Text

Issuance of Amendment Regarding Revision to Watts Bar Unit 2 Technical Specification 4.2.1 Fuel Assemblies, and Watts Bar Units 1 and 2 Technical Specifications Related to Fuel Storage
	ML18347B330
Person / Time
Site:	Watts Bar
Issue date:	05/22/2019
From:	Robert Schaaf; Plant Licensing Branch II
To:	James Shea, Skaggs M; Tennessee Valley Authority
	Schaaf R, NRR/DORL/LPL2-2, 415-6020
References
	EPID L-2017-LLA-0427
	Download: ML18347B330 (94)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D,C 20555-0001

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2- ISSUANCE OF AMENDMENT REGARDING REVISION TO WATTS BAR NUCLEAR PLANT, UNIT 2, TECHNICAL SPECIFICATION 4.2.1, "FUEL ASSEMBLIES," AND WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2, TECHNICAL SPECIFICATIONS RELATED TO FUEL STORAGE (EPID L-2017-LLA-0427)

Dear Mr. Shea:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 125 to Facility Operating License No. NPF-90 and Amendment No. 27 to Facility Operating License No. NPF-96 for the Watts Bar Nuclear Plant (WBN), Units 1 and 2, respectively. These amendments are in response to your application dated December 20, 2017, as supplemented February 15, April 9, and October 4, 2018.

The amendments revise Technical Specification 4.2.1, "Fuel Assemblies," for WBN, Unit 2, to allow up 1,792 tritium producing burnable absorber rods in the reactor; and revisions to the WBN, Units 1 and 2, TSs related to fuel storage.

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

Sincerely,

µJj~

Robert G. Schaaf, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-390 and 50-391

Enclosures:

1. Amendment No. 125 to NPF-90

2. Amendment No. 27 to NPF-96

3. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O,C, 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 125 License No. NPF-90

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

A. The application for amendment by Tennessee Valley Authority (TVA, the licensee) dated December 20, 2017, as supplemented February 15, April 9, and October 4, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-90 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 125 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance, and shall be implemented prior to startup from the outage where any number of tritium producing burnable absorber rods is inserted in the Watts Bar Nuclear Plant, Unit 2 reactor core, not to exceed December 31, 2022.

FOR THE NUCLEAR REGULATORY COMMISSION Undine Shoop, Chief Plant Licensing Branch 11-2 Division of operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: May 22, 2019

ATTACHMENT TO AMENDMENT NO. 125 WATTS BAR NUCLEAR PLANT, UNIT 1 FACILITY OPERATING LICENSE NO. NPF-90 DOCKET NO. 50-390 Replace the following pages of the Facility Operating License No. NPF-90 and the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Facility Operating License No. NPF-90 REMOVE INSERT Technical Specifications REMOVE INSERT 3.7-31 3.7-31 3.7-39 3.7-40 3.9-16 3.9-16 4.0-2 4.0-2 4.0-3 4.0-3 4.0-9 4.0-10 5.0-25a 5.0-25a

(4) TVA, 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, instrument calibration; or other activity associated with radioactive apparatus or components; and (5) TVA, 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 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 TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 125 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Safety Parameter Display System (SPDS) (Section 18.2 of SER Supplements 5 and 15)

Prior to startup following the first refueling outage, TVA shall accomplish the necessary activities, provide acceptable responses, and implement all proposed corrective actions related to having the Watts Bar Unit 1 SPDS operational.

(4) Vehicle Bomb Control Program (Section 13.6.9 of SSER 20)

During the period of the exemption granted in paragraph 2.D.(3) of this license, in implementing the power ascension phase of the approved initial test program, TVA shall not exceed 50% power until the requirements of 10 CFR 73.55(c)(7) and (8) are fully implemented. TVA shall submit a letter under oath or affirmation when the requirements of 73.55(c)(7) and (8) have been fully implemented.

Facility License No. NPF-90 Amendment No. 125

Spent Fuel Pool Assembly Storage 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Assembly Storage LCO 3.7.15 The initial enrichment of each fuel assembly stored 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 LCO A.1 ------------NOTE---------------

not met. LCO 3.0.3 is not applicable.

Initiate action to move the Immediately noncomplying fuel assembly.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify by administrative means the initial enrichment Prior to storing the of the fuel assembly is in accordance with fuel assembly.

Specification 4.3.1.1.

Watts Bar-Unit 1 3.7-31 Amendment 6, 40, 1 25

Fuel Storage Pool Boron Concentration 3.7.18

3. 7 PLANT SYSTEMS

3. 7 .18 Fuel Storage Pool Boron Concentration LCO 3. 7 .18 The fuel storage pool boron concentration shall be :::: 2300 ppm APPLICABILITY: When fuel assemblies are stored in the fuel storage pool ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool --------------------NO TE--------------------

boron concentration LCO 3.0.3 is not applicable.

not within limit.

A.1 Initiate action to restore fuel Immediately storage pool boron concentration to within limit.

Watts Bar-Unit 1 3.7-39 Amendment . 1 2 5

Fuel Storage Pool Boron Concentration 3.7.18 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify the fuel storage pool boron concentration is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months within limit.

Watts Bar-Unit 1 3.7-40 Amendment 1 2 5

Spent Fuel Pool Boron Concentration 3.9.9 3.9 REFUELING OPERATIONS 3.9.9 Spent Fuel Pool Boron Concentration LCO 3.9.9 Boron concentration of the spent fuel pool shall be ~ 2300 ppm.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration A.1 Initiate action to restore fuel Immediately not within limit. storage pool boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.9.1 Verify boron concentration in the spent fuel pool is ~ 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 2300 ppm.

Watts Bar-Unit 1 3.9-16 Amendment 125

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks (shown in Figure 4.3-1) are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent (wt%) (nominally 4.95 +/- 0.05 wt% U-235);

b. kett :5 0.95 if fully flooded with 2300 ppm borated water, which, includes an allowance for uncertainties as described in Sections 4.3.2.7 and 9.1 of the FSAR, and a kett less than critical when flooded with unborated water;

c. Distances between fuel assemblies are a nominal 10.375 inch center-to-center spacing in the twenty-four flux trap rack modules. *

(continued)

Watts Bar Unit 1 4.0-2 Amendment 6, 40, 95, 1 2 5

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage (continued)

A water cell is less reactive than any cell containing fuel and therefore a water cell may be used at any location in the loading arrangements. A water cell is defined as a cell containing water or non-fissile material.

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

a. Fuel assemblies having a maximum enrichment of 5.0 weight percent U-235 and shall be maintained with the arrangement of 120 storage locations shown in Figure 4.3-2;

b. kett s 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the FSAR;

c. kett s 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 9.1 of the FSAR; and

d. A nominal 21-inch center to center distance between fuel assemblies placed in the storage racks.

(continued)

Watts Bar Unit 1 4.0-3 Amendment 6, 40, 1 2 5

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.20 Control Room Envelope Habitability Program (continued)

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate defined in the Ventilation Filter Testing Program (VFTP), at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

5.7.2.21 Spent Fuel Storage Rack Neutron Absorber Monitoring Program This Program provides controls for monitoring the condition of the neutron absorber used in the spent fuel pool storage racks to verify the Boron-10 areal density is consistent with the assumptions in the spent fuel pool criticality analysis. The Program shall be in accordance with NEI 16-03-A, "Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools," Revision 0, May 2017.

Watts Bar-Unit 1 5.0-25a Amendment 70, 78, 1 2 5

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-391 WATTS BAR NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 27 License No. NPF-96

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

A. The application for amendment by Tennessee Valley Authority (TVA, the licensee) dated December 20, 2017, as supplemented February 15, April 9, and October 4, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) and 2.C.(12) of Facility Operating License No. NPF-96 are hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 27 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. Accordingly, by Amendment No. 27, Facility Operating License No. NPF-96 is amended to authorize revision to the Updated Final Safety Analysis Report (UFSAR), as set forth in the application dated December 20, 2017, as supplemented by letters dated February 15, April 9, and October 4, 2018. The licensee shall update the UFSAR to incorporate the replacement of the containment isolation thermal relief check valves on the Unit 2 supply lines to the containment for the Component Cooling Water System and Essential Raw Cooling Water System with simple relief valves prior to loading tritium producing burnable absorber rods (TPBARs) in the Unit 2 reactor core.

4. This license amendment is effective as of the date of its issuance, and shall be implemented prior to startup from the outage where any number of TPBARs is inserted in the Watts Bar Nuclear Plant, Unit 2 reactor core not to exceed December 31, 2022.

FOR THE NUCLEAR REGULATORY COMMISSION

~

Undine Shoop, Chief ~

Plant Licensing Branch 11-2 Division of operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: May 22, 201 9

ATTACHMENT TO AMENDMENT NO. 27 WATTS BAR NUCLEAR PLANT, UNIT 2 FACILITY OPERATING LICENSE NO. NPF-96 DOCKET NO. 50-391 Replace the following pages of the Facility Operating License No. NPF-96 and the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Facility Operating License No. NPF-96 REMOVE INSERT Technical Specifications REMOVE INSERT 3.7-30 3.7-30 3.7-37 3.7-38 3.9-12 3.9-12 4.0-1 4.0-1 4.0-2 4.0-2 4.0-3 4.0-3 4.0-8 4.0-9 5.0-27 5.0-27

C. The 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 TVA is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 27 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) TVA shall implement permanent modifications to prevent overtopping of the embankments of the Fort Loudon Dam due to the Probable Maximum Flood by June 30, 2018.

(4) PAD4TCD may be used to establish core operating limits until the WBN Unit 2 steam generators are replaced with steam generators equivalent to the existing steam generators at WBN Unit 1.

(5) By December 31, 2019, the licensee shall report to the NRC that the actions to resolve the issues identified in Bulletin 2012-01, "Design Vulnerability in Electrical Power System," have been implemented.

(6) The licensee shall maintain in effect the provisions of the physical security plan, security personnel training and qualification plan, and safeguards contingency plan, and all amendments made pursuant to the authority of 10 CFR 50.90 and 50.54(p).

(7) TVA shall fully implement and maintain in effect all provisions of the Commission approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The TVA approved CSP was discussed in NUREG-0847, Supplement 28, as amended by changes approved in License Amendment No. 7.

(8) TVA shall implement and maintain in effect all provisions of the approved fire protection program as described in the Fire Protection Report for the facility, as described in NUREG-0847, Supplement 29, subject to the following provision:

Facility Operating License No. NPF-96 Amendment No. 27

Spent Fuel Pool Assembly Storage 3.7.15 3.7 PLANT SYSTEMS

3. 7 .15 Spent Fuel Pool Assembly Storage LCO 3.7.15 The initial enrichment of each fuel assembly stored 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 LCO A.1 -------NOTE------

not met. LCO 3.0.3 is not applicable.

Initiate action to move the Immediately noncomplying fuel assembly.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 7 .15.1 Verify by administrative means the initial enrichment Prior to storing the of the fuel assembly is in accordance with fuel assembly.

Specification 4.3.1.1.

Watts Bar- Unit 2 3.7-30 Amendment 2 7

Fuel Storage Pool Boron Concentration 3.7.18

3. 7 PLANT SYSTEMS 3.7.18 Fuel Storage Pool Boron Concentration LCO 3.7.18 The fuel storage pool boron concentration shall be~ 2300 ppm APPLICABILITY: When fuel assemblies are stored in the fuel storage pool ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool ------------NOTE------------

boron concentration LCO 3.0.3 is not applicable.

not within limit.

A.1 Initiate action to restore fuel Immediately storage pool boron concentration to within limit.

Watts Bar-Unit 2 3.7-37 Amendment 2 7 I

Fuel Storage Pool Boron Concentration 3.7.18 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 7 .18.1 Verify the fuel storage pool boron concentration is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months within limit.

Watts Bar-Unit 2 3.7-38 Amendment 2 7

Spent Fuel Pool Boron Concentration 3.9.9 3.9 REFUELING OPERATIONS 3.9.9 Spent Fuel Pool Boron Concentration LCO 3.9.9 Boron concentration of the spent fuel pool shall be ~ 2300 ppm.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration A.1 Initiate action to restore Immediately not within limit. fuel storage pool boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.9.1 Verify boron concentration in the spent fuel pool is

~ 2300 ppm. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Watts Bar - Unit 2 3.9-12 Amendment 27

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site 4.1.1 Site and Exclusion Area Boundaries The site and exclusion area boundaries shall be as shown in Figure 4.1-1.

4.1.2 Low Population Zone (LPZ)

The LPZ shall be as shown in Figure 4.1-2 (within the 3-mile circle).

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zirlo fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. For Unit 2, Watts Bar is authorized to place a maximum of 1792 Tritium Producing Burnable Absorber Rods into the reactor in an operating cycle.

4.2.2 Control Rod Assemblies The reactor core shall contain 57 control rod assemblies. The control material shall be silver indium cadmium as approved by the NRC.

(continued)

Watts Bar - Unit 2 4.0-1 Amendment 2 7

Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks (shown in Figure 4.3-1) are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent (wt%) (nominally 4.95 +/- 0.05 wt% U-235);

b. kett s 0.95 if fully flooded with 2300 ppm borated water, which, includes an allowance for uncertainties as described in Sections 4.3.2.7 and 9.1 of the FSAR, and a ketr less than critical when flooded with unborated water;

c. Distances between fuel assemblies are a nominal 10.375 inch center-to-center spacing in the twenty-four flux trap rack modules.

(continued)

Watts Bar - Unit 2 4.0-2 Amendment 2 7

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued)

A water cell is less reactive than any cell containing fuel and therefore a water cell may be used at any location in the loading arrangements.

A water cell is defined as a cell containing water or non-fissile material.

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

a. Fuel assemblies having a maximum enrichment of 5.0 weight percent U-235 and shall be maintained with the arrangement of 120 storage locations shown in Figure 4.3-2;

b. ke1t::;; 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the FSAR;

c. keff::;; 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 9.1 of the FSAR; and

d. A nominal 21-inch center to center distance between fuel assemblies placed in the storage racks.

4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below Elevation 747 feet - 1 1/2 inches.

4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1386 fuel assemblies in 24 flux trap rack modules.

Watts Bar - Unit 2 4.0-3 Amendment 2 7

Procedures, Programs, and Manuals 5.7

5. 7 Procedures, Programs, and Manuals 5.7.2.20 Control Room Envelope Habitability Program (continued)

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate defined in the Ventilation Filter Testing Program (VFTP), at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE.

These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences.

Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

5.7.2.21 Spent Fu~I Storage Rack Monitoring Program This Program provides controls for monitoring the condition of the neutron absorber used in the spent fuel pool storage racks to verify the Boron-10 areal density is consistent with the assumptions in the spent fuel pool criticality analysis. The Program shall be in accordance with NEI 16-03-A, "Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools," Revision 0, May 2017.

Watts Bar - Unit 2 5.0-27 Amendment 2 7

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 125 AND 27 TO FACILITY OPERATING LICENSE NOS. NPF-90 AND NPF-96 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-390 AND 50-391

1.0 INTRODUCTION

By letter dated December 20, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML173548282), as supplemented February 15, April 9, and October 4, 2018 (ADAMS Accession Nos. ML18047A181, ML18100A953, and ML18283A107, respectively),

the Tennessee Valley Authority (TVA or the licensee), submitted a license amendment request (LAR) for changes to the Watts Bar Nuclear Plant (WBN), Units 1 and 2, Technical Specifications (TSs). The requested changes would revise WBN, Unit 2, TS 4.2.1, "Fuel Assemblies," to authorize up to 1,792 tritium producing burnable absorber rods (TPBARs) that can be irradiated in the reactor. In addition, the requested changes would revise the WBN, Unit 1 and 2, TSs related to fuel storage.

Specifically, the requested changes would revise WBN, Units 1 and 2, TS 3. 7 .15, "Spent Fuel Assembly Storage," to simplify the fuel storage limitations on fuel assemblies by eliminating the burnup-related criteria. The requested changes would add WBN, Units 1 and 2, TS 3.7.18, "Fuel Storage Pool Boron Concentration," to specify the minimum fuel storage pool boron concentration when fuel is stored in the pool. The requested changes would revise WBN, Units 1 and 2, TS 3.9.9, "Spent Fuel Pool Boron Concentration," to modify the minimum fuel storage pool boron concentration during refueling operations when fuel is stored in the pool. The requested changes would revise WBN, Units 1 and 2, TS 4.3, "Fuel Storage," to replace the storage limitations on fuel assembly burnup and storage with a single requirement to maintain a specified boron concentration in the spent fuel pool (SFP). The requested changes would add WBN, Units 1 and 2, TS 5.7.2.21, "Spent Fuel Storage Rack Neutron Absorber Monitoring Program."

Enclosure 3

The letter dated December 20, 2017, contained a proprietary report titled, Hl-2177876, "Licensing Report for the Criticality Safety Analysis of the Watts Bar Nuclear Plant Spent Fuel Pool," dated September 25, 2017. This proprietary report was withheld from public disclosure. A nonproprietary version of the report is contained in Enclosure 4 of the letter dated December 20, 2017.

The supplement dated October 4, 2018, provided additional information that clarified the application, and did not expand the scope of the application as originally noticed in the Federal Register on June 8, 2018 (83 FR 26709).

2.0 REGULATORY EVALUATION

The primary regulatory requirement being used to evaluate this LAR can be found in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46, "Acceptance criteria for emergency core cooling systems [(ECCS)] for light-water nuclear power reactors." This regulation requires adequate core cooling following a loss-of-coolant accident (LOCA) such that specified acceptance criteria are satisfied. In particular, the core temperature must be maintained at an acceptably low value by appropriate removal of decay heat for the extended cooling period required by the long-lived radioactive nuclides in the core.

The post-LOCA long-term core cooling analysis for WBN, Unit 2, requires that the core remain subcritical when all sources of liquid are injected and mixed in the containment sump. This includes the liquid from: (1) the reactor coolant system (RCS), (2) the cold-leg accumulators (CLAs), (3) the reactor water storage tank (RWST), (4) the melted liquid from the ice condenser, and (5) any other possible sources of liquid. The resulting boron concentration must be sufficient to preclude criticality in the core, assuming cold conditions and the most reactive time in the cycle.

The increase in the number of TPBARs loaded in the core will affect the criticality of the core, because an increased U-235 loading in the core is necessary to support the increase in TPBARs, due to the strong neutron-absorbing properties of the TPBARs.

Section 50.61 of 10 CFR, "Fracture toughness requirements for protection against pressurized thermal shock events," contains requirements to prevent potential failure of the reactor vessel as a result of postulated pressurized thermal shock events. In particular, this rule discusses the use of a screening criterion evaluated at the projected fluence experienced by the reactor vessel during its service lifetime.

Section 50.68 of 10 CFR, "Criticality Accident Requirements," states that if the licensee does not credit soluble boron in its criticality analysis, the k-effective (kett) of the SFP storage racks must not exceed 0.95 at a 95 percent probability, 95 percent confidence level. The kett is defined as the effective neutron multiplication factor.

Section 50.44 of 10 CFR, "Combustible gas control for nuclear power reactors," contains requirements for monitoring and controlling the concentration of combustible gases in containment during a design-basis accident (OBA).

Section 50.36 of 10 CFR, "Technical specifications," includes requirements for the contents of the TS. This may include design features limiting the number of TPBARs that can be loaded into the

core, as well as surveillance requirements (SRs) that ensure that the limiting conditions for operation (LCOs) will be met.

Section 50.36(c)(3) of 10 CFR states that 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 LCOs will be met.

Section 50.34(b )(3) of 10 CFR is applicable as it pertains to describing "the kinds and quantities of radioactive materials expected to be produced in the operation and the means for controlling and limiting radioactive effluents and radiation exposures within the limits set forth in [1 O CFR]

part 20 .... "

Part 20 of 10 CFR is applicable as it pertains to ensuring that radiation doses are within the dose limits for occupational workers and members of the public, and are as low as is reasonably achievable (ALARA).

Part 50 of 10 CFR, Appendix I, is applicable as it pertains to ensuring that the routine radioactive effluent releases are within the design objectives to meet the ALARA criterion.

Section 50.120 of 10 CFR, "Training and qualification of nuclear power plant personnel," states that each holder of an operating license shall establish, implement, and maintain a training program that is derived from a systems approach to training and provides for the training and qualification of nuclear power plant personnel.

Section 100.11 of 10 CFR, "Determination of exclusion area, low population zone, and population center distance," requires, in part, that the licensee determine:

(1) An exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem [roentgen equivalent man] or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

(2) A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

Section 50.67 of 10 CFR, "Accident source term," requires, in part, that:

(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent {TEDE), (ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release {during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose

equivalent {TEDE), and (iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent {TEDE) for the duration of the accident.

Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements," specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.

Appendix H to 10 CFR Part 50, "Reactor Vessel Material Surveillance Program," requires licensees to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of light water nuclear power reactors which result from exposure of these materials to neutron irradiation and the thermal environment. Under the program, fracture toughness test data are obtained from material specimens exposed in surveillance capsules, which are withdrawn periodically from the reactor vessel.

Appendix A to 10 CFR Part 50, "General Design Criteria [GDC] for Nuclear Power Plants,"

establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety. According to Section 3.1.1 of the WBN, Unit 2, Updated Final Safety Analysis Report (UFSAR), the WBN plant was designed to meet the intent of the "Proposed General Design Criteria for Nuclear Power Plant Construction Permits" published in July 1967. The WBN construction permit was issued in January 1973. The WBN plant, in general, meets the intent of the NRC GDC published as Appendix A to 10 CFR Part 50 in July 1971, as discussed in UFSAR Section 3.1.2. The following GDCs were used to evaluate this LAR.

GDC 10, "Reactor Design," requires the reactor core and associated reactor coolant, control, and protection systems to 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 (AOOs).

GDC 15, "Reactor coolant system design," requires the RCS and associated auxiliary, control, and protection systems to be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including AOOs.

GDC 19, "Control room," states, in part:

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of

5 rem [0.05 Sv] whole body, or its equivalent to any part of the body, for the duration of the accident.

GDC 27, "Combined reactivity control system capability," requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system (ECCS), of reliably controlling reactivity changes under postulated accidents conditions.

GDC 61, "Fuel Storage and Handling and Radioactivity Control," which states that "These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety."

GDC 62, "Prevention of Criticality in Fuel Storage and Handling," which states that "Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations."

Section 9.1, "Fuel Storage and Handling," of the WBN UFSAR describes spent fuel storage, fuel pool cooling, fuel handling, and TPBAR consolidation activities. The SFP is a reinforced concrete, Seismic Category I structure that is located in the auxiliary building and shared between Units 1 and 2. The SFP contains 24 stainless steel storage rack structures providing 1,386 storage locations. The SFP cooling and cleanup system (SFPCCS) provides two cooling trains to maintain acceptable pool temperatures by removing decay heat generated in the stored fuel assemblies and other equipment. The SFPCCS piping is arranged such that failure of any piping segment would not drain the SFP below the water level necessary for radiation shielding. Fuel assemblies are removed from the reactor during refueling using the refueling machine and transferred to the fuel transfer system for horizontal movement through the transfer tube from the containment refueling canal to the auxiliary building fuel transfer canal. The SFP bridge crane moves fuel assemblies within the SFP and, during refueling, from the upender in the fuel transfer canal to storage locations in the SFP. The SFP bridge crane removes irradiated TPBAR assemblies from fuel assemblies in the SFP and transfers the TPBAR assemblies underwater to the cask loading area for consolidation into canisters. Consolidation canisters are stored in the SFP racks until the canisters are loaded into TPBAR transport casks in the cask loading area.

The auxiliary building overhead crane is used to move the TPBAR transport casks from the cask loading area to the transportation vehicle for shipment.

The following sections of NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," were used to evaluate this LAR.

Chapter 4, "Reactor," provides guidance for the review of fuel rod cladding materials, the fuel system, the design of the fuel assemblies and control systems, and the thermal and hydraulic design of the core.

Chapter 4, Section 4.2, "Fuel System Design," Revision 3 (ADAMS Accession No. ML070740002), provides guidance for the review to provide assurance that:

o the fuel system is not damaged as a result of normal operation and AOOs,

o fuel system damage is never so severe as to prevent control rod insertion when it is required, o the number of fuel rod failures is not underestimated for postulated accidents, and o coolability is always maintained.

