ML25268A077

Attachment 5: Cobalt-60 Supplemental Information to Support the Use of Cobalt Burnable Absorber (Coba) Assemblies in Pressurized Water Reactors

ML25268A077
Person / Time
Site:	Salem
Issue date:	09/24/2025
From:	Westinghouse
To:	Office of Nuclear Reactor Regulation
Shared Package
ML25268A071	List: LR-N25-0074, Attachment 6: Site Specific Information in Support of Cobalt Absorber (Coba) Assemblies in Salem Generating Station, Units 1 and 2 ML25268A072 ML25268A076 ML25268A077
References
LR-N25-0074, LAR S25091
Download: ML25268A077 (1)

Text

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 1

Cobalt-60 Supplemental Information To Support the use of Cobalt Burnable Absorber (COBA) Assemblies in Pressurized Water Reactors Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, PA 16066

© 2025 Westinghouse Electric Company LLC All Rights Reserved

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 2

Table of Contents 1

Purpose and Summary 2

Description of the Cobalt-60 Production Process 3

Cobalt Burnable Absorber (COBA) Assembly Design 3.1 Design Considerations 3.2 Description of Design 3.3 COBA Fabrication Process 4

Mechanical Design Evaluation 4.1 Design Criteria 4.2 Cobalt Slug/Capsule Configuration 4.3 Capsule/COBA Mechanical Integrity 4.4 Creep Analysis of the COBA Capsule 4.5 Capsule Production Leak Test 4.6 Capsule Wear Evaluation 4.7 ISO 2919 Testing 4.8 Baseplate Connections 4.9 Assessment of COBA on Operation Risk 4.10 Impact of COBA Capsule Breach on Reactor Coolant Chemistry 5

Thermal and Hydraulic Evaluation 5.1 DNB and Core Components 5.2 Core Bypass Flow and Control Rod Function 5.3 COBA Rodlet Design Thermal-Hydraulic Analysis 6

Materials 6.1 Capsules, [

]a,c 6.2 COBA Rodlet / [

]a,c and Other Associated Parts 6.3 Baseplate Connection and Associated Parts 6.4 Nickel Plated Cobalt Slugs 7

Nuclear Design 7.1 Methodology 7.2 Benchmarking of Nuclear Methods 7.3 Reactor Core Design Approach 7.4 Implications on Reload Safety Evaluation 8

Safety Analyses 8.1 Design Parameters and Best Estimate Flows 8.2 Design Transients and Control Systems 8.3 Non-LOCA Transients 8.4 Loss of RHR at Midloop / Natural Circulation Cooldown 8.5 FLEX Operation, Shutdown Cooling, and SFP Cooling and Boron Dilution

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 3

8.6 Loss of Coolant Accident 8.7 Steam Generator Tube Rupture 8.8 Steamline Break M&E and Steam Release for Dose 8.9 Radiological Consequences (Doses) of Accidents 9

Effects on Spent Fuel Pool 9.1 Spent Fuel Pool Criticality 9.2 Spent Fuel Pool Time to Boil 9.3 Gamma Heating of the SFP Concrete Structure 9.4 Spent Fuel Pool Cooling 10 Harvesting 10.1 Harvesting Process Overview 10.2 Harvesting Workstation Tooling 10.3 Electrical and Software 10.4 Foreign Material Control 10.5 COBA Waste Disposal 10.6 Effects Within Spent Fuel Pool 10.7 Exposure to Personnel 11 Measurement of Cobalt-60 Activity 11.1 Hardware 11.2 Process Activities 11.3 Software and Data Capture 12 Byproduct Material Tracking and Reporting

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 4

Acronyms AMS Activity Measurement System ANS American Nuclear Society ASD After Shutdown ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials BAPC Boric Acid Precipitation Control BTP USNRC Branch Technical Position CANDU Canadian Pressurized Heavy-Water Reactor CFD Computational Fluid Dynamics CNSC Canadian Nuclear Safety Commission COBA Cobalt-60 Burnable Absorber COMS Cold Overpressure Mitigation System CRE Control Rod Ejection CRSB COBA Rodlet Storage Basket CTE Coefficient of Thermal Expansion DHR Decay Heat Removal DNBR Departure from Nucleate Boiling Ratio ECCS Emergency Core Cooling System ELAP Extended Loss of AC Power EOL End of Life EQ Equipment Qualification FEA Finite Element Analysis FA Fuel Assembly FHA Fuel Handling Accident FLEX Flexible Coping Strategies FMEA Failure Modes and Effects Analysis HLSO Hot Leg Switchover IAEA International Atomic Energy Agency IFBA Integral Fuel Burnable Absorber ISO International Organization for Standardization LBLOCA Large Break Loss of Coolant Accident LAR License Amendment Request LEP Lower End Plug LOCA Loss of Coolant Accident LTOPS Low Temperature Overpressure System M&E Mass and Energy MSLB Main Steamline Break MWt Megawatts thermal NCC Natural Circulation Cooldown NSSS Nuclear Steam Supply System NSTS National Source Tracking System PCT Peak Clad Temperature PPSA Peripheral Power Suppression Assembly PWR Pressurized Water Reactor

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 5

Acronyms (cont.)

QMS Quality Management Program RBMK Russian Graphite-Moderated Reactor RCCA Rod Cluster Control Assembly RCS Reactor Coolant System RHR Residual Heat Removal RSAC Reload Safety Analysis Checklist RSE Reload Safety Evaluation RV Reactor Vessel RVI Reactor Vessel Internals SBLOCA Small Break Loss of Coolant Accident SDC Shutdown Cooling SFP Spent Fuel Pool SGTR Steam Generator Tube Rupture SIL Safety Integrity Level SLB Steamline Break SLBOC Small Line Break Outside Containment SPD Self-Powered Detector SRBT Single Rod Burst Test SS Stainless Steel T-H Thermal Hydraulic TPI Thimble Plugs Installed TPR Thimble Plugs Removed TTB Time to Boil U.S. DOT U.S. Department of Transportation WABA Wet Annular Burnable Absorber ZIRLO is a trademark or registered trademark of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 6

1 Purpose and Summary Westinghouse and Nordion (Canada) are developing an innovative isotope production technology, and with our partner utilities, plan to produce cobalt-60 in Westinghouse-designed Pressurized Water Reactors (PWRs). Nordion is a major supplier of cobalt-60 sources for industrial and medical purposes. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation.

Cobalt-60 is produced by subjecting cobalt-59 material to neutron irradiation in a nuclear reactor core.

Nickel-plated slugs of cobalt-59 target material would be provided to Westinghouse by Nordion. The target material is placed in [

]a,c and loaded into fuel component inserts, called Cobalt Burnable Absorber (COBA) assemblies, at the Westinghouse Columbia Fuel Fabrication Facility (CFFF).

These COBAs are then inserted into Westinghouse fuel assemblies, shipped to the plant site and loaded into the reactor core as part of the fuel assembly, much like other fuel assembly components such as wet annular burnable absorbers (WABA). When subjected to the neutron flux in the core, cobalt-59 is converted to cobalt-60. In order to reach the desired activity level, the COBA will be shuffled into fresh fuel assemblies and further irradiated for multiple cycles.

