ML20224A285

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Issuance of Amendment Nos. 231 and 228 to Revise Technical Specifications 3.3.1 and 3.3.7
ML20224A285
Person / Time
Site: Farley  Southern Nuclear icon.png
Issue date: 10/13/2020
From: Shawn Williams
Plant Licensing Branch II
To: Gayheart C
Southern Nuclear Operating Co
Williams S A-NRR/DORL 301-415-1009
References
EPID L-2019-LLA-0278
Download: ML20224A285 (21)
v • d • e


Text

October 13, 2020 Ms. Cheryl A. Gayheart Regulatory Affairs Director Southern Nuclear Operating Company 3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 231 AND 228 TO REVISE TECHNICAL SPECIFICATIONS 3.3.1 AND 3.3.7 (EPID L-2019-LLA-0278)

Dear Ms. Gayheart:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 231 to Renewed Facility Operating License No. NPF-2 and Amendment No. 228 to Renewed Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant, Units 1 and 2, respectively. The amendments are in response to your application dated December 12, 2019, as supplemented by letter dated May 20, 2020.

The amendments modify the Technical Specification (TS) 3.3.1, Reactor Trip System (RTS)

Instrumentation, and TS 3.3.7, Control Room Emergency Filtration/Pressurization System (CREFS) Actuation Instrumentation.

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

Sincerely,

/RA/

Shawn A. Williams, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosures:

1. Amendment No. 231 to NPF-2
2. Amendment No. 228 to NPF-8
3. Safety Evaluation cc: Listserv

ML20224A285 *email concurrence OFFICE DORL/LPL2-1/PM* DORL/LPL2-1/PM* DORL/LPL2-1/LA* DEX/EICB/BC*

NAME SDevlin-Gill SWilliams KGoldstein MWaters DATE 08/17/2020 08/17/2020 08/17/2020 07/24/2020 OFFICE DRA/ARCB/BC* DSS/SNSB/BC* DSS/STSB/BC* OGC/NLO*

NAME JDozier for KHsueh SKrepel VCusumano STurk DATE 07/16/2020 07/29/2020 08/21/2020 09/18/2020 OFFICE DORL/LPL2-1/BC* DORL/LPL2-1/PM*

NAME MMarkley SWilliams DATE 10/13/2020 10/13/2020 SOUTHERN NUCLEAR OPERATING COMPANY ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 231 Renewed License No. NPF-2

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

A. The application for amendment to the Joseph M. Farley Nuclear Plant, Unit 1 (the facility), Renewed Facility Operating License No. NPF-2 (the license) filed by Southern Nuclear Operating Company (the licensee), dated December 12, 2019, as supplemented by letter dated May 20, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (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 Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment. Paragraph 2.C.(2) of the license is hereby amended to read as follows:

2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 231, are hereby incorporated in the renewed license.

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

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

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Michael T. Michael T. Markley Date: 2020.10.13 Markley 12:44:57 -04'00' Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: October 13, 2020

SOUTHERN NUCLEAR OPERATING COMPANY ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 228 Renewed License No. NPF-8

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

A. The application for amendment to the Joseph M. Farley Nuclear Plant, Unit 2 (the facility), Renewed Facility Operating License No. NPF-8 (the license) filed by Southern Nuclear Operating Company (the licensee), dated December 12, 2019, as supplemented by letter dated May 20, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment. Paragraph 2.C.(2) of the license is hereby amended to read as follows:

2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 228, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Michael T. Michael T. Markley Date: 2020.10.13 Markley 12:45:34 -04'00' Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: October 13, 2020

ATTACHMENT TO JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT NO. 231 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 AND LICENSE AMENDMENT NO. 228 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert License License NPF-2, page 4 NPF-2, page 4 NPF-8, page 3 NPF-8, page 3 TSs TSs 3.3.1-9 3.3.1-9 3.3.7-4 3.3.7-4

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 231, 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;
5) Procedures defining corrective actions for off control point chemistry conditions; and Farley - Unit 1 Renewed License No. NPF-2 Amendment No. 231

(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 2821 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 228, are hereby incorporated in the renewed license.

