ML20196L929

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Issuance of Amendment Nos. 229 and 226 Regarding Spent Fuel Pool Criticality Safety Analysis
ML20196L929
Person / Time
Site: Farley  Southern Nuclear icon.png
Issue date: 10/06/2020
From: Shawn Williams
Plant Licensing Branch II
To: Gayheart C
Southern Nuclear Operating Co
Williams S
References
EPID L-2019-LLA-0212
Download: ML20196L929 (31)
v • d • e


Text

October 6, 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. 229 AND 226 REGARDING SPENT FUEL POOL CRITICALITY SAFETY ANALYSIS (EPID L-2019-LLA-0212)

Dear Ms. Gayheart:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 229 to Renewed Facility Operating License No. NPF-2 and Amendment No. 226 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 September 30, 2019, as supplemented by letter dated April 13, 2020.

The amendments modify the Technical Specification (TS) 3.7.15, Spent Fuel Assembly Storage, and TS 4.3, Fuel Storage, and update the spent fuel pool criticality safety analysis to account for the impact on the spent fuel for a measurement uncertainty recapture power uprate.

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. 229 to NPF-2
2. Amendment No. 226 to NPF-8
3. Safety Evaluation cc: Listserv

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. 229 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 September 30, 2019, as supplemented by letter dated April 13, 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 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. 229, 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 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Michael T. Michael T. Markley Date: 2020.10.06 Markley 08:00:23 -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 6, 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. 226 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 September 30, 2019, as supplemented by letter dated April 13, 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 are hereby amended to read as follows:

2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 226, 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 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Michael T. Michael T. Markley Date: 2020.10.06 Markley 08:00:59 -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 6, 2020

ATTACHMENT TO JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT NO. 229 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 AND LICENSE AMENDMENT NO. 226 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 3.7.15-1 3.7.15-1 3.7.15-2 Deleted 4.0-2 4.0-2 4.0-5 4.0-5 4.0-6 4.0-6 4.0-7 4.0-7 4.0-8 4.0-8 4.0-9 4.0-9 4.0-10 4.0-10 4.0-11 4.0-11 4.0-12 4.0-12

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 229, 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. 229

(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. 226, 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. 226

Spent Fuel Assembly Storage 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Assembly Storage LCO 3.7.15 The combination of initial enrichment and burnup of each spent fuel assembly stored in the spent fuel storage pool shall be in accordance with Specification 4.3.1.1.

APPLICABILITY: Whenever any fuel assembly is stored in the spent fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 -------------NOTE-------------

LCO not met. LCO 3.0.3 is not applicable.

Initiate action to move the Immediately noncomplying fuel assembly to an acceptable storage location.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify by administrative means the initial enrichment Within 7 days and burnup of the fuel assembly is in accordance following the with Specification 4.3.1.1. relocation or addition of fuel assemblies to the spent fuel storage pool.

Farley Units 1 and 2 3.7.15-1 Amendment No. 229 (Unit 1)

Amendment No. 226 (Unit 2)

Design Features 4.0 4.0 DESIGN FEATURES 4.3.1.1 (continued)

b. keff < 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 4.3.2.7.2 of the FSAR;
c. keff 0.95 if fully flooded with water borated to 400 ppm, which includes an allowance for uncertainties and biases as described in Section 4.3.2.7.2 of the FSAR;
d. A nominal 10.75 inch center to center distance between fuel assemblies placed in the fuel storage racks;
e. New or partially spent fuel assemblies that must be stored according to their combination of discharge burnup and nominal enrichment, decay time since operation, required Integral Fuel Burnable Absorber (IFBA) (if applicable), and must comply with Figure 4.3-1, Table 4.3-1, and Tables 4.3-3 through 4.3-5 (as applicable). Each assembly shall be stored in an appropriate storage configuration according to its fuel category as specifically described in Table 4.3-1 and geometry based on Figure 4.3-1;
f. Fuel assemblies that are stored in accordance with every applicable storage array as shown in Figure 4.3-1 of which they are a part (i.e., one fuel assembly can be part of up to four different storage arrays, each storage array shall be in accordance with Figure 4.3-1); and
g. Unit 1 only Damaged fuel assemblies F02, F05, F06, F15, F17, F18, F19, F20, F30, F31, and F32 shall be stored in accordance with Figure 4.3-2.

4.3.1.2 The new fuel pit storage racks are designed and shall be maintained with:

a. Fuel assemblies with Standard Fuel Assembly fuel rod diameters having a maximum nominal U-235 enrichment of 4.25 weight percent; (continued)

Farley Units 1 and 2 4.0-2 Amendment No. 229 (Unit 1)

Amendment No. 226 (Unit 2)

Design Features 4.0 Any 2x2 array of storage cells containing fuel shall comply with the requirements of Array A, Array B, or Array C, as applicable.

A. Fuel is divided into two Groups, based on Fuel Type (Standard Fuel Assembly (STD)/Robust Fuel Assembly (RFA) or Optimized Fuel Assembly (OFA)).

B. Arrays A, B and C designate the pattern of fuel which may be stored in any 2x2 Array.

C. Fuel Categories 1-4 are defined in Table 4.3-1.

Array A 1 X Two Category 1 assemblies with two empty storage locations. The Category 1 fuel assemblies must only be face adjacent to an empty storage location. X 1 4 4 Array B One Category 2 assembly with three Category 4 assemblies.

4 2 3 3 Array C Four Category 3 assemblies.