After removal from the core, the now-activated capsules are harvested (removed from the COBA rodlets), loaded into a Nordion F-231 transport package, and transferred to Nordions Canada facility.

Nordions F-231 Type B(U) transport package has been designed to meet the requirements of the IAEA transport regulations. The F-231 is certified by the Canadian Nuclear Safety Commission (CNSC) and endorsed by the U.S. Department of Transportation (U.S. DOT). At Nordion, the cobalt-60 slugs are removed from the Westinghouse irradiated capsules, and [

]a,c so that Nordion can create finished sources of specific activity levels to meet various end-use specifications. Nordion then uses the cobalt-60 slugs to manufacture sealed sources for distribution to their customers for sterilization and medical end-use applications.

Activities requiring a byproduct material license are covered under the requirements specified in 10 CFR Part 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material," Section 30.3, "Activities requiring license." This section states that, except as noted in 30.3(a),"...no person shall manufacture, produce, transfer, receive, acquire, own, possess, or use byproduct material except as authorized in a specific or general license issued in accordance with the regulations in this chapter." Further, 10 CFR Part 31, General Domestic Licenses For Byproduct Material, Section 31.1, Purpose and scope, states: This part establishes general licenses for the possession and use of byproduct material and a general license for ownership of byproduct material. Therefore, since either a specific or a general license is required, and since the provisions of the general license do not address production or transfer of byproduct material, it is concluded that a specific byproduct license is required. The requirements for a specific byproduct material license have been reviewed as defined in 10 CFR Parts 32, 33, 34, 35, 36, and 39 to determine if any of these requirements are applicable to the generation of Co-60. This review has determined that there are no requirements that are applicable to this situation. Therefore, the requirement for Part 30 byproduct material license

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 7

will be addressed by amendment to the existing Part 50 Facility Operating License.

The Cobalt-60 byproduct material that is generated in the reactor core is radioactive byproduct material sealed in a capsule. However, based on the proposed source manufacturing process, and because these sources are not in a final form for end use, these sources do not fall within the scope of the Nationally Source Tracking System (NSTS) requirements in 10 CFD 20.2207.

Consequently, they would also not be subject to the serialization requirements in 10 CFR 32.201.

This report presents a description of the COBA assemblies, as well as descriptions of the analyses and evaluations of the mechanical design, thermal hydraulic design, materials, nuclear design and safety analyses, harvesting, measurement, marking and tracking, and transportation.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 8

2 Description of the Cobalt-60 Production Process For each core reload, [

]a,c COBA assemblies are loaded into the reactor core. Each COBA is loaded into a fresh fuel assembly to maximize its irradiation. The capsules containing the cobalt-59 are fabricated from [

]a,c that are filled with nickel-plated cobalt-59 slugs. The capsules are then loaded into [

]a,c which are then attached to a baseplate to form a single COBA assembly (see Figures 2-1 and 2-2). The existing WABA platform is used as a design basis for the COBA assembly design.

Prototype capsules undergo ISO 2919 testing, which is a requirement of the CNSC and U.S. DOT Certification for the Nordion transport package to demonstrate that the structural integrity requirements of the capsules are met.

Each COBA is inserted into its host fuel assembly and loaded into the reactor core with its fuel assembly. The COBA assembly will be repositioned into other fresh fuel assemblies after each cycle to optimize the irradiation to the capsules. After the requisite number of activation cycles in the core, the COBAs are removed from the fuel assembly and moved to the harvesting area. During harvesting, the cumulative activity is tracked to ensure that the total activity in each shipping cask does not exceed the activity limits defined by Nordion in the cask certification.

The capsules now containing cobalt-60 are transported in the licensed transport package to Nordions Canada facility. The activated slugs are removed from the capsules and [

]a,c for the manufacture of finished sealed sources for distribution to their end users.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 9

Figure 2-1 COBA Assembly Figure 2-2 COBA Assembly a,c a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 10 3

Cobalt Burnable Absorber (COBA) Assembly Design 3.1 Design Considerations The COBA insert is designed to hold cobalt-59 slugs during irradiation in PWR conditions. To accomplish this, the cobalt targets are inserted in the thimble tubes of a fuel assembly similar to other core components. The COBA rodlets are mounted on a baseplate that can be handled from above using existing tooling, and that can easily fit into the thimble tubes and be withdrawn at the end of the irradiation cycle. All of the components of the COBA insert are designated as safety related, including the capsule and the nickel-plated cobalt slugs. All aspects of the COBA development program and all COBA products to be provided are controlled under the Westinghouse Quality Management System (QMS). The QMS is approved by the Nuclear Regulatory Commission (NRC).

In order to isolate the cobalt slugs from the reactor coolant, a robust barrier encapsulates the slugs.

The encapsulation is designed to withstand the reactor environment such that the slugs can be removed from the capsule and used post-irradiation. In order to adequately irradiate the slugs and create favorable neutronic conditions in the core, [

]a,c The configuration of the capsules [

]a,c around the capsules. After irradiation, the encapsulated slugs are sent to Nordion for further processing using an existing shipping container design.

3.2 Description of Design Because PWR conditions consist of much higher temperature and pressure than other reactors where cobalt is currently irradiated (CANDU and RBMK reactors), the capsule designed for PWR application is made of [

]a,c A stack of [

]a,c is contained within a [

]a,c outer tube or rodlet. The rodlet holds the capsule stack in place in the fuel assembly. [

]a,c rodlets are attached to a baseplate, with thimble plugs in the remaining thimble locations, to form the COBA insert, [

]a,c material of the rodlet allows neutrons to pass through to the cobalt to activate the cobalt-59, converting it to cobalt-60.

The rodlet length is determined such that existing handling tools at reactor sites that currently interface with WABA assemblies can be used to handle the COBA inserts. [

]a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 11 3.3 COBA Fabrication Process The cobalt slugs are manufactured by Nordions slug supplier, where they are fabricated and then nickel-plated. The slugs are sent to the capsule supplier, who manufactures the [

]a,c capsules and places the slugs inside. The capsule is backfilled with a mixture of [

]a,c and the end caps are welded onto the capsule. The capsule end plug welds have non-destructive examination and visual inspections, ensuring the quality of the welds and no lack of penetration. The capsules are then shipped to CFFF.

At CFFF, the bottom end plug is welded to the bottom of the rodlet, then the lower standoff tube is inserted into the bottom of the rodlet. [

]a,c capsules are then loaded into the rodlet and rest on the lower standoff tube. A [

]a,c is used to maintain the capsule stack in the proper axial location relative to the active core region and to minimize movement of the capsules during shipment and refueling operations. The top end plug is welded to the top of the rodlet. A [

]a,c top connector. [

]a,c The capsules are located axially in the rodlet by the [

]a,c on the bottom end plug and top end plug such that the capsule stack is axially centered with respect to the active fuel. The interior surface of the bottom end plug has [

]a,c is seated correctly on the bottom end plug and centered in the rodlet.