Chapter 9, Section 9.1.1, "Criticality Safety of Fresh and Spent Fuel Storage and Handling," Revision 3 (ADAMS Accession No. ML070570006), provides guidance regarding the specific acceptance criteria and review procedures to ensure that the proposed changes satisfy the requirements in 10 CFR 50.68 and GDC 62.

Chapter 9 Section 9.1.2, "New and Spent Fuel Storage," Revision 4 (ADAMS Accession No. ML070550057), provides guidance regarding the specific acceptance criteria and review procedures to ensure that the proposed changes satisfy the requirements in 10 CFR 50.68.

Chapter 11, "Radioactive Waste Management," provides guidance and acceptance criteria for determining the impact of the proposed change on plant effluent treatment systems and whether the ALARA design criteria of 10 CFR Part 50, Appendix I, are met.

Chapter 12, "Radiation Protection," provides guidance and acceptance criteria for determining whether radiation protection design feature, and programs, are sufficient to ensure that requirements of 10 CFR Part 20 are met such that there is reasonable assurance that occupational doses, and doses to members of the public, will be maintained within the limits, and will be ALARA.

Chapter 13, Section 13.2.1, "Reactor Operator Requalification Program; Reactor Operator Training," Revision 4 (ADAMS Accession No. ML15006A035), provides guidance for reviewing the impact of the proposed change on the operator training program. In addition, Section 13.5.2.1, "Operating and Emergency Operating Procedures," Revision 2 (ADAMS Accession No. ML070100635), provides guidance for reviewing the impact of the proposed change on operating and emergency operating procedures.

Chapter 15, "Introduction - Transient and Accident Analysis," Revision 3 (ADAMS Accession No. ML070710376), provides guidance for verifying that the proposed change is bounded by previous license amendments approved by the NRC with respect to the remaining analysis acceptance criteria for transient and accident analyses.

Chapter 15, Section 15.1.5, Appendix A, "Radiological Consequences of Main Steam Line Failures Outside Containment of a PWR [Pressurized-Water Reactor]," Revision 2 (ADAMS Accession No. ML052350118), provides guidance regarding the specific acceptance criteria and review procedures to ensure that the proposed changes comply with GDCs 13, 17, 27, 28, 31, and 35.

Chapter 15, Section 15.6.3, "Radiological Consequences of Steam Generator Tube Failure (PWR)," Revision 2 (ADAMS Accession No. ML052350149),

provides guidance regarding the specific acceptance criteria and review procedures to ensure that the proposed changes satisfy the requirements in 10 CFR Part 100.

Chapter 15, Section 15.6.5, Appendix A, "Radiological Consequences of a Design Basis Loss-of-Coolant Accident including Containment Leakage Contribution,"

Revision 1 (ADAMS Accession No. ML052350158), provides guidance regarding the specific acceptance criteria and review procedures to ensure that the proposed changes satisfy the requirements in 10 CFR 100, 10 CFR 50.46 and compliance with GDCs 13 and 35.

Chapter 18, "Human Factors Engineering," Revision 3 (ADAMS Accession No. ML16125A114), provides guidance for Human Factors Engineering Program reviews.

The final Division of Safety Systems (DSS) Interim Staff Guidance (ISG), DSS-ISG-2010-01, Revision 0, "Staff guidance regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools (ADAMS Accession No. ML110620086), provides updated guidance to the NRC staff reviewer to address the increased complexity of recent spent fuel pool (SFP) nuclear criticality analyses and operations. The ISG DSS-ISG-2010-01 references NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology (J.C. Dean, R.W. Tayloe, Jr., "Guide for Validation of Nuclear Criticality Safety Calculational Methodology," U.S. Nuclear Regulatory Commission, Science Applications International Corporation, January 2001) (ADAMS Accession No. ML050250061 ).

NUREG/CR-6698 states, in part, that:

In general, the critical experiments selected for inclusion in the validation must be representative of the types of materials, conditions, and operating parameters found in the actual operations to be modeled using the calculational method. A sufficient number of experiments with varying experimental parameters should be selected for inclusion in the validation to ensure as wide an area of applicability as feasible and statistically significant results.

The NRC staff used the following regulatory guidance documents in its review:

Regulatory Guide (RG) 1.24, "Assumptions used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure,"

Revision O (ADAMS Accession No. ML083300020).

RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Revision O (ADAMS Accession No. ML003716792).

RG 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," Revision O (ADAMS Accession No. ML031490640).

RG 8.32, "Criteria for Establishing a Tritium Bioassay Program," Revision O (ADAMS Accession No. ML003739479).

NUREG-1764, "Guidance for the Review of Changes to Human Actions," Revision 1 (ADAMS Accession No. ML072640413).

NUREG-0711, "Human Factors Engineering Program Review Model," Revision 3 (ADAMS Accession No. ML12324A013).

NUREG-1672, "Safety Evaluation Report Related to the Department of Energy's Topical Report

[(TR)] on the Tritium Production Core," March 1999 (ADAMS Accession No. ML992810127).

TR Nuclear Energy Institute (NEI) 16-03-A, "Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools" (ADAMS Accession No. ML17263A133). This TR provides guidance endorsed by the NRC staff to the industry for developing adequate monitoring programs for fixed neutron absorbers in SFPs.

3.0 TECHNICAL EVALUATION

3.1 Technical Evaluation Introduction The LAR proposed to revise WBN, Unit 2, TS 4.2.1, "Fuel Assemblies," to authorize irradiation of up to 1,792 TPBARs in the Unit 2 core per cycle.

The LAR also proposed to change WBN, Units 1 and 2, TSs related to fuel storage. Specifically, the requested changes are to revise WBN, Units 1 and 2, TS 3.7.15, "Spent Fuel Assembly Storage," to simplify the fuel storage limitations on fuel assemblies by eliminating the burnup-related criteria. The requested changes add WBN, Units 1 and 2, TS 3. 7.18, "Fuel Storage Pool Boron Concentration," to specify the minimum fuel storage pool boron concentration when fuel is stored in the pool. The requested changes revise WBN, Units 1 and 2, TS 3.9.9, "Spent Fuel Pool Boron Concentration," to modify the minimum fuel storage pool boron concentration during refueling operations when fuel is stored in the pool. The requested changes revise WBN, Units 1 and 2, TS 4.3, "Fuel Storage," to replace the storage limitations on fuel assembly burnup and storage with a single requirement to maintain a specified boron concentration in the SFP. The requested changes add WBN, Units 1 and 2, TS 5.7.2.21, "Spent Fuel Storage Rack Neutron Absorber Monitoring Program."

3.2 Technical Analyses This LAR is justified by the licensee based on analysis, testing, and evaluation of the TPBARs as reported previously by the DOE. DOE has previously submitted a classified/proprietary version of the Tritium Production Core (TPC) Topical Report NDP-98-153, Revision 1, and an unclassified/non-proprietary version, NDP-98-181, Revision 1 (ADAMS Accession No. ML16077A093) for NRC review. The NRC staff reviewed these TPC Topical Reports and issued NUREG-1672, "Safety Evaluation Report Related to the Department of Energy's Topical Report on the Tritium Production Core." The plant-specific interface items from NUREG-1672 are addressed below for WBN, Unit 2, as well as the accident analyses to support the TS changes in Section 3.3 below.

3.2.1 Specific Assessment of Hydrogen Source and Timing of Recombiner Operation The WBN hydrogen mitigation system (HMS) is designed to enhance the containment's capability to accommodate combustible gases that could be released during a OBA. This system has been engineered to be redundant, capable of remaining operational in a post-accident environment, and capable of actuation by the operators from the main control room. In addition, the HMS is designed to have an ample number of igniters distributed throughout the containment to mitigate the effects of combustible gases being released in containment during a severe accident as required by 10 CFR 50.44.

The accumulation of combustible gas in the containment building following an accident can be the result of production from several sources. The potential sources of combustible gases are zirconium-water reaction, corrosion of construction materials, and radiolytic decomposition of the emergency core cooling solution.

The NRC staff reviewed the HMS description in the WBN, Unit 2, UFSAR Section 6.2.5 to verify that the licensee's plant-specific assessment has been completed. The HMS provides adequate coverage by distributing 68 igniters throughout the various regions of the containment in which combustible gases could be released or flow in significant quantities. Eight of the HMS igniters are scattered on the reactor cavity wall exterior and crane wall interior at a midpoint elevation to ensure the partial burning that accompanies upward flame movement.

The NRC staff have determined that due to the igniter type and locations, redundancy, main control room actuation, and remote surveillance the HMS performs its intended function in a manner that provides adequate safety margins. The NRC staff verified that the containment building can survive the effects of a OBA when hydrogen hazards are mitigated by HMS.

Based on the conservative assessment of the TPBARs contribution to the combustible gas concentrations in containment following a loss-of-coolant accident (LOCA), the combustible gas control systems are not expected to be affected by the proposed increase in TPBARs. As discussed above, the combustible gas control requirements of 10 CFR 50.44 will continue to be met.

3.2.2 Anticipated Transient Without Scram (ATWS)

An ATWS is an anticipated operational occurrence during which an automatic reactor scram is required, but fails to occur because of a common mode fault in the reactor protection system (RPS).

The final NRC ATWS rule requires that Westinghouse-designed plants install ATWS mitigation system circuitry (AMSAC) to initiate a turbine trip and actuate auxiliary feedwater flow independent of the RPS. The WBN AMSAC design is described in WBN UFSAR Section 7.7.1.12. To meet the requirements of 10 CFR 50.62, the AMSAC must be designed to perform its function in a reliable manner and be independent of the existing reactor trip system (from sensor output to interruption of power to the control rods). The existing reactor trip system is composed of the Westinghouse Eagle 21 process protection system, and the Westinghouse Solid State Protection System.

The Westinghouse licensing implementation strategy that addresses the requirements in 10 CFR 50.62 was documented in topical report WCAP-15831-NP-A, Revision 2 (ADAMS Accession No. ML072550560). In this topical report, operating plants were placed into three separate groups sorted according to their ATWS Configuration Management Program. The WBN, Units 1 and 2, are within Group 2, which states that they do not have a Diverse Scram System, but instead have a Moderator Temperature Coefficient {MTC) that is consistent with the basis for the ATWS rule. As such, the implementation of the NRG-approved ATWS analysis methodology is not required. Instead, the full power MTCs will continue to be maintained within the limits specified in the Core Operating Limits Report and LCO 3.1.4 which requires the maximum upper limit to be

s; 0 L\k/k°F (upper limit) for all times in core life.

The NRC staff determined that WBN, Unit 2, will meet the 10 CFR 50.62 AMSAC requirements and lies within the analysis basis of the ATWS rule. The licensee will verify that WBN, Unit 2, will continue to remain within the class of generic ATWS analyses by verifying that 95-percent of the MTC value is sufficiently negative to preclude the RCS from exceeding the acceptance limit during an ATWS event in accordance with NUREG-0847, Supplement 24, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant Unit 2," September 2011 (ML11277A148).

Based on a review of the supporting topical reports and confirmation that the licensee has satisfied the regulatory commitments of 10 CFR 50.44 and 10 CFR 50.46, the NRC staff concludes that reasonable assurance exists that the WBN, Unit 2, will, with the proposed TS modifications, continue to meet the applicable regulatory requirements.

3.2.3 Cold Overpressure Mitigating System (COMS)

The COMS is designed to prevent violations of the limits of Appendix G to 10 CFR Part 50 during low temperature operating conditions. This system is also known as the "low temperature overpressure protection system." The pressurizer power-operated relief valves and/or the residual heat removal (RHR) relief valves are used to protect the pressure/temperature limits during an overpressure transient caused either by a mass addition or heat addition to the RCS.

TS 5.9.6.b, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," states that the analytical methods used to determine the RCS pressure and temperature limits and COMS setpoints shall be those previously reviewed and approved by the NRC. The NRC staff has reviewed the licensees COMS methodology that was provided in Section 6 of WCAP-18191-NP, Revision O (ADAMS Accession No. ML17289A327).

WCAP-18191-NP is the NRG-approved Westinghouse methodology for developing RCS heatup and cooldown curves and COMS setpoints. This methodology may be referenced by licensees to implement the PTLR. The licensee will update the COMS methodology documents that are specified in TS 5.9.6.b prior to Cycle 4; the P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (ADAMS Accession No. ML050120209). Based on this review, the NRC staff has concluded that the requirements of 10 CFR 50.61 will continue to be met.

3.2.4 Nuclear Criticality Safety (NCS) 3.2.4.1 Background The WBN has one SFP that is shared for fuel storage between Units 1 and 2. In order to ensure that there are no adverse impacts on SFP NCS from either unit, the NCS analyses are performed using a design basis fuel assembly that bounds all fuel expected to be used in either unit. The SFP racks currently utilize the BORALTM neutron-absorbing material for criticality control, which is a composite material consisting of boron carbide particles and sintered aluminum, surrounded by an aluminum cladding. The SFP rack configuration and BORALTM material remain unchanged relative to the prior licensing basis for WBN. This LAR was submitted to revise the licensing basis for the WBN SFP to replace burnup and configuration controls with soluble boron as one of the primary criticality controls. Upon approval and implementation of this license amendment, WBN would no longer have to maintain controls to restrict how fuel can be loaded in the SFP; instead, they would be required to maintain a minimum soluble boron concentration in the SFP at all times.

Currently, they are only required to maintain a minimum soluble boron concentration in the SFP during refueling.

3.2.4.2 SFP NCS Analysis Method There is no comprehensive, NRG-approved generic methodology for performing NCS analyses for fuel storage and handling. The methods used for the NCS analysis for fuel in the WBN SFP are described in Attachment 4 to the letter dated December 20, 2017. A previous version of the report documenting the computer code benchmarking analyses supporting use of the MCNPS, Version 1.51 code (described below) for this application was reviewed and approved by the NRC (ADAMS Accession No. ML18204A286). One potential nonconservatism was identified during the review, but as will be discussed below, the NRC staff determined that sufficient margin exists to the regulatory limit to accommodate any expected reactivity impacts.

Computational Methods For the NCS calculation, the licensee used MCNPS, Version 1.51, with continuous energy cross-section data based on the Evaluated Nuclear Data File, Version 7 (ENDF/B-VII) neutron cross section library. MCNPS is a state-of-the-art Monte Carlo criticality code developed and maintained by Los Alamos National Laboratory for use in performing reactor physics and NCS analyses for nuclear facilities and transportation/storage packages. The use of MCNP has previously been approved for use in SFP NCS analyses for licensing purposes.

For all MCNPS, Version 1.51, calculations, the licensee used reasonable values for the following calculational parameters: number of histories per cycle, number of cycles skipped before averaging, and total number of cycles. The initial source distribution was only specified as being "in the fueled regions (assemblies)," however, given the uniformity of the fuel in the base models and number of skipped cycles, the exact source distribution is relatively unimportant as long as the initial neutrons are seeded within the active fuel. More importantly, the licensee confirmed that all calculations converged using appropriate checks.

Based on the pedigree of the computer codes and the use of appropriate inputs and verification methods to ensure that an accurate eigenvalue was determined, the NRC staff determined that the computational methods implicit in the codes used for the NCS analyses are acceptable.

Computer Code Validation The purpose of the criticality code validation is to ensure that appropriate code bias and bias uncertainty are determined for use in the NCS calculation based on available benchmarking data from critical experiments and other tests.

The NRC staff used NUREG/CR-6698 as guidance for review of the code validation methodology presented in the application. The basic elements of validation are outlined in NUREG/CR-6698, including identification of operating conditions and parameter ranges to be validated, selection of critical benchmarks, modeling of benchmarks, statistical analysis of results, and determination of the area of applicability.

The NRC has previously reviewed and accepted the results from this version of the criticality benchmarking report. However, the licensee also needed to demonstrate how the validation report's findings are applied for this specific analysis. The licensee identified the applicable operating conditions for the validation (e.g., fuel assembly materials and geometry, enrichment of fissile isotope, fuel density, types of neutron absorbers, moderators and reflectors, rack material, and physical configurations). The licensee compared the spectral parameters (e.g., energy of average lethargy causing fission, spectrum type) between the benchmarks and the WBN SFP conditions to demonstrate that the selected benchmarks are applicable.

The NRC staff noted that the validation data set contains experiments that include plutonium as a fissile material. The licensee performed their NCS analyses using fresh, unpoisoned fuel, so the data from these experiments are not applicable. However, the selection of the limiting bias and bias uncertainty included consideration of the subsets within the data set which only contain uranium as a fissile material. Therefore, the final values are verified to bound the design basis fuel assembly utilized by the licensee in the SFP NCS analyses documented in Attachment 4 of the letter dated December 20, 2017.

Based on the NRC staffs review of the validation database and its applicability to the compositions, geometries, and methodologies used in the licensee's NCS analyses, the NRC staff found the code validation acceptable with all identified biases and uncertainties propagated appropriately.

3.2.4.3 SFP and Fuel Storage Racks SFP Water Temperature The SFP water temperature was treated in a bounding manner. The design basis calculations were run using the minimum SFP temperature, and follow-up calculations were performed to verify that the maximum SFP temperature did not result in a higher kett value. In order to obtain appropriate cross sections for some of the intermediate temperatures, the NJOY code was used to adjust the standard cross sections to a specified temperature and to apply a molecular energy adjustment for certain materials. Other analyses approved by the NRC use a similar methodology

to account for the fact that the standard MCNP cross section libraries only contain data at specific temperatures for each nuclide, so this approach is an acceptable way to ensure that the cross sections used in the evaluation are reasonably accurate for the given temperatures. Since cross sections are used for the minimum SFP temperature MCNP calculations without the need to use NJOY, the NRC staff finds the elevated temperature calculations using cross sections generated with NJOY to be acceptable to confirm that a different temperature will not be more limiting.

The SFP water temperature was treated in a bounding manner, so the NRC staff finds the licensee's approach to be acceptable.

SFP Storage Rack Models WBN has multiple fuel storage racks clustered in the SFP, with water gaps between adjacent racks. In the NCS analysis, the licensee chose to use a bounding approach in which a fresh, unpoisoned lattice is identified as the design basis lattice. Further discussion about this lattice and how it was determined to bound all fuel stored in the SFP racks can be found below. The SFP is then assumed to be fully loaded with fuel assemblies that have this lattice along its entire axial length. Most of the calculations are performed utilizing a 2x2 storage rack model with periodic boundary conditions. In effect, the base model is that of a laterally infinite SFP. Water is modeled above and below the fuel assembly in lieu of structural material, which is conservative because the structural material typically absorbs some neutrons. Some modeling simplifications are discussed in Section 2.3. 7 of Attachment 4 to the letter dated December 20, 2017, but they all have a conservative or negligible impact on reactivity.

The licensee performed a series of sensitivity studies to evaluate various combinations of the fuel assembly position within the SFP cell, and the maximum reactivity difference was applied as a bias and the 95/95 uncertainty on this reactivity difference was statistically combined with the other uncertainties.

Since BORALTM is known to develop blisters, which displaces moderator and can cause an increase in reactivity, the licensee performed calculations to determine the potential reactivity impact due to displacement of moderator within the space in which the BORALTM panels are installed. A range of conditions was modeled, varying from no blistering to full displacement of moderator in the gap between the BORAL'. The results showed that the most limiting reactivity impact was from a situation where the moderator surrounding the BORALTM was fully displaced by void. The licensee incorporated this modeling into all design basis calculations. Operating experience has shown that when BORALTM blisters reach the SFP rack wall material, the aluminum is not strong enough to push it outward. Instead, the blisters flatten and grow radially to displace more moderator in the gap between the BORAL' and the SFP rack wall. Therefore, assuming that all moderator in the gap is displaced by void represents a very conservative bounding scenario given that operating experience has not shown blistering to the extent that it would cover the entire BORALTM panel. As such, this approach is acceptable.

The licensee also addressed the reactivity of a fuel assembly moving outside an SFP storage cell and a scenario where non-fuel hardware was stored in an SFP storage cell instead of fuel, since both are considered to be part of the normal operating conditions of the SFP. The former scenario was explicitly analyzed with a model of a single design basis fuel assembly surrounded by water.

The latter scenario was addressed by a qualitative discussion that the reactivity effect of a design

basis fuel assembly in an SFP cell on the adjacent SFP cells would be much greater than that of increased moderator or other materials in the SFP cell. Therefore, the design basis analysis bounds empty SFP cells or storage of non-fuel hardware in SFP cells. The NRG staff finds this disposition to be acceptable.

In order to analyze the mislocated fuel accident, an 8x8 fuel rack model is used. Section 2.6.5 of to the letter dated December 20, 2017, describes the model. The SFP wall is not modeled, however, the amount of water that would have to exist in order to fit a fuel assembly between the SFP rack and the SFP wall is sufficient to neutralize any neutron reflection effect that the wall may have. The small effect that the wall may have on the mislocated fuel assembly can readily be accommodated by the margin to the regulatory limit. Other than this modeling simplification, the model used for the mislocated fuel accident is an accurate representation of the local conditions for the postulated event. The modeling simplifications due to use of reflective boundary conditions only affect regions of the model that are too far away to have any significant impact on the calculated reactivity due to the mislocated fuel assembly.

The licensee has an acceptable Neutron Absorber Material (NAM) monitoring program, as described in Section 3.2.5 of this SE. Therefore, reasonable assurance exists that any significant deviation from the parameters as modeled in the NGS analysis will be identified and evaluated by the licensee.

The NRG staff has evaluated the relevant aspects of the storage rack modeling and found them to be modeled conservatively or using appropriate parameters. Other potential normal SFP operating conditions were adequately addressed. As a result, the storage rack modeling is acceptable.

SFP Storage Rack Models Manufacturing Tolerances and Uncertainties The manufacturing tolerances of the storage racks contribute to SFP reactivity.

DSS-ISG-2010-01 does not explicitly discuss the approach to be used in determining manufacturing tolerances, but past practice has been consistent with the Kopp letter (referenced in DSS-ISG-2010-01) that determination of the maximum kett should consider either: (1) a worst-case combination with mechanical and material conditions set to maximize kett, or (2) a sensitivity study of the reactivity effects of tolerance variations to determine reactivity uncertainties.

If used, a sensitivity study should include all possible significant tolerance variations in the material and mechanical specifications of the racks. The licensee chose to utilize the former approach with the key BORALTM parameters by utilizing minimum certified values for the boron-10 areal density, panel width, and panel thickness as the inputs to the design basis model. The licensee chose the second approach for the rack manufacturing tolerances by performing sensitivity studies to determine reactivity uncertainties based on bounding parameters based on the manufacturing tolerances for the relevant SFP storage rack modeling parameters: storage cell inner diameter, storage rack wall thickness, storage cell sheathing thickness, storage cell - BORALTM gap thickness, and storage cell flux trap. The NRG staff notes that the possible variations in storage cell pitch would be implicitly captured through consideration of these uncertainties. The reactivity uncertainties were then statistically combined and included in the final kett summation.

As a result of evaluating the licensee's treatment of the manufacturing tolerances, the NRG staff has determined that the licensee treated all relevant BORALTM parameters in a bounding (i.e,

conservative) manner. Furthermore, the uncertainties associated with the rack manufacturing tolerances were treated in a manner consistent with approaches previously approved by the NRC.

Therefore, all manufacturing tolerances and uncertainties in the SFP storage rack models have been addressed in an acceptable manner.

SFP Storage Rack Interfaces There are multiple SFP rack modules of the same design installed in the WBN SFP. They are typically separated by small gaps which may become smaller if the racks move for some reason.

However, the licensee states that the storage cell racks are separated by at least 1 inch due to the rack baseplate extensions and bumper bars at the top. If the SFP racks are moved as close as physically possible, the rack baseplate extensions and bumper bars at the top would prevent the outside of the peripheral SFP cells from being less than 1 inch apart. This configuration would mimic that of the flux trap between adjacent SFP storage cells in an SFP rack, except that the flux trap is 0.973 inches wide. A smaller flux trap results in an increase in reactivity, therefore, the gap between SFP racks is bounded by the flux traps in the laterally infinite SFP array of the design basis model.

The possible reactivity impacts of the SFP storage rack interfaces were not explicitly evaluated by the licensee, but sufficient information was provided to allow the NRC staff to determine that the reactivity effect of the gap between adjacent SFP racks is bounded by the reactivity effect of the flux traps in the laterally infinite array of SFP cells for the design basis model.