3 3 Notes:

1. Any storage array location designated for a fuel assembly may be replaced with non-fissile material.
2. Empty locations designated with an X must remain completely empty.

Figure 4.3-1 Spent Fuel Pool Loading Restrictions Page 1 of 3 Farley Units 1 and 2 4.0-5 Amendment No. 229 (Unit 1)

Amendment No. 226 (Unit 2)

Design Features 4.0 Notes Continued:

3. Other Fuel Categories are determined as follows:
a. For STD/RFA assemblies, determine the fitting coefficients A1 - A4 using Table 4.3-3.
b. For OFA assemblies, determine the fitting coefficients A1 - A4 using Table 4.3-4.
c. For assemblies with Initial Enrichment (En) values greater than or equal to the values in Table 4.3-2, the required Minimum Burnup value (in MWd/MTU) for each Fuel Category is calculated based on initial enrichment, decay time, and the appropriate fitting coefficients. If the fuel assembly burnup is greater than the calculated Minimum Burnup value, then the fuel may be classified into this Fuel Category.

The equation for Minimum Burnup is:

Minimum Burnup (MWd/MTU) = 1,000 x [A1 x En3 + A2 x En2 + A3 x En + A4]

Note: If the computed Minimum Burnup value is negative, zero shall be used.

The equation for Minimum IFBA required for Fuel Category 2 assemblies as a function of enrichment between 3.2 and 5.0 weight percent Uranium-235 is:

Minimum IFBA (rods) = A1 x En2 + A2 x En + A3 Note: The Minimum IFBA should be rounded up to the next whole number.

Note: Below 3.2 weight percent U-235, IFBA is not required.

d. Decay time is measured in years. For decay times between the values in Tables 4.3-3 and 4.3-4, linear interpolation or the lower decay time value may be used. If interpolation is used, linear interpolation based on actual decay time should be performed between calculated values of Minimum Burnup associated with tabulated Decay Times greater and less than the actual Decay Time. No extrapolation beyond 20 years is permitted.
e. Initial enrichment (En) is the nominal U-235 enrichment of the central zone region of fuel, excluding axial blankets. If the fuel assembly contains axial regions with different U-235 enrichment values, such as axial blankets, the maximum enrichment value should be utilized. If the computed Minimum Burnup value is negative, zero shall be used.

Figure 4.3-1 Spent Fuel Pool Loading Restrictions Page 2 of 3 Farley Units 1 and 2 4.0-6 Amendment No. 229 (Unit 1)

Amendment No. 226 (Unit 2)

Design Features 4.0 Notes Continued:

4. An empty (water-filled) cell may be substituted for any fuel-containing cell in all storage arrays.
5. Fuel Category 2 fuel which has been operated must have at least 10,000 MWd/MTU of burnup.

Figure 4.3-1 Spent Fuel Pool Loading Restrictions Page 3 of 3 Farley Units 1 and 2 4.0-7 Amendment No. 229 (Unit 1)

Amendment No. 226 (Unit 2)

Design Features 4.0 Table 4.3-1 Fuel Categories Ranked by Reactivity Fuel Category 1 High Reactivity Fuel Category 2 Fuel Category 3 Fuel Category 4 Low Reactivity Notes:

1. Assembly storage is controlled through the storage arrays defined in Figure 4.3-1.
2. Fuel Categories are ranked in order of decreasing reactivity, e.g., Fuel Category 2 is less reactive than Fuel Category 1, etc.
3. Each storage cell in an array can only be populated with assemblies of the fuel category defined in the array definition or a lower reactivity fuel category.
4. Fuel Category 1 contains fuel with an initial maximum enrichment up to 5 weight percent U-235.

Neither burnup nor IFBA is required.

5. Fuel Category 2 contains fuel with an initial maximum enrichment up to 5 weight percent U-235.

Storage of fresh fuel is determined from the minimum IFBA equation and coefficients provided in Table 4.3-5. To be eligible for Fuel Category 2, fuel which has been operated in the reactor requires at least 10,000 MWd/MTU of burnup.

6. Fuel Categories 3 and 4 are determined from the minimum burnup equation and coefficients provided in Table 4.3-3 for STD/RFA fuel and in Table 4.3-4 for OFA fuel.

Table 4.3-2 Maximum Enrichment allowed with 0.0 MWd/MTU Burnup Fuel Category RFA/STD OFA 1 5.0 5.0 2 5.01 5.01 3 2.15 2.15 4 1.70 1.75 Notes:

1. Requires IFBA credit for greater than 3.2 weight percent U-235.
2. For assemblies with an Initial Enrichment below the values listed above, no burnup is required Farley Units 1 and 2 4.0-8 Amendment No. 229 (Unit 1)

Amendment No. 226 (Unit 2)

Design Features 4.0 Table 4.3-3 Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Decay Time and Initial Enrichment (En) for STD/RFA Fuel Fuel Decay Time Coefficients Category (years)

A1 A2 A3 A4 0 0.2251 -2.5199 21.4065 -36.6115 5 0.3002 -3.4376 24.0978 -38.9002 3 10 0.1856 -2.3309 20.2704 -34.6503 15 0.0892 -1.3905 17.0683 -31.1550 20 0.0388 -0.9253 15.5082 -29.4500 0 -0.6112 4.6655 6.7127 -21.8911 5 -0.3326 2.0713 12.8468 -26.1880 4 10 -0.1305 0.0505 18.3242 -30.7080 15 0.1360 -2.6856 26.5239 -38.3300 20 0.2321 -3.7177 29.5977 -41.1200 Farley Units 1 and 2 4.0-9 Amendment No. 229 (Unit 1)