[

]a,c The COBAs are shipped horizontally with the fuel assemblies in the Westinghouse Traveller transport package. The Traveller transport package is instrumented with trip accelerometers to assure that excessive loads are not experienced by the fuel assembly during shipment. Also, the COBA rodlets are shipped with the rodlets inside the fuel assembly thimbles, thus precluding any damage to the rodlets that may impede capsule movement.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 12 4

Mechanical Design Evaluation 4.1 Design Criteria The following design criteria were identified as affected for the COBA design, based on similar core components:

Conditions I and II Loads -- Meets stress criteria during operating loads.

Conditions III and IV Loads -- Meets stress criteria during seismic and LOCA blowdown.

Shipping and handling loads - Meets all shipping and handling requirements.

Rod retention -- Rodlets and capsules (absorber material) retain proper position to maintain designed reactivity Interface requirements - COBA inserts are compatible with fuel assemblies, shipping containers, handling equipment, and harvesting equipment.

COBA Rodlets lateral positioning -- Accurate lateral positioning is required for fit up with the fuel assembly.

Material requirements -- Materials are satisfactory in strength, corrosion resistance, and stability throughout the COBA design life.

Dimensional clearances - Components maintain clearance and ensure no mechanical interference at any point during design life.

4.2 Cobalt Slug/Capsule Configuration Each cobalt slug is [

]a,c The capsule is [

]a,c 4.3 Capsule/COBA Mechanical Integrity Core components are evaluated for Conditions I & II loads and shipping and handling loads. The structural members of the COBA components are qualified in accordance with the criteria of NG-3000 of the ASME Code,Section III.

Seismic events, as well as LOCA blowdown (seismic generally bounds LOCA lateral loads), tend to apply significant lateral loads to the fuel assembly. However, lateral loads on the core component rods due to seismic events are small since core component rods and thimbles plugs are supported along their entire length by the surrounding guide thimbles. Therefore, the core component rodlets are protected against damage from these loads. Thus, lateral loads are experienced by the core component assemblies, but the resulting stresses are considered to be small and further analysis for them is not warranted.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 13 Maximum system temperature and pressure are considered in the analysis of the [

]a,c capsules. The COBA assembly is subjected to shipping and handling loads to assure that the appropriate limits are satisfied. The rodlet to baseplate connection is qualified by performing pull tests and bend tests to assure that the connections will withstand all shipping and handling, and operational forces, including insertion into and withdrawal from the fuel assembly. The survivability of the

[

]a,c capsules has been evaluated and is presented in Section 8.6.1.

4.4 Creep Analysis of the COBA Capsule During the nuclear irradiation process, the Cobalt Burnable Absorber (COBA) inserts will be exposed to [

]a,c The creep-down effects on the COBA capsules have been analyzed to determine [

]a,c demonstrates sufficient dimensional clearance for removal of the irradiated slugs from the capsules under all expected operating conditions.

4.5 Capsule Production Leak Test As described in Section 3, after the cobalt-59 slugs are placed in the capsule, the capsule is backfilled with [

]a,c All production capsules are required to pass a leak test using a [

]a,c The leak check ensures that each capsule is leak-tight and will not allow any coolant to enter the capsule.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 14 4.6 Capsule Wear Evaluation The design of the COBA assembly is such that the individual capsules (and the slugs within them) have a limited ability to move within the outer tube. Although the flow through the fuel assembly thimble is somewhat limited, flow-induced vibration can occur. This combined with known reactor internals vibration creates the potential for the capsules and slugs to wear against their surrounding components. Conservative analyses of the capsules and slugs suggested that wear is not a concern, however the analysis was challenged due to the lack of corroborating benchmark testing.

To support the conclusions drawn from analysis, a prototypic wear test featuring a production COBA rodlet in a typical Westinghouse thimble tube was performed. Assuming a COBA operating life of

[

]a,c 4.7 ISO 2919 Testing The Nordion material transport package (transportation cask) to be used to transfer the irradiated capsules to the Nordion facility is certified by the Canadian Nuclear Safety Commission and endorsed by the U.S. DOT. The certification specifies that the capsules must meet the requirements of ISO 2919, Radiological Protection Sealed Radioactive Sources General Requirements and Classification.

ISO 2919 requires several types of tests to confirm capsule integrity during shipment. The first three, bending, puncture, and impact, are performed on the empty tubing that makes up the capsule, which is considered a conservative condition as it is not being supported by end caps which are welded to the tubing or by the cobalt slugs which will be contained within the tubing in the production capsules.

Successful prototype testing has been performed on all thicknesses of tube wall, including two that were thinner than the final capsule design. No defects were observed. Testing of the production capsule design will also be performed.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 15 The other three ISO 2919 tests, vibration, pressure, and temperature, are performed on completed capsules, as the slugs and end caps may influence the results. In all cases, the capsules are visually inspected after the tests are completed, in accordance with ISO 2919.

4.8 Baseplate Connections The baseplate forms the structural base for the COBA insert. The [

]a,c rodlets containing the capsules with cobalt targets are attached to the baseplate. The lateral and vertical position of the rodlets is maintained by being mounted on the holddown assembly baseplate. The holddown assembly baseplate is held down against the fuel assembly top nozzle adapter plate by the coil spring which is an integral part of the holddown assembly.

[

]a,c For the thimble tube locations that do not contain a COBA rodlet, a modified thimble plug is attached to the baseplate, [

]a,c Pull and Bending tests were performed on the baseplate connection to confirm that the connection will withstand all expected handling loads.

4.9 Assessment of COBA Insertion on Operation Risk The release of cobalt-60 into the RCS is controlled through encasing the nickel coated-cobalt slugs within a [

]a,c capsule. This capsule provides a high reliability protective barrier preventing cobalt ingression into the reactor coolant. [

]a,c is compatible with operation in a reactor environment and has been shown to be resistant to corrosion and wear. To understand the robustness and integrity of the capsule, a Failure Modes and Effects Analysis (FMEA) was conducted on the COBA assembly. A review of the FMEA identified the following potential mechanisms for capsule breach:

1. [

2.

3.

]a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 16

4. [

5.

6.

]a,c

[

]a,c The relevant set of failure modes important to COBA operation are briefly discussed and dispositioned. Mechanism specific capsule failure probabilities were established based on a review of past experience with similar rodlet installations, and a combination of expert judgement and analysis.

Overall disposition of failure modes is summarized in Table 4.9-1. All failure mechanisms were judged to be of low likelihood and would not be expected to result in capsule failure given the capsule operating environment and residence time.

Based on results of the assessment of the COBA capsule failure modes, the overall probability of capsule failure during the [

]a,c Thus, the likelihood of cobalt release from the addition of COBAs into the core is estimated at [

]a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 17 Table 4.9-1 Summary Assessment of COBA Capsule Failure Modes Resulting in Capsule Breach Capsule Breach Failure Mechanism Quantification of Basic Events

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 18 4.10 Impact of COBA Capsule Breach on Reactor Coolant Chemistry The COBA assembles, and the capsules in particular, are designed to provide a leak-tight barrier to prevent the release of cobalt into the reactor coolant. The preceding sections describe the various testing and analyses that are performed to confirm the integrity of the capsules, including stress and creep analyses, and production leak testing, wear testing, and ISO 2919 testing. In addition, an assessment has demonstrated the low risk of capsule failure.