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

(3) Deleted per Amendment 144 (4) Deleted per Amendment 149 (5) Deleted per Amendment 144 Farley - Unit 2 Renewed License No. NPF-8 Amendment No. 228

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 ----------------------------NOTE--------------------------------

Not required to be performed for source range instrumentation until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after THERMAL POWER is < P-6.

Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.2 ----------------------------NOTE------------------------------

Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is 15% RTP.

Compare results of calorimetric heat balance In accordance with calculation to power range channel output. Adjust the Surveillance power range channel output if calorimetric heat Frequency Control balance calculation results exceed power range Program channel output by more than +2% RTP.

SR 3.3.1.3 ----------------------------NOTES------------------------------

1. Not required to be performed until 7 days after THERMAL POWER is 50% RTP.
2. Performance of SR 3.3.1.9 satisfies this SR.

Compare results of the incore detector In accordance with measurements to Nuclear Instrumentation System the Surveillance (NIS) AFD. Adjust NIS channel if absolute difference Frequency Control is 3%. Program Farley Units 1 and 2 3.3.1-9 Amendment No. 231 (Unit 1)

Amendment No. 228 (Unit 2)

CREFS Actuation Instrumentation 3.3.7 Table 3.3.7-1 (page 1 of 1)

CREFS Actuation Instrumentation FUNCTION APPLICABLE REQUIRED SURVEILLANCE TRIP SETPOINT MODES OR OTHER CHANNELS REQUIREMENTS SPECIFIED CONDITIONS

1. Manual Initiation 1,2,3,4, (a), (b) 2 trains SR 3.3.7.6 NA
2. Automatic Actuation Logic 1,2,3,4 2 trains SR 3.3.7.3 NA and Actuation Relays SR 3.3.7.4 SR 3.3.7.5
3. Control Room Radiation 1,2,3,4 1 SR 3.3.7.1 1.0 X 10-5 Control Room Air Intake (a), (b) 2 SR 3.3.7.2 µCi/cc (c)

(R-35A, B) SR 3.3.7.7

4. Containment Isolation - Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 3.a., for all initiation functions and Phase A requirements.

(a) During CORE ALTERATIONS.

(b) During movement of irradiated fuel assemblies.

(c) Above background with no flow.

Farley Units 1 and 2 3.3.7-4 Amendment No. 231 (Unit 1)

Amendment No. 228 (Unit 2)

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 231 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO. 228 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364

1.0 INTRODUCTION

By letter dated December 12, 2019, as supplemented by letter dated May 20, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML19346E959 and ML20141L530, respectively), Southern Nuclear Operating Company (SNC, the licensee) requested changes to the Technical Specifications (TSs) for the Joseph M. Farley Nuclear Plant, Units 1 and 2 (Farley). The licensee proposed two unrelated changes to the TSs:

1) TS 3.3.1, Reactor Trip System (RTS) Instrumentation. The licensee proposed to delete the measurement unit RTP (rated thermal power) from the 3% absolute difference acceptance criterion specified in surveillance requirement (SR) 3.3.1.3; and,
2) TS 3.3.7, Control Room Emergency Filtration/Pressurization System (CREFS) Actuation Instrumentation. The licensee proposed to change the unit of measure associated with the trip setpoint and added a footnote to clarify that the value represents radiation above background with no system flow.

During the week of April 13, 2020, the U.S. Nuclear Regulatory Commission (NRC, the Commission) staff conducted a regulatory audit with regards to the proposed revision to TS 3.3.7. The audit summary can be found at ADAMS Accession No. ML20115E453.

The supplemental dated May 20, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on February 25, 2020, 85 FR 10730.

Enclosure 3

2.0 REGULATORY EVALUATION

2.1 TS 3.3.1, Reactor Trip System (RTS) Instrumentation 2.1.1 Description of System The licensee proposed a change to SR 3.3.1.3 regarding the measurement units for the axial flux difference (AFD) acceptance criterion. TS 1.1 Definitions, states that AFD is the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector. AFD limits prevent a highly top- or bottom-skewed axial power distribution and minimizes the potential for xenon transients. SR 3.3.1.3 compares incore detector measurements to the Nuclear Instrumentation System (NIS) AFD value.