Amendment No. 226 (Unit 2)

Design Features 4.0 Table 4.3-4 Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Decay Time and Initial Enrichment (En) for OFA Fuel Fuel Decay Time Coefficients Category (years)

A1 A2 A3 A4 0 0.1692 -1.8852 18.5219 -32.7830 5 0.0191 -0.4154 13.4482 -27.1777 3 10 -0.0705 0.4300 10.5987 -24.0722 15 -0.1420 1.1146 8.2825 -21.5440 20 -0.1959 1.6375 6.5093 -19.6130 0 0.4957 -6.0715 37.2851 -49.1282 5 0.7476 -8.7581 45.3241 -56.5172 4 10 0.9041 -10.4334 50.3246 -61.0800 15 1.0799 -12.2326 55.7508 -66.1820 20 1.2541 -13.9154 60.5977 -70.5720 Farley Units 1 and 2 4.0-10 Amendment No. 229 (Unit 1)

Amendment No. 226 (Unit 2)

Design Features 4.0 Table 4.3-5 Fuel Category 2 Coefficients to Calculate the Minimum IFBA Required as a Function of IFBA Thickness and Fuel Type Coefficients Fuel Type IFBA Thickness A1 A2 A3 1.00X 5.2750 8.3325 -79.9546 STD/RFA 1.25X 3.7476 10.8046 -72.0974 1.50X 1.8593 19.8050 -81.5075 1.00X 6.2658 0.8890 -65.4949 OFA 1.25X 3.9144 9.3963 -68.9414 1.50X 1.5898 21.8436 -84.9630 Farley Units 1 and 2 4.0-11 Amendment No. 229 (Unit 1)

Amendment No. 226 (Unit 2)

Design Features 4.0 F31 Empty F30 F06 F18 F17 F19 F02 F15 F20 F05 F32 Water Note: All Assemblies are 3.0 w/o 235U nominal enrichment Figure 4.3-2 Damaged Fuel Assembly Configuration (Unit 1 Only)

Farley Units 1 and 2 4.0-12 Amendment No. 229 (Unit 1)

Amendment No. 226 (Unit 2)

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 229 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO. 226 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 September 30, 2019 (Agencywide Documents and Access Management System (ADAMS) Package Accession No. ML19275E393) as supplemented by letter dated April 13, 2020 (ADAMS Accession No. ML20104C140), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) that proposed changes to the Technical Specifications (TSs) for the Joseph M. Farley Nuclear Plant, Units 1 and 2 (FNP or Farley).

The amendments modify TS 3.7.15, Spent Fuel Assembly Storage, and TS 4.3, Fuel Storage, and update the spent fuel pool (SFP) criticality safety analysis to account for the impact on the spent fuel to support a measurement uncertainty recapture (MUR) power uprate.

The supplemental letter dated April 13, 2020, provided additional information that clarified the application, did not expand the scope of the application as noticed, and did not change the U.S.

Nuclear Regulatory Commission (NRC) staff proposed no significant hazards consideration determination as published in the Federal Register on January 7, 2020 (85 FR 733).

Enclosure 3

2.0 REGULATORY EVALUATION

2.1 System Description Farley Unit 1 and Unit 2 contain their own SFPs. A description of Farleys SFPs is in UFSAR, Revision 29, Section 9.1.2, Wet Spent-Fuel Storage (ADAMS Accession No. ML20125A210).

The SFPs are made up of one fuel storage rack design that maintains 10.75 inch center-to-center spacing between spent fuel assemblies. The Farley SFPs each consist of two 6 x 7, nineteen 7 x 7, and seven 7 x 8 storage racks. The SFP storage capacity is 1,407 fuel assemblies. The actual storage capacity is limited by TS 3.7.15, Spent Fuel Assembly Storage, and is dependent upon the fuel characteristics of the Farley fuel inventory.

2.2 Description of Proposed Changes The updated SFP nuclear criticality safety analysis (NCS) evaluates the SFP storage racks for the placement of fuel within the storage arrays defined in the technical specifications. Proposed revision to the TSs would include:

TS 3.7.15, Spent Fuel Assembly Storage The proposed revision to TS 3.7.15 would include:

  • delete within the Acceptable Burnup Domain of Figure 3.7.15-1 or from LCO 3.7.15,
  • delete TS Figure 3.7.15-1, Fuel Assembly Limit Burnup Requirements For All Cell Storage. This figure will be replaced by Figure 4.3-1, Table 4.3-1, and Tables 4.3-3 through 4.3-5 in TS 4.3.

TS 4.3, Fuel Storage The proposed revision to TS 4.3 would include:

  • Current TS 4.3.1.1.e. is replaced in its entirety. Revised TS 4.3.1.1 would state, New or partially spent fuel assemblies that must be stored according to their combination of discharge burnup and nominal enrichment, decay time since operation, required Integrated Fuel Burnable Absorber (IFBA) (if applicable), and must comply with Figure 4.3-1, Table 4.3-1, and Tables 4.3-3 through 4.3-5 (as applicable). Each assembly should be stored in an appropriate storage configuration according to its fuel category as specifically described in Table 4.3-1 and geometry based on Figure 4.3-1;
  • Current TS 4.3.1.1.f. is replaced in its entirety. Revised TS 4.3.1.1.f. would state, Fuel assemblies that are stored in accordance with every applicable storage array as shown in Figure 4.3-1 of which they are a part (i.e., one fuel assembly can be part of up to four different storage arrays, each storage array shall be in accordance with Figure 4.3-1); and
  • Current TS 4.3.1.1.g. is revised to replace Figure 4.3-6 with Figure 4.3-2
  • Current Figures 4.3-1 through 4.3-5 are replaced with revised/new Figure 4.3-1 and Tables 4.3- 1 through 4.3-5.