3.2.4.4 Fuel Assembly Bounding Fuel Assembly Design The fuel stored in the WBN SFPs may be one of three different Westinghouse PWR fuel designs:

V5H, V+/P+, or RFA-2. The licensee performed a series of calculations that were essentially design basis models utilizing different fuel designs, all of which were modeled as fresh, unpoisoned fuel assemblies with 5.0 weight percent (w/o) U-235 enrichment (no radial or axial variation in enrichment). The bounding fuel density was utilized in all cases. The licensee selected the V+/P+ fuel assembly as the design basis fuel assembly design, which was not the case that yielded the highest reactivity. However, the case that did result in the highest reactivity, the V5H fuel assembly, was only 0.0001 delta-k more reactive than the V+/P+, which is statistically insignificant and overwhelmed by the underlying conservatism in the overall fuel assembly modeling.

Fuel rod reconstitution can have an impact on reactivity, however, the licensee did not address this impact in their analysis (and marked as part of the NEI 12-16 checklist in Appendix B to of the letter dated December 20, 2017, that fuel reconstitution is not included as part of the normal condition of the NCS analysis). Reconstituted fuel assemblies with replacement rods manufactured with lower U-235 enrichments or stainless steel will have a negative reactivity effect relative to a fresh 5.0 w/o U-235 rod. However, any other situations, such as missing fuel rods, would not necessarily be bounded by this NCS analysis. Since the licensee did not request approval for storage of reconstituted fuel in the SFP, these types of scenarios were not considered by the NRC staff as part of the review of this LAR.

The selection and modeling of the fuel lattice used in the NCS analyses was determined to bound all fuel intended for storage in the WBN SFP. This fuel lattice was used along the full active fuel length for a bounding design basis fuel assembly, with structural materials neglected. This is conservative, and therefore, the NRC staff finds it acceptable for use in an SFP NCS analysis intended to bound all fuel in the WBN SFP.

Fuel Assembly Manufacturing Tolerances and Uncertainties The manufacturing tolerances of the fuel assemblies contribute to SFP reactivity.

DSS-ISG-2010-01 does not explicitly discuss the approach to be used in determining manufacturing tolerances, but past practice has been consistent with the Kopp letter (referenced in DSS-ISG-2010-01) that determination of the maximum kett should consider either: ( 1) a worst-case combination with mechanical and material conditions set to maximize kett, or (2) a sensitivity study of the reactivity effects of tolerance variations. If used, a sensitivity study should include all possible significant tolerance variations in the material and mechanical specifications of the fuel assemblies. The licensee chose to utilize the former approach for the U-235 enrichment and the theoretical density for the fuel. The limiting direction for variation in the parameter is well-established, so the licensee simply used the upper bound for the manufacturing tolerance range in all calculations. For the other fuel manufacturing tolerances-the cladding thickness, fuel rod pitch, fuel pellet outer diameter, and the guide/instrument tube thickness-the licensee performed sensitivity studies to determine the reactivity uncertainties associated with each parameter. In each case, both extremes for the manufacturing tolerance range were evaluated to determine which value would result in the maximum reactivity. The resulting combination of worst-case reactivity increases were statistically combined and applied as an uncertainty to the final kett summation.

As a result of evaluating the licensee's treatment of the manufacturing tolerances, the NRC staff has determined that the licensee treated all relevant fuel assembly parameters in a bounding (i.e.,

conservative) manner. Furthermore, the uncertainties associated with the fuel manufacturing tolerances were treated in a manner consistent with approaches previously approved by the NRC.

Therefore, the NRC staff finds that all manufacturing tolerances and uncertainties in the fuel assembly models have been addressed in an acceptable manner.

Spent Fuel Characterization The licensee elected to utilize a fresh, unpoisoned fuel lattice to construct the design basis fuel assembly for evaluation of the WBN SFP. As discussed above, the uranium-235 enrichment is sufficient to bound all fuel stored in the WBN SFP. The cumulative reactivity impact due to burnup related uncertainties will be much smaller than the reactivity decrease due to the depletion of U-235, so no burnup related uncertainties need to be evaluated for the approach that the licensee selected.

Integral Burnable Absorbers The licensee used a design basis fuel assembly with no burnup or burnable poison. Since the reactivity of fresh fuel with no poison is much larger than the reactivity of depleted fuel with poison after accounting for any uncertainties related to fuel depletion and gadolinia (whether they are

caused by manufacturing or depletion), no further disposition is necessary for the burnable absorbers.

3.2.4.5 Analysis of Abnormal Conditions Section 2.6 of Attachment 4 to the letter dated December 20, 2017, presents the abnormal conditions considered in the analysis. The licensee considered the following abnormal conditions:

SFP temperature exceeding the normal range

Dropped fuel assembly

Misloaded fuel assembly (fuel assembly loaded in the wrong location in a storage rack)

Mislocated fuel assembly (fuel assembly positioned outside the storage rack)

Rack movement due to seismic activity Three of these accidents are bounded by the design basis analysis. The licensee demonstrated that the most reactive temperature for the SFP is the minimum temperature, therefore, increases in SFP temperature will reduce reactivity relative to the design basis analysis. The design basis model includes a fresh, unpoisoned fuel assembly that bounds all fuel assemblies stored in the SFP. This assembly is loaded in all SFP cells, therefore, any misloaded fuel assemblies would also be bounded by the design basis model. Finally, the design basis model bounds the minimum physically possible gap size between adjacent racks; therefore, a seismic event cannot move the racks close enough together to cause the design basis model to become nonconservative.

The licensee states that the dropped fuel assembly scenarios are bounded by the design basis model, given that they generally involve only one fuel assembly and the resulting fuel arrangements would be bounded by the design basis model. In the horizontal dropped fuel scenario, a fuel assembly lying on top of the SFP racks would not be neutronically coupled to the fuel in the SFP storage cells, so no significant reactivity impact would be expected. A vertically dropped fuel assembly that enters a storage cell would cause, at most, a slight compression of any fuel assembly already stored in the storage cell which is not expected to have a significant effect on reactivity. The fuel assembly structural material above and below the active fuel length would prevent the fuel assemblies from moving close enough together to become neutronically coupled.

As a result, the NRC staff agrees that the dropped fuel assembly scenarios are bounded by the design basis model.

The licensee evaluated a mislocated fuel scenario where a design basis fuel assembly (which bounds all fuel stored in the WBN SFP) is located outside the fuel storage racks, along the periphery of the SFP racks. This model did not include the concrete SFP wall, but any reactivity effect due to neutron reflection from the SFP wall would be expected to be small enough to be accommodated by the available margin to the regulatory limit. The reactivity increase due to this scenario is mainly a consequence of the fact that no flux trap exists along the periphery of the SFP rack and there is only a single BORALTM panel between fuel stored in the SFP racks and the mislocated fuel assembly. The licensee found that 500 parts per million (ppm) of soluble boron in the SFP was sufficient to ensure that the kett for the SFP remains below 0.95 in this scenario. The TS will require a much higher soluble boron concentration in the SFP at all times (2300 ppm), and the double contingency principle means that a mislocated fuel assembly accident concurrent with a severe boron dilution accident does not need to be considered. Therefore, this analysis is

acceptable to demonstrate regulatory compliance for the only accident condition that is not bounded by the design basis analysis for normal conditions.

The NRC staff finds that the licensee performed a thorough evaluation of all potential accident conditions and dispositioned or explicitly calculated all possible reactivity impacts, so the assessment for postulated accident conditions is acceptable.

3.2.4.6 Disposition of Potential Nonconservatisms/Margins The neutron reflection effect from the concrete walls surrounding fuel storage racks are known to have a significant effect on reactivity for storage configurations that primarily depend on neutron leakage for criticality control, such as new fuel vaults where the fuel assemblies are separated by significant distances. For high density SFP racks with a uniform loading of fuel, this effect is not important because the region driving the reactivity of the configuration is the interior regions of the rack, where fuel assemblies are surrounded by other fuel assemblies on all sides. The mislocated fuel assembly accident is a configuration where the reactivity is driven by a region with high neutron leakage (since this configuration was calculated to have a higher reactivity than the design basis model, so the geometry around the mislocated fuel assembly is higher reactivity than the rack interior). Therefore, neutron reflection due to the SFP wall may have some impact on the reactivity. The licensee did not include the SFP wall in the modeling of the mislocated fuel assembly accident or justify the omission of the SFP wall. However, the overall reactivity of this configuration is driven by local conditions rather than global conditions. Therefore, only a small portion of the SFP wall would be involved and the reactivity impact is expected to be small. The licensee shows 0.0073 delta-k margin to the regulatory limit, which is sufficient to offset any small potential reactivity impact due to neutron reflection from the SFP wall.

The licensee included some studies to determine the reactivity margins due to some of the assumptions incorporated in the design basis model. These additional margins were not necessary to make a finding on regulatory compliance. However, the NRC staff notes that any reactivity change due to use of nominal values instead of values that bound manufacturing tolerances would still need to be included in the analysis uncertainty. Additionally, the amount of conservatism inherent in the approach used to bound any blistering on the BORALTM material would be dependent on the actual condition of the BORALTM panels. The licensee's margin evaluation assumes that there is no blistering on the BORALTM panels, which is a non-conservative assumption. However, the results from the margin evaluation were not included in the criticality analysis to demonstrate regulatory compliance and do not affect the conclusions of the criticality analysis. Therefore, the potential non-conservatisms in the assumptions underlying the margin evaluation do not affect the NRC staffs safety finding.

3.2.5 Neutron Absorbing Material At WBN the SFP racks currently use BORALTM as the neutron absorbing material (NAM).

BORALTM is comprised of a boron carbide and aluminum matrix core sandwiched in between aluminum cladding. It is currently commonly used throughout the operating fleet of commercial reactors. The use of BORALTM by the operating fleet has provided operating experience on the chemical compatibility of BORALTM and the SFP environment. The NRC staff has previously determined that the use of BORALTM in SFPs is acceptable and that BORALTM is generally chemically compatible with the SFP environment.

3.2.5.1 Neutron Absorber Program Licensee Description The licensee has proposed a NAM Monitoring Program, which would serve to monitor the condition of the NAMs installed in the SFP storage racks. This would help to ensure that subcriticality margin in the SFP is maintained. The program would be based on NRG-approved TR NEI 16-03-A, "Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools."

The NRC staff has reviewed the NAM Monitoring Program proposed by the licensee and has determined it is acceptable. This is because the program will follow the guidance found in NRC approved TR NEI 16-03-A. This TR contains acceptable guidance for the development of an effective NAM monitoring program.

3.2.6 Loss-Of-Coolant Accident (LOCA)

The core inventory used in the current WBN, Unit 2, licensing basis has been revised to reflect a Tritium Producing Core (TPC); see Table 4.1-3 in the letter dated December 20, 2017. In the letter dated December 20, 2017, the licensee stated that the changes made to WBN, Unit 2, core inventory are the same as the WBN, Unit 1, core inventory that was approved in license Amendment No. 40; it consisted of calculating the core inventory utilizing ORIGEN 2.1, and assumes that all tritium in the TPBARs is released to the environment. The licensee assumed that all of the tritium content of 2,304 TPBARs is released to the containment atmosphere after occurrence of a LOCA. This is based on a design inventory 1.2 grams of tritium per TPBAR and results in a total of 2.68E+07 curies of tritium in the core. The rest of the core inventories were determined based on a 96 feed equilibrium cycle, which consisted of 96 once-burned assemblies, 96 twice-burned assemblies, and one thrice-burned assembly. The inventory of each set of fuel assemblies was then summed together to determine the total core inventory. The TPC inventory is based on 102-percent of the licensed power (3,411 megawatts thermal (MWt)) and it includes ECCS uncertainty. The licensee stated that all other parameters remain the same as that documented in Section 15.4.1 of NUREG-0847, Supplement 25 with the exception of the control room isolation time.

The licensee proposed to use the control room parameters and the atmospheric dispersion factors in their current licensing basis as reflected in NUREG-0847, Supplement 25, except for the control room isolation time, which has been corrected to account for an error. The licensee stated:

The control room radiation monitor loops utilize the RP-30AM analog rate meter.

A time constant of 7.17E-3 minutes was previously used to determine the rate meter response time, which would be appropriate for a count rate between 1E4 and 1E5 counts per minute (cpm). However, the setpoint for these monitors is 400 cpm; thus, a time constant of 4.34E-1 minutes should have been used. This resulted in an increase in the rate meter response time from 0.86 seconds to 52.08 seconds. Combined with the response times determined for the remainder of the loop, the total loop response time increased from 6.6 seconds to 57.8 seconds.

The analyses rounded this to 60 seconds. The isolation damper response time remains unchanged.

The licensee evaluated two cases. One case analyzes a single failure, such that one train of the Emergency Gas Treatment System (EGTS) fails from the beginning of the accident. The second case analyzes a single failure in the controls of the EGTS, such that one set of EGTS dampers is assumed to be in the full exhaust position.

The licensee evaluated the radiological consequences resulting from the postulated LOCA and concluded that the radiological consequences at the exclusion area boundary, the low population zone, and the control room comply with the limits provided in 10 CFR 100.11 and 10 CFR Part 50, Appendix A, GDC 19. The NRC staff review found that the licensee used analyses, assumptions, and inputs consistent with applicable regulatory guidance. The other assumptions previously found acceptable to the NRC staff are presented in NUREG-0847, Supplement 25. The licensee's calculated dose results are given below in Table 1. The NRC staff performed independent, confirmatory dose evaluations as necessary to ensure a thorough understanding of the licensee's methods. The NRC staff finds, with reasonable assurance, that the licensee's estimates of the dose consequences of a design basis LOCA will comply with the requirements of 10 CFR 100.11, 10 CFR Part 50 Appendix A, GDC 19, and the accident specific dose guidelines specified in Regulatory Guide (RG) 1.195, and are, therefore, acceptable.

Table 1 Radiological Consequences for LOCA SinQle Emeri ency Gas Treatment Failure System Case Regulatory Exclusion Low Regulatory Control Limit for Area Population Limit for Dose (rem)

Room Control Boundary Zone EAB and Room (EAB) (LPZ) LPZ Whole Body 0.887 5 2.07 1.89 25 Beta 7.49 30 1.14 2.26 300 Thyroid 3.62 30 38.7 13.8 300 PCO Control Failure Case Whole Body 1.07 5 2.42 2.30 25 Beta 9.10 30 1.38 2.61 300 Thyroid 3.09 30 30.3 11.9 300 3.2. 7 Fuel Handling Accident (FHA)

The core inventory used in the current WBN, Unit 2, licensing basis has been revised to reflect a TPC by using the activity per fuel assembly determined for the TPC; see Table 4.1-3 in the letter dated December 20, 2017. The core inventories were determined based on a 96 feed equilibrium cycle, which consisted of 96 once-burned assemblies, 96 twice-burned assemblies, and 1 thrice-burned assembly. The TPC inventory is based on 102-percent of the 3411 MWt licensed power, and it includes ECCS uncertainty. In the letter dated December 20, 2017, the licensee stated that the tritium activity assumed to be released to the environment (21,122.5 curies) is the same that was assumed for WBN, Unit 1, as approved in license Amendment No. 107, which assumes that 25-percent of all the tritium content of 24 TPBARs (84,490 curies) is released into the surrounding water after an FHA. This is the maximum number of TPBARs that would be in one fuel assembly based on WBN, Unit 1, Amendment No. 92, "Issuance of Amendment to Allow Selective Implementation of Alternate Source Term to Analyze the Dose Consequences

Associated with Fuel-Handling Accidents (TAC No. ME8877)" dated June 19, 2013 (ADAMS Accession No. ML13141A564). This maximum number of TPBARs in one fuel assembly is the same for WBN, Unit 2.

Section 15.4.5 of NUREG-0847, Supplement 25, the licensing basis for WBN, Unit 2, describes two cases for the FHA that are evaluated using the assumptions from RG 1.183. The first case considered an FHA inside the SFP area located in the auxiliary building. The scenario assumes that the reactor building purge ventilation system is operating and releases through the shield building exhaust vent until it isolates at 12. 7 seconds, and then the remainder of the release is through the auxiliary building vent. The second case is an open containment case for an FHA inside containment where there is open communication between the containment and the auxiliary building. This scenario assumes the release is through the auxiliary building vent with no credit for any filtration systems.

For WBN, Unit 2, the licensee stated that all other parameters remain the same as that documented in Section 15.4.5 of NUREG-0847, Supplement 25 with the exception of the control room isolation time. The licensee proposed a new control room isolation delay time, which is increased from 40 seconds to 74 seconds to correct an error in how the delay time was determined. The corrected control room isolation delay time of 74 seconds is used in both the FHA in t.he SFP area and FHA in containment analyses. The NRC staff reviewed the FHA with regard to the new tritium source term and the corrected control room isolation delay time.

During the review the NRC staff noticed that NUREG-0847, Supplement 25, Table 15.5, assumptions for FHA inside closed containment, and Table 15.6, assumptions for FHA in the auxiliary building or in open containment, stated that the control room atmospheric dispersion factors are 2.59E-3 seconds per cubic meter and that these control room atmospheric dispersion factors are different from those stated in the letter dated December 20, 2017, for the FHA.

Therefore, the NRC staff asked the licensee to explain the differences in the control room atmospheric dispersion factors. In letter dated October 4, 2018, the licensee stated:

The atmospheric dispersion factor of (2.59E-3) listed in NUREG-0847 Supplement 25, Tables 15.5 and 15.6, is incorrect, because it reflects the main steam valve vault release point. The correct value for Table 15.5 is 1.09E-03, because the release point for the FHA scenario is the shield building exhaust vent. The correct value for Table 15.6 is 2.56E-03, because the release point for that FHA scenario is the auxiliary building vent. The information provided to the NRC by TVA letter dated September 23, 2011 (Reference 2 [ADAMS accession number ML11269A064]), as referenced in Section 15.4.5.1 of NUREG-0847 Supplement 25, correctly identifies the release points in Sections 15.5.6.1 and 15.5.6.2 of Reference 2. However, Table 15.5-14 of Reference 2 did not include the information for the auxiliary building release point. Table 4.1-7 of Reference 1

[ADAMS Accession No. ML17354B282] provided the correct information (i.e., the last column was properly labeled "WGDT/FHA" and two footnotes were added).

The NRC staff reviewed the licensee's response and agrees that information in NUREG-0847, Supplement 25, Tables 15.5 and 15.6, is incorrect and did not reflect the control room atmospheric dispersion factors for the correct release points and that the control room

atmospheric dispersion factors stated in the letter dated December 20, 2017, are correct for the FHA release points.

The licensee evaluated the radiological consequences resulting from the postulated FHA and concluded that the radiological consequences at the exclusion area boundary, the low population zone, and the control room comply with the reference values provided in 10 CFR 50.67 and the.

accident specific dose guidelines specified in RG 1.183. The NRC staff's review found that the licensee used analyses, assumptions, and inputs consistent with applicable regulatory guidance.

The other assumptions previously found acceptable to the NRC staff are presented in the Section 15.4.5 of NUREG-0847, Supplement 25 as corrected by the above control room atmospheric dispersion factors. The licensee's calculated dose results are given below in Table 2.

The NRC staff performed independent, confirmatory dose evaluations as necessary to ensure a thorough understanding of the licensee's methods.

The NRC staff finds, with reasonable assurance, that the licensee's estimates of the dose consequences of a design basis FHA will comply with the requirements of 10 CFR 50.67 and the accident specific dose guidelines specified in RG 1.183, and are, therefore, acceptable.

Table 2 Radiological Consequences for FHA Auxiliary Building FHA Regulatory Regulatory Control Limit for Limit for Dose (rem) EAB LPZ Room Control EAB and Room LPZ TEDE 2.39 5 2.83 0.792 6.25 Containment FHA TEDE 2.33 5 2.83 0.792 6.25 TPBAR Only FHA TEDE 1.16 5 0.288 0.0806 6.25 3.2.8 Main Steam Line Break (MSLB) and Steam Generator Tube Rupture (SGTR)

The current licensing basis primary and secondary coolant concentrations for WBN, Unit 2, are based on ANSI/ANS-18.1-1984. In the letter dated December 20, 2017, the licensee stated that TVA is correcting errors in its calculation of the primary and secondary coolant concentrations; see Tables 4.1-4 and 4.1-5 in the letter dated December 20, 2017. The following errors are corrected for WBN, Unit 2:

1. The reactor coolant system (RCS) volume was corrected from 11,375 cubic feet to 12,708.4 cubic feet.

2. The specific volume previously used to determine the RCS weight was based on a temperature outside the normal operating range.

3. The weight of RCS water previously included the volume of vapor space in the pressurizer.

4. The weight of water in the steam generator (SG) previously included the weight of water in the primary side of the steam generator instead of just the secondary side.

5. The condensate demineralizer was previously assumed to be in operation but is not typically used and should not have been credited.

In addition to the errors above, the MSLB and SGTR analyses for WBN, Unit 2, are revised to reflect a TPC by increasing the average tritium concentration in the primary and secondary coolant to .that associated with two TPBAR failures. The licensee proposed that the WBN, Unit 2, analyses assume 1,792 total TPBARs with an assumed permeation rate of 5 curies per TPBAR per year and two failed TPBARs. This results in a tritium concentration of 120 micro curies per gram.

For WBN, Unit 2, the licensee stated that all other parameters remain the same as that documented in Section 15.4.2 and 15.4.3 of NUREG-0847, Supplement 25 with the exception of the control room isolation time. The licensee proposed a new control room isolation delay time, which is increased from 40 seconds to 74 seconds to correct an error in how the delay time was determined. The corrected control room isolation delay time of 74 seconds is used in both the MSLB and SGTR analyses. The NRC staff reviewed these analyses with regard to the new tritium source term and the corrected control room isolation delay time.

The licensee evaluated the radiological consequences resulting from the postulated MSLB and SGTR accidents and concluded that the radiological consequences at the exclusion area boundary, the low population zone, and the control room comply with the limits provided in 10 CFR 100.11 and 10 CFR Part 50, Appendix A, GDC 19. The NRC staff's review has found that the licensee used analyses, assumptions, and inputs consistent with applicable regulatory guidance. The other assumptions previously found acceptable to the NRC staff are presented in Section 15.4.2 and 15.4.3 of NUREG-0847, Supplement 25. The licensee's calculated dose results are given below in Tables 3 and 4. The NRC staff performed independent, confirmatory dose evaluations as necessary to ensure a thorough understanding of the licensee's methods.

The NRC staff finds, with reasonable assurance, that the licensee's estimates of the dose consequences of a design basis MSLB and SGTR will comply with the requirements of 10 CFR 100.11, 10 CFR Part 50, Appendix A, GDC 19, and the accident specific dose guidelines specified in RG 1.195, and are, therefore, acceptable.

Table 3 Radiological Consequences for MSLB Pre-Accident Iodine Spike Regulatory Regulatory Control Limit for Limit for Dose (rem) EAB LPZ Room Control EAB and Room LPZ Whole Body 3.68E-03 5 0.025 0.0105 25 Beta 0.036 30 8.15E-03 4.03E-03 300 Thyroid 7.51 30 2.41 1.21 300 Coincident Iodine Spike Whole Body 7.85E-3 5 0.112 0.139 2.5 Beta 0.0645 30 0.027 0.0334 30 Thyroid 10.9 30 3.34 5.32 30

Table 4 Radioloaical Consequences for SGTR Pre-Accident Iodine Spike Regulatory Regulatory Control Limit for Limit for Dose (rem) EAB LPZ Room Control EAB and Room LPZ TEDE 0.0647 5 0.411 0.121 25 Whole Body 0.723 30 0.237 0.0726 300 Beta 13.1 30 14.4 4.13 300 Coincident Iodine Spike Whole Body 0.0627 5 0.639 0.188 2.5 Beta 0.728 30 0.285 0.0875 30 Thyroid 2.45 30 8.51 2.52 30 3.2.9 Loss of Offsite Power (LOOP)

The loss of offsite power (LOOP) radiological consequence analysis for WBN, Unit 2, is revised to reflect a TPC by increasing the average tritium concentration in the primary and secondary coolant to that associated with two TPBAR failures. The licensee proposed that the WBN, Unit 2, analyses assume 1,792 total TPBARs with an assumed permeation rate of 5 curies per TPBAR per year and two failed TPBARs. This results in a tritium concentration of 120 micro curies per gram.