Co-60 is a common radioisotope generated in the reactor coolant system (RCS) from elemental cobalt in primary system materials, which circulates throughout the system and is a dominant contributor to personnel dose at PWRs. Plant personnel routinely measure the concentrations of gamma isotopes in the coolant, including Co-60, as part of diagnostic chemistry monitoring. Although the COBA assemblies are designed to prevent the release of Co-60 to the reactor coolant, it is prudent to consider guidance on monitoring Co-60 in the RCS and actions to take if levels increase, in the unlikely event of a capsule breach.

Experimental studies have been conducted by Westinghouse to estimate the rate of cobalt release into the RCS in the unlikely event of a capsule breach. [

]a,c Given the low likelihood of the failure of an individual capsule coupled with the lack of a significant release from the cobalt slug even when the benefit of the nickel coating is not credited, the release of a significant amount of cobalt into the reactor coolant is determined to be very unlikely. [

]a,c

• • • Additional details are provided in the Site-Specific attachment to this LAR.

a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 19 5

Thermal and Hydraulic Evaluation 5.1 DNB and Core Components COBA Rod/Thimble Cooling Standard design methods are typically employed to evaluate cooling of fuel assembly components such as control rods and WABAs. To model the flow and heat transfer characteristics of the COBA rod/thimble two flow path configuration, both the standard methods and CFD analysis techniques were used as discussed in Section 5.3. The predicted flow and temperature distributions with heat transfer considerations [

]a,c to be below the water saturation temperature. As such, bulk boiling was shown to be precluded in all flow paths of COBA rod/thimble configuration. Additionally, CFD based temperature and flow distributions were employed to calculate cobalt, [

]a,c COBA rod component metal temperatures, which were below the respective metal melting temperatures to confirm the no centerline melt criterion for COBA rods.

DNB Assessment Assessments have been made to determine the impact of COBA inserts on the minimum Departure from Nucleate Boiling Ratio (DNBR) of a hot assembly containing COBA inserts. The overall impact would be similar to the impact of including wet annular burnable absorber (WABA) as a secondary burnable absorber in addition to an integral fuel burnable absorber (IFBA).

The flow characteristics of a COBA rod placed inside a guide tube do not directly impact the T/H parameters used in the DNB safety calculations. Any impact to DNBR would be from variations in power distributions, which have been shown to be overall minor and within the variations typically seen on a cycle-to-cycle basis. This alone would indicate that DNBR is not impacted, Further, the COBA rodlets release gamma energy of their own by neutron activation. The amount of energy is so small compared to the nuclear power from the fuel rods that it is negligible for DNB calculations. Thus, the insertion of COBA rodlets into guide thimble tubes has minimal impact on DNBR, similar to variations seen between reload cycles. The effects are adequately addressed using standard Westinghouse reload methodology.

COBA Capsule Lift Force Although vertical movement of capsules within the rodlet is limited, the COBA capsules could lift off within the COBA rod outer tube if the weight of the capsule stack is lower than the lift force acting on the capsules. The lift force due to pressure drop of the capsules and cross-sectional area inside the outer tube is calculated. [

]a,c The resulting maximum lift force acting on the capsules is found to be lower than the weight of the capsules. Therefore, the COBA capsules will not experience a lift off during operation and will maintain the capsule stack position inside the outer tube.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 20 5.2 Core Bypass Flow and Control Rod Function Evaluation of the reactor internals involves the calculation of various hydraulic flow information, primarily the reactor vessel (RV) pressure losses and RV Internals (RVI) associated core bypass flows. In addition, the elapsed time for complete, gravity induced rod cluster control assembly (RCCA) insertion into the core is also determined.

In support of the Non-LOCA Transient Analyses (Section 8.3), several assumptions have been identified that required further assessment regarding the potential impact of the addition of COBAs on the Non-LOCA Analyses. Three of those assumptions (Items 1, 2, and 6, listed below) are addressed by the thermal hydraulic evaluation of the reactor core.

1. No adverse effect on the rod drop time

2. No adverse effect on the reactor trip rod position versus time curve.

6. No adverse effect on the core bypass flow fraction and core pressure drop.

Assumptions 1 and 2 are related to the DROP computer code and assumption 6 is related to the THRIVE computer code. The DROP computer code calculates the elapsed time for complete, gravity induced, RCCA drop (insertion) into the core. The Thermal Hydraulic Reactor Internals Vessel Evaluation (THRIVE) computer code models the reactor vessel internals system and computes various hydraulic flow information, primarily the reactor vessel (RV) pressure losses and RV Internals (RVI) associated core bypass flows.

Assessments have been performed to evaluate the corresponding impact of the COBA assemblies on THRIVE core bypass flows. It was determined in these analyses that the implementation of COBA assembly inserts will not increase the total design core bypass flow calculated in THRIVE. Therefore, assumption 6 remains valid, and the implementation of COBAs has no adverse effect on the core bypass flow fraction and core pressure drop.

Regarding assumptions 1 and 2, further investigation was performed. Bypass flows, fuel assembly flow areas, and specific thimble tube dimensions are inputs to the DROP computer code. Since these parameters are not impacted by the implementation of COBA assemblies, no impact to the DROP code results is expected. Therefore, assumptions 1 and 2 remain valid, and the change has no adverse effect on either the rod drop time or the reactor trip rod position versus time curve.

5.3 COBA Rodlet Design Thermal-Hydraulic Analysis 5.3.1 Design Criteria The following thermal-hydraulic (TH) design criteria were evaluated for the COBA rodlet design:

I.

The maximum temperature of the core component rods shall not exceed the melting temperature of the material used for Condition I and II operation.

II.

There will be no surface boiling from the core component rod within the dashpot region of the fuel assembly guide thimble for Condition I operation.

III.

There will be no bulk boiling in the fuel assembly guide thimble for Condition I operation.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 21 IV.

The sum of the bypass flows through all the different types of thimble-core component rod combinations and the instrumentation tubes in the core shall not exceed the core design limit for Condition I and II operation.

V.

The capsules shall remain seated in the rodlet during normal operation to prevent damage and blocking of flow paths inside COBA rodlet.

In addition to Thermal-Hydraulic core component design criteria mentioned above, inside the COBA rodlet design, bulk boiling shall not occur in the gap between the outer tube and capsule (Region 1 as shown in Figure 5-1) for Condition I operation. Condition I operation was modeled, which means normal steady-state operating conditions of pressure, temperature, and flow.

Figure 5-1 Cross-Section View of COBA Rodlet/Guide Thimble Design Standard design methodology for Westinghouse core components was followed in demonstrating that Criteria I through V are met for the COBA rodlet design.