2.1.2 Description of Proposed Changes Current SR 3.3.1.3 states, in part:

Compare results of the incore detector measurements to Nuclear Instrumentation System (NIS) AFD. Adjust NIS channel if absolute difference is 3% RTP.

Revised SR 3.3.1.3 would delete the measurement unit RTP from SR 3.3.1.3, so the second sentence in the SR would state Adjust NIS channel if absolute difference is 3%.

2.1.3 Description of Regulatory Requirements The NRC staff reviewed the licensees submittal relative to the following regulations and industrial standards:

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, Section 50.36, Technical Specifications, establishes the requirements for TSs. Specifically, 10 CFR 50.36(c)(3), SRs, states that: Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operations will be met.

The regulation in 10 CFR Part 50, Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, Criterion 10 (GDC10), Reactor design, states that: The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

The regulation in 10 CFR Part 50, Appendix A, GDC 20, Protection system functions, states that: The [reactor] protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

The RTS instrumentation must comply with the regulations in 10 CFR 50.55a(h)(2), Protection systems, which states, in part, that: For nuclear power plants with construction permits issued after January 1, 1971, but before May 13, 1999, protection systems must meet the requirements

in [Institute of Electrical and Electronics Engineers (IEEE)] Std 279-1968, Proposed IEEE Criteria for Nuclear Power Plant Protection Systems, or the requirements in IEEE Std 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations, or the requirements in IEEE Std 603-1991, Criteria for Safety Systems for Nuclear Power Generating Stations, and the correction sheet dated January 30, 1995.

2.2 TS 3.3.7, Control Room Emergency Filtration/Pressurization System (CREFS)

Actuation Instrumentation 2.2.1 Description of System The Farley Unit 1 and 2 common control room must be kept habitable for the operators during normal operation, anticipated transients, and design-basis accidents. Control room habitability is supported by the Control Room Emergency Filtration/Pressurization System (CREFS). The CREFS initiates filtered ventilation and pressurization of the control room on receiving an actuation signal.

The Control Room Air Intake Radiation Monitors (R-35A, B) continuously measure the radioactivity concentrations in the ventilation system air intake to the Control Room and compare the concentrations to the trip setpoint. Upon exceedance of the R-35A or R-35B trip setpoint, the CREFS system isolates the control room air intake and actuates the filtered ventilation and pressurization system.

TS Table 3.3.7-1, Function 3, Control Room Radiation Control Room Air Intake (R-35A, B),

provides a trip setpoint value for actuation of the CREFS. The normal unfiltered outside air supply is isolated from the control room upon receipt of a high radiation signal from one of these detectors.

The setpoint is selected to limit the radionuclide concentration in the main control room to ensure that the radiological dose to control room occupants remains below the 10 CFR 50.67(b)(2)(iii) limit, 10 CFR 50, Appendix A, GDC 19 limit, and below 10% of the applicable 10 CFR Part 20 Appendix B derived air concentration (DAC) value.

2.2.2 Description of Proposed Changes The licensee is requesting to revise TS 3.3.7 to change the unit of measurement associated with the trip setpoint of TS Table 3.3.7-1, CREFS Actuation Instrumentation, Function 3, Control Room Radiation Control Room Air Intake (R-35A, B), from 800 cpm [counts per minute] to an equivalent value of 1.0 x 10-5 µCi/cc and add a footnote (c) clarifying that the value represents radiation Above background with no flow.

The licensee is requesting this change to provide a universal trip setpoint value that allows for flexible changes in radiation monitoring equipment without the need for an associated TS change.