For WBN, Unit 2, the licensee stated that all other parameters remain the same as that documented in Section 15.4.7 of NUREG-0847, Supplement 25 with the exception of the control room isolation time. The licensee proposed a new control room isolation delay time, which is increased from 40 seconds to 74 seconds to correct an error in how the delay time was determined. The NRC staff reviewed these analyses with regard to the new tritium source term and the corrected control room isolation delay time.

The licensee evaluated the radiological consequences resulting from the postulated LOOP event and concluded that the radiological consequences at the exclusion area boundary, the low population zone, and the control room comply with the reference values provided in 10 CFR 100.11 and 10 CFR Part 50, Appendix A, GDC 19 limits. The NRC staff's review has found that the licensee used analyses, assumptions, and inputs consistent with applicable regulatory guidance. The other assumptions previously found acceptable to the NRC staff are presented in Section 15.4.7 of NUREG-0847, Supplement 25. The licensee's calculated dose results are given below in Table 5. The NRC staff performed independent, confirmatory dose evaluations as necessary to ensure a thorough understanding of the licensee's methods. The NRC staff finds, with reasonable assurance, that the licensee's estimates of the dose consequences of a design basis LOOP event will comply with the requirements of 10 CFR 100.11, 10 CFR Part 50, Appendix A, GDC 19, and the accident specific dose guidelines specified in RG 1.195, and are, therefore, acceptable.

Table 5 Radiological Consequences for LOOP Regulatory Regulatory Control Limit for Limit for Dose (rem) EAB LPZ Room Control EAB and Room LPZ Whole Body 1.25E-04 5 3.74E-04 2.14E-04 25 Beta 1.72E-03 30 2.11 E-04 1.21E-04 300 Thyroid 3.58E-02 30 4.74E-02 2.71E-02 300 3.2.10 Waste Gas Decay Tank (WGDT) Rupture The WGDT rupture radiological consequence analysis is revised to reflect a TPC by using a tritium source term based on the bounding case of 2,500 TPBARs with a permeation rate of 10 curies per TPBAR per year and two TPBAR failures as described in NUREG-1672. As discussed above in NUREG-0847, Supplement 25, the licensee is correcting errors in their calculation of the primary and secondary coolant concentrations. For WBN, Unit 2, the licensee stated that all other parameters remain the same as that documented in Section 15.4.8 of NUREG-0847, Supplement 25 with the exception of the control room isolation time. The licensee proposed a new control room isolation delay time, which is increased from 40 seconds to 74 seconds to correct an error in how the delay time was determined. The NRC staff reviewed these analyses with regard to the new tritium source term and the corrected control room isolation delay time.

The licensee evaluated the radiological consequences resulting from the postulated WGDT rupture and concluded that the radiological consequences at the exclusion area boundary, the low population zone, and the control room comply with the reference values provided in 10 CFR 100.11, 10 CFR Part 50, Appendix A, GDC 19 limits, and the accident specific dose guidelines specified in RG 1.24. The NRC staff's review found that the licensee used analyses, assumptions, and inputs consistent with applicable regulatory guidance. The other assumptions previously found acceptable to the NRC staff are presented in the Section 15.4.8 of NUREG-084 7, Supplement 25. The licensee's calculated dose results are given below in Table 6. The NRC staff performed independent confirmatory dose evaluations as necessary to ensure a thorough understanding of the licensee's methods. The NRC staff finds, with reasonable assurance, that the licensee's estimates of the dose consequences of a postulated WGDT rupture will comply with the requirements of 10 CFR 100.11, 10 CFR Part 50, Appendix A, GDC 19, and the accident specific dose guidelines specified in RG 1.24 and are therefore, acceptable.

Table 6 Radiological Consequences for Waste Gas Decay Tank Rupture Regulatory Regulatory Control Limit for Limit for Dose (rem) EAB LPZ Room Control EAB and Room LPZ Whole Body 0.944 5 0.596 0.167 2.5 Beta 8.17 30 1.62 0.452 30 Thyroid 0.0108 30 0.0129 3.60E-03 30

3.2.11 Rod Ejection Accident The rod ejection accident is defined as the mechanical failure of a control rod mechanism pressure housing resulting in the ejection of a rod cluster control assembly and drive shaft. The consequence of this mechanical failure is a rapid, positive reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage. NUREG-0847, Supplement 25, Section 15.4.4, "Control Rod Ejection Accident," states:

The NRC staff evaluated the control rod ejection accident in Section 15.4.4 of the

. SER (NUREG-0847), and there have been no supplements to this section. The current WBN, Unit 2 FSAR Section 15.5, "Environmental Consequences of Accidents," does not include a detailed evaluation of this accident except to state that it is bounded by the LOCA. The LOCA dose consequence results for WBN, Unit 2 are less than 25 percent of the reference values in 10 CFR 100.11. Since the source term for a control rod ejection accident is considerably less than for a LOCA, and the dose consequence results for the WBN, Unit 2 LOCA are less than the SRP acceptance criteria for a rod ejection accident (25 percent of the reference values in 10 CFR 100.11 ), the staff concludes that the dose consequence for the rod ejection accident will be bounded by the LOCA for WBN, Unit 2.

In the letter dated September 25, 2015, TVA states:

As discussed in NUREG-0847, Supplement 25 (Reference 22), the source term for a rod ejection accident is considerably less than for a LOCA. Because the dose consequence results for the WBN, Unit 2 LOCA are less than the SRP acceptance criteria for a rod ejection accident (25 percent of the values in 10 CFR 100), the rod ejection accident is not explicitly analyzed.

The NRC staff reviewed NUREG-0847 and it does not include a detailed evaluation of the control rod ejection accident except to state that it is bounded by the LOCA. The LOCA dose consequence results for WBN, Unit 2, are less than 25 percent of the reference values in 10 CFR 100 .11. Since ( 1) fuel damage is not postulated to occur during the rod ejection accident, (2) the source term for a control rod ejection accident is considerably less than for a LOCA, and (3) the dose consequence results for the WBN, Unit 2, LOCA are less than the SRP acceptance criteria for a rod ejection accident (25 percent of the reference values in 10 CFR 100.11 ), the NRC staff concludes that the dose consequence for the rod ejection accident is bounded by the LOCA for WBN, Unit 2.

3.2.12 Source Term The letter dated December 20, 2017, describes two radioactive source terms used in evaluating the radiological impact on normal operations, and anticipated operational occurrences (AOOs), of TPC operations with 1,792 TPBARs.

The first of these is the conservative or design-basis source term. The design-basis source term assumes the maximum allowable release of radioactive material from the core to the reactor coolant due to reactor operations with fuel defects. This source term is used to evaluate the adequacy of plant design features such as shielding, ventilation, and radwaste system processing

capacities. The letter, dated December 20, 2017, assumed that the contribution to the design-basis source term of tritium permeating from the TPBARs will be 5 Ci/year/TPBAR. The assumption of 5 Ci/year/TPBAR is a conservative assumption of tritium permeation, because historical TPC operation at WBN, Unit 1, has never exceeded approximately 3.5 Ci/year/TPBAR.

The second source term, commonly called the realistic source term, is used to evaluate the plant effluents against the design criteria in 10 CFR Part 50, Appendix I, and to demonstrate compliance with the liquid and gaseous effluent concentration limits and dose limits to members of the public in 10 CFR Part 20. The licensee's realistic source term was based on guidance in NUREG-0017, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents for Pressurized Water Reactors," Revision 1, with an additional allowance for the tritium produced during TPC operation. The realistic source term does not assume maximum TS fuel leakage.

However, the licensee elected to have the realistic source term conservatively assume a tritium contribution from TPBAR operations of 5 Ci/year/TPBAR from 1,792 TPBARs loaded in the TPC. The realistic annual tritium value from NUREG-0017 was determined to be 1,392 Ci with an additional 8,960 Ci from TPC operation (i.e. 1,792 TPBARs at 5 Ci/year/TPBAR); this analysis yields a total average annual value of 10,352 Ci of tritium. In May 1999, the NRC staff issued NUREG-1672, "Safety Evaluation Report related to the DOE's topical report on the tritium production core" (ADAMS Legacy Accession No. 9907210095). Consistent with previous WBN, Unit 1, licensing actions, the licensee's realistic source term differs from that which was assumed in the DOE Topical Report in that the failure of two TPBARs is not considered to be a credible event during the plant's lifetime. Therefore, the failure of two TPBARs event does not fall within the scope of the definition of an AOO.

The NRC staff finds that the licensee's use of these source terms is adequate because it is consistent with prior licensing actions and reasonably conservative. Specifically, the use of 5 curies/year/TPBAR of tritium permeation is bounding of the 3.5 Ci/year/TPBAR maximum the licensee has measured during previous TPC operations, and the addition of 5 Ci/year/TPBAR to the realistic source term adds conservatism to effluent calculations. Finally, the licensee's approach of omitting an analysis of the failure of two TPBARs is adequate, because the NRC staff has found in prior licensing actions that this event is not credible during a plant's lifetime, thus it does not meet the description of an AOO, and the NRC staff finds this acceptable.

Based on the above, the NRC staff finds that there is reasonable assurance that WBN, Unit 2, can be operated at or below 10 CFR Part 20 limits while operating with the proposed maximum core loading of 1,792 TPBARs and, therefore, the NRC staff finds such operation is acceptable.

3.2.13 Evaluation of Occupational Doses To estimate the impact on dose rates in containment from the operation of a TPC containing 1,792 TPBARs, the licensee applied non-TPC operating experience from WBN, Unit 1, Cycle 3 and Sequoyah Nuclear Plant, Unit 1, Cycle 10. These two cycles consisted of long-term, "breaker-to-breaker," runs that enabled conditions for establishment of a strong correlation between RCS tritium concentration and the airborne tritium in containment. The average RCS tritium concentration value for the cycle was 1 micro-curie/gram. The licensee determined that the calculated tritium release to the RCS with a TPC containing 1,792 TPBARs will result in about a factor of 7 increase over the non-TPC tritium production rate. Consistent with the WBN, Unit 1, tritium production rate during Cycles 11 and 12 with 544 TPBARs, the licensee estimates that the

average RCS tritium concentration will be approximately 12 micro-curies/gram at a permeation rate of 5 Curies/TPBAR/year for 1,792 TPBARs. This average value results in an estimated containment tritium derived air concentration (DAC)-fraction of 0.96 which equates to a containment dose rate of about 2.4 mrem/hour.

The licensee estimated the impact of TPC operation on collective dose by adjusting the historic site average collective, committed effective dose equivalent (CEDE) by a ratio of the expected RCS tritium concentration of 12 micro-curies/gram to the average RCS tritium concentration of 1 micro-curie/gram from extended cycle operation with a non-TPC. The licensee conservatively attributed all historical CEDE to tritium exposure, even though, realistically, tritium is just one of many nuclides that contributed to historic CEDE. The licensee determined that TPC operation with 1,792 TPBARs will result in an additional 24 person-rem of CEDE. The additional CEDE combined with an additional estimated 1.3 person-rem/year for handling activities, based on WBN, Unit 1, operating experience, yields a total increase in collective CEDE of 25.3 person-rem/year for operation of a TPC with 1,792 TPBARs.

While collective radiation exposure is not a radiation safety standard, it is a useful indicator of the impact that TPC operation will have on the occupational aspects of the WBN radiation protection program. Consequently, even with the addition of 25 person-rem/year of collective radiation exposure, WBN will remain within the nominal range of contemporary industry performance as it relates to collective radiation exposure, as discussed in NUREG-0713, "Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities."

To ensure that individual occupational doses are maintained as-low-as-reasonably-achievable

{ALARA), the licensee has implemented a tritium control program through procedure RCl-137, "Radiation Protection Tritium Control Program." Page E1-70 of 116 of Enclosure 1 of the letter dated December 20, 2017, provides a table describing the program thresholds, survey requirements, actions and bases. Since tritium is primarily an internal exposure concern, the licensee's radiation protection program provides a graded-approach for bioassay based on risk and radiological conditions. The licensee's program establishes criteria for performing in-vitro bioassay: (1) based on process water concentrations; (2) for exposures greater than or equal to four DAG-hours in seven consecutive days; and (3) for cases of skin contact with water exceeding 0.01 micro-curies/milliliter of tritium. Tritium DAC-hr tracking is initiated whenever a worker is exposed to concentrations of 0.3 DAC of tritium or greater. The licensee's procedures cover (1) types of appropriate bioassay; (2) selection of individuals for bioassay; (3) bioassay collection; ( 4) sample volume, storage packaging and shipping; (5) analytical equipment detection limits; and (6) internal dose calculation methods. The program requires a minimum detectable activity of less than 1E+4 pico-curie/liter (i.e. 0.01 micro-curie/liter) tritium for urine bioassays. The program described in the letter dated December 20, 2017, is acceptable to NRC staff because it satisfies the guidance of RG 8.32, "Criteria for Establishing a Tritium Bioassay Program."

The NRC staff finds that the licensee's estimates of the impact of TPC operation with 1,792 TPBARs on occupational dose are reasonable and within the nominal range of contemporary industry performance. Additionally, the NRC staff concludes that through the licensee's radiation protection program, as described in the letter dated December 20, 2017, sufficient radiological survey and monitoring capabilities will be available to provide reasonable assurance that occupational doses at WBN, Unit 2, with a TPC can be maintained within the occupational dose

limits of 10 CFR Part 20 and that the licensee has implemented procedures to achieve doses that are ALARA.

3.2.14 Evaluation of Public Doses and Effluents To ensure public doses and radioactive effluent discharges will be within the limits of 10 CFR Part 20, the licensee applied the approach described in 10 CFR 20.1302(b )(2). The licensee calculated the sum-of-ratios of each isotope concentration (C) to its corresponding effluent concentration limit (ECL}-as listed in 10 CFR 20, Appendix B, Table 2, Columns 1 and 2 for gaseous and liquid effluents, respectively-(i.e. C/ECL). A C/ECL sum of less than 1.0 indicates that the annual average effluent release is less than 50 mrem (or SO-percent of the public dose limit of 10 CFR 20.1301 ). As described in 10 CFR 20.1302(b )(2), the additional SO-percent of the dose limit is reserved for the external exposure that is limited to 50 mrem per year by 10 CFR 20.1302(b )(2)(ii).

The licensee controls effluents through WBN, Unit 2, TS 5.7.2.7, "Radioactive Effluent Controls Program," which imposes the requirement that the licensee maintain an Offsite Dose Calculation Manual. Under this program, the release of radioactive liquids from the liquid waste system is made after laboratory analysis of the material to be discharged. The licensee has procedural and engineering controls in place to ensure that inadvertent discharges do not occur and that discharges are performed when sufficient dilution flow is available. WBN, Unit 2, has sufficient storage to enable holdup, dilution and adequate timing of radioactive releases. Tables 4.1-28 and 29 and 4.1-31 and 32 in the letter dated December 20, 2017, provide the results of licensee's analyses and confirm that liquid and gaseous effluents will be within applicable limits.

The licensee evaluated the total site dose from operation of TPCs at WBN, Units 1 and 2, to verify compliance with 10 CFR 20.1301(e}-which enforces the U.S. Environmental Protection Agency's 40 CFR Part 190 limits. Table 4.1-36 provides the results of this comparison and shows that calculated offsite doses originating from a site with two TPCs are within the limits of 40 CFR Part 190.

To verify that public radiation doses resulting from radioactive effluent releases will be maintained ALARA, the licensee compared the annual projected impact of TPC operation with 1,792 TPBARs to the numerical guides found in 10 CFR 50, Appendix I. Table 4.1-35 provides the results of this comparison and shows that the increase in tritium reactor coolant activity and resultant effluent releases will yield public doses that are within NRC design objectives and thus are considered ALARA.

The NRC staff finds that the licensee adequately demonstrated that WBN, Unit 2, TPC can be operated with 1,792 TPBARs, and maintain gaseous and liquid effluents and the resultant public doses within the limits of 10 CFR Part 20.

3.2.15 Operator Manual Actions and Human Factors 3.2.15.1 Description of Operator Action(s) Added/Changed/Deleted The proposed change will eliminate the following operator manual actions: manual isolation of the six-inch essential raw cooling water (ERCW) supply to the lower containment cooler Group D from

the main supply Header 28 by closing valve 2-ISV-67-5238 and manual isolation of the six-inch component cooling water (CCW) system supply and return lines for the reactor coolant pump oil cooler penetrating containment by closing 2-ISV-70-156 and 2-ISV-70-700. As a result of its review of operator actions, the licensee has identified that a reduction in human error probability can be accomplished by replacing the containment isolation thermal relief check valves on the lower compartment supply lines to the containment CCW and ERCW system with simple relief valves (i.e., passive devices). This will reduce both the time required and the operator workload involved in isolating the CCS and ERCW as potential sources of unwanted boron dilution during execution of Emergency Operating Instruction 2-E-O, "Reactor Trip of Safety Injection."

The licensee also stated in Section 4.1.14 of the LAR that TVA will review and modify actions, action levels, and sample frequencies, as necessary, associated with TPBAR Liquid Waste Management Systems.

3.2.15.2 NRC Staff Review of Added/Changed/Deleted Operator Action(s)

NUREG-1764 is the guidance used by the NRC to review changes to human actions. In accordance with the generic risk categories established in Appendix A to NUREG-1764, the elimination of operator actions to isolate an unborated dilution source reviewed herein is not considered "risk-important" due to the fact that it reduces the operators' workload during an accident, thereby reducing overall risk. Because of its low risk importance, the NRC staff performed a "Level Three" review, i.e., the least stringent of the graded reviews possible under the guidance of NUREG-1764. In addition, the guidance contained in NUREG-0711 was utilized to evaluate the licensee's design control process described in the LAR dated December 20, 2017.

3.2.15.2.1 Design Control WBN uses TVA procedure NPG-SPP-09.3, "Plant Modifications and Engineering Change Control," to manage the content, impact review, and implementation of design changes. From the human factors perspective, this process provides assurance that the affected Disciplines and Departments review the change as it is developed for impact to items under their control, such as procedures, training, control room interfaces, and the simulator. The NRC staff finds this process consistent with the criteria of NUREG-0711.

3.2.15.2.2 Human-System Interface Design The licensee stated in its December 20, 2017, submittal that Human-System Interface (HSI}

design of the control room and the simulator, including the design of the safety parameter display system, will not be affected by the proposed LAR. Based on the fact that no changes are needed to the current, approved HSI design, the NRC staff finds the WBN proposal acceptable.

3.2.15.2.3 Procedure Design In its December 20, 2017, submittal, the licensee stated that Emergency Operating Instruction 2-E-O, "Reactor Trip of Safety Injection," will be revised to remove the above described operator manual actions. In addition, plant-specific TPBAR Interface: "Liquid Waste Management Systems," of NUREG-1672, was addressed by the licensee in Section 4.1.14 of the LAR submittal.

The TPBAR Interface includes consideration of plant-specific tritium monitoring and surveillance

programs and procedures for operator actions on an abnormal tritium release event. The licensee's evaluation determined that the effect of potential chemical contaminant releases into the RCS or the SFP due to leaching effects from TPBARS will not require any changes to existing WBN sampling frequencies. However, the licensee notes that procedures were revised prior to TPBAR irradiation to require liquid sampling in the SFP for tritium while moving and storing irradiated TPBARs. The licensee will review and modify actions, action levels, and sample frequencies, as necessary, based on tritium production core operating experience.

The NRC staff finds that WBN has correctly identified procedure impact and has stated that revisions will be made as required. Per Section 4.2.2 of the LAR dated December 20, 2017, required procedure revisions will be controlled per TVA procedure NPG-SPP-09.3, "Plant Modifications and Engineering Change Control." The NRC staff finds that implementation of necessary procedure changes per the licensee's normal plant configuration control processes per TVA procedure NPG-SPP-09.3 is appropriate. The NRC staff finds the procedure design acceptable per the review guidance provided in NUREG-1764 and NUREG-0711.

3.2.15.2.4 Training Program Design The licensee stated in its December 20, 2017, submittal that operator training on the deletion of the operator manual actions will be required. Based on its incorporation of the deleted actions into the training program, the NRC staff finds the WBN proposal acceptable per review guidance in NUREG-1764.

3.2.15.2.5 Human Factors Verification and Validation No verification and validation of operator tasks or procedures is necessary, because there are no operator manual actions being changed or added; only eliminated.

3.2.15.2.6 Operator Manual Actions and Human Factors Conclusion The NRC staff finds that appropriate administrative controls are being applied to revise procedures and training, and that the HSI design is not affected by the proposed change.

Therefore, the NRC staff concludes that the proposed LAR is acceptable with regard to human factors considerations per the review guidance provided in NUREG-1764 and NUREG-0711.

3.2.16 Quality Assurance The DOE manages the tritium production program, including issuance of major procurements.

DOE procures TPBAR design, fabrication, irradiation, and transportation services for delivery of irradiated TPBARs to the DOE Tritium Extraction Facility. The major DOE suppliers for the TPBAR project include Pacific Northwest National Laboratory (PNNL), WesDyne International LLC (WesDyne), and TVA.

The NRC staff reviewed the quality relationships among the contracting parties, focusing on the TPBAR supplier procurement and fabrication activities of PNNL and WesDyne. The overall flow of quality assurance (QA) program requirements for the tritium production program is illustrated in Figure 4.1-4 in the letter dated December 20, 2017.

3.2.16.1 Tritium Production Program TPBARs are supplied to TVA as "Government Furnished Property" per DE-AI02-00DP00315, "lnteragency Agreement between the United States Department of Energy National Nuclear Security Administration (NNSA) Contracts and Procurement Division and Tennessee Valley Authority for irradiation Services," dated May 2, 2014, between TVA and the NNSA. The interagency agency agreement provides a means for imposing TVA requirements directly on DOE TPBAR suppliers. TVA has no direct procurement document with any of the material, service, or component suppliers of TPBARs. TPBARs are classified as safety-related components that shall be supplied in accordance with a QA program that complies with requirements of 10 CFR Part 50, Appendix B. To accomplish this, NNSA has established unique protocol to implement the TVA QA requirements that are applicable to TPBARs in their procurement documents.

TVA has an interagency agreement with the NNSA that requires NNSA to flow down TVA requirements to suppliers and requires the NNSA direct suppliers to be on the TVA approved supplier list (ASL). The main TVA document establishing these QA requirements is TVA-TPPR-99-01, "Tritium Production Program Requirements Technical, Functional, & Quality Requirements for TPBARs," Revision 4, dated July 5, 2017, which defines the technical, functional, and quality requirements associated with design, analysis, materials, fabrication, and delivery of TPBARs. TVA-TPPR-99-01 requires NNSA to flow down TVA QA requirements to their respective suppliers. It also requires direct suppliers to TVA maintain a TVA-accepted QA program. Other requirements included in TVA-TPPR-99-01 are TVA acceptance of deviation resolution, interface controls, reporting requirements, document submittal requirements, and TPBAR functional requirements. Activities associated with TPBAR design, material and services procurements, fabrication, and delivery are performed under the auspices of TVA's NRG-approved nuclear QA program (NQAP), TVA-NQA-PLN89-A, Revisions 35 and 36, dated May 8, 2018 (ADAMS Accession No. ML18129A317), as approved via safety evaluation dated November 8, 2018 (ADAMS Accession No. ML17291A547). TVA-TPPR-99-01 states that although NNSA manages the Tritium Production Program procurement activities, all safety-related materials, items, and services are to be procured from TVA-accepted suppliers and must comply with TVA-specified technical, functional, and quality requirements. The NRC staff approved TVA's NQAP in a safety evaluation dated November 8, 2018.

3.2.16.2 Tritium Production Program Requirements The activities associated with TPBAR design, material and service procurements, fabrication, and delivery are performed in accordance with TVA's NQAP. The licensee's NQAP includes provisions for TVA to specify the applicable QA requirements for items or services supplied by others. Per the DOE/TVA interagency agreement, TVA has elected to qualify TPBAR suppliers as though these suppliers were direct suppliers to TVA The DOE direct suppliers (PNNL and WesDyne) have submitted their QA programs to TVA and have been placed on TVA's ASL in accordance with the NQAP. Suppliers are maintained on the ASL through annual evaluations and triennial audits.