For the additional COBA criterion of no bulk boiling [

]a,c available thermal-hydraulic codes typically used for core component cooling analyses are not suitable for [

]a,c To take advantage of the flexibility of handling complex geometries and modeling multiple physics, such as conjugate heat transfer, Computational Fluid Dynamics (CFD) was used to model the flow and heat transfer characteristics of

[

]a,c The full three-dimensional geometry of a COBA rodlet inside a thimble tube was a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 22 modeled. The CFD model results showed that no boiling occurs in all flow paths of the COBA rodlet/thimble tube configuration with considerable margin. Therefore, it is concluded that the boiling criterion is met for the COBA rodlet. Benchmarking of the CFD analysis is discussed in Section 5.3.2.

5.3.2 Thermal Hydraulic Analysis Benchmarking As discussed in Section 5.3.1, CFD was used to assess the criterion of no boiling inside of the COBA rodlet at normal steady-state operating conditions. Different criteria are established for transient conditions and are not evaluated using CFD. For example, for Condition I and II transient events, the criterion is no melting of the materials, and a large margin is demonstrated.

Westinghouse is confident in the CFD results based on the use of Westinghouse processes: CFD methodology (which includes industry best practices), verification of the analysis process, and the design review process. Nevertheless, a benchmark test has been performed to demonstrate that the CFD methodology (meshing, physical modeling and numerical algorithm) for the COBA rodlet thermal hydraulic analyses is appropriate.

[

]a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 23 6

Materials The following materials are used in the fabrication of the COBA assemblies. The use of materials is based on Westinghouse extensive experience and the performance history in PWR environments.

[

]a,c Two design criteria are evaluated: The first criterion is to keep cobalt out of the coolant and the coolant outside of the capsule. This is addressed in Section 6.1. The second criterion is to maintain the integrity of the nickel coating (addressed in Section 6.4).

6.1 Capsules, Capsule [

]a,c

[

]a,c The risk of a capsule breach is addressed in Section 4.9. Section 4.10 addresses the impact of cobalt on the surface of irradiated slugs in the unlikely event of a capsule failure.

[

]a,c Westinghouse material specifications provide the chemical composition for the standard [

]a,c Westinghouse material properties manual provides the mechanical and physical properties at room temperature and high temperature for unirradiated and irradiated materials. These properties were used in creep analysis and provide assurance that the [

]a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 24 6.2 COBA Rodlet / [

]a,c and Other Associated Parts - [

]a,c

[

]a,c have a long history of proven performance in PWR environments. Westinghouse material specifications and provide material properties and heat treatment requirements in details. [

]a,c 6.3 Baseplate Connection and Associated Parts - [

]a,c

[

]a,c Increase in yield and ultimate strengths is expected combined with a lower elongation and reduction of area. Similar to [

]a,c under PWR operating temperature and neutronic flux conditions, swelling and embrittlement will not be significant and represent no concerns. Since the chemical composition of the alloy provides the wear and galling resistance, any changes in microstructure due to irradiation should not have deteriorating effects on wear and galling resistance of the alloy.

6.4 Nickel Plated Cobalt Slugs

[

]a,c With cobalt lattice structure being hexagonally close-packed (hcp) at temperature

< 417°C, the expansion along the a and c crystallographic directions are different from each other.

From 438C - 455C, there is a phase transformation where the crystal structure changes to face centered cubic (fcc). [

]a,c The phase transformation results in a volume increase of about 0.3%, which is insignificant. Evaluation of the thermal expansion between the slugs and capsule has shown that there is sufficient clearance between the slug stack and the capsule for hot conditions.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 25 The radioactive cobalt composition is likely to be a mixture of cobalt-58, cobalt-60, and nickel-60.

Cobalt-60 formed in production will start to decay immediately and form nickel-60. This slight and slow increase of nickel content in the cobalt as a decay product is not a concern. Irradiation has an effect on the physical and mechanical properties of iron and nickel and is expected to have a similar effect on cobalt. Cobalt, Iron, and Nickel are named the Iron Triad, and they share similar chemical and physical properties. [

]a,c Similarly, irradiation growth for pure cobalt is likewise not expected.

Irradiation effects on mechanical properties, including hardening, will not be a concern as the capsules [

]a,c The decay products of Co-60 are Ni-60 and then Fe-56 and He. The Co/Ni/Fe ratio has no impact on the safety basis of the capsules, and He content has been conservatively accounted for using conservative assumptions for gas retention and capsule/slug dimensions.

The second design criterion is addressed via laboratory testing to evaluate corrosion of unirradiated cobalt and unirradiated nickel plated cobalt in autoclave water and steam tests, and to model the cobalt diffusion into the nickel coating under normal operating conditions.

[

]a,c This oxidation impacts the integrity of the coating but it only occurs following a breach of the capsule. Maintaining capsule integrity eliminates that degradation mechanism.

[

]a,c Radiation further enhances the diffusion.

Experimental studies of thermal diffusion and analytical calculations of radiation enhanced diffusion conducted by Westinghouse provide a unified understanding of the anticipated diffusion behavior in reactor over 3 cycles. [

]a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 26 7

Nuclear Design 7.1 Methodology Nuclear design calculations for COBA inserts in fuel assemblies are based on unit assembly results and homogenized group cross-sections generated by the PARAGON lattice code (Reference 7-1).

PARAGON is a 2D multi-group transport theory code which utilizes a 70 energy group structure derived primarily from ENDF/B-VI data. PARAGON performs a 2D 70-group flux calculation which couples the individual sub-cell regions (e.g., pellet, clad and moderator) as well as the surrounding rods using the collision probability technique and an interface current method. It provides the capability for cell lattice modeling on a fuel or unit assembly (UA) level and can generate two energy group homogenized cross-sections for use in the 3D core simulator code ANC.

Specific to COBA, PARAGON has the capability to explicitly model all the detailed regions of the cobalt rod within a fuel assembly guide thimble. The absorber slug is modeled with [

]a,c within the cobalt region followed by [

]a,c The remainder of the COBA cell geometry consists of [

]a,c and the fuel assembly guide tube as well as the gap regions between each tube as illustrated in Figure 7-1.

PARAGON can also model the [

]a,c The group constants generated by PARAGON are coupled to the spatial few-group code using the NEXUS nuclear data methodology (Reference 7-2).

COBA inserts can also be modeled using the PARAGON2 lattice code (Reference 7-3) for the purpose of generating homogenized group cross-section for in-core modeling. PARAGON2 uses the same physics theory as PARAGON for all modules (flux solver, depletion, etc.) except the resonance calculation, where PARAGON2 is based on ultra-fine energy mesh. The other new feature in PARAGON2 is the use of the resonance scattering model for the elastic scattering cross-section matrices. All engineering modeling capabilities of PARAGON code are also supported by PARAGON2.