Proposed Figure 4.3-1 contains graphical and verbal descriptions of the four allowed storage arrays.

Proposed Table 4.3-1 provides interpretation of the fuel categories included in Figure 4.3-1 and references to the applicable table that provides the fitting coefficients that are to be used to calculate the minimum required fuel assembly burnup or the minimum IFBA requirement.

Proposed Table 4.3-2 provides the maximum enrichment allowed for each fuel category with 0.0 MWd/MTU burnup.

Proposed Table 4.3-3 provides the fitting coefficients to calculate the minimum required fuel assembly burnup for fuel categories 3 and 4 for Standard Fuel Assembly (STD)/Robust Fuel Assembly (RFA) fuel.

Proposed Table 4.3-4 provides the fitting coefficients to calculate the minimum required fuel assembly burnup for fuel categories 3 and 4 for Optimized Fuel Assembly (OFA) fuel.

Proposed Table 4.3-5 provides the fitting coefficients to calculate the minimum IFBA requirements for fuel category 2.

  • Renumbering Figure 4.3-6 as Figure 4.3-2. TS that reference Figure 4.3-6 are being changed to reference Figure 4.3-2.

2.3 Description of Regulatory Requirements The regulatory requirements and guidance documents that the NRC staff used in the review of the LAR are listed below.

Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix A, General Design Criteria for Nuclear Power Plants (GDC) Criterion 62, Prevention of criticality in fuel storage and handling, requires that, Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

The regulations in 10 CFR 50.68, Criticality accident requirements, requires, in part, under 10 CFR 50.68(a), each holder of an operating license shall comply with either 10 CFR 70.24 or the requirements in 10 CFR 50.68(b). SNC has not elected to comply with the requirements of 10 CFR 50.68(b); therefore, it must comply with the requirements of 10 CFR 70.24. FNP, Unit 1 and 2, was granted an exemption to 10 CFR 70.24 on July 31, 1996 (ADAMS Accession No. ML20116D649). The 10 CFR 70.24 exemption remains the FNP, Unit 1 and 2, licensing basis for new and spent fuel storage.

The regulations in 10 CFR 50.36, Technical specifications, contain the requirements for the content of TSs. The regulations in 10 CFR 50.36(b) require TSs to be derived from the analyses and evaluation included in the safety analysis report and amendments thereto. In accordance with 10 CFR 50.36(c)(2), Limiting conditions for operation, (LCOs) are the lowest functional capability or performance levels of equipment required for safe operation of the facility. In accordance with 10 CFR 50.36(c)(3), Surveillance requirements, (SRs) 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. The regulation in 10 CFR 50.36(c)(4),

Design features, requires that the TSs include design features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2),

and (3) of 10 CFR 50.36.

The NRC staff review was performed consistent with Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling (ADAMS Accession No. ML070570006), and Section 9.1.2, New and Spent Fuel Storage (ADAMS Accession No. ML070550057), of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants.

The NRC staff also used an internal memorandum dated August 19, 1998, containing guidance for performing the review of SFP NCS analysis (NRC memorandum from L. Kopp to T. Collins, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants (ADAMS Accession No. ML003728001) (hereafter referred to as the Kopp memo). While the Kopp memo does not specify a methodology, it does provide guidance on the more salient aspects of an NCS analysis, including computer code validation. The guidance is germane to boiling-water reactors and PWRs for both borated and unborated fuel storage pools. The NRC staff used the Kopp memo for its review of this application.

On September 29, 2011, the NRC staff issued Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01, Revision 0, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, dated September 2011 (ADAMS Accession No. ML110620086); notice of availability published in the Federal Register on October 13, 2011 (76 FR 63676), for review of SFP criticality analyses. The guidance in DSS-ISG-2010-01 is used by the NRC staff to review NCS analyses for the storage of new and spent nuclear fuel as it applies to: (i) future applications for construction and/or operating licenses, and (ii) future applications for license amendments and requests for exemptions from compliance with applicable requirements that are approved after the date of this interim staff guidance. The NRC staff used DSS-ISG-2010-01 for its review of this application.

3.0 TECHNICAL EVALUATION

3.1 Background

The licensees NCS analyses describes the methodology and analytical models used to show that the SFP storage rack maximum Keff will be less than 1.0 when flooded with unborated water for normal conditions, and less than or equal to 0.95 when flooded with borated water for normal and credible accident conditions at a 95-percent probability, 95-percent confidence level.

3.2 SFP NCS Analysis Method There is no generic or standard NRC-approved methodology for performing NCS analyses for fuel storage and handling. The plant-specific methods used for the NCS analyses for fuel in the Farley SFP are described in the LAR as WCAP-18414-P, J.M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis, (Attachment 4, proprietary) and WCAP-18414-NP (Attachment 5, non-proprietary), as supplemented by letter dated April 13, 2020.