2.2.3 Description of Regulatory Requirements The NRC staff reviewed the licensees submittal relative the following regulations and industrial standards:

The regulation in 10 CFR Part 50, Appendix A, GDC 13, Instrumentation and control, states that: Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

The regulation in 10 CFR 50.36(a)(1), Technical specifications, states, in part, that: Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section The regulation in 10 CFR 50.36(c)(1)(ii)(A) states, in part, that: Limiting safety system settings

[LSSS] for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded.

If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor.

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

Appendix B to 10 CFR Part 20, Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage, in discussing Table 1, Occupational Values, states, in part, that:

The derived air concentration (DAC) values are derived limits intended to control chronic occupational exposures. The DAC values for occupational dose via inhalation are listed in Table 1, Column 3. This table indicates that the DAC value for Xenon (Xe)-133 and Krypton (Kr)-85 is 1.0 x 10-4 µCi/ml (µCi/cc).

The CREFS control room isolation instrumentation designs must comply with the regulations in 10 CFR 50.55a(h)(2), Protection systems, which states, in part, that: For nuclear power plants with construction permits issued after January 1, 1971, but before May 13, 1999, protection systems must meet the requirements in [Institute of Electrical and Electronics Engineers (IEEE)]

Std 279-1968, Proposed IEEE Criteria for Nuclear Power Plant Protection Systems, or the requirements in IEEE Std 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations, or the requirements in IEEE Std 603-1991, Criteria for Safety Systems for Nuclear Power Generating Stations, and the correction sheet dated January 30, 1995.

NRC Information Notice (IN) 82-49, Correction for Sample Conditions for Air and Gas Monitoring, dated December 16, 1982, informed the addressees of potential errors in radioactive gaseous effluent.

3.0 TECHNICAL EVALUATION

3.1 TS 3.3.1, Reactor Trip System (RTS) Instrumentation 3.1.1 Licensee Evaluation The licensee stated that the proposed change is needed to correct the TS acceptance criterion from Adjust NIS channel if absolute difference is 3% RTP to Adjust NIS channel if absolute difference is 3% because the absolute difference is not expressed in the unit of RTP. The proposed change will also align the Farley TS to the equivalent surveillance acceptance criterion specified in SR 3.3.1.3 of NUREG-1431, Revision 4, Standard Technical Specifications, Westinghouse Plants (ADAMS Accession No. ML12100A222).

The licensee stated that the proposed change does not alter the intent or meaning of the requirement but rather corrects the error introduced in Farley license Amendment Nos. 203 and 199, issued on August 3, 2016 (ADAMS Accession No. ML15233A448). The licensee considers the change administrative in nature with no technical impact.

3.1.2 NRC Staff Evaluation The NRC staff reviewed the licensees proposed change to SR 3.3.1.3 regarding the measurement units for the AFD acceptance criterion. The NRC staff determined that a direct comparison between the incore thermal power detector measurements versus NIS excore AFD expressed in % units would be appropriate, and the RTP is not relevant to such a dynamic setpoint. The NRC staff finds that SR 3.3.1.3 will continue to require the overtemperature T NIS channels to be adjusted when the comparison between the incore thermal power detector measurements and the NIS excore AFD meets the greater than or equal to 3% acceptance criterion. The NRC staff finds that the correct unit for comparison is % of absolute deviation, not % of RTP. Therefore, the proposed change of deleting the unit RTP from the 3% deviation adjustment acceptance criterion does not change the overtemperature T reactor trip function of the instrumentation. The NRC staff confirmed the proposed change is consistent with SR 3.3.1.3 of NUREG-1431, Revision 4. The NRC staff concludes that this change is administrative in nature and is, therefore, acceptable.

3.2 TS 3.3.7, Control Room Emergency Filtration/Pressurization System (CREFS)

Actuation Instrumentation 3.2.1 Proposed Change to TS Table 3.3.7-1 Trip Setpoint for Function 3 3.2.1.1 Licensee Evaluation In the license amendment request (LAR), the licensee stated that the cpm unit is specific to the Victoreen model radiation monitors (Beta Scintillation detectors) that are used currently by the licensee. The licensee used the Victoreen detectors gaseous effluent sensitivity curve from its operations and maintenance manual (on page 19 of the Calculation SM-SNC972214-001) to calculate and indicate that 800 cpm (the existing applicable trip setpoint of Function 3, Control Room Radiation Control Room Air Intake (R-35A, B) in TS Table 3.3.7-1) is equivalent to

1.0 x 10-5 µCi/cc (the proposed applicable Trip Setpoint for Function 3 in TS Table 3.3.7-1).

