

SUBJECT:

Dear Mr. Pierce:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

Enclosures:

Navigation menu

Search