3.2.1 Computational Methods The licensees NCS analyses consider the decrease in fuel reactivity typically seen in PWRs, as the fuel is depleted during reactor operation. This approach is frequently used in PWR NCS analyses and is sometimes referred to as burnup credit. Burnup credit NCS analysis requires a two-step process. The first step relates to depletion where a computer code simulates the reactor operation to calculate the changes in the fuel composition of the fuel assembly. The second step is a modeling of the depleted fuel assembly in the SFP storage racks and the determination of the system Keff. The validation of the computer codes in each step is a significant portion of the analysis. Since the licensees NCS analyses credit fuel burnup, it is necessary for the NRC staff to consider validation of the computer codes and data used to calculate burned fuel compositions, and the computer code and data that utilize the burned fuel compositions to calculate Keff for systems with burned fuel.

The methodology in the licensees NCS analyses employs the two-dimensional (2-D) transport lattice code PARAGON Version 1.2.0, as documented in WCAP-16045-P-A, Revision 0, Qualification of the Two-Dimensional Transport Code PARAGON, August 2004 (ADAMS Accession No. ML030760103, non-proprietary version) for the purposes of performing reactor core simulation analyses in the depletion step.

The methodology in the licensees NCS analyses employs SCALE 6.2.3 as documented in ORNL/TM-2005/39, SCALE: A Modular Code System for Performing Standard Computer Analyses for Licensing Evaluation with the ENDF/B-VIl 238-group cross-section library.

3.2.2 Depletion Computer Code Validation The licensees NCS analyses calculated the depletion uncertainty as five percent of the reactivity difference between fresh unpoisoned fuel and the burnup of interest. The method used by the licensee is consistent with NRC guidance documents DSS-ISG-2010-01, Revision 0, and the Kopp Memo and is, therefore, acceptable.

3.2.3 SFP Keff Computer Code Validation The study used to support validation of Keff calculations using the SCALE 6.2.3 CSAS5 sequence was documented in Appendix A of Attachments 4 (proprietary) and 5 (non-proprietary) to the licensees letter dated September 30, 2019, as supplement by letter dated April 13, 2020. The validation set includes critical configurations from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and French Haut Taux de Combustion (HTC) critical experiments from NUREG/CR-6979, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data (ADAMS Accession No. ML082880452).

The validation was performed in a manner consistent with NUREG-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology (ADAMS Accession No. ML050250061) and included an evaluation for temperature bias. Therefore, the NRC staff finds that this validation of SCALE 6.2.3 CSAS5 acceptable.

3.3 SFP and Fuel Storage Racks 3.3.1 SFP Water Temperature The NRC guidance provided in the Kopp memo states the NCS analysis should be done at the temperature corresponding to the highest reactivity. The licensees analysis calculated a base Keff using a constant temperature. The analysis performed a sensitivity analysis for each array to determine if another temperature might result in an increase in reactivity. Where reactivity increases were identified, the licensee included a temperature bias in the final estimation of Keff.

3.3.2 SFP Storage Rack Models The Farley SFP storage racks utilize a flux trap design. The storage cells are constructed of stainless-steel boxes that are attached to a frame with a separation between the cells, which is the flux trap. The storage racks were constructed with Boraflex positioned on the outside of each storage cell face. Because of Boraflex degradation, the licensee does not credit Boraflex to meet their SFP subcriticality requirements (Farley Amendments 133 and 125, respectively (ADAMS Accession No. ML013130226)).

3.3.3 SFP Storage Rack Models Manufacturing Tolerances and Uncertainties The licensees analyses of SFP storage rack models manufacturing tolerances and uncertainties are included in WCAP-18414-P, Revision 0 (Attachment 4 to the LAR) Section 5.2.3.1 of the updated criticality safety analysis.

NRC guidance in DSS-ISG-2010-01, Revision 0, and the Kopp Memo allows for the statistical combination of uncertainties using the root sum of the squares (RSS). When using the RSS statistical method to combine uncertainties the largest individual uncertainties dominate the result with smaller uncertainties having lesser impact on the final answer. The licensees SFP NCS analyses used the RSS statistical method to combine uncertainties. The licensee used the largest individual uncertainties. Based on its review, the NRC staff determined that the licensees SFP NCS analyses has sufficient margin to its licensing basis Keff requirements to accommodate not including minor uncertainties. Therefore, the NRC staff finds that the SFP storage rack models manufacturing tolerances and uncertainties are acceptable in this analysis.

3.4 Fuel Assembly 3.4.1 Bounding Fuel Assembly Design The fuel assemblies used at Farley are all 17x17 assemblies manufactured by Westinghouse of either the OFA or STD/RFA design. The Farley SFP NCS analyses did not determine a single bounding fuel assembly design. Instead, the licensee performed an analysis to determine separate design basis for each of the OFA or STD/RFA designs. Separate storage requirements were established for each design. Based on its review, the NRC staff finds this approach acceptable.

3.4.2 Fuel Assembly Manufacturing Tolerances and Uncertainties The licensees analyses of fuel assembly manufacturing tolerances and uncertainties are included in WCAP-18414-P, Revision 0 (Attachment 4 to the LAR) Section 5.2.3.1 of the updated criticality safety analysis. As discussed in Section 3.3.3 of this SE, the licensee used the RSS statistical method to combine uncertainties and used the larger uncertainties. The result of the RSS method is dominated by the larger uncertainties. Based on its review, the NRC staff determined that the licensees SFP NCS analysis has sufficient margin to its licensing basis Keff requirements to accommodate not including minor uncertainties. Therefore, the NRC staff finds that the fuel assembly manufacturing tolerances and uncertainties are acceptable in this analysis.