ANC (References 7-4 and 7-5) is an advanced nodal code capable of two energy group 3D core model simulations. Data required for solving the coarse and fine mesh flux and power distributions are obtained by various PARAGON or PARAGON2 lattice models used to represent all the different fuel segments and operating conditions in the core. User input provides a detailed description of the axial and radial features of the fuel and core inserts used in the core model including COBA. In a 3D core model, the fuel assemblies are divided into a 2x2 mesh configuration radially. The axial mesh structure is defined [

]a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 27 The following summarizes the 59Co depletion chain:

60Co 60Ni + +

60Co + 1n 61Ni Thermal neutron capture in 59Co results in the production of 60Co, 54.5% of which is in the isomeric or metastable state 60mCo. The isomer 60mCo decays with a 10.467 minute half-life, with 0.2486% by beta decay to 60Ni and the remainder by isomeric transition (IT) decay to the ground state 60Co (Reference 7-6). As a result, the cobalt activation cross section resulting in 60Co is 0.136% (about 0.05 barns) less than the total neutron absorption cross section of 59Co. Also, the cross sections for 60mCo isotope are not available in the basic nuclear data libraries used and the absorption by this isotope is accounted for by assuming that all 59Co neutron capture yields to the 60Co ground state.

[

]a,c In addition to the 60Co losses due to neutron absorption and decay during irradiation in the core, ANC will also track the decay of 60Co during any refueling outage or other reactor shutdown. ANC will explicitly track the 60Co production and loss [

]a,c 7.2 Benchmarking of Nuclear Methods The PARAGON 70-group neutron cross-section library was updated to include cobalt and nickel data from ENDF/B-VII.1 and JEFF 3.2, using the same methodology described in Reference 7-1. The primary neutron absorbing isotope in a PWR cobalt target slug during its irradiation lifetime is cobalt-59. During a typical irradiation period of [

]a,c of the cobalt-59 is converted into cobalt-60 and nickel isotopes. Historical measurements of the cobalt-59 2200 m/s neutron capture cross-section can be obtained from Reference 7-6. The 24 measurements discussed in Reference 7-6 are summarized in Table 7-1 and Figure 7-2. The weighted average of all 24 measurements is 37.204+/-0.075 barns. Reference 7-6 states that the last 5 measurements in Table 7-1 are the most precise and documented measurements and uses these data points to propose the recommended cobalt-59 2200 m/s neutron capture cross section of 37.178+/-0.056 barns. The ENDF/B-VII.1 and JEFF 3.2 2200 m/s (0.0253 eV) cobalt-59 neutron capture cross section is 37.184 barns, nearly identical to the value recommended in Reference 7-2.

45.5%

0.25%

60Ni 60mCo 60Co 59Co + 1n

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 28 The ENDF/B-VII.1 and JEFF 3.2 cobalt-59 neutron capture resonance integral is 75.87 barns, similar to the Reference 6 value of 74.2+/-2 barns derived from the measured data summarized in Table 7-2.

The ability of PARAGON and PARAGON2 to model and accurately deplete the COBA fuel assembly insert was benchmarked against the industry standard depletable Monte Carlo code SERPENT2 (Reference 7-7). Comparisons were made of intra-assembly rod power distribution, effects of COBA on assembly reactivity and cobalt-60 production for fresh and irradiated COBA in Westinghouse 17x17 fuel assemblies with 0.374 inch and 0.360 inch fuel clad outer diameter.

Results of the reactivity comparisons of PARAGON and PARAGON2 to SERPENT2 are given in Tables 7-3 and 7-4 and Figures 7-3 through 7-6. Results are presented as the absolute reactivity differences between PARAGON or PARAGON2 and SERPENT2, as well as the difference in the COBA inserted reactivity for inserts containing [

] a,c The results in Tables 7-3 and 7-4 and Figures 7-3 and 7-4 provide confirmation that PARAGON and PARAGON2 have implemented the cobalt cross sections and depletion chain correctly and can predict assembly reactivity with accuracy comparable to a fuel assembly without a COBA insert. It is expected that the nodal code ANC using PARAGON or PARAGON2 based homogenized few group constants will predict the reactivity of a fuel assembly hosting a COBA with the same level of accuracy.

Comparisons of the intra-assembly pin power distributions from PARAGON, PARAGON2 and SERPENT2 for fuel assemblies with [

] a,c fresh and irradiated COBA rods are given in Figures 7-7 through 7-14. These results provide confirmation that PARAGON and PARAGON2 can accurately predict the intra-assembly flux distribution for a fuel assembly hosting a COBA insert. It is expected that the nodal code ANC using PARAGON or PARAGON2 based homogenized few group constants will also predict the pinwise power distribution of a fuel assembly hosting a COBA with the same level of accuracy as a fuel assembly without COBA. Therefore, the existing calculational and measurement power distribution uncertainties remain applicable to core designs with COBA.

Comparisons of the cobalt-60 production from PARAGON, PARAGON2 and SERPENT2 for fuel assemblies with [

]a,c COBA rods are given in Figures 7-15 and 7-16. These results provide confirmation that PARAGON and PARAGON2 can accurately model the cobalt depletion chain discussed in Section 7.1. It is expected that the nodal code ANC using PARAGON or PARAGON2 based homogenized few group constants will also model the cobalt depletion chain with the same level of accuracy.

7.3 Reactor Core Design Approach As stated previously, irradiated cobalt slugs will be created to manufacture finished sources for ultimate use in various gamma emitting devices. To optimize the strength of the finished sources and the loading of these devices, it is desirable to maximize the cobalt-60 produced within the COBA slugs over the cycles they are irradiated in core.

Creating the radioisotope cobalt-60 in a PWR core neutron flux level and energy spectrum is a relatively slow reaction process. During a typical 18-month fuel cycle, about [

]a,c of the cobalt-59 is converted into the gamma emitting isotope cobalt-60. The amount of cobalt-60 produced

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 29 each fuel cycle will depend on the specific cycle length and depletion strategy, core operating conditions and fuel lattice design features that can impact the neutron energy spectrum. [

]a,c PWRs operating under an 18-month refueling cycle strategy will typically irradiate each COBA insert

[

]a,c For example, quarter-core loading patterns are illustrated in Figure 7-17 for a Westinghouse 4-loop core without COBA and for the same core with a [

]a,c loading of COBA inserts. The top half of Figure 7-17 indicates the placement of the [

]a,c fuel assemblies as well as the locations of the RCCAs. Since it is desirable to maximize the cobalt-60 produced within the COBA slugs during each cycle of irradiation, [

]a,c The bottom half of Figure 7-17 illustrates a loading pattern for the same core now with a [

]a,c loading of COBA inserts. In this particular example, a [

]a,c. Also, in this example it was determined that a COBA harvest of [

]a,c would yield a sufficient volume of cobalt slugs at a desired specific activity. To allow for a steady production and harvest of cobalt-60 every cycle, the total number of COBA inserts in the core is [

]a,c as illustrated in Figure 7-17.

Notice in Figure 7-17 that [

]a,c The numbers of other boron or gadolinia based burnable absorbers are [

]a,c.

The fuel and COBA loading shown in Figure 7-17 is strictly an example irradiation strategy and is not meant to be indicative of all future use of COBA inserts. The [

]a,c will be determined on an ongoing basis considering the market requirements for cobalt-60 production.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 30 7.4 Implications on Reload Safety Evaluation As discussed in the preceding sections, the implementation of COBA or the cobalt depletion chain does not impact the methodology utilized for PARAGON, PARAGON2 or ANC (References 7-1, 7-3, and 7-4, respectively). The cross sections added to track the cobalt depletion chain have no impact on the ability of the codes to predict the reactivity and power distribution results of the core. COBA implementation does not result in any significant change to the reload core design other than [

]a,c. Therefore, the standard reload safety evaluation (RSE) methodology documented in Reference 7-8 and standard calculational and measurement uncertainties will continue to be applied to core designs implementing COBA. To summarize:

The modeling of COBA does not require any change to the licensed codes documented in References 7-1, 7-3 and 7-4.