The licensee requires that all TPBAR safety-related materials, items, and services comply with the technical, functional, and quality requirements of TVA-TPPR-99-01. This document requires that TPBARs be designed, fabricated, and delivered in accordance with the methods of the basic and supplementary requirements of American Society of Mechanical Engineers (ASME)

NQA-1-1994 and comply with the regulatory positions of RG 1.28, Revision 3. The NRC staff has reviewed the DOE/TVA interagency agreement and TVA-TPPR-99-01 and found that they establish an effective method for controlling the TPBAR procurement and fabrication process in accordance with applicable NRC regulatory procurement and QA requirements.

3.2.16.3 Quality Requirements (Direct Suppliers)

DOE has selected two direct suppliers of TPBARs:

PNNL, a DOE Office of Science site operated by the Battelle Memorial Institute, performs TPBAR design and procurement activities.

WesDyne, a wholly owned subsidiary of Westinghouse Electric Company LLC, performs TPBAR procurement and fabrication activities.

The PNNL scope includes design evolution and fabrication process improvements associated with supporting full-scale TPBAR fabrication and material and subcomponent procurements.

PNNL has been maintained on TVA's ASL and this has been supported by performing program reviews, source surveillances, and audits. To verify TVA's assessment of the quality program of PNNL, the NRC staff evaluated the most recent triennial audit of PNNL's QA program performed by TVA ("TVA Sequoyah and Watts Bar Nuclear Plants- Design, Procurement, Fabrication, and Test & Inspection Activities for Tritium Production and Associated Safety Related Components

- TVA Supplier Audit 2017V-13," dated August 2, 2017). The NRC staff found TVA's conclusion that PNNL QA program, as implemented for the activities associated with the TPBAR, met the requirements of 10 CFR Part 50, Appendix B, and 10 CFR Part 21 to be acceptable.

The TPBAR design interface agreement between TVA and PNNL is documented in TVA-TPPR-99-02 ("TPBAR Design Interface Agreement between Tennessee Valley Authority and Pacific Northwest National Laboratory," Revision 1, dated September 21, 2017). This agreement identifies the controls associated with PNNL obtaining and using TVA and TVA fuel vendor technical information in TPBAR design activities. The TTP-7-065 ("Tritium Technology Program - Interface Agreement between Pacific Northwest National Laboratory and WesDyne International, LLC for Production TPBARs," Revision 5, dated December 18, 2017) is the PNNL controlled document that describes requirements for TPBARs for the interface between PNNL (the TPBAR design authority) and WesDyne (the TPBAR fabricator). The TTP-7-065 flows down requirements of TVA-TPPR-99-01 as well as addressing additional interface requirements, test articles, and services that may be exchanged. It identifies the roles and responsibilities of WesDyne, PNNL, and TVA in the evaluation of nonconforming items. The NRC staff reviewed these documents as part of this evaluation and determined that they provided the necessary detail to satisfy the applicable requirements of 10 CFR Part 50, Appendix B, for establishing measures for the identification and control of design interfaces and for coordination among participating design organizations.

WesDyne WesDyne's scope includes procurement of materials and services, assembly, fabrication of final TPBARs, and delivery of certified TPBARs to TVA or TVA's nuclear fuel manufacturer(s) for use in TVA reactor cores. WesDyne has been maintained on TVA's ASL and this has been supported by performing program reviews, source surveillances, and audits. To verify TVA's assessment of the quality program of WesDyne, the NRC staff evaluated the most recent triennial audit of WesDyne's QA program performed by TVA ("Tritium Producing Burnable Absorber Rod (TPBAR) Project- TVA Sequoyah and Watts Bar Nuclear Plants -Audit of WesDyne International, LLC-TVA Supplier Audit 2015V-13," dated August 4, 2015). The NRC staff found TVA's conclusion that WesDyne QA program, as implemented for the activities associated with the TPBAR, met the requirements of 10 CFR Part 50, Appendix B, and 10 CFR Part 21 to be acceptable.

The TPBAR interface requirements between WesDyne and Westinghouse Global Quality is documented in WD-TP-23.1 ("WesDyne/Global Quality Program Interface Agreement for TPBAR and Related Activities," Revision 0.0, dated April 6, 2016). WD-TP-23.1 also defines the responsibilities between WesDyne and Westinghouse Global Quality for evaluation of TPBAR suppliers and related QA activities. WesDyne is ultimately responsible for TPBAR fabrication and provides management and QA oversight of all TPBAR fabrication activities.

WesDyne performs the work for the TPBAR program to a Project Quality Plan (PQP). Per the PQP, the Westinghouse Quality Management System procedures allow WesDyne to utilize the Westinghouse Quality Suppliers List. WesDyne subcontracts a significant portion of the fabrication of its work at the Westinghouse Columbia Fuel Fabrication Facility to Westinghouse Nuclear Fuels. WesDyne PQP commits to compliance with the QA program requirements using the latest revision of the NRG-approved Westinghouse Quality Management System (ADAMS Accession No. ML14336A487) that meets the requirements of 10 CFR Part 50, Appendix B, and the NRC staff finds this acceptable.

3.2.16.4 Quality Requirements (Material/Service Subcontracts)

Quality oversight (such as program reviews, source surveillances, and audits) of material, service, and subcomponent suppliers are the responsibility of the procuring organization (i.e., PNNL or WesDyne) with periodic participation by a TVA observer. Suppliers currently producing parts or providing services to be used in the production of TPBARs have established and implemented QA programs that meet the requirements of 10 CFR Part 50, Appendix B.

These suppliers have been placed on the procuring organization (i.e., PNNL or WesDyne)

Qualified Supplier List). As part of this LAR, the licensee submitted the overall results of the most recent triennial audits performed by PNNL or WesDyne for some of the suppliers listed in Table-1.

For those suppliers that did not have a triennial audit submitted as part of this LAR, the NRC staff was able to evaluate the triennial audits of those suppliers during an inspection performed at the Westinghouse Office located in Cranberry Township, PA The results of the inspection are documented in inspection report No. 99900404/2018-201, dated July 6, 2018 (ADAMS Accession No. ML18176A395). WesDyne is a wholly owned subsidiary of Westinghouse. The NRC staff evaluated the triennial audits and found that the conclusion from the procuring

organization to be acceptable, that the suppliers' QA programs as implemented for the activities associated with the TPBAR meet the requirements of 10 CFR Part 50, Appendix B, and 10 CFR Part 21.

PNNL QA Program was approved by the NRC staff in Amendment No. 40, "Watts Bar Nuclear Plant, Unit 1 - Issuance of Amendment to Irradiate Up to 2304 Tritium-Producing Burnable Absorber Rods in the Reactor Core (TAC No. MB1884)," dated September 23, 2002 (ADAMS Accession No. ML022540925). PNNL is using the same QA Program for WBN, Unit 2, TPBARs; therefore, the NRC staff finds it acceptable.

A listing of major suppliers for TPBARs is shown in Table 4.1-1 in the letter dated December 20, 2017.

3.2.16.5 Summary of Quality Requirements The licensee's Tritium Production Program Requirements document (TVA-TPPR 99-01) describes requirements established by TVA, which are applied to the procurement and fabrication of TPBARs to be irradiated in the WBN, Unit 2, reactor core. This document provides the technical, functional, and quality requirements associated with design, analysis, materials, fabrication, and delivery of TPBARs.

The DOE/TVA interagency agreement identifies TPBARs as safety-related basic components.

As such, TPBARs and related services furnished under the agreement must be supplied in accordance with a QA program that complies with 10 CFR Part 50, Appendix B, in accordance with RG 1.28, Revision 3. In addition, the reporting requirements of 10 CFR Part 21 are imposed on suppliers of TPBARs and related services.

The DOE has selected PNNL and WesDyne as direct suppliers of TPBARs. The activities of these suppliers are conducted in accordance with QA programs that the NRC has determined to be acceptable in that they conform to the requirements of 10 CFR Part 50, Appendix B. The licensee has audited the supplier QA programs for compliance and maintains them on its ASL. In accordance with contractual requirements imposed by the direct TPBAR suppliers, their subcontractors and suppliers of components and services have implemented QA programs that meet the requirements of 10 CFR Part 50, Appendix B, and the reporting requirements of 10 CFR Part 21.

The overall commercial light-water reactor (CLWR) project structure provides for effective control of all supplier activities in compliance with applicable regulatory QA and procurement requirements. The QA programs of the DOE direct suppliers have been reviewed by TVA and qualified as having QA programs in conformance with the requirements of 10 CFR Part 50, Appendix B. Through its procurement process, DOE has imposed Part 21 requirements on all TPBAR suppliers. The plant-specific interface issue related to TPBAR procurement and fabrication has been satisfactorily addressed by the licensee.

The NRC staff approved the quality requirements for the DOE direct supplies in Amendment No.

40 for WBN, Unit 1, and the DOE direct suppliers are the same for WBN, Unit 2. Therefore, the NRC staff finds that TPBARs suppliers QA programs meet the requirements of 10 CFR Part 50, Appendix B, and the reporting requirements of 10 CFR Part 21.

The overall commercial light-water reactor (CLWR) project structure provides for effective control of all supplier activities in compliance with applicable regulatory QA and procurement requirements. The QA programs of the DOE direct suppliers have been reviewed by TVA and qualified as having QA programs in conformance with the requirements of 10 CFR Part 50, Appendix B. Through its procurement process, DOE has imposed Part 21 requirements on all TPBAR suppliers. The plant-specific interface issue related to TPBAR procurement and fabrication has been satisfactorily addressed by the licensee, and the NRC staff finds this acceptable.

3.2.17 Nuclear Performance and Code Review 3.2.17.1 Compliance with Departure from Nucleate Boiling {DNB) Criterion NUREG-1672, Section 2.4.4 describes thermal-hydraulic design evaluation of a generic WBN core with TPBARs using the acceptance criteria outlined in Section 4.4 of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Report for Nuclear Power Plants." The analysis has shown that the bypass flow remains within the design limit of 8.4-percent and that the DNB criterion continues to be met with no feature of the TPBAR component affecting the coolability of the core. The licensee assessed the continued compliance with the DNB criterion as an interface item for which plant-specific information is submitted to the facility operating license complying with the criteria that requires the demonstration that DNB would not occur on the most limiting fuel rod on at least a 95 probability at a 95-percent confidence level.

The methodology used to determine the acceptability of operating with TPBARs in the WBN, Unit 2, reactor core is consistent with the current Westinghouse methods for inserting new components in to Westinghouse cores as per the NRG-approved Topical Report, NDP-98-153, "Tritium Production Core (TPC)." For the WBN, Unit 1, 1,792 TPBAR Tritium Production equilibrium cycle, the normal thermal-hydraulic DNB related reload analyses were performed using VIPRE-01 (ADAMS Accession No. ML993160096). Since the power level, fuel type and other plant parameters used in the thermal-hydraulic and DNB analysis, are similar between WBN, Units 1 and 2, the analysis process at Unit 1 is applicable to Unit 2. Specifically the following detailed thermal-hydraulic evaluations at WBN, Unit 1, are:

Axial power shape study performed for the TPBAR case has confirmed that the power distributions used in the core design would still be valid in the presence of the TPBAR.

This axial power shape study compares power shapes due to depletion during operation of the cycles to reference shapes used as the basis for thermal-hydraulic design analysis.

The steamline break with rod withdrawal at power transient was analyzed to demonstrate the continued acceptability of the Departure from Nucleate Boiling Ratio (DNBR) design basis for this transient.

The zero-power hypothetical steamline break to demonstrate that the DNB design basis was met.

The axial power shape comparison showed that the reference power shapes assumed in the current safety analysis for WBN would remain bounding and the TPBARs do not present any excessive power distribution changes beyond those which are bounded within the thermal-hydraulic design bases. Westinghouse considered dimensional tolerances of the thimble and TPBAR in the input to the model. Bounding values of enthalpy rise and axial power distributions were used for computations of the fuel rod side boundary conditions and for heat

generation within the absorber. Westinghouse used generic power distribution profiles for thermal and hydraulic DNBR analyses. Standard inputs were used in explicit core component models to determine loss coefficients and orifice flow paths.

To prevent excessive heat and corrosion, Westinghouse imposed a temperature limit on the TPBAR to keep it below its melting temperature. This limit will prevent surface boiling along the TPBAR within the dashpot region of the thimble and prevents bulk boiling along the length of the thimble. Also, the sum of the flow through all the thimbles and the TPBAR combinations must be less than that allowed by the bypass flow limits that are used to ensure adequate flow for core cooling.

The thermal-hydraulic DNBR compliance analysis has demonstrated the presence of TPBARs in the reload core design does not challenge the DNB criterion. The results indicated that having TPBARs in the core would not affect the power distributions beyond those already bounded by the thermal and hydraulic design bases. Finally, the results of the DNB analyses showed that the DNBR design basis will continue to be met with the presence of the TPBARs.

The NRC staff finds the presence of TPBARs in the reload core design does not challenge the DNB criterion.

3.2.17 .2 Reactor Vessel Integrity Section 2.5.3 of the NUREG-1672 safety evaluation report discussed TVA's analyses of reactor vessel integrity for a typical Westinghouse reactor. As explained in the NUREG, reactor vessel integrity analyses depend on specific reactor vessel materials used, and the actual neutron fluence experienced, in a reactor. In the TPC topical report, DOE concludes, and the NRC staff finds, that the reference plant's pressure/temperature limits report (PTLR) and final safety analysis report (FSAR) would need to be updated to reflect the change to the PTS value and include the updated P-T curves for the applicable effective-full-power-years (EFPYs), because the reactor vessel integrity analyses are dependent upon the plant-specific materials properties and neutron fluence. The NRC staff concludes that a licensee participating in DOE's program for the CLWR production of tritium must present the material properties for its reactor vessel and perform analyses that demonstrate it will meet the requirements of Appendices G and H to 10 CFR Part 50 and of 10 CFR 50.61.

Neutron Fluence Methodology and Calculations This sub-section describes the methodology and results from neutron fluence calculations that are required to support the reactor vessel integrity analyses and the updated P-T curves for the applicable EFPYs. The WCAP-18191-NP report provides the methodology and results of the generation of heatup and cooldown P-T curves for normal operation of the WBN, Unit 2, reactor vessel. This analysis considers the implementation of TPBARs at the beginning of Cycle 4. In order to support the heatup and cooldown P-T curves generation, discrete ordinates (Sn) calculations were performed for WBN, Unit 2, to determine the neutron environment within the beltline and extended beltline regions of the reactor pressure vessel (RPV). The discrete ordinates calculations were based on nuclear cross section data derived from Evaluated Nuclear Data File (ENDF), specifically ENDF/BVIII. The neutron transport methodologies meet the guidelines and requirements set forth in the Regulatory Guide 1.190. Also, the methods used to

determine the pressure vessel neutron exposure are consistent with the NRG-approved methodology described in WCAP-14040-A. The neutron exposure evaluations for the WBN, Unit 2, RPV were performed using the three-dimensional flux synthesis technique which used the axial power distributions. The flux synthesis technique obtained the three-dimensional neutron flux distribution which is a combination of neutron flux distribution in (r, 9) direction, neutron flux in the (r, z) direction and the neutron flux in the radial (r) direction.

The core power distributions used in the plant-specific transport analysis for each of the first seven projected fuel cycles at WBN, Unit 2, included cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions. Cycles 1-6 at WBN, Unit 2, are based on the expected core design for these cycles and Cycle 7 is a representative equilibrium fuel cycle that is then extended to 32 and 40 EFPYs. These calculations take into account implementation of TPBARs at the beginning of Cycle 4. The TPBAR core design was reflected in the radial power distributions that are used as critical inputs into the RPV fluence calculation. TPBAR core designs show a higher relative power in the peripheral assemblies than for a low-leakage core design which causes an increase in the RPV fast neutron fluence exposure in those TPBAR cycles. The vessel integrity calculations consider the TPBAR core design and justify safe operation through 32 EFPYs with respect to RPV integrity. The information from these calculations was used to develop spatial- and energy-dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle-averaged neutron fluence rate, which, when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle.

The data tabulations include both plant- and fuel-cycle-specific calculated neutron exposures at the end-of-cycle (EOC) 1 through 7 and at further projections to 32 and 40 EFPY. These calculations are for an expected core power of 3,411 MWth in Cycle 1 and 3,459 MWth in Cycles 2 through 7. The projections after Cycle 7 are based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 7 are representative of future plant operations.

Results from neutron transport analyses are provided in Table 2-2 through Table 2-10 in WCAP-18191-NP, Revision 0. The calculated fast neutron exposure rate and total exposure values at the geometric center of the surveillance capsule test specimen are provided in Table 2-2 and Table 2-3, respectively. Table 2-4 through 2-6 of WCAP-18191-NP, Revision 0, show the maximum calculated exposure and exposure rate at the clad/base metal interface. Azimuthally, angles 0.00, 20.00 through 22.00 at 0.25 degrees intervals, 23.00, 30.00, and 45.00 were analyzed. Table 2-7 through Table 2-10 of WCAP-18191-NP, Revision 0, provide maximum neutron exposure at the pressure vessel clad/base interface at various points for the projected Cycle 1 through 7 and projections for an operating time extending to 32 and 40 EFPYs. The calculated fast neutron fluence for neutron energy greater than 1 MeV (million electron-volt) at pressure vessel intermediate to lower shell circular weld are found to be 1.861 E+10 n/cm 2 and 2.344 n/cm 2 at 32 and 40 EFPY, respectively.

The NRC staff in the letter dated February 15, 2018, sought clarification for the difference in the limiting ins.ide surface neutron fluence for the RPVat 32 EFPY reported in WCAP-18191-NP, Revision O (1.861 E+19 n/cm 2 ) and in WCAP-17035-NP (3.17E+19 n/cm 2 ). The licensee responded that WCAP-17035 was created prior to the licensing of WBN, Unit 2, and the neutron fluence values provided in WCAP-17035-NP were based on a design basis core power

distribution that assumed an out/in loading strategy for the entire operating lifetime of the reactor.

The calculations performed in WCAP-18191-NP for WBN, Unit 2, are based on actual power distributions, which are available since it is currently an operating plant. In the case of WBN, Unit 2, the actual power in the peripheral fuel assemblies calculated for WCAP-18191 is less than that used in the WCAP-17035-NP calculation. This reduction in peripheral power, relative to the design basis, results in a decrease in the neutron flux at the inside surface of the RPV. The reduction in peripheral power overrides the effect of the inclusion of the TPBARs. Further, the licensee clarified why the average neutron flux for Cycles 5 through 7 increases as the number of TPBARs are increased. The clarification is that the RPV flux calculation is sensitive to the relative power produced from the peripheral assemblies, the enrichment of the peripheral assemblies, the coolant temperatures in the peripheral fuel assemblies and ex-core regions, and the axial power shape. In fact, there is no direct effect on RPV neutron flux as a result of inclusion or removal of TPBARs. The observed increase of neutron flux reported in WCAP-18191 is a result of changes to the core loading patterns, i.e., higher core average enrichment, that are implemented in order to offset the reactivity penalty of irradiating TPBARs.

The NRC staff reviewed the methodology and results for the neutron flux and neutron fluence calculations in the WBN, Unit 2, RPV and determined that the methodology and results are acceptable for RV integrity calculations for WBN, Unit 2.

3.2.17.3 New and Spent Fuel Storage New Fuel Storage Vault The TPBARs are neutron poison that are different from what was previously used in WBN, Unit 2.

The current New Fuel Storage Vault criticality analysis has shown that un-poisoned fuel assemblies (without either discrete or integral poison) containing nominal enrichments up to 5.0 weight percent U-235 can be stored in the fresh fuel rack array utilizing 120 specific cells of the 130 available storage locations. Fresh fuel with TPBARs stored in the new fuel rack will have a lower reactivity than un-poisoned fresh fuel assemblies. Since the fuel assembly design did not change from the current new fuel storage criticality analysis with TPBARs, the current criticality analysis and the new fuel storage configuration remains conservative and valid for storage of fresh fuel assemblies with TPBARs. Since the new fuel storage vault is common to both WBN, Units 1 and 2, the current new fuel storage criticality analysis approved in WBN, Unit 1, Amendment No. 15 is still bounding for fuel with TPBARs at WBN, Unit 2.

Spent Fuel Storage Pool The licensee reanalyzed the criticality analysis for region 1. of the spent fuel storage racks to support the TPBAR irradiation with fuel assemblies of nominal enrichments up to 5.0 weight percent U-235 containing TPBARs and addressed other neutron poisons including Wet Annular Burnable Absorbers and Integral Fuel Burnable Absorbers (IFBAs). The reanalysis demonstrated that sufficient conservatism was present in the previous analysis of the region 1 storage racks to account adequately for the effects of operating with TPBARs.

Updated SFP Criticality Analysis The criticality safety analysis of record for the WBN SFP was developed in 2001 and relied, in part, on analyses performed in 1996. The purpose of the updated analysis is to provide a complete up-to-date criticality safety evaluation for the WBN SFP based on the latest methodologies consistent with current NRC guidance. The updated SFP analysis was performed by Holtec International in Hl-2177876 Report, September 2017. This report replaces the analysis of record (AOR).

The WBN SFP contains a single type of BORALTM racks with flux traps designed for storage of pressurized-water reactor (PWR) fuel. Criticality control in the WBN BORALTM storage racks rely on the following:

fixed neutron absorbers: BORALTM panels

storage cell spacing, i.e., flux traps between storage cells

soluble boron The criticality calculations qualify the BORALTM storage racks uniformly loaded with fresh fuel assemblies with an initial enrichment up to 5 weight percent U-235.

Each design basis analysis calculation considers fresh fuel with a uniform enrichment.

The same bounding enrichment is considered along the entire active length for each fuel pin. Lower enriched blankets are neglected. Therefore, there is no axial or radial variation in fuel along the entire active length.

A bounding fuel density of all types of fuel assemblies is considered. This bounding approach provides analysis simplicity and margin.

Acceptance Criteria The objective of the analysis is to demonstrate that the effective neutron multiplication factor (keff) of the SFP is less than 1.0 with the pool flooded with un-borated water, and will not exceed the regulatory limit of 0.95 with credit for 500 ppm soluble boron all for 95-percent probability at a 95-percent confidence level.

General Approach for Methodology The calculations are performed using either the worst case bounding approach or the statistical analysis approach with respect to the various calculation parameters. These calculations are used to determine the final keff used to show compliance with the regulatory limits for both normal and accident conditions. The accident calculations are essentially modifications of the design basis cases for the normal conditions but do not introduce a new fuel or rack design change. The uncertainty and bias calculations for the normal conditions are applicable and do not need to be repeated for the accident calculations.

Computer Codes Used in Analysis MCNPS-1.51 is used for the criticality analyses. MCNPS-1.51 is a three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory. MCNP was selected, because it has a long history of successful use in fuel storage criticality analyses and has all the necessary features for the analysis to be performed for the WBN SFP. MCNPS-1.51 calculations use continuous energy cross-section data predominantly based on ENDF/B-VII. These cross sections are adjusted for temperature dependence using a continuous energy cross section data processed with NJOY 99.36 using the ENDF/B-VII cross section library.

Benchmarking of MCNPS-1.51 for criticality calculations is based on NRC guidance and includes calculations for critical experiments with fresh U02 fuel, fresh MOX fuel, and fuel with simulated actinide composition of spent fuel. The statistical treatment used to determine those values considered the variance of the population about the mean and used appropriate confidence factors and trend analysis.

Analysis Methods The calculations are performed using either the worst case bounding approach or the statistical analysis approach with respect to the various calculation parameters. The bounding inputs and assumptions for the fuel and storage rack model consist of:

Bounding fuel designs and fuel assembly parameters,

Bounding storage rack parameters, and

Bounding SFP moderator temperature Specifically, to address regulatory concerns related to the BORALTM neutron absorber:

The minimum BORALTM B-1 O areal density, BORALTM panel width, and BORALTM panel thickness are used in all calculations because it is a well-known effect that minimum values of these parameters will increase reactivity.

An uncertainty of 5-percent (a typical value) on the minimum areal density of the B-10 content in the BORALTM is included in the uncertainty analysis to account for the measurement uncertainty associated with the B-10 content.