3.4.3 Spent Fuel Characterization Characterization of fresh fuel is based primarily on U-235 enrichment and various manufacturing tolerances. The manufacturing tolerances are typically manifested as uncertainties, as discussed above, or are bounded by values used in the analysis. These tolerances and bounding values would also apply to the spent nuclear fuel. Common industry practice has been to treat the uncertainties as unaffected by the fuel depletion. As discussed below, the NRC staff has previously accepted the practice. Because of the margin available in the analysis, the NRC staff finds this practice acceptable in the Farley analysis. The characterization of spent nuclear fuel is complex. Its characterization is based on the specifics of its initial conditions and its operational history in the reactor. That characterization has three

main areas: depletion uncertainty, the axial and radial apportionment of the burnup, and the core operation that achieved that burnup. These characteristics are evaluated below.

3.4.3.1 Depletion Uncertainty With respect to depletion uncertainty, the Kopp memo states, In the absence of any other determination of the depletion uncertainty, an uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest is an acceptable assumption. This method represents the engineering judgment of the memos author, based on his experience with the ability of approved NRC depletion and reactor operation simulation codes and methods to accurately predict the post-irradiated isotopic concentrations of nuclear fuel. The licensees NCS analyses used this method to determine the depletion uncertainty. As noted above, the licensees NCS analyses used the NRC-approved code PARAGON for the depletion portion of the analysis and calculated the depletion uncertainty consistent with NRC guidance. Therefore, the NRC staff considers determination of the depletion uncertainty acceptable.

3.4.3.2 Axial Apportionment of the Burnup or Axial Burnup Profile Another important aspect of fuel characterization is the selection of the axial burnup profile. At the beginning of life, a PWR fuel assembly will be exposed to a near-cosine axial-shaped flux, which will deplete fuel near the axial center at a greater rate than at the ends. As the reactor continues to operate, the cosine flux shape will flatten because of the fuel depletion and fission-product buildup that occurs near the center. Near the fuel assembly ends, burnup is suppressed due to neutron leakage. If a uniform axial burnup profile is assumed, the burnup at the ends is over-predicted. Analysis discussed in NUREG/CR-6801, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis (ADAMS Accession No. ML031110292), has shown that, at assembly burnups above about 10 to 20 gigawatt-days per metric ton of uranium (GWd/MTU), the use of a uniform axial burnup profile results in an under-prediction of Keff.

Generally, the under-prediction becomes larger as burnup increases. This is what is known as the end effect. Proper selection of the axial burnup profile is necessary to ensure Keff is not under-predicted due to the end effect.

The DSS-ISG-2010-01 recommends using the bounding axial burnup profiles from NUREG/CR-6801 or using limiting site-specific profiles. The licensees NCS analyses used the site-specific approach. The analyses evaluated numerous site axial burnup profiles based on burnup in the regions of the fuel where the end effect is prominent and ran calculations on the axial burnup profiles deemed most likely to be limiting. The analyses then used the limiting axial burnup profile from those calculations, consistent with the recommendation in DSS-ISG-2010-

01. Therefore, the NRC staff considers the licensees treatment of axial burnup profiles to be acceptable.

3.4.3.3 Radial Burnup Distribution Due to the neutron flux gradients in the reactor core, assemblies can show a radially tilted burnup distribution (i.e., differences in burnup between portions or quadrants of the cross section of the assembly). The licensees NCS analyses did not consider the effect of radial burnup distribution on reactivity.

The impact of radial burnup gradients may be estimated by comparing the distribution of radial burnup tilt information provided in Figure 3-4 of the U.S. Department of Energy document, DOE/RW-0496, Horizontal Burnup Gradient Datafile for PWR Assemblies, with information on

the sensitivity of Keff to radial burnup tilt provided in Section 6.1.2 of NUREG/CR-6800, Assessment of Reactivity Margins and Loading Curves for PWR Burnup-Credit Cask Designs (ADAMS Accession No. ML031110280). From DOE/RW-0496, the maximum quadrant deviation from assembly average burnup had been observed to be less than 25 percent at low assembly average burnups (burnup < 20 GWd/MTU) and was observed to decrease with burnup, generally being less than 10 percent at burnups above 20 GWd/MTU. Combining these radial tilt bounding estimates with the Keff sensitivity information provided in NUREG/CR-6800, the NRC staffs review of these radial burnup tilts indicate that Keff could increase by as much as 0.002 k. Based on the above, the NRC staff finds that its potential impact is small. The NRC staff also finds that the licensees analysis indicates sufficient margin to its licensing basis Keff to accommodate the small effect of radial burnup distribution on reactivity, and is, therefore, acceptable.

3.4.3.4 Burnup History/Core Operating Parameters The analysis in NUREG/CR-6665, Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel (ADAMS Accession No. ML003688150), provides some discussion on the treatment of depletion analysis parameters that determine how the burnup was achieved.

While NUREG/CR-6665 is focused on NCS analysis in storage and transportation casks, the basic principles with respect to the depletion analysis apply generically, since the phenomena occur in the reactor as the fuel is being depleted. The results have some applicability to the licensees NCS analyses. The basic strategy for this type of analysis is to select parameters that maximize the Doppler broadening/spectral hardening of the neutron field, resulting in maximum plutonium-239/241 production. NUREG/CR-6665 discusses six parameters affecting the depletion analysis: core fuel temperature, core moderator temperature, core soluble boron concentration, specific power and operating history, fixed burnable poisons, and integral burnable poisons. While the mechanism for each is different, the effect is similar (i.e., Doppler broadening/spectral hardening of the neutron field resulting in increased plutonium-239/241 production). NUREG/CR-6665 provides an estimate of the reactivity worth of these parameters.