The implementation of COBA does not result in a significant perturbation of the core fuel management strategy. Any changes to the loading pattern to include COBA are within typical cycle by cycle variation. The only change to a typical loading pattern will be [

]a,c Core loading patterns that include COBA can be developed that are very similar to those without COBA. Therefore, impacts on control rod worth, shutdown margin and trip reactivity are expected to be small and within the typical cycle to cycle variations. In addition, depletion performance of the COBA loading patterns is expected to be similar to loading patterns without COBA with respect to boron concentrations, peaking factors and axial offset behavior, again within typical cycle to cycle variations.

COBA is inserted into the core much like a discrete burnable absorber or other core component. COBA does not require any modifications to the fuel lattice design or uranium loading other than [

]a,c such that the impacts on [

]a,c.

The steady state axial power shape and the power shape distribution assumed in loss of coolant accident (LOCA) analyses are largely unimpacted by the inclusion of COBA. Overall, no reload safety analysis checklist (RSAC) items (i.e. key safety parameters applied as part of the RSE) are expected to be significantly impacted by the inclusion of COBA in a core design.

The safety evaluation of a core containing COBA will be performed following the existing analysis methodology documented in Reference 7-8 and apply existing calculational and measurement uncertainties.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 31 References 7-1 Ouisloumen, M., et al., Qualification of the Two-Dimensional Transport Code PARAGON, WCAP-16045-P-A (Proprietary), August, 2004.

7-2 Zhang, B., et al., Qualification of the NEXUS Nuclear Data Methodology, WCAP-16045-P-A Addendum 1-A (Proprietary), August, 2007 7-3 Ouisloumen, M., et al., Qualification of the Two-Dimensional Transport Code PARAGON2, WCAP-18443-P-A (Proprietary), July, 2021.

7-4 Liu, Y. S., et al., ANC: A Westinghouse Advanced Nodal Computer Code, WCAP-10965-P-A (Proprietary), September, 1986.

7-5 Zhang, B., et al., Qualification of the New Pin Power Recovery Methodology, WCAP-10965-P-A Addendum 2-A (Proprietary), September, 2010.

7-6 Holden, N., Neutron Capture Cross Section Standards for BNL 325 Fourth Edition, BNL-NCS-51388, January, 1981.

7-7 Leppnen J., et. al., The Serpent Monte Carlo Code: Status, Development and Applications in 2013, Annals of Nuclear Energy (82), 142-150, 2015.

7-8 Bordelon, F. M., et.al, Westinghouse Reload Safety Evaluation Methodology, WCAP-9272-P-A (Proprietary), July 1985.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 32 Table 7-1 Summary of 59Co (n,) Thermal Cross Section Measurement Data Meas. #

Citation in Ref. 7-2 (n,)

(barns) 1 20 36+/-3 2

21 41.5+/-3.9 3

22 36.1+/-4 4

23 36.6+/-1.5 5

24 35.7+/-1.8 6

25 36.9+/-1 7

26 38.4+/-1.2 8

27 38.5+/-1.6 9

28 36.4+/-1.5 10 29 38.5+/-0.7 11 30 38.4+/-0.5 12 31 36.3+/-0.6 13 32 35.51+/-0.87 14 33 34.9+/-1 15 34 37.5+/-1.9 16 35 35.1+/-0.7 17 41 38+/-2 18 37 38+/-0.3 19 38 37.7+/-0.4 20*

36 37.3+/-0.3 21*

39 37.16+/-0.24 22*

40 37.19+/-0.48 23*

42 37.245+/-0.11 24*

8 37.145+/-0.07

• Most precise measurements.

Weighted Average:

All 24 Measurements = 37.204+/-0.075 barns Most Precise Measurements = 37.178+/-0.056 barns

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 33 Table 7-2 Summary of 59Co (n,) Resonance Integral Measurement Data Meas. #

Citation in Ref. 7-2 (n,)

(barns) 1 29

~70 2

48 75.2+/-5 3

49 74.8+/-7.2 4

50 72.3+/-4.2 5

34 73.3+/-7.9 6

51 74.0+/-8.3 7

37 70.4+/-7.0 8

52

~69 9

40 75.8+/-3.7 10 41 73.6+/-7.0 11 35

~70.3 12 53 76.2+/-8.1 Weighted Average = 74.2+/-2.0 barns

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 34 Table 7-3 Summary of PARAGON and SERPENT2 Assembly Reactivity Differences Value PARAGON - SERPENT2 (pcm)

COBA - No COBA (pcm) 0.374 inch fuel rod OD Average RMS Max.

Min.

0.360 inch fuel rod OD Average RMS Max.

Min.

Table 7-4 Summary of PARAGON2 and SERPENT2 Assembly Reactivity Differences Value PARAGON2 - SERPENT2 (pcm)

COBA - No COBA (pcm) 0.374 inch fuel rod OD Average RMS Max.

Min.

0.360 inch fuel rod OD Average RMS Max.

Min.

a,c a,c a,c a,c a,c a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 35 Figure 7-1 Illustration of COBA Rod As Modeled in PARAGON a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 36 Figure 7-2 Summary of 59Co (n,) Thermal Cross Section Measurement Data

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 37 Figure 7-3 Comparison of PARAGON and SERPENT2 Assembly Reactivity for Various COBA Loadings (0.374 in. fuel rod OD) a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 38 Figure 7-4 Comparison of PARAGON and SERPENT2 Assembly Reactivity for Various COBA Loadings (0.360 in. fuel rod OD) a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 39 Figure 7-5 Comparison of PARAGON2 and SERPENT2 Assembly Reactivity for Various COBA Loadings (0.374 in. fuel rod OD) a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 40 Figure 7-6 Comparison of PARAGON2 and SERPENT2 Assembly Reactivity for Various COBA Loadings (0.360 in. fuel rod OD) a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 41 Figure 7-7 Comparison of PARAGON, PARAGON2 and SERPENT2 Intra-Assembly Pin Power Distributions with [ ]a,c Fresh COBA, 0.374 inch Fuel Rod OD a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 42 Figure 7-8 Comparison of PARAGON, PARAGON2 and SERPENT2 Intra-Assembly Pin Power Distributions with [ ]a,c Irradiated COBA, 0.374 inch Fuel Rod OD a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 43 Figure 7-9 Comparison of PARAGON, PARAGON2 and SERPENT Intra-Assembly Pin Power Distributions with

[

]a,c Fresh COBA, 0.374 inch Fuel Rod OD a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 44 Figure 7-10 Comparison of PARAGON, PARAGON2 and SERPENT2 Intra-Assembly Pin Power Distributions with [