The licensee used Appendix B of 10 CFR Part 20 to verify that the proposed Trip Setpoint (1.0 x 10-5 µCi/cc) is based upon a release during an FHA in which Kr-85 and Xe-133 are the predominant radionuclides. The DAC limits are listed in Table 1, Column 3 of Appendix B to 10 CFR Part 20.

The licensee stated that the proposed change of the trip setpoint of Function 3 in TS Table 3.3.7-1 is administrative in nature since the proposed change is a change in the units of radioactive measurement.

3.2.1.2 NRC Staff Evaluation Using the applicable regulatory requirements in Section 2.2.3 of this safety evaluation (SE), the NRC staff reviewed the LAR to:

Verify that the licensees proposed trip setpoint value is equivalent to the existing trip setpoint value; and, Verify that the proposed trip setpoint value continues to assure that the required protective actions will be initiated before the associated plant process parameter exceeds its analytical limit.

TS Table 3.3.7-1, CREFS Actuation Instrumentation, establishes a trip setpoint of 800 cpm for Function 3, Control Room Radiation Control Room Air Intake (R-35A, B). The current trip setpoint units of measurement are in cpm, which are specific to the Victoreen model radiation monitors. A change to the TS would be required if or when the Victoreen radiation monitors are replaced with a different detector model with a different calibration factor. The proposed change in units of measurement to microcuries per cubic centimeter (µCi/cc) would make any associated TS setpoint value independent of any specific radiation monitor type. This proposed change would also make the trip setpoint units of measurement consistent with the setpoint units of measurement for Function 3 of TS Table 3.3.6-1, Containment Radiation Gaseous (R-24A, B), and Function 3 of TS Table 3.3.8-1, Spent Fuel Pool Room Radiation Gaseous (R-25A, B).

During a telephone call conducted by the NRC staff to clarify the equivalency of the existing trip setpoint value and the proposed trip setpoint value, the licensee explained that it reached its equivalency conclusion based on the Victoreen detectors gaseous effluent sensitivity curve contained in Southern Nuclear Design Calculation Number SM-SNC972214-001, Bases for R-24, R-25, & R-35 Tech Spec Setpoints. Subsequently, during the week of April 13, 2020, the NRC staff conducted a regulatory audit with the purpose of reviewing the information contained in SM-SNC972214-001 to assess the equivalency of the existing and proposed setpoint units of measurement and determine the need for any requests for additional information (RAI). The NRC staff issued an RAI on April 24, 2020 (ADAMS Accession No. ML20115E392). The licensee responded to the RAI in its supplemental letter dated May 20, 2020.

The NRC staff reviewed SM-SNC972214-001 to evaluate the proposed change. On Sheet 19 of SM-SNC972214-001, the licensee provided the Victoreen detectors gaseous effluent sensitivity curve. This figure provided a description of the equivalency of units of measurement for the trip setpoint, as shown below in Figure 1 of this SE. The licensee noted that the dotted red line (corresponding to 800 cpm) intersects the Kr-85 sensitivity line at 1.0 x 10-4 µCi/cc.

Figure 1: Victoreen Detectors Gaseous Effluent Sensitivity Curve On Sheet 20 of SM-SNC972214-001, the licensee stated that the 800 CPM setpoint line intersects the Kr-85 sensitivity line at 1.0 x 10-5 µCi/mL, or one tenth of its DAC. This is reasonable: MCR [Main Control Room] ventilation intake is isolated well below the regulatory limit.