The largest effect appears to be due to moderator temperature. NUREG/CR-6665 approximates the moderator temperature effect in an infinite lattice of high burnup fuel to be 90 percent mille per degree Kelvin (°K). Thus, a 10°F change in moderator temperature used in the depletion analysis would result in 0.005 k. The effects of each core operating parameter typically have a burnup or time dependency. The NUREG/CR-6665 parameters affecting the depletion analysis are discussed below.

Core Fuel Temperature The licensees NCS analyses used the FIGHTH code documented in Westinghouse WCAP-9522, FIGHTH - A Simplified Calculation of Effective Temperatures in PWR Fuel Rods for Use in Nuclear Design May 1979 (FIGHTH) (ADAMS Accession No. ML20269A413) to determine the core fuel temperature during the depletion portion of the analysis. FIGHTH calculates the steady state radial temperature distribution at each burn up, given the local value of the heat generation rate in the rod, the moderator temperature, and coolant flow rate. The licensee used the limiting core operating temperature profiles discussed above and the minimum RCS flow.

These would produce reasonably limited core fuel temperature profiles. Therefore, the NRC staff considers core moderator temperature profiles used to be acceptable for this application.

Core Moderator Temperature The licensees analytical method used to determine its limiting core moderator temperature profiles is similar to the method used to determine limiting axial burnup profiles. The derived core moderator temperatures are reasonably limiting for Farley current operation and fuel designs. Therefore, the NRC staff considers core moderator temperature profiles used to be acceptable for this application.

Core Soluble Boron Concentration A common practice in the depletion portion of SFP NCS analyses has been to use a constant soluble boron concentration rather than a time-dependent soluble boron letdown curve. This simplifies the modeling significantly. The use of a constant cycle average soluble boron for the depletion modeling has been shown to be conservative in the paper by J. C. Wagner, Impact of Soluble Boron Modeling for PWR Burnup Credit Criticality Safety Analyses, Transactions of the American Nuclear Society, 89, pp. 120 (2003). The work in the Wagner paper models three identical complete cycles and compares the effect of modeling a constant cycle-average soluble boron with use of a linear time-dependent soluble boron letdown curve. The work shows that using a constant cycle average soluble boron for the depletion modeling vs a time dependent soluble boron letdown curve is conservative. The licensee is using the constant soluble boron concentration in a manner consistent with the Wagner study. Therefore, the NRC staff finds the licensees use of a constant cycle average soluble boron for the depletion modeling is acceptable.

Fixed and Integral Burnable Absorbers In the past, the licensee used the Westinghouse Integral Fuel Burnable Absorber (IFBA) and two types of fixed burnable absorbers: Pyrex burnable absorbers and the Westinghouse Wet Annular Burnable Absorber (WABA). The licensee analyzed the integral and fixed absorbers separately. Based on the staffs review, the NRC staff finds that the analysis for each is reasonably conservative, and is, therefore acceptable.

3.4.4 Rod Cluster Control Assembly (RCCA) Usage If RCCAs are present in assemblies for significant amounts of time in the reactor, the associated spectral hardening can increase plutonium generation, leading to higher fuel reactivity for the same burnup. The licensees NCS analyses did not model RCCA inserts explicitly for their impact on post-irradiation reactivity. The LAR states, Any assemblies incurring significant rodded operation going forward must not credit the rodded burnup. However, the analysis does not fully define what would constitute significant rodded operation. In its supplement dated April 13, 2020, the licensee clarified that significant rodded operation would be outside its historical normal practice of operating with all rods out, with exceptions for necessary short durations during normal plant operations such as startup and shutdowns and short unplanned transients. Significant rodded operation would include Flexible Power Operations. The NRC finds this clarification acceptable.

3.4.5 Credited Isotopes Modeling every isotope from the chart of the nuclides is impracticable and unnecessary as most would have no impact on the estimation of criticality. The importance of a particular isotope to criticality depends on its macroscopic cross-section for neutron absorption and/or fission. While

they are all physically present, there is a diminishing impact on the estimation of keff as those with lesser importance are included in the analysis. NUREG CR-7108, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety AnalysesIsotopic Composition Predictions (ADAMS Accession No. ML12116A124), and NUREG CR-7109, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety AnalysesCriticality (keff ) Predictions (ADAMS Accession No. ML12116A128) list 28 isotopes as the most important. Table 2.1 of WCAP-18414-P (Attachment 4 of the LAR) contains the list of isotopes that are included in the licensees analysis.

The NRCs evaluation determined that licensees analysis was sufficient in considering those isotopes most important in NUREG/CR-7108 (ADAMS Accession No. ML12116A124) and NUREG/CR-7109 (ADAMS Accession No. ML12116A128). The list could have included hundreds more. Omission from the list represent a slight conservatism as those not listed are primarily weak neutron absorbers. Therefore, the NRC staff considers the isotopes used in the licensees analysis to be acceptable.

3.4.6 Eccentricity of the Fuel Within the Storage Cell The base Keff calculation models all fuel assemblies in the center of their respective storage cell.