All design basis calculations are analyzed by assuming a void with thickness of 0.09 inch in the gap between the BORALTM and the steel sheathing to account for the reactivity effect of BORALTM blistering. While blisters would only be expected locally, they are conservatively modeled over the entire length and width of all panels in the pool.

The WBN SFP contains various Westinghouse 17x17 PWR fuel designs, such as, V5H, V+/P+,

and RFA-2 which are selected for potential storage in the BORALTM racks.

The following cases are evaluated:

Reactivity effect of fuel design parameters such as, enrichment with tolerances, minimum and maximum cladding thickness, minimum and maximum fuel rod pitch, minimum and maximum fuel pellet outer diameter, minimum and maximum guide tube/instrument tube thickness.

Reactivity effect of BORALTM rack parameters, such as, minimum and maximum storage cell dimensions.

Reactivity effect of SFP water temperature at the most reactive temperature and density.

Reactivity effect of fuel radial positioning.

Reactivity effect of BORALTM blistering to account for the loss of neutron absorber at various thickness of void in the gap, The SFP criticality analyses were performed for normal conditions as well as for accident conditions. The accident conditions at which the criticality analyses were performed include the effect of SFP temperature exceeding the normal range, dropped fuel assembly, misloaded fuel assembly (in the wrong location), and rack movement due to seismic activity.

Results and Safety Findings The results from criticality calculations for the WBN SFP BORALTM storage racks show that kett of the PWR BORALTM racks loaded with fuel of the highest anticipated reactivity at a temperature corresponding to the highest reactivity, is less than 1.00 with no credit for soluble boron for normal conditions, and less than 0.95 with credit for soluble boron for both normal and accident conditions, all of the results with 95-percent probability at a 95-percent confidence level.

The summary of the results is listed in Table 4.1-23 in the letter dated December 20, 2017.

3.2.17.4 Use of LOCTA JR Code for LOCA Analysis In order to assess the interaction of the TPBARs with the Loss of Coolant Accident (LOCA) transients, it is necessary to first estimate the response of the TPBARs to the design basis LOCAs, both large and small breaks (LB and SBLOCA). The TPBAR generates significantly less heat than the fuel during a LOCA. TPBARs are heated primarily by radiation from the fuel rods to the fuel assembly guide thimble and radiation from the thimble across the gap to the TPBAR. To estimate the impact of TPBARs on LOCA, Westinghouse used one-dimensional radial heat conduction code, LOCTA_JR, which is developed by Westinghouse based on the original LOCTA code series (Topical Report NDP-98-i 81, Revision 1).

The LOCTA_JR code performed one-dimensional radial heat conduction calculations for a fuel rod. In June 2000, TVA submitted Westinghouse Topical Report WCAP-15409, "Description of the Westinghouse LOCTA_JR 1-D Heat Conduction Code for LOCA Analysis of Fuel Rods," to the NRC for review and approval. The NRC staff accepted the LOCTA_JR code for referencing in licensing analysis. Although LOCTA_JR contains models for fuel rod performance (such as, clad oxidation, rod internal pressure, creep, burst and blockage), they are not used to analyze TPBARs and were, therefore, not considered in the original NRC review.

The TPBARs are modeled with the thimble tube heatup model, which solves the one-dimensional transient conduction and radiation equations. The thimble tube model allows the specification of one average material in the thimble tube, so it only applies to transients with slow heatup rates, where one can assume that there is no significant temperature variation within the thimble tube insert. This code will only be used to predict the thermal behavior of TPBARs during a LOCA.

The licensee no longer uses the LOCTA_JR code to predict the thermal-mechanical behavior of TPBARs during a LOCA. The TPBAR temperatures used in LOCTA_JR are obtained using boundary conditions from Westinghouse's Best Estimate Large-Break LOCA (LBLOCA) and Appendix K (Small-Break LOCA (SBLOCA)) analyses of record for the WBN plant. The LOCTA_JR code is used only to model the fuel assembly thimble and the TPBAR.

Modeling assumptions for the analysis include: ( 1) steam flow in the annulus between TPBAR and the thimble will be minimal due to low heat generation rate in the TPBAR, (2) thimble and TPBAR temperature calculations are performed one-dimensionally at the elevations of high fuel temperatures, (3) heat transfer to the outer surface of the thimble includes radiant heat transfer from the fuel rods and convective cooling from the core steam, and due to high thermal conductivity of gases within the TPBAR and low heatup rates the temperature gradient inside the TPBAR is minimal. The fuel rod temperatures and the core steam and entrained liquid convective heat transfer coefficients temperatures used in the previous LOCTA_JR predictions are found to bound those in WBN, Unit 2.

In summary, the LBLOCA and SBLOCA analyses were performed with NRG-approved Westinghouse methodologies, in conjunction with the LOCTA_JR code for TPBARs and instrument thimble. Introduction of TPBARs into WBN, Unit 2, has demonstrated compliance with the LOCA criteria defined in 10 CFR 50.46(b ). However, the TPBARs have been observed to have a slight effect on core axial power distributions. The fuel rod temperatures and the core steam and entrained liquid convective heat transfer coefficients temperatures used in the previous LOCTA_JR predictions bound those in WBN, Unit 2. The NRC staff has reviewed the LOCA methodology in conjunction with the LOCTA_JR code used for LOCA analysis and the results from these analyses and has determined that the WBN, Unit 2, LOCA analysis with TPBARs is acceptable.

3.2.17.5 Post-LOCA Subcriticality Evaluation The post-LOCA subcriticality analysis supports evaluations for each reload core to demonstrate that the core will remain subcritical during the reflood phase and during the cold-leg and hot-leg sump recirculation phases of ECCS operation. TVA performed an additional analysis for the long-term cooling phase of hot-leg recirculation. The post-LOCA reflood phase has previously been evaluated for WBN, Unit 1, and shown to be non-limiting for subcriticality in Amendment No. 107 for WBN, Unit 1, dated July 29, 2016 (ADAMS Accession No. ML16159A057). Because the WBN, Unit 2, design is the same, a separate WBN, Unit 2, analysis is not necessary.

The analysis of record considered a whole range of break sizes up to and including a double-ended guillotine rupture of the main coolant loop piping. Additionally, the analysis takes credit for the isolation of the potential unborated dilution source that would have entered the containment at a maximum rate of 40 gallons per minute (gpm), and a conservative lithium leaching assumption for TPBARs assumed to fail.

During the sump recirculation phase, the boron concentration of the sump water must be sufficient to keep the core subcritical when the sump water is delivered to the RV during the cold-leg and hot-leg recirculation phases. The sump mixed mean boron concentration calculations are used to develop a post-LOCA subcriticality boron limit that is confirmed on a cycle-specific basis as part of the Westinghouse Reload Safety Evaluation Methodology.

Lithium leach tests were performed in unstirred stainless steel pressure vessels by taking samples periodically, one day and 14 days. The average lithium leaching seen at one day with 95-percent confidence upper bound was 2.05 weight percent per day. The total lithium leached after 14 days with 95-percent confidence upper bound was 8.73 weight percent, or an average of 0.62 weight percent per day. The maximum TPBAR temperature after reflood is found to be less than 240°F based on post-LOCA coolant temperature. The boron concentration in the RWST and the safety injection accumulators used in the analysis were 3,100 ppm minimum RWST boron concentration and 3,000 ppm for minimum accumulator boron concentration, respectively. The evaluation assumed no borated essential raw cooling water (ERCW) or component cooling system (CCS) water leakage in to the containment sump co-incident with the LOCA.

The lithium leach rate assumption for the WBN, Unit 2, cold leg break scenario is the same assumption that was used for WBN, Unit 1, Amendment No. 107. The lithium leach rate assumption is a time dependent leaching rate of 3-percent for the first day and a maximum of SO-percent thereafter. The cold leg break scenario differs from the hot leg break scenario due to the assumption of TPBAR failure and the potential for sump dilution only being applicable for the cold leg break location. At hot-leg switchover (HLSO), the TPBAR failure assumption of an instantaneous loss of 3-percent lithium inventory is conservative because leaching of the TPBARs is not instantaneous. The bounding leaching rate is 3-percent per day; therefore, some value less than 3-percent of the lithium would have leached at the time of HLSO (3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ).

The NRC conducted an audit of the TPBAR burst and lithium leaching test programs at PNNL (ADAMS Accession No. ML15345A424). As a result of the audit, NRC requested that TVA submit certain portion of the TPBAR burst test document. The Areva Small Array Test (SMART) facility is an open-loop, high-temperature, low-pressure test facility built in 1997. TVA submitted information on statistical analysis for TPBAR test pellet length loss, pellet length effect on TPBAR burst testing, description of the SMART facility, test method and test article burst test summaries.

The test information is also applicable to the WBN, Unit 2, TPC license amendment.

The core characteristics for fuel management are nearly identical between WBN, Unit 1, and WBN, Unit 2. Both plants have been analyzed (for the purpose of tritium production) at the same power level, utilize the same fuel product, and have similar inlet temperature and flow values. As a result, the same loading pattern developed to support the WBN, Unit 1, Amendment No. 107 is also applicable to WBN, Unit 2. The same analysis that led to the limiting boron concentration for the WBN, Unit 1, post-LOCA evaluation can be used and compared to the different limit for WBN, Unit 2.

The NRC staff verified the post-LOCA subcriticality evaluation and the post-LOCA lithium leaching analysis and assumption and found that the sub-criticality evaluation for WBN, Unit 2, is acceptable.

3.2.18 Reactor Pressure Vessel and Reactor Pressure Boundary Integrity 3.2.18.1 Reactor Pressure Vessel Material Surveillance Program By letter dated September 5, 2017 (ADAMS Accession No. ML17248A420), the licensee assessed the impacts of this LAR on the Reactor Pressure Vessel (RPV) Material Surveillance

Program and RPV surveillance capsule withdrawal schedule for the unit. In that submittal, the licensee proposed a change to the withdrawal schedule removal time for RPV Capsule U, such that the capsule would be removed from the RPV during the end-of-Cycle 2 refueling outage (RFO), when the capsule is projected to achieve a cumulative neutron fluence exposure of 0. 771 x 10 19 n/cm 2 (E > 1.0 MeV). The NRC staff approved the change to the RPV surveillance capsule withdrawal schedule in a letter and safety evaluation dated November 20, 2017 (ADAMS Accession No. ML17312A260).

Thus, the LAR's impact on the RPV surveillance program and withdrawal schedule for WBN, Unit 2, has already been evaluated and found to be acceptable by the NRC staff in the safety evaluation, dated November 20, 2017.

3.2.18.2 Pressure-Temperature (P-T) Limit Curve Analysis The scope of the NRC staff's review of the LAR in relation to the P-T limit curves for the RPV and reactor coolant pressure boundary is limited only to the NRC staff's review for confirming: (a) that the new P-T limit curves for 32 effective-full-power-years (EFPY) have been calculated in accordance with the methodology that is required to be used for P-T limit calculations by TS Section 5.9.6.b, and (b) that the proposed P-T limit curves for 32 EFPY will meet the requirements in 10 CFR Part 50, Appendix G for incorporating conservatisms and required safety margins in P-T limits for U.S. light water reactors. For this objective, the NRC staff performed its review in accordance with the guidelines in Sections 5.3.2 of NUREG-0800 and NRC Branch Technical Position (BTP) 5-3.

In the letter dated December 20, 2017, the licensee assessed the impacts of the LAR on the licensed P-T limit curve analyses for WBN, Unit 2. Specifically, the applicant addressed these matters in Sections 1 - 9 and Appendices A - C of Westinghouse Report W CAP-18191-N P, Revision O (ADAMS Accession No. ML17289A327). The new P-T limit curves effective to 32 EFPY were provided in Chapter 8 of WCAP-18191-NP, Revision 0. On February 9, 2018 (ADAMS Accession No. ML18040A434), TVA submitted an updated Pressure-Temperature Letter Report (PTLR) for WBN, Unit 2. In Appendix B of WCAP-18191-NP, Revision 0, TVA developed P-T limit curves for the RPV inlet and outlet nozzles under Service B Cooldown Conditions in order to demonstrate the RPV cooldown curves for the nozzles at 32 EFPY would be bounded by those calculated for the RPV beltline region under steady state, -20

~F/hr, -40 °F/hr, -60 °F/hr and -100 °F/hr cooldown rates.

The NRC staff confirmed that, with the exception of the topic that is addressed in the next two paragraphs, the licensee calculated the new P-T limit curves for 32 EFPY in accordance with the approved analytical methodology in WCAP-14040-NP-A, Revision 4. The NRC staff also verified thatthe updated P-T limits for 32 EFPYin Chapter 8 ofWCAP-18191-NP, Revision 0, were based on the 1/4T and 3/4T RT NOT values for the most limiting RPV component for RT NOT, which is RPV intermediate shell forging 05 (as fabricated from heat of material No. 527828).

In NRC Regulatory Information Summary (RIS) No. 2014-11 (ADAMS Accession No. ML14149A165), the NRC staff informed the industry that the assessment of plant-specific P-T limits must account for additional stress intensities imparted from discontinuities in the RPV design, including those imparted to the RPV shell from nozzles adjoined to the RPV. Specifically, in this RIS, the NRC staff stated that the RPV material with the highest reference temperature may

not always produce the most limiting P-T limits, because the consideration of stress levels from structural discontinuities (including nozzles joined to the RPV) may produce lower allowable pressures in the P-T limit calculations. Thus, the RIS informs the industry that the existence of structural discontinuities in the design of an RPV component with a lower reference temperature (i.e., RT NOT value) may result in more restrictive P-T limits than those calculated for the RPV beltline shell component with the most limiting upper-bound reference temperature.

The NRC staff reviewed the contents of WCAP-18191-NP, Appendix B, and verified that the licensee performed confirmatory P-T limit curve calculations of the RPV inlet and outlet nozzles in order to address the information in RIS No. 2014-11. The NRC staff noted that the licensee performed the nozzle calculations solely for the purpose of verifying that: (a) the P-T limits provided for the RPV beltline in Chapter 8 of the report would bound any P-T limit curves established for the design of the RPV inlet and outlet nozzles in Appendix B of the report, and (b) the P-T limit curves for 32 EFPY that are given in Chapter 8 of the report would not need to include the P-T limit curves for the RPV inlet and outlets nozzles provided in Appendix B of the report. The NRC staff verified that the licensee used the staff-developed methodology in Oak Ridge National Laboratory (ORNL) Report No. ORNL/TM-2010/246, "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles" (ADAMS Accession No. ML110060164), to calculate the P-T limits of these nozzles because the scope of methodology in WCAP-14040-NP-A, Revision 4, does not include any methods for performing P-T limit calculations of RPV nozzle locations. The NRC staff did not find any technical reason to reject the ORNL report as a basis for performing the P-T limit calculations of the RPV inlet and outlet nozzles. Thus, the NRC staff finds that the licensee has: (a) addressed the information in RIS 2014-11 for assessing regions of RPV discontinuities, (b) provided sufficient demonstration that the updated P-T limit curves for the beltline region of RPV in Chapter 8 of W CAP-18191-N P, Revision 0, are bounding for those that the licensee has calculated for the RPV inlet and outlet nozzles through 32 EFPY, and (c) provided adequate demonstration that the P-T limit curves provided for 32 EFPY in Chapter 8 of WCAP-18191-NP, Revision 0, are the proper P-T limit curves for 32 EFPY that the licensee will need to incorporate into the next update and submittal of the PTLR for the facility.

During the NRC staffs review, the NRC staff also observed that, in WCAP-18191-NP, Revision 0, the licensee set the LlRT NOT values and the associated cra values used in the margin term calculations of some components to values of O°F if the calculated or projected adjusted reference temperature shifts for the components were less than 25 °F. The licensee applied this basis to the RT NDT calculations of the following RPV extended beltline components: (a) upper shell forging 06, (b) upper-to-intermediate shell circumferential weld W06, (c) lower shell to bottom head ring weld seam W04, and (d) bottom head ring 03. The licensee cited NRC Technical Letter Report {TLR) No. TLR-RES/DE/CIB-2013-01, "Evaluation of the Beltline Region for Nuclear Reactor Pressure Vessels," as the basis for setting the LlRTNOT values and cra values of these components to a value of O °F.

The NRC staff noted that the licensee's use of the TLR to discount LlRTNOT values, and associated cra values of these RPV extended beltline components would not be consistent with the information in RIS No. 2014-11, because the design of the components could include areas of regional discontinuities. For these components, the information in the RIS would call for the

licensee to account for both the impacts of neutron fluence exposure and applied stresses on the adequacy of the P-T limit curves for the facility. However, the NRC staff also verified, that for the

licensee's development of the P-T limits for the facility, the evaluation of these RPV extended beltline components would not cause any of these extended beltline components to be limiting for the RT NOT inputs in the P-T limit calculations even if the 6RT NOT values and cra values of the components were calculated in accordance with the guidelines in RG 1.99, Revision 2. 1 Thus, the NRC staff finds that the licensee's use of the TLR for this type of calculational objective would not impact the conservatisms or margins in the P-T limit curves for 32 EFPY that were included in Chapter 8 of the WCAP-18191-NP, Revision O report. However, the NRC staff is not using the scope of this safety evaluation as a basis for accepting or endorsing the alternative 6RT NOT and cra basis in TLR-RES/DE/CIB-2013-01 for use in RPV component-specific RT NOT calculations.

Therefore, the NRC staff finds that TVA's P-T limit curves for 32 EFPY in Chapter 8 of WCAP-18191-NP, Revision O have been calculated in accordance with the requirements of TS Section 5.9.6.b, because the NRC staff has determined that the P-T limits have been calculated in accordance with staff-approved P-T limit methodology in WCAP-14040-NP, Revision 4. The NRC staff also finds that the proposed P-T limit curves for 32 EFPY are in compliance with the requirements in 10 CFR Part 50, Appendix G, because the NRC staff has determined that the curves are at least as conservative as those that would be generated if the methods of analysis in the 2010 edition of ASME Code Section XI, Appendix G were used for calculations.

Based on this review, the NRC staff concludes that the licensee has adequately addressed the impacts of this LAR on the acceptability of the P-T limits for the facility.

3.2.18.3 Upper Shelf Energy (USE) Analysis The scope of the NRC staffs review of the LAR in relation to the USE analysis for the facility is limited only to the NRC staffs confirming that the updated end-of-life (EOL) USE values for forging and weld materials in the beltline of the RPV will meet the lower bound acceptance criterion of 50 ft-lbs set for EOL USE values in 10 CFR Part 50, Appendix G. For this objective, the NRC staff performed its review in accordance with the guidelines in Sections 5.3.2 of NUREG-0800 and NRC Branch Technical Position (BTP) 5-3.

In the letter dated December 20, 2017, the licensee assessed the impacts that the LAR may have on the existing USE analysis for WBN, Unit 2. Specifically, the licensee addressed these matters in Appendix D of WCAP-18191-NP, Revision 0, which was included as an enclosure in TVA Letter, dated December 20, 2017. In this analysis, the licensee provided updated EOL 1/4T USE values (i.e., USE values for 32 EFPY) for the RPV beltline and extended beltline forging, ring and weld components using the methodology for performing these calculations in Position 1.2 of RG 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials" (ADAMS Accession No. ML003740284 ). For its assessment, the licensee identified that the USE assessment for the RPV is limited by the 1/4T EOL USE calculation performed for RPV intermediate shell forging 05 (as fabricated from heat of material No. 527828). For this component, the licensee reported a limiting EOL USE of 72.0 ft-lb at 32 EFPY. Based on its assessment, the licensee concluded that 1 That is, both the NRC staff and the licensee have independently verified that the P-T limits for 32 EFPY are limited by the 1/4T and 3/4T RT Nor calculations for RPV beltline shell forging #05 (which is made from Forging Heat No. 527828) and not by the 1/4T and 3/4T RTNDT calculations for RPV upper shell forging 06, RPV upper-to-intermediate shell circumferential weld W06, RPV lower shell-to-bottom head ring weld seam W04, or bottom head ring 03.

the LAR would not have any adverse impact on the USE assessments for the ferritic components in the beltline and extended beltline regions of the RPV because the 1/4T EOL USE values for these components will all remain above the 50 ft-lb EOL USE criterion specified in 10 CFR Part 50, Appendix G.

The NRC staff performed independent 1/4T EOL USE calculations for all of the ferritic forging, ring, and weld components that were reported as comprising the beltline and extended beltline regions of the RPV. Similar to the calculation performed by the licensee, the NRC staff verified that the USE assessment is limited by the USE assessment for RPV intermediate shell forging 05. For this component, the NRC staff calculated an USE value of 72.0 ft-lb at 32 EFPY using Position 1.2 in RG 1.99, Revision 2. For the limiting RPV circumferential weld (i.e., for lower shell-to-bottom ring weld 04), the NRC staff calculated a 1/4T EOL USE value of 80.90 ft-lb.

The NRC staff also performed an additional 1/4T EOL USE calculation for the sole RPV ferritic weld or forging component that has sufficient sources of RPV surveillance data available for incorporation into the USE assessment (i.e., for the USE assessment of RPV lower shell-to-intermediate shell circumferential weld 05 [Weld Heat 8950751). For this weld component, the NRC staff calculated a 1/4T EOL USE value of 108.6 ft-lb based on the use of Regulatory Position 2.2 in RG 1.99, Revision 2, and the availability of credible sister plant data for the component's weld heat, as compiled from the Catawba Nuclear Station, Unit 1, McGuire Nuclear Station, Unit 2, and WBN, Unit 2, RPV materials surveillance programs. Based on this assessment, the NRC staff finds the licensee's 1/4T EOL USE analysis is valid and provides adequate demonstration that the USE values for the RPV beltline and extended beltline forgings, rings, and weld components will remain above the 50 ft-lb lower bound acceptance criterion established in 10 CFR Part 50, Appendix G, for USE values that are projected to the end of the current licensed operating period. Thus, the NRC staff concludes that the LAR will not have an adverse impact on the 1/4T EOL USE analysis for the facility (i.e., the licensee's USE evaluation or 32 EFPY).

3.2.18.4 Pressurized Thermal Shock (PTS) Analysis The scope of the NRC staff's review of the LAR in relation to the PTS analysis for WBN, Unit 2, is limited only to the NRC staff's review for confirming that the updated RT PTs values for the RPV beltline component materials will meet the PTS screening criteria limits that are specified in the 10 CFR 50.61 rule. 2 For this objective, the NRC staff performed its review in accordance with the guidelines in Section 5.3.2 of NUREG-0800 and Branch Technical Position (BTP) 5-3.

In the TVA letter dated December 20, 2017, the licensee assessed the impacts of the LAR on the PTS analysis for WBN, Unit 2. Specifically, the licensee addressed these matters in Appendix E, "Pressurized Thermal Shock and Emergency Response Guidelines Limits Evaluation," of the Westinghouse Report WCAP-18191-NP, Revision O (ADAMS Accession No. ML17289A327). In this analysis, the licensee provided updated RTPTS values (i.e., RT PTs values for 32 EFPY) for the RPV beltline and extended beltline forging, ring and weld components in Appendix E of the 2 As stated in 10 CFR 50.61 (b)(2), the PTS screening criteria are 270 °F for ferritic RPV forging, plate, or axial weld materials (longitudinal weld materials) in the beltline region of the RPV, and 300 °F for ferritic RPV circumferential weld materials (girth weld materials) in the beltline of the RPV.

Westinghouse Report WCAP-18191-NP, Revision 0, in the TVA letter dated December 20, 2017, using the methodology for performing these calculations as specified in 10 CFR 50 .61 .

The NRC staff confirmed that, with the exceptions of the topics that follow in this paragraph, the licensee calculated the RTPTs values for 32 EFPY in accordance with the required analytical methodology for performing RT PTs calculations in the 10 CFR 50.61 rule. Both the licensee and the NRC staff verified that the analysis is limited by the RT PTs calculation for RPV intermediate shell forging 05 (made from forging Heat No. 527828). The NRC staff verified that, for the purposes of the calculations, the licensee used the initial RT NDT value and the copper (Cu) and nickel (Ni) alloying content values previously reported and approved for this forging material in WCAP-17035-NP, Revision 2 (ADAMS Accession No. ML100550651 ). 3 For this component, both the licensee and NRC staff calculated an EOL RT PTS value of 84.3 °F based on use of the projected 32 EFPY neutron fluence reported for this material in WCAP-18191-NP, Revision 0, and use of the chemistry factor table for plate or forging materials in the 10 CFR 50.61 rule. Thus, based on this review, the NRC staff confirmed that the RT PTs value for this limiting material in the RPV will meet the NRC staffs screening criterion of 270 °F specified in 10 CFR 50.61 for RPV forging materials at the end of the current licensed operating period.