The licensee applied the detector's Kr-85 sensitivity (7.8 x 107 cpm/µCi/mL) to the one tenth of its DAC to calculate the trip setpoint (800 cpm). The licensee calculated the 800 cpm value by using the following equation:

Setpoint (cpm) = 7.8 x 107 cpm/µCi/mL x 1.0 x 10-5 µCi/mL

= 780 cpm

Thus: 1.0 x 10-5 µCi/mL = 780 cpm / 7.8 x 107 cpm/µCi/mL where 1 milliliter (mL) is equal to 1 cubic centimeter (cc)

Therefore: 1.0 x 10-5 µCi/cc = 780 cpm / 7.8 x 107 cpm/µCi/cc (780 cpm was rounded up to 800 cpm because the analog logarithmic scale display has one significant figure of accuracy and 780 would be difficult to read on this scale.)

The NRC staff evaluated the figure on sheet 19 of the supplement dated May 20, 2020. The dotted red line corresponding to 800 cpm intersects the Kr-85 sensitivity curve at 1.0 x 10 5 µCi/cc. This intersection is reasonable because the MCR ventilation intake is isolated well below the regulatory limit (1.0 x 10-4 µCi/cc). In addition, the result of the calculation shown on Sheet 20 of SM-SNC972214-001 confirmed that the proposed Trip Setpoint of 1.0 x 105 µCi/cc is equivalent to 800 cpm. Therefore, the NRC staff finds that the proposed change from 800 cpm to 1.0 x 10-5 µCi/cc is acceptable. The NRC staff concludes that the proposed Trip Setpoint of Function 3 will provides reasonable assurance that the licensee will maintain these variables and systems within the prescribed operating ranges to satisfy the requirement of GDC 13.

The dotted red line corresponding to 800 cpm also intersects the Xe-133 sensitivity curve at approximately 0.3 x 10-4 µCi/cc, which is less than the DAC limit of 1 x 10-4 µCi/cc. This intersection demonstrates that with a proposed instrument setpoint value of 1.0 x 10-5 µCi/cc, MCR ventilation intake isolation will occur well before the DAC limits are exceeded. The DAC limits of Xe-133 and Kr-85 are 1.0 x 10-4 µCi/cc, as listed in Table 1, Column 3, of Appendix B to 10 CFR Part 20. Therefore, the NRC staff finds that the proposed Trip Setpoint satisfies the DAC limit values for occupational dose required in Table 1 of the Appendix B to 10 CFR Part 20.

The margin between the 1.0 x 10-4 µCi/cc and 1.0 x 10-5 µCi/cc DAC and 1/10th DAC limits respectively (two horizontal purple dashed lines) is 0.9 x 10-4 µCi/cc. This margin ensures that the proposed trip setpoint has been chosen to assure that a trip or safety actuation will occur significantly before radiation doses reach the DAC limit values for both Xe-133 and Kr-85.

Therefore, the NRC staff finds that the proposed trip setpoint setting satisfies the LSSS requirements of 10 CFR 50.36(c)(1)(ii)(A) and is, therefore, acceptable.

The NRC staff reviewed the TS SRs 3.3.7.1, Perform Channel Check; 3.3.7.2, Perform COT; and 3.3.7.7, Perform Channel Calibration, that are applicable to Function 3 in TS Table 3.3.7-1. The NRC staff verified that the proposed change will not impact these SRs because these SRs are independent of the units of measurement of the Function 3 trip setpoint.

Therefore, these SRs continue to assure that the necessary quality of systems and components are maintained and that facility operation will be within safety limits. Therefore, the NRC staff concludes that the proposed change satisfies the SR requirements of 10 CFR 50.36(c)(3).

The NRC staff also reviewed the proposed TS and verified that the proposed change for the trip setpoint unit of Function 3 in TS Table 3.3.7-1 is consistent with the setpoint units of Function 3 of TS Table 3.3.6-1, Containment Radiation Gaseous (R-24A, B), and Function 3 of TS Table 3.3.8-1, Spent Fuel Pool Room Radiation Gaseous (R-25A, B).