However, the fuel assemblies can be anywhere within their respective storage cell. The eccentricity portion of the analysis is intended to determine the reactivity effect of the fuel assemblies being in other than the center of their storage cell. While the number of locations a fuel assembly could be within its storage cell is numerous, there is no practical difference between most. From an analytical perspective, the NRC staff considers it reasonable to assume the fuel assembly could be in five locations within the storage cell; the center and each corner.

WCAP-18414-P, Revision 0, (Attachment 4 to the LAR) Section 5.2.3.1.13, Eccentric Fuel Assembly Positioning Bias, describes the analysis the licensee performed to consider the eccentricity of position of fuel assemblies within a storage cell. The licensee stated that, the fuel assemblies are assumed to be nominally located in the center of the storage rack cell however, it is recognized that an assembly could in fact be located eccentrically within its storage cell. However, the licensee did not fully describe the models used or the number of cases considered. In its supplement dated April 13, 2020, the licensee provided more details regarding its eccentric positioning analysis. The licensees supplement provided sufficient analysis to determine its eccentric positioning bias. Therefore, the NRC staff finds the licensees eccentric positioning bias acceptable.

3.5 Non-Standard Fuel Configurations/Reconstituted Fuel The Farley SFPs has several non-standard fuel configurations. These are discussed in WCAP-18414-P, Revision 0, (Attachment 4 to the LAR) Section 5.4, Normal Conditions, and its subsections. The licensee considered reconstitution of fuel where a fuel rod is replaced one at a time with either inert rod or another fuel rod. PWR fuel assemblies are under moderated; removing a fuel rod increases the moderation of the fuel assembly. Having one fuel rod location empty will have a virtually imperceptible impact on reactivity; having several fuel rod locations empty will have a significant impact on reactivity. The licensees submittal confirms only one fuel rod is removed at a time. The licensee also has failed fuel rod storage cannisters (FRSC).

The FRSC hold a limited number of fuel rods and due to the limited number of fuel rods these components are over-moderated and typically non-limiting with regard to storage in the SFP.

The licensees analysis for the licensee showed conservatism for Farley, Unit 1 and 2. In

addition, the licensee has loose pellet transport cannisters. As with the FRSC, these components are typically over-moderated and non-limiting. The licensees analysis showed conservatism for Farley, Unit 1 and 2, for these components as well. The licensees treatment of these aspects of SFP criticality are reasonable and conservative. Therefore, the NRC staff finds the licensees proposed control of non-standard fuel configurations and reconstituted fuel acceptable.

3.6 Determination of Soluble Boron Requirements The licensees analysis for the determination of soluble boron requirements for normal conditions is in WCAP-18414-P, Revision 0 (Attachment 4 to the LAR), Section 5.5.1. The licensee determined from linear interpolation that the minimum soluble boron concentration to maintain Keff < 0.95 is 270 parts per million (ppm) of soluble boron. The licensee maintains a conservative administrative margin of 320 ppm of soluble boron. Therefore, NRC staff finds the minimum soluble boron concentration of 320 ppm continues to be acceptable in normal conditions.

The licensees analysis for the determination of soluble boron requirements for accident conditions is in WCAP-18414-P, Revision 0 (Attachment 4 to the LAR), Section 5.5.2. The licensee considered several potential accident scenarios. The licensee found the multiple misloading of fuel assemblies to be limiting. The Farley TS 3.7.14 require the SFP to have at least 2000 ppm of soluble boron whenever fuel assemblies are stored in the SFP. The licensee showed 2000 ppm of soluble boron in the SFP is sufficient to maintain Keff less than or equal to 0.95 at a 95-percent probability, 95-percent confidence level, if flooded with borated water under credible accident scenarios. Therefore, the NRC staff finds that the minimum soluble boron concentration of 2000 ppm continues to be acceptable for accident conditions.

3.7 Conclusion Based on the above, the NRC staff concludes that there is reasonable assurance that the Farley, Unit 1 and 2, SFP meets the applicable regulatory requirements in GDC 62 and 10 CFR 70.24. The July 31, 1996, exemption to 10 CFR 70.24 continues to be Farleys licensing basis.

Additionally, the NRC staff determined that the proposed TSs would continue to be based on the analyses and evaluations included in the UFSAR and amendments thereto in accordance with 10 CFR 50.36(b). The NRC staff determined the LCO for TS 3.7.15 would continue to require lowest functional capability or performance levels of equipment required for safe operation of the facility in accordance with 10 CFR 50.36(c)(2). The NRC staff determined SR 3.7.15.1 would continue 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, in accordance with 10 CFR 50.36(c)(3). The NRC staff also determined that the proposed TSs will continue to include required design features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety, in accordance with 10 CFR 50.36(c)(4). Therefore, the NRC staff has determined that the proposed TSs 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 7, 2020. On July 16, 2020, the State official confirmed 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, and there has been no public comment on such finding (85 FR 733) January 7, 2020. 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: K. Wood Date: October 6, 2020

ML20196L929 *via e-mail OFFICE DORL/LPL2-1/PM* DORL/LPL2-1/LA* DSS/SFNB/BC* DNRL/NCSG/BC*

NAME SWilliams KGoldstein RLukes SBloom DATE 08/10/2020 08/07/2020 08/03/2020 08/10/2020 OFFICE DSS/STSB/BC (A)* OGC-NLO* DORL/LPL2-1/BC* DORL/LPL2-1/PM*

VCusumano (MHamm NAME AGoshNaber MMarkley SWilliams for)

DATE 08/10/2020 09/29/2020 10/06/2020 10/06/2020

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