The NRC staff noted that there was one review matter that necessitated further evaluation by the staff when assessed in relation to PTS requirements stated in 10 CFR 50.61. Specifically, the NRC staff observed that, in WCAP-18191-NP, Revision 0, the licensee set the aRTPTs values and the associated Ot:i. values used in the margin term calculations of some components to values of 0 °F if the calculated or projected adjusted reference temperature shifts (i.e., a RTPTs values) for the components were less than 25 °F. This basis was applied to the RT PTs calculations for the following RPVextended beltline components: (a) upper shell forging 06, (b) upper-to-intermediate shell circumferential weld W06, (c) lower shell to bottom head ring weld seam W04, and (d) bottom head ring 03. The licensee cited NRC TLR No. TLR-RES/DE/CIB-2013-01, "Evaluation of the Beltline Region for Nuclear Reactor Pressure Vessels", as the basis for setting the aRTPTs values and associated Ot:i. values of these components to a value of O °F.

The NRC staff noted that the TLR's methodology for discounting aRT PTs and associated oa values when the predicted temperature shifts are less than 25°F may not be consistent with the criteria for performing RT PTS calculations in the 10 CFR 50.61 rule. The rule requires the calculations of RTPTs for RPV beltline materials must account for the effects of neutron radiation. Similarly, Ot:i.

values are always established as non-zero °F values in accordance with prescribed methods stated in the 10 CFR 50.61 rule. However, the NRC staff also verified, that for the licensee's updated PTS assessment of the facility, the evaluation of extended beltline components would not cause any of the RT PTS calculations for the components to become limiting for the 3 In WCAP-17035-NP, Revision 2, the licensee reported an initial RTNor (i.e., RTNDT(UJ) value of 14 °F and Cu and Ni alloying contents of0.05 Wt.% and 0. 78 Wt.% for this material. The staff verified that, although the heat of material for RPV Forging 05 (Heat No. 527828) is represented in the site-specific RPV surveillance program for WBN, Unit 2, the licensee has not yet removed a sufficient number for surveillance capsules from its RPV material surveillance program to warrant performance of a surveillance data-based RTPrs calculation of this RPV forging material in the LAR submittal.

The requirements in 10 CFR 50.61 will require TVA to perform a surveillance data-based RTPrscalculation for this material after the licensee removes the second RPV surveillance capsule (Capsule W} in accordance with the licensee's site-specific surveillance program and has compiled a sufficient amount of surveillance data for this forging material. For more information, refer to the requirements in 10 CFR Part 50, Appendix H, and the staffs approval of the RPV surveillance program withdrawal schedule that was granted in a letter and safety evaluation dated November 20, 2017 (ADAMS Accession No. ML17312A260). This is the surveillance program withdrawal schedule that is provided in Appendix F of the WCAP-18191-NP, Revision O report.

RT PTs evaluation even if the tiRTNDT values and 01:,. values of the extended beltline components were calculated in accordance with the specified requirements for calculating these parameters in the 10 CFR 50.61 rule. 4 Thus, the NRC staff finds that the use of the TLR does not impact the component-specific margin for the most limiting RPV component for RT PTS in Appendix E of the WCAP-18'191-NP, Revision O report. Thus, the NRC staff concludes that the LAR will not have an adverse impact on the PTS analysis for the facility. However, the NRC staff is not using the scope of this safety evaluation as a basis for accepting or endorsing the alternative flRT NDT and 01:,. basis in TLR-RES/DE/CIB-2013-01 for use in RPV component-specific RT PTs calculations.

3.2.19 Plant Systems Review License Amendment No. 40 to the WBN, Unit 1, operating license, dated September 23, 2002 (ADAMS Accession No. ML022540925), authorized irradiation up to 2304 TPBARs in the WBN, Unit 1, core,. The licensee identified the number of TPBARs to be irradiated in the safety evaluation performed for each reload core and noted the number in the Core Operating Limits Report for each fuel cycle. The licensee operated with TPBARs installed within the core at numbers well below the original authorized maximum through several operating cycles.

Subsequent license amendments reduced the number of TPBARs authorized for irradiation due to technical issues and concerns with tritium permeation through the TPBAR cladding during operation. After a design change reduced the concern with tritium permeation and the technical issues had been resolved, License Amendment No. 107 to the WBN, Unit 1, operating license, dated July 29, 2016 (ADAMS Accession No. ML16159A057) authorized irradiation of up to 1792 TPBARs in the WBN, Unit 1, core. Thus, the shared spent fuel and TPBAR storage and handling activities and equipment have been evaluated as part of the WBN, Unit 1, license amendment, leaving only the effects of changes in the number of irradiated TPBARs present in the SFP resulting from TPBAR irradiation in the WBN, Unit 2, core unevaluated.

3.2.19.1 Handling of TPBARs In NUREG-1672, the NRC staff identified three main issues pertaining to TPBAR handling.

These issues included:

activities required to remove the TPBARs from the fuel assemblies,

activities required to prepare the TPBARs for shipment, and

post-irradiation movement of the TPBARs outside of the fuel assemblies.

As described by the licensee in Enclosure 1 of the LAR dated December 20, 2017, TPBARs are inserted into new fuel assemblies or placed in specially-designed transport containers prior to shipment to the site. The physical design of the TPBAR and the burnable poison rod assemblies (BPRA) are nearly identical in design and weight, which permits the use of the BPRA tool in transferring TPBARs to selected fuel assemblies. The fuel assembly with the TPBAR is then transferred te> the reactor vessel for placement in the core using the same methods, procedures, and equipment as non-TPBAR fuel. Following irradiation in the reactor, the spent fuel assemblies 4 As stated earlier, both the NRC staff and the licensee have independently verified that the PTS evaluation for the RPV is limited by the RT PTS calculation for RPV beltline shell forging #05 (which is made from Forging Heat No. 527828),

with a projected RTPTs value of 84.3 °F. The licensee has adequately demonstrated that the RTPTS value for RPV Forging 05 will meet the PTS screening criterion of 270 °F at the end of the current licensed operation period for the unit (i.e. at 32 EFPY).

containing the TPBARs are returned to the SFP and stored in the spent fuel racks. The licensee states that approximately 30 days after refueling is complete, TPBAR consolidation begins. The time to commence consolidation is not limited by any safety issues (e.g., decay heat), but rather is based upon scheduling. The 30-day estimate corresponds to when the licensee expects to be finished with all outage related activities and can begin consolidation efforts.

For the consolidation process, the licensee uses a specially-designed TPBAR Consolidation Fixture (TCF) installed in the cask loading pit. During the consolidation process, the TPBARs are withdrawn from their storage location and transferred from the SFP to the TCF using the TPBAR Assembly Handling Tool suspended from the SFP bridge crane. A specially-designed release tool is used to detach individual TPBARs from the baseplate. Once released, the TPBAR slides into a specially designed canister, which has a storage capacity of 300 TPBARs. In the event of a stuck TPBAR within the fixture, several design features allow the licensee to ensure the safe placement of the TPBAR in the canister. Loaded canisters will be returned to the spent fuel racks, until they are removed from the site. The spent fuel bridge crane is used to handle the canisters, and the NRC staff's evaluation of this activity is addressed with the light load handling system.

The consolidation and transportation of irradiated TPBARs involves the movement of heavy loads using the auxiliary building crane. In preparation for shipment, the loaded consolidation canisters are transferred from the SFP to a specially-designed cask that will be placed in the cask loading pit. Once loaded, the cask will be removed from the cask loading pit and prepared for transportation. These activities involve heavy load lifts. The initial installation of the TCF in the cask pit and movement of the TCF to the cask laydown area to allow for cask handling and loading also involve heavy load lifts.

The licensee provided the following statement regarding safety of the heavy load lifts in Section 4.1.1 of Enclosure 1 of the LAR:

For TPBAR associated heavy load lifts, the auxiliary hoist of the Auxiliary Building 125/10-Ton crane meets NUREG-0612, Section 5.1.2, option number one by complying with Section 5.1.6, specifically Appendix C of NUREG-0612 for existing cranes, except for the load hang-up protection and associated testing. Lifts are controlled by site procedures, that require pre-lift briefings, trained operators, etc.

Therefore, lifts will be adequately monitored to help preclude load hang-ups.

Certain items of compliance are contingent upon the fact that the loads for TPBAR associated lifts are less than half of the hook capacity, thereby yielding increased safety factors for the structural/wear-related requirements.

Lifting devices and interfacing lift points for TPBAR-related heavy loads are required to meet the requirements of NUREG-0612 Section 5.1.6, either by redundant paths or increased safety factors, as delineated in ANSI/ASME N14.6.

The licensee clarified the specific qualifications of the handling systems used for each lift. For the TPBAR transportation cask, the maximum cask weight is 52,000 pounds, and it will be handled by the main 125-ton single-failure-proof hoist. The licensee described that the TCF would be handled in two parts, each of which was less than half the rated capacity of the 10-ton auxiliary hoist. For the TCF lifts, the licensee considered the auxiliary hoist equivalent to single-failure-proof because the hoist complies with the guidelines of Appendix C to NUREG-0612

for upgrades of existing cranes, with the exception of load hang-up protection and the conduct of associated testing. Compliance with these guidelines was based in part on the increased safety factors for crane components provided by the TCF lifts weighing less than half the rated capacity of the auxiliary hoist. To compensate for the absence of load-hang-up protection, TVA will ensure that all lifts are controlled by site procedures. These procedures require pre-lift briefings and trained operators to adequately monitor the lift operation to preclude load hang-up. Lifting devices and the interfacing lift points for TPBAR heavy loads comply with the redundancy and increased-safety factors specified in Sections 5.1.1 and 5.1.6 of NUREG-0612 for single-failure-proof handling systems.

The consolidation activities described above and the measures for control of heavy loads are essentially unchanged from those approved for WBN, Unit 1, operations and use the same facilities and equipment. The NRC staff concludes that the equipment and administrative controls that TVA states will be applied to handling TPBAR assemblies provide reasonable assurance of safety. The combination of the physical separation of the TCF from the stored fuel, along with the proposed personnel training, equipment inspections, and procedural controls provide adequate defense-in-depth to assure an extremely small probability of a load drop during TCF handling operations that would damage spent fuel, which satisfies NUREG-0612 guidelines. The described handling of the TPBAR cask is fully in conformance with the guidelines of NUREG-0612 for single-failure-proof handling systems. Therefore, the NRC staff finds the additional heavy load handling activities associated with loading TPBARs into the WBN, Unit 2, core to be acceptable.

3.2.19.2 Assessment of Combustible Gas Control with a Tritium Production Core Criteria for combustible gas control is provided in relevant requirements of 10 CFR 50.44, "Combustible gas control for nuclear power reactors." This regulation was modified in 2003, and now specifies, in part, that pressurized-water reactors (PWR) with ice condenser containments must have the capability for controlling combustible gas generated from a metal-water reaction involving 75 percent of the fuel cladding surrounding the active fuel region so that there is no loss of containment structural integrity and equipment necessary to establish and maintain safe shutdown. Previous requirements for hydrogen recombiners were removed from the regulation.

The NRC provided updated guidance for implementation of the regulation in RG 1. 7, "Control of Combustible Gas Concentrations in Containment," Revision 3.

The licensee implemented the updated guidance at the WBN facility. Section 15.4.1.2 of the previous version of the WBN UFSAR contained an evaluation of post-LOCA hydrogen generation and recombiner initiation timing, but this section of the current WBN UFSAR (applicable to Units 1 and 2) now credits deliberate hydrogen ignition systems to control the hydrogen that may be released from the core during or following a severe accident. The additional hydrogen present in the core as tritium in the TPBARs would be easily accommodated by the deliberate hydrogen ignition system present within the WBN containment structures because the additional hydrogen (in the form of tritium) present due to operation with a TPC is a small fraction of the postulated hydrogen release specified by 10 CFR 50.44 and the hydrogen ignition system is not capacity-limited.

3.2.19.3 Light-Load Handling System The TPBAR consolidation process uses a light load handling system consisting of the SFP bridge crane, tooling, and a specialized consolidation fixture. During the consolidation process, the irradiated TPBAR assembly will be transported from its location in the SFP to the consolidation fixture located in the cask loading pit. The consolidation process is comprised of releasing the TPBAR rods from the baseplate, consolidating the individual rods in a specially designed canister, dispose of empty baseplates, transport the canisters for storage in the SFP, and load the canisters into shipping casks. Each canister is designed to accommodate a maximum of 300 TPBARs. The loaded canister is then transported to the designated SFP cell location using a canister handling tool suspended from the SFP bridge crane.

The SFP bridge crane will be used to handle TPBAR assemblies and consolidation canisters within the pool. The weight of a loaded canister submerged in water is less than 700 lbs. Since the SFP bridge has a rated load capacity of 4,000 lbs, the bridge has a large structural safety factor when handling the consolidation canister. An additional safety measure is provided by administrative controls requiring the use of a safety lanyard on the canister handling tool. This lanyard limits the canister descent in the fuel pool, prevents canister tipping, and is sized to stop the canister from the maximum hook speed of 40 ft per minute. An analysis completed by PNNL has demonstrated that no TPBAR cladding failures are expected to occur during an accidental impact with a rigid structure at that hook speed. Therefore, the TPBARs are adequately protected from damage during routine handling.

Although lacking many features specified in NUREG-0612 for single-failure-proof cranes, the SFP bridge and associated canister lifting tool provide a reliable means of handling the TPBAR consolidation canisters. The large safety factor resulting from the low weight of the consolidation canister relative to the rated load of the SFP bridge provides assurance that critical load supporting components, such as the hoist brakes, the load block attachment point, the wire rope, the reeving system, and the hoist drum bearings, would have an extremely low probability of failure. In addition, the SFP bridge does have some features common to single- failure-proof cranes, such as interlocks to prevent simultaneous motion of the bridge and hoist, redundant and diverse upper limit switches, and a load monitoring device. The canister lifting tool is designed in accordance with the requirements of American National Standards Institute (ANSI) N14.6, "Standard for Special Lifting Devices for Shipping Container Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials," such that it will also have increased safety factors. The tool has been and will continue to be used for handling of consolidation canisters containing TPBARs irradiated in WBN, Unit 1, and, consistent with the ANSI N14.6 standard, the tool has been subject to periodic testing and inspection to ensure safe operation. The tool has an air-actuated, fail closed safety latch to ensure the tool hook will not inadvertently disengage from the canister lifting bail.

The licensee states that the light-load handling system complies with the intent of NUREG-0612 for consolidation canister handling. Since each consolidation canister holds up to 300 TPBARs, and since the proposed maximum core inventory of TPBARs is 1,792 TPBARs, the NRC staff expects the total number of consolidation canister handling evolutions per operating cycle to be small. Considering the expected number of consolidation canister handling evolutions using the shared spent fuel storage and light load handling equipment for both WBN, Unit 1 and Unit 2, TPBARs handling, and the features identified to reduce the potential for loss of control during

handling, the NRC staff concludes that the potential for a drop of a consolidation canister is extremely small. This outcome satisfies the guidelines of Section 5.1 of NUREG-0612.

Therefore, the NRC staff finds that the equipment and administrative control measures proposed for handling of the consolidation canisters continues to be acceptable considering the potential small increase in the frequency of handling activities resulting from WBN, Unit 2, operation of a TPC.

3.2.19.4 Station Service Water System In the NRC staff's safety evaluation contained in NUREG-1672, the NRC staff identified the need to quantitatively evaluate the effect of increased SFP heat load resulting from placement of irradiated TPBARs in the pool on affected systems, including the service water system. The design basis function of the station service water system, which is the shared essential raw cooling water (ERCW) system for WBN, includes providing a cooling loop for heat removal from component cooling system (CCS). The ERCW supplies water from the ultimate heat sink (UHS) to cool the CCS. The CCS intermediate cooling loop, in turn, provides a heat sink to the SFP CCS and the residual heat removal (RHR) system.

The licensee's analysis showed the TPC impact on decay heat loads was approximately 0.3 MW(t), or 1 MBTU/hr. The estimated heat load considered both the decay heat generated by freshly discharged fuel assemblies during a RFO and the additional residual decay heat from the increased discharge rate of fuel assemblies into the pool when operating a TPC. The licensee determined that the net increase of 1 MBTU/hr of decay heat due to the TPC was a small fraction of the design basis limiting heat load imposed on the ERCW system and within available margin.

The licensee's ERCW analysis showed that the increase of 1 MBTU/hr of decay heat due to TPC produced an insignificant increase (less than 0.1 °F) in the ERCW temperature leaving the plant site.

Based upon the review of the licensee's analysis for the increased decay heat due to TPC, a comparison of the ERCW design capacity with the increase in heat load, and the resultant temperature increase of less than 0.1°F, the NRC staff concludes that the ERCW system has adequate cooling capacity and margin to perform its safety functions with the additional heat loads imposed by TPC activities at WBN, Unit 2, and that tritium production activities will not have an adverse impact on the ERCW heat removal capabilities.

3.2.19.5 Ultimate Heat Sink In the NRC staff's safety evaluation contained in NUREG-1672, the NRC staff identified the need to quantitatively evaluate the effect of increased SFP heat load resulting from placement of irradiated TPBARs in the pool on affected systems, including the UHS. The design basis function of the UHS is to provide an uninterrupted source of cooling water for decay heat removal.

The net increase in decay heat associated with operation of a TPC is approximately 1 MBTU/hr.

The additional increase in the decay heat load to the SFP CCS was decay heat shifted from RH Rs to the SFP CCS as a result of earlier start of core-offload and does not represent a net increase in CCS heat load on the UHS. Because the Tennessee River flow is large relative to the ERCW flow, the increased heat rejected to the river has no effect on inlet ERCW temperature. The licensee's analysis showed that the increase of 1 MBTU/hr in decay heat due to operation of a

TPC produced an insignificant increase (less than 0.1 °F) in the ERCW temperature leaving the plant site. Therefore, the additional heat loads imposed by the WBN, Unit 2, TPC operation will not have an adverse impact on the UHS heat removal capabilities.

3.2.19.6 New and Spent Fuel Storage In the NRC staff's safety evaluation contained in NUREG-1672, the NRC staff identified the need to evaluate the effect of activities to remove irradiated TPBARs from fuel assemblies and prepare them for shipment on the spent fuel storage racks. These activities have been conducted in the shared cask loading pit for the WBN, Unit 1, TPCs. These activities are planned to be conducted in the cask loading pit for the WBN, Unit 2, TPCs as well. Therefore, the only significant change affecting the racks would be the increased frequency of storage of loaded TPBAR consolidation canisters in some rack locations.

As part of TPC operation at WBN, Unit 1, the licensee evaluated the heat production from a fully loaded canister and its potential effect on the spent fuel racks. A TPBAR rod will only produce approximately 3 watts of heat 30 days after reactor shutdown. Assuming a fully loaded canister contains a maximum of 300 TPBAR rods, this translates into a maximum heat load of 900 watts per canister. This heat load is small given that adequate circulation is provided through the open topped canister and the drainage/cooling holes located on the sides and bottom of the canisters.

The NRC staff concluded that the described configuration will provide adequate natural circulation cooling to preclude adverse effects on the TPBARs, the fuel storage racks, and adjacent stored fuel. Since the equipment is common to WBN, Units 1 and 2, this assessment applies equally to WBN, Unit 2, operation of a TPC.

3.2.19. 7 Spent Fuel Pool Cooling and Cleanup System In the NRC staff's safety evaluation contained in NUREG-1672, the NRC staff identified the need to quantitatively evaluate the effect of increased SFP heat load resulting from placement of irradiated TPBARs in the pool on affected systems, including the SFP CCS. The WBN licensing basis for the SFP CCS provides for cycle-specific analysis to establish the minimum decay time prior to discharge of fuel assemblies to the SFP based on plant and environmental conditions at the time of the outage. The licensee's current analysis established 50.2 MBTU/hr as the bounding heat load for planned RFOs, considering the alternating discharges from the two WBN units and the combined effects of below-design cooling water temperatures and heat exchanger fouling conditions. This method provides assurance that the maximum SFP water temperature would not exceed established limits.

The NRC staff concludes that the effect of the WBN, Unit 2, TPC operation on SFP heat load would be acceptable based on administrative controls ensuring SFP total heat load and temperature remain within the current licensing basis values. In combination with the existing provision of an SFP cooling system having reliability consistent with the importance of decay heat removal at high heat loads and the provision of makeup system capable of maintaining SFP water level under accident conditions, fuel storage design capabilities would remain consistent with the requirements of GDC 61, of Appendix A to 10 CFR Part 50.

3.2.19.8 Component Cooling Water System The design basis function of the CCS includes providing an intermediate cooling loop for heat removal from several, safety-related heat exchangers and several non-safety-related components. Two of the highest heat loads placed on the CCS include SFP CCS and RHR system. These two heat removal systems are the primary means for cooling the plant and removing residual decay heat during later stages of plant cooldown and during outages.

The licensee's analysis showed that the CCS has adequate capacity and cooling margin to perform its safety and non-safety related functions with the additional heat loads imposed by tritium production activities for Unit 2. Section 9.2.2, "Component Cooling System," of the WBN UFSAR states that the highest design heat removal demand of approximately 192 MBtu/hr under normal operation occurs with one unit in Hot Shutdown and the other unit in Cold Shutdown. Thus, the decay heat load imposed by the TPBARs in the TPC core of the unit in Cold Shutdown represents approximately 0.5 percent of the total shared CCS heat removal demand and, since the Cold Shutdown heat removal is dominant, less than 1 percent of the Cold Shutdown unit heat removal demand, which is not considered significant. During the transition to refueling from cold shutdown, the heat load imposed by the TPBARs would transfer from the RHR system to the SFP CCS but remain a part of the heat load removed by CCS.

Based on the review of the licensee's analysis for the increased decay heat, the NRC staff found that the CCS has adequate cooling capacity to perform its safety and non-safety functions with the additional heat loads imposed by WBN, Unit 2, tritium production activities.

3.2.19.9 Demineralized Water Makeup System Operation with a TPC may increase tritium levels in the reactor coolant system (RCS) due to normal reactor tritium production plus tritium permeation from the TPBARs. To maintain the RCS tritium levels at conventional core levels, additional feed and bleed operations may be necessary.

Any increase in feed and bleed operations would place increased demands on the demineralized water makeup system (DWMS) for the required makeup.

The licensee performed an analysis to determine the adequacy of the existing DWMS to meet increased feed demands which may result from TPC operation. The licensee determined that the plant's current storage and purified water production capabilities are adequate to handle the potential increase in RCS tritium levels. This analysis also evaluated a potential operating occurrence involving the failure of two TPBARs at end of fuel cycle, which would raise the tritium concentration in the RCS. The existing demineralized water makeup system has a capacity to produce demineralized makeup water at a nominal 400 gallons per minute rate. The licensee's evaluation concluded that the demineralized water makeup system capacity is adequate for plant operation with the TPC.

Since the necessary dilution for potential TPC operational impacts is within the normal capacity of the system and the purpose of the dilution is to maintain monitored releases within applicable limits and as low as reasonably achievable per 10 CFR Part 20, the NRC staff concludes that sufficient storage and water makeup capacity is available to adequately meet any additional feed and bleed demands from WBN, Unit 2, TPC operation.

3.2.19.10 Liquid Waste Management System Primary coolant discharge volumes have the potential to rise with TPC operations, if increased feed and bleed is required. As stated in Interface Issue 13, operation with a TPC may increase RCS tritium levels due to normal reactor tritium production plus tritium permeation from the TPBARs. Under these normal TPC operating conditions, the licensee performed an evaluation of the major source of liquid waste to the systems and determined that the normal reactor RCS feed and bleed operation for boron control will be maintained throughout the plant's operating cycle. Therefore, the anticipated coolant discharges with the TPC will be comparable to current plant practice. In the event increased RCS feed and bleed is required to reduce RCS tritium concentration, it may be necessary to temporarily store the increased volume of tritiated liquid onsite in order to allow discharge of other plant liquid waste or to dilute the tritiated liquid to ensure that 10 CFR Part 20 discharge limits are met.