Based on the above, the NRC staff finds that the proposed trip setpoint of Function 3 in the TS Table 3.3.7-1 (1.0 x 10-5 µCi/cc) is acceptable because the proposed trip setpoint (1) is equivalent to the existing trip setpoint, (2) maintains sufficient margin to the Xe-133 and Kr-85

DAC limits, and, (3) continues to satisfy the applicable regulatory requirements as set forth in Section 2.2.3 of this SE.

3.2.2 Proposed Addition of Footnote (c) to Table 3.3.7-1 The licensee proposes to add footnote (c) to Table 3.3.7-1, CREFS Actuation Instrumentation.

Footnote (c) states Above background with no flow. This footnote would clarify that the value represents radiation above background with no system flow.

3.2.2.1 Licensee Evaluation The licensee stated that the addition of proposed footnote (c) represents a clarification that the Trip Setpoint value is that value above the normal background radiation level with no adjustment for flow in the system. The addition of the clarifying footnote is consistent with the existing footnote (b) for Function 3 of TS Tables 3.3.6-1 and 3.3.8-1 associated with the trip setpoint.

The proposed footnote (c) to Table 3.3.7-1 is considered an administrative change.

3.2.2.2 NRC Staff Evaluation The NRC staff issued a RAI on April 24, 2020, requesting the licensee to explain the meaning of the proposed footnote. The licensee responded to the RAI in its supplemental letter dated May 20, 2020.

The first part of the of the footnote above background annotates that the setpoint is established above the ambient background radiation level. The second part of the footnote with no flow address the fact that 10 CFR Part 20 Appendix B limits are at static (atmospheric) conditions. These are the same atmospheric conditions experienced by the public and main control room operators.

The R-35A, B radiation monitor nominal setpoints do not include a correction factor to account for the pressure differential that occurs as a result of the instrument sample line taking suction on the ventilation system in order to draw an air sample into the R-35A, B radiation monitoring instruments. The need for this correction factor is further described in NRC IN 82-49, Correction for Sample Conditions for Air and Gas Monitoring, dated December 16, 1982. As described in IN 82-49, the radionuclide concentration measured by a radiation monitor that draws an air sample through a detector chamber would be lower than the actual concentration at atmospheric conditions. Therefore, the with no flow portion of the footnote ensures that the nominal setpoint is adjusted, if needed, to account for the pressure difference identified in IN 82-49.

The licensee has evaluated the impact of this correction factor in its response to IN 82-49, entitled, REA99-2094-02, Re-Evaluation of Information Notice 82-49 (January 19, 2001), with respect to the effect of this adjustment on various air-sampling radiation monitors. The licensee stated that the information to explain this footnote will be added to the licensees Technical Specification Bases.

The NRC staff reviewed the proposed addition of footnote (c) to TS Table 3.3.7-1 and concludes that the addition of clarifying footnote (c) to Table 3.3.7-1 is acceptable because the footnote satisfies applicable regulatory requirements as described in Section 2.2.3 of this SE and maintains the same level of radiological protection for the control room envelope.

3.

2.3 NRC Staff Conclusion

for TS 3.3.7 Based on the above, the NRC staff concludes that the licensees proposed changes to the TS Table 3.3.7-1 trip setpoint for Function 3 are acceptable because the proposed trip setpoint (1) is equivalent to the existing trip setpoint, (2) maintains sufficient margin to the Xe-133 and Kr-85 DAC limits, and, (3) continues to satisfy applicable regulatory requirements as described in Section 2.2.3 of this SE. The NRC staff further concludes the proposed addition of footnote (c) to TS Table 3.3.7-1 is acceptable because the footnote maintains the same level of radiological protection for the control room envelope. Therefore, the NRC staff concludes that there is reasonable assurance that the regulatory requirements will continue to be met and the proposed changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Alabama State official was notified of the proposed issuance of the amendments on July 8, 2020. On July 16, 2020, the State official stated that the State of Alabama had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. 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 (85 FR 10730, February 25, 2020), and there has been no public comment on such finding. 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 Commissions 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: Dawnmathews Kalathiveettil Steven Garry Syed Haider Date of Issuance: October 13, 2020

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