To accommodate this liquid waste resulting from RCS feed and bleed operations, WBN has three large tanks in the liquid radwaste system. These tanks include the 500,000 gallon capacity tritiated water storage tank, which has a capacity many times that of the RCS volume. The licensee determined that these tanks can be used for liquid effluent holdup, dilution, and timing of releases to ensure that the 10 CFR Part 20 effluent concentration limit values are met.

Based on the above, the NRC staff concludes that WBN has sufficient liquid radioactive waste storage capacity to adequately meet any additional feed and bleed demands that may result from WBN, Unit 2, TPC operations and to ensure that compliance with all regulatory limits on tritium discharge will be maintained.

3.2.19.11 Process and Effluent Radiological Monitoring and Sampling System The acceptance criteria for the process and effluent radiological monitoring and sampling system are based on meeting the relevant requirements specified in 10 CFR 20.106 as it relates to radioactivity in effluents to unrestricted areas, GDC 60 as it relates to waste management design and the control of releases of radioactive materials to the environments, and GDC 63 and GDC 64 as they relate to the radioactive waste system design for monitoring radiation levels and leakage.

The licensee had reviewed the WBN process and effluent monitoring and sampling equipment prior to operation with a TPC. Although no additional sampling points were necessary, the licensee determined that modifications to effluent monitoring capability were necessary to support TPC operations. These changes involved modifications to the auxiliary building and shield building heating, ventilating, and air conditioning (HVAC) exhaust sampling from periodic grab samples to continuous effluent sampling with fixed monitoring systems. These systems do not initiate any automatic actions, and the licensee developed plant-specific procedures to address conditions related to TPC operation.

With the described enhancements implemented to monitor the shared structures for WBN, Unit 1, TPC operations, the NRC staff concludes that the WBN process and effluent radiological monitoring and sampling systems would continue to be sufficient to adequately monitor potential tritium release paths arising from WBN, Unit 2, TPC operations.

3.2.20 Technical Analyses Conclusion The NRC staff finds that the new operating and accident conditions associated with the licensee's proposed increase in the maximum number of TPBARs allowed to be loaded in the WBN, Unit 2, core were analyzed through the appropriate use of NRG-approved methodologies in accordance with NRC guidelines. The NRC staff further finds that the licensee used methods consistent with the regulatory requirements and guidance identified in Section 2.0 above.

3.3 Technical Specifications Changes TS 4.2.1, "Fuel Assemblies" contains a description of the fuel assembly materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety which are not covered in safety limits, LCOs or SRs.

3.3.1 WBN, Unit 2, TS 4.2.1, "Fuel Assemblies" Current WBN, Unit 2, TS 4.2.1 The current WBN, Unit 2, TS 4.2.1 states:

The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zirlo fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

Proposed WBN, Unit 2, TS 4.2.1 The licensee proposed the following for WBN, Unit 2, TS 4.2.1:

The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zirlo fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. For Unit 2, Watts Bar is authorized to place a maximum of 1,792 Tritium Producing Burnable Absorber Rods into the reactor in an operating cycle.

Revisions to WBN, Unit 2, TS 4.2.1, "Fuel Assemblies" Review The NRC staff reviewed the proposed change and determined it is acceptable because the additional sentence, "For Unit 2, Watts Bar is authorized to place a maximum of 1792 Tritium Producing Burnable Absorber Rods into the reactor in an operating cycle.", is necessary to ensure TS 4.2.1, as amended, continues to contain a description of the fuel assembly materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety which are not covered in safety limits, LCOs or SRs.

The NRC staff finds the proposed change to TS 4.2.1 acceptable based on the above review.

3.3.2 WBN, Units 1 and 2, TS 3.7.15, "Spent Fuel Assembly Storage" CurrentWBN, Units 1 and 2, TS 3.7.15 The current WBN, Units 1 and 2, TS 3.7.15, in part, states:

3.7.15 Spent Fuel Assembly Storage LCO 3. 7.15 The combination of initial enrichment and burnup of each spent fuel assembly stored shall be in accordance with Specification 4.3.1.1.

SR 3.7.15.1 Verify by administrative means the initial enrichment and burnup of the fuel assembly is in accordance with Specification 4.3.1.1.

Proposed WBN, Units 1 and 2, TS 3.7.15 The licensee proposed the following change~ to WBN, Units 1 and 2, TS 3.7.15:

3.7.15 Spent Fuel Pool Assembly Storage LCO 3. 7 .15 The initial enrichment of each fuel assembly stored shall be in accordance with Specification 4.3.1.1.

SR 3. 7 .15.1 Verify by administrative means the initial enrichment of the fuel assembly is in accordance with Specification 4.3.1.1.

Revisions to TS 3.7.15, "Spent Fuel Assembly Storage" Review The licensee proposed revising the TS 3. 7 .15 title from "Spent Fuel Assembly Storage" to "Spent Fuel Pool Assembly Storage" to align it with NUREG-1431. The NRC staff determined this proposed change is administrative and acceptable.

The NRC staff reviewed the proposed changes to the TS 3.7.15 LCO and SR and determined them to be acceptable, because the new criticality analysis is based on the initial enrichment of the new and spent fuel assembles in the SFP, the revised LCO statement lists the lowest functional capability or performance level of equipment required for safe operation of the facility, and the revised SR will ensure that the revised LCO will be met.

The NRC staff finds the proposed change to TS 3. 7.15 acceptable based on the above review.

3.3.3 WBN, Units 1 and 2, TS 3.9.9, "Spent Fuel Pool Boron Concentration" Current WBN, Units 1 and 2, TS 3.9.9 The current WBN, Units 1 and 2, TS 3.9.9, in part, states:

LCO 3.9.9 Boron concentration of the spent fuel pool shall be .::2000 ppm.

A.1 Suspend fuel movement.

SR 3.9.9.1 Verify boron concentration in the spent fuel pool is .::2000 ppm.

FREQUENCY Prior to movement of fuel in the spent fuel pool AND 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months thereafter Proposed WBN, Units 1 and 2, TS 3.9.9 The licensee proposed the following changes to WBN, Units 1 and 2, TS 3.9.9:

LCO 3.9.9 Boron concentration of the spent fuel pool shall be .::2300 ppm.

A.1 Initiate action to restore fuel storage pool boron concentration to within limit.

SR 3.9.9.1 Verify boron concentration in the spent fuel pool is .::2300 ppm.

FREQUENCY 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Revisions to TS 3.9.9, "Spent Fuel Pool Boron Concentration" Review The licensee proposed revising the TS 3.9.9, "Spent Fuel Pool Boron Concentration" LCO statement from "Boron concentration of the spent fuel pool shall be .:: 2000 ppm." to "Boron concentration of the spent fuel pool shall be .:: 2300 ppm." The applicability would be changed from "During fuel movement in the flooded spent fuel pool" to "Whenever any fuel assembly is stored in the flooded spent fuel pool." Required Action A.1 would be changed from "Suspend fuel movement" with a Completion Time of Immediately to "Initiate action to restore fuel storage pool boron concentration to within limit" with a Completion Time of Immediately.

Section SR 3.9.9.1.1 would be revised from "Verify boron concentration in the spent fuel pool is

.:: 2000 ppm" with a frequency of "Prior to movement of fuel in the spent fuel pool AND 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months thereafter to "Verify boron concentration in the spent fuel pool is .:: 2300 ppm" with a frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The NRC staff reviewed the proposed changes and determined them to be acceptable, because the new minimum value for boron concentration and applicability ensure the initial condition assumption in the boron dilution analysis will be preserved and the revised LCO statement lists the lowest functional capability or performance level of equipment required for safe operation of the facility, the TS contains appropriate remedial actions to be taken in the event the LCO is not met, and the SR will ensure that the new LCO will be met.

The NRC staff finds the proposed changes to TS 3.9.9 acceptable based on the above review.

3.3.4 WBN, Units 1 and 2, TS 4.3.1, "Criticality" Current WBN, Units 1 and 2, TS 4.3.1 The current WBN, Units 1 and 2, TS 4.3.1 states:

4.3.1.1 The spent fuel storage racks (shown in Figure 4.3-1) are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent (wt%);

b. keff s 0.95 if fully flooded with unborated water, which, includes an allowance for uncertainties as described in Sections 4.3.2. 7 and 9.1 of the FSAR;

c. Distances between fuel assemblies are a nominal 10.375 inch center-to-center spacing in the twenty-four flux trap rack modules.

d. Fuel assemblies with initial enrichments less than a maximum of 5 wt%

U-235 enrichment (nominally 4.95 +/- 0.05 wt% U-235) may be stored in the spent fuel racks in any one of four arrangements with specific limits as identified below:

1. Fuel assemblies may be stored in the racks in an all cell arrangement provided the burnup of each assembly is in the acceptable domain identified in Figure 4.3-3, depending upon the specified initial enrichment.

2. New and spent fuel assemblies may be stored in a checkerboard arrangement of 2 new and 2 spent assemblies, provided that each spent fuel assembly has accumulated a minimum burnup in the acceptable domain identified in Figure 4.3-4.

3. New fuel assemblies may be stored in 4-cell arrays with 1 of the 4 cells remaining empty of fuel (i.e. containing only water or water with up to 75 percent by volume of non-fuel bearing material).

4. New fuel assemblies with a minimum of 32 integral fuel burnable absorber (IFBA) rods may be stored without further restriction, provided the loading of ZrB2 in the coating of each IFBA rod is minimum of 1.25x (1.9625mg/in).

A water cell is less reactive than any cell containing fuel and therefore a water cell may be used at any location in the loading arrangements. A water cell is defined as a cell containing water or non-fissile material with no more than 75 percent of the water displaced.

See ADAMS Accession Nos. ML052930169 and ML15301A140 for Figures 4.3-3, and 4.3-4.

Proposed WBN. Units 1 and 2, TS 4.3.1 The licensee proposed the following changes to TS 4.3.1:

4.3.1.1 The spent fuel storage racks (shown in Figure 4.3-1) are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent (wt%) (nominally 4.95 .:t 0.05 wt% U-235);

b. keff s 0.95 if fully flooded with 2300 ppm borated water, which, includes an allowance for uncertainties as described in Sections 4.3.2.7 and 9.1 of the FSAR, and kett less than critical when flooded with unborated water;

c. Distances between fuel assemblies are a nominal 10.375 inch center-to-center spacing in the twenty-four flux trap rack modules.

A water cell is less reactive than any cell containing fuel and therefore a water cell may be used at any location in the loading arrangements. A water cell is defined as a cell containing water or non-fissile material.

The licensee proposed to delete Figures 4.3-3 and 4.3-4 Revisions to TS 4.3.1.1, "Criticality" Review TS 4.3.1.1, "Criticality" contains a description of the materials of construction and geometric arrangements for fuel storage, which, if altered or modified, would have a significant effect on safety which are not covered in safety limits, LCOs or SRs. The licensee proposed modifying TS 4.3.1.1.a by adding a fuel enrichment limit of nominally 4.95 +/- 0.05 wt% U-235. This limit exists in current TS 4.3.1.1.d. The licensee also proposed deleting TS 4.3.1.1.d along with Figures 4.3-3 and 4.3-4 to reflect the new SFP criticality controls that are no longer based on spent fuel assembly burnup or integral burnable absorber credit. The licensee proposed revising TS 4.3.1.1.b to add the minimum boron concentration requirement of 2,300 ppm to preserve the initial condition assumption in the boron dilution analysis, and adding the requirement that keff be less than critical when the fuel pool is flooded with unborated water.

Finally, the licensee proposed changing the last paragraph of TS 4.3.1 to reflect the conditions of the new criticality analysis. The final paragraph would be changed from:

A water cell is less reactive than any cell containing fuel and therefore a water cell may be used at any location in the loading arrangements. A water cell is defined as a cell containing water or non-fissile material with no more than 75 percent of the water displaced.

to: A water cell is less reactive than any cell containing fuel and therefore a water cell may be used at any location in the loading arrangements. A water cell is defined as a cell containing water or non-fissile material.

The NRC staff reviewed the proposed changes and determined they are acceptable, because they are necessary to ensure TS 4.3.1.1 continues to contain a description of the materials of constr_uction and geometric arrangements for fuel storage, which, if altered or modified, would have a significant effect on safety which are not covered in safety limits, LCOs or SRs.

The NRC staff finds the proposed changes to TS 4.3.1.1 acceptable based on the above review.

3.3.5 WBN, Units 1 and 2, TS 3.7.18, "Fuel Storage Pool Boron Concentration" New Proposed WBN, Units 1 and 2, TS 3.7.18 See the letter dated December 20, 2017, for the formatting of TS 3.7.18. The licensee proposed the following new TS 3. 7.18:

3. 7 .18 Fuel Storage Pool Boron Concentration LCO 3.7.18 The fuel storage pool boron concentration shall be 2:

2300 ppm APPLICABILITY: When fuel assemblies are stored in the fuel storage pool ACTIONS CONDITION A Fuel storage pool boron concentration not within limit.

REQUIRED ACTION

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

LCO 3.0.3 is not applicable.

A.1 Initiate action to restore fuel storage pool boron concentration to within limit.

COMPLETION TIME

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3. 7.18.1 Verify the fuel storage pool boron concentration is within limit.

FREQUENCY 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months New TS 3. 7 .18, "Fuel Storage Pool Boron Concentration" Review The licensee proposed adding a new TS, TS 3.7.18, "Fuel Storage Pool Boron Concentration" for WBN, Units 1 and 2. The new TS is based on TS 3.7.16 in NUREG-1431. LCO 3.7.18 would state "The fuel storage pool boron concentration shall be ~ 2300 ppm" with an applicability of "When fuel assemblies are stored in the fuel storage pool." When the LCO is not met for the Condition when the Fuel storage pool boron concentration is not within limit, the licensee would be required to immediately Initiate action to restore fuel storage pool boron concentration to within limit by Required Action A.1. Required Action A.1 would be modified by a Note which states "LCO 3.0.3 is not applicable." The new TS would contain SR 3. 7.18.1 which would state "Verify the fuel storage pool boron concentration is within limit." The frequency for the new SR would be 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The NRC staff reviewed the proposed new TS and the licensee's justification and compared the new TS to the format and content guidance in NUREG-1431. The NRC staff determined that the new LCO statement lists the lowest functional capability or performance level of equipment required for safe operation of the facility, the TS contains appropriate remedial actions to be taken in the event the LCO is not met, and the SR will ensure that the new LCO will be met.

The N RC staff finds the proposed change to add new TS* 3. 7 .18 acceptable based on the above review.

3.3.6 WBN, Units 1 and 2, TS 5.7.2.21, "Spent Fuel Storage Rack Neutron Absorber Monitoring Program" New Proposed WBN, Units 1 and 2, TS 5.7.2.21 The licensee proposed the following new TS 5.7.2.21:

This Program provides controls for monitoring the condition of the neutron absorber used in the spent fuel pool storage racks to verify the Boron-10 areal density is consistent with the assumptions in the spent fuel pool criticality analysis.

The Program shall be in accordance with NEI 16-03-A, "Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools," Revision 0, May 2017.

New TS 5.7.2.21, "Spent Fuel Storage Rack Neutron Absorber Monitoring Program" Review

The purpose of the program is to ensure the boron-10 neutron absorber areal density assumed in the SFP storage rack nuclear criticality analyses remains conservative with respect to the actual plant conditions.

The TS Section 5.7.2 21 program imposes a requirement to have a licensee-controlled program that is in accordance with NEI 16-03-A, "Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools," Revision 0, May 2017. In the safety evaluation for NEI 16-03, dated March 3, 2017, the N RC approved and accepted the document for referencing in licensing applications for nuclear power plants. The NEI 16-03-A topical report and the NRC's safety evaluation for NEI 16-03 provide the technical justification for the proposed program.

The purpose of a NAM monitoring program is to verify that the NAM installed in SFPs continues to perform its safety function (i.e., criticality control) as assumed in the AOR. The guidance provided in NEI 16-03 for a NAM monitoring program, relies on periodic inspection, testing, monitoring, and analysis of the NAM to ensure that the required subcriticality margin is maintained in accordance with 10 CFR 50.68 requirements. To accomplish this purpose, the guidance document states that a monitoring program must be capable of identifying unanticipated changes in the absorber material and determining whether anticipated changes can be verified. The guidance recommends a combination of coupon testing, in-situ measurement, and SFP water chemistry monitoring as a means to monitor potential changes in characteristics of the NAM.

The NRC staff reviewed the proposed guidance for what constitutes an acceptable monitoring program and its ability to ensure that potential degradation of SFP NAM will be detected, monitored, and mitigated. The NRC staff determined that an appropriate combination of the three methods listed above (coupon testing, in situ measurement, and SFP water chemistry monitoring) can comprise an effective NAM monitoring program. Section 3.4 of the NRC's Safety Evaluation of NEI 16-03 states that in order for a NAM program to be acceptable, a licensee must perform neutron attenuation testing to verify the 8-10 areal density. The proposed program allows TVA to incorporate an acceptable NAM monitoring program into the WBN TS.

Based on its review of the proposed TS changes and previous approval of NEI 16-03, the NRC staff has determined that TVA's proposed NAM monitoring program meets the provisions in NEI 16-03-A. Therefore, the NRC staff finds that the ability of the NAM to perform its safety function, as assumed in the AOR, is maintained, thus demonstrating compliance with the subcriticality requirements of 10 CFR 50.68.

WBN TS Section 5.7.2, as modified by addition of the new program, will continue to contain provisions relating to procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. The NRC staff determined that the TS, as modified, would continue to comply with the requirements of 10 CFR 50.36. Therefore, the NRC staff determined that the proposed change is acceptable.

The NRC staff finds the proposed change to add new TS 5.7.2.21 acceptable based on the above review.

3.3.7 Technical Specifications Changes Conclusion The regulation at 10 CFR 50.36(b) requires TS to be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto. The licensee proposed revising WBN, Units 1 and 2, TS associated with preventing criticality of fuel assemblies outside the reactor to align the TS with the new SFP criticality analysis to support the allowance of TVA use of TPBARs at WBN, Units 1 and 2.

The NRC staff reviewed the proposed changes as well as the licensee's justifications for the changes. The NRC staff determined that each new and modified LCO, and its associated Actions, will continue to meet the regulatory requirements of 10 CFR 50.36(c)(2). Likewise, the NRC staff determined that the new and modified SRs will continue to meet the regulatory requirements of 10 CFR 50.36(c)(3). The changes made to the design features TSs allow the TSs to continue to meet the regulatory requirements of 10 CFR 50.36(c)(4). Finally the addition of the NAM program allow the TSs to continue to meet the regulatory requirements of 10 CFR 50.36(c)(5). Therefore, the NRC staff determined the proposed changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendment on February 27, 2019. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Register on February 11, 2019 (84 FR 3259).

Accordingly, based upon the environmental assessment, the Commission has determined that issuance of these amendments will not have a significant effect on the quality of the human environment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: John Lamb, Steve Jones, David Garmon-Candelaria, Alexander Chereskin, John Hughey, James Medoff, Scott Krepel, Terrence Brimfield, Matthew Hamm, Jonathan Oretga-Luciano, and Mathew Panicker.

Da~: May 22, 2019

LIST OF ACRONYMS ADAMS Agencywide Document and Access Management System AOO anticipated operational occurrence AMSAC ATWS mitigation system circuitry ANSI American National Standards Institute ALARA as low as is reasonably achievable APT Accelerator Production of Tritium ASME American Society of Mechanical Engineers B-10 boron 10 BPRA burnable poison rod assemblies ccs component cooling system CEDE committed effective dose equivalent C/ECL Concentration per Effluent Concentration Limit CFR Code of Federal Regulations Cu copper CLA cold-leg accumulator CLWR commercial light-water reactors cpm counts per minute DAC derived air concentration OBA design-basis accident DOD Department of Defense DOE Department of Energy DNB departure from nucleate boiling EAB exclusion area boundary ECCS emergency core cooling system ECL Effluent Concentration Limit ERCW essential raw cooling water system FHA fuel-handling accident FR Federal Register FY fiscal year GDC general design criterion / criteria gpm gallons per minute HLSO hot leg switchover HSI human-system interface IFBA integrated fuel burnable absorbers keff k-effective LAR license amendment request

LCO limiting condition of operation LOCA loss-of-coolant accident LOOP loss of offsite power LTA lead test assemblies LPZ low population zone LWR light water reactor MDA minimum detectable activity mrem millirem MSLB main steam line break NAM neutron absorbing material NEI Nuclear Energy Institute NNSA National Nuclear Security Agency NRC Nuclear Regulatory Commission NUREG NRC technical report PEIS Programmatic Environmental Impact Statement PNNL Pacific Northwest National Laboratories ppm parts per million PWR pressurized-water reactor RAI request for additional information RCS reactor coolant system rem roentgen equivalent man RFO refuel outage RG Regulatory Guide RWST reactor water storage tank SFP spent fuel poolTR SFPCCS SFP cooling and cleanup system SR surveillance requirement SRM Staff Requirements Memorandum SRP Standard Review Plan SRS Savannah River Site SGTR steam generator tube rupture Sv Sievert TEDE total effective dose equivalent TPBAR tritium producing burnable absorber rod TCD thermal conductivity degradation TPC tritium production core TR topical report

TS technical specification TVA Tennessee Valley Authority U-235 uranium 235 UFSAR updated final safety analysis report WBN Watts Bar Nuclear Plant WGDT waste gas decay tank

µC micro-Curie

µCi/gm micro-Curie per gram

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT REGARDING REVISION TO WATTS BAR NUCLEAR PLANT, UNIT 2, TECHNICAL SPECIFICATION 4.2.1, "FUEL ASSEMBLIES," AND WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2, TECHNICAL SPECIFICATIONS RELATED TO FUEL STORAGE (EPID L-2017-LLA-0427)

DATED MAY 22, 2019 DISTRIBUTION:

PUBLIC PM File Copy RidsACRS_MailCTR Resource RidsNrrDorlLpl2-2 Resource RidsNrrPMWattsBar Resource RidsNrrLABClayton Resource RidsNrrLAIBetts RidsRgn2MailCenter Resource Steve Jones, NRR David Garmon-Candelaria, NRR Alex Chereskin, NRR John Hughey, NRR Jim Medoff, NRR Scott Krepel, NRR Terrence Brimfield, NRR Matt Hamm, NRR Mathew Panicker, NRR ADAMS Access1on No.: ML183478330 *b>y e-ma1 **b>Y memo OFFICE NRR/DORL/LSPB/PM NRR/DORL/LPL2-2/LA NRR/DSS/SRXB/BC**

NAME JLamb BClayton (!Betts) JWhitman DATE 10/4/2018 1/28/2018 10/5/2018 OFFICE NRR/DRA/ARCB/BC* NRR/DSS/STSB/BC** NRR/DMLR/MCCB/BC(A)*

NAME KHsueh VCusumano MYoder DATE 11/2/2018 11/5/2018 10/16/2018 OFFICE NRR/DMLR/MVIB/BC* NRR/DRA/APOB/BC** NRO/DCIP/BC**

NAME DAIiey CJFong KKavanagh DATE 12/3/2018 11/9/2018 11/26/2018 OFFICE N RR/DSS/SBPB/BC** NRR/DSS/SNPB/BC** OGC- NLO NAME SAnderson RLukes ERuesch*

jDATE I 12/3/2018 j 12/18/2018 I 3/29/2019 I

jOFFICE INRR/DORL/LPL2-2/BC INRR/DORL/LPL2-2/PM I I jNAME j UShoop IRSchaaf I I jDATE j s11712019 j s12212019 I I OFFICIAL RECORD COPY

v t e Letter Sequence Approval
EPID:L-2017-LLA-0427, Application to Revise Watts Bar Unit 2 Technical Specification 4.2.1, Fuel Assemblies, and Watts Bar Units 1 and 2 Technical Specifications Related to Fuel Storage (WBN-TS-17-028) (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance Supplement Administration Questions 27 August 2018 Answers 9 April 2018 4 October 2018 Results Approval Other: ML18128A293, ML18128A306, ML18332A014
MONTHYEAR2018-03-01 [Table View]

ML18347B330

Text

SUBJECT:

Dear Mr. Shea:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

SUBJECT:

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