]a,c Irradiated COBA, 0.374 inch Fuel Rod OD a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 45 Figure 7-11 Comparison of PARAGON, PARAGON2 and SERPENT2 Intra-Assembly Pin Power Distributions with [ ]a,c Fresh COBA, 0.360 inch Fuel Rod OD a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 46 Figure 7-12 Comparison of PARAGON, PARAGON2 and SERPENT2 Intra-Assembly Pin Power Distributions with [ ]a,c Irradiated COBA, 0.360 inch Fuel Rod OD a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 47 Figure 7-13 Comparison of PARAGON, PARAGON2 and SERPENT2 Intra-Assembly Pin Power Distributions with [

]a,c Fresh COBA, 0.360 inch Fuel Rod OD a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 48 Figure 7-14 Comparison of PARAGON, PARAGON2 and SERPENT2 Intra-Assembly Pin Power Distributions with [

]a,c Irradiated COBA, 0.360 inch Fuel Rod OD a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 49 Figure 7-15 Comparison of PARAGON, PARAGON2 and SERPENT2 60Co Production with [

]a,c COBA, 0.374 inch Fuel Rod OD a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 50 Figure 7-16 Comparison of PARAGON, PARAGON2 and SERPENT2 60Co Production with [

]a,c COBA, 0.360 inch Fuel Rod OD a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 51 Figure 7-17 Example Quarter-Core Loading Pattens Without COBA and With Steady-State COBA Loading a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 52 8

Safety Analyses 8.1 Design Parameters and Best Estimate Flows The implementation of COBA assemblies was determined to have no impact on the design core bypass (Section 5.1). Accordingly, the implementation of COBA assemblies would not impact the current Design Parameters and Best Estimate Flows.

• • • Additional details are provided in the Site-Specific attachment to this LAR.

8.2 Design Transients and Controls Systems The impact of the implementation of cobalt-60 production in PWRs on the NSSS design transients, plant operability margin to trip analyses, and low temperature overpressure mitigation is discussed in the following sections.

Design Transients Evaluation The NSSS design transients represent operational events postulated to occur during the lifetime of the nuclear power plant. Depending on the magnitude and frequency of occurrences, these postulated events could be significant to the structural integrity of the RCS and components.

Therefore, the design transients are used to analyze and evaluate the cyclic behavior of the major components, which comprise the NSSS. The NSSS design transients are the fluid systems transients and are described by the variations in fluid pressure, fluid temperature and flow.

The impacts that cobalt-60 has on the nuclear fuel parameters have been evaluated. A reload safety analysis checklist (RSAC) assessment has been performed (see Section 7.4). While some minor nuclear fuel kinematic changes have been identified, the deviations are not out of line with the cycle-by-cycle variation and are expected to be small and in line with typical violations seen during typical reload evaluations. Additionally, it has been concluded COBA assembly implementation does not directly impact the thermal-hydraulic (T-H) parameters used in the departure from nucleate boiling (DNB) safety calculations (Section 5.1). Therefore, no fuel kinematic items are expected to be significantly impacted by the implementation of COBA assemblies.

The design parameter impacts from COBA are discussed in Section 8.1. It was determined that the implementation of COBA rods have no impact on design core bypass. Therefore, there are no significant fluence changes expected that will impact the P-T limits.

Control Systems Operability / Margin to Trip / Component Sizing Evaluation The insertion of cobalt-59 target material, which is used to produced cobalt-60, in fuel assembly thimbles does not impact the control systems operability since the control systems logic, setpoints and time constants, and the NSSS capacities and performance capabilities are not being revised.

Furthermore, this activity does not affect the full power design conditions and has negligible impact on the nuclear fuel kinetics parameters. Therefore, the control systems performance during steady state, and during and following the design basis operational transients is not affected. Since the control and protection system setpoints and time constants and the capacities do not change, the cobalt-59 target

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 53 material will have no effect on the available plant operational margins. The control system setpoints do not need to be revised.

Cold Overpressure Mitigation System / Low Temperature Overpressure Protection System (COMS /

LTOPS) Evaluation The COMS/LTOPS protects the reactor vessel from potentially being exposed to conditions of fast propagating brittle fracture caused by the design basis mass and heat input transients at low temperature operation. The insertion of cobalt-59 target material does not impact the design basis COMS/LTOPS transients since there are no significant changes to the best estimate RCS flows (BEFs) and core P. Furthermore, this activity does not affect the COMS/LTOPS design basis transients mitigation capabilities. The key input assumptions and limits that are used to generate the COMS/LTOPS setpoints are also not impacted. Thus, no COMS/LTOPS setpoint changes are needed and the current setpoints remain valid.

Decay Heat From COBAs The COBA rod component heating created by the gamma radiation produced by the decay of cobalt-60 has been determined for use in the safety analyses. All heat generated by the cobalt-60 decay is assumed to deposit into the core as an additional component of decay heat. A conservative total value is calculated which bounds the decay heat which is added by the inclusion of COBA in the core design. [

]a,c The added decay heat component from cobalt-60 decay is assumed to remain constant for all safety analyses due to the relatively long cobalt-60 half life. The total decay heat for each core will be ensured to remain below the calculated conservative total value by calculating the total cobalt-60 Curie production in a given cycle and comparing it to the calculated total decay heat from COBA that is included in safety analysis evaluations.

8.3 Non-LOCA Transients The cobalt-60 production process utilizes the fuel assembly (FA) thimbles where the cobalt-59 is positioned inside the FA thimbles using specially designed capsules. The non-LOCA safety analyses do not model the FA thimbles. As such, there is no impact on the non-LOCA safety analyses from the cobalt-60 production contingent on the following:

1. No adverse effect on the rod drop time

2. No adverse effect on the reactor trip rod position versus time curve.

3. No adverse effect on the rod reactivity insertion versus position curve.

4. No adverse effect on the RCCA-related RSAC parameters, including the core axial power shapes, peaking factors, and axial offset limits.

5. No adverse effect on the decay heat generated in the core following a reactor trip. The applicable decay heat model for a given plants reload cycle is an RSAC item, and must be confirmed to remain valid for plants using the cobalt-59 capsules.

6. No adverse effect on the core bypass flow fraction and core pressure drop.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 54 These six items were reviewed. It has been confirmed that Items 1, 2, and 6 (Section 5.2) are met.

Westinghouse has confirmed the assumptions in Item 4 above with respect to peaking factors and axial offset limits and expects no significant impact on the trip reactivity curve (Item 3 above) or on the general core axial power shape (Item 4 above) (Section 7). A significant impact on RSAC items, which would include Items 3 and 4 above, is not expected from the inclusion of COBA in a core design and any deviations from the current limits are expected to be small and in line with typical violations seen during typical reload evaluations that are dealt with during cycle development.

It has also been confirmed that no Thermal-Hydraulics (T-H) RSAC items are expected to be significantly impacted by the inclusion of COBA in a loading pattern design. Any deviations from current limits, if any, are expected to be small and in line with typical violations seen during typical reload evaluations that are dealt with during cycle development.

As noted in Sections 5.2 and 8.1, it has been confirmed that the implementation of COBA rods would not impact the current Design Parameters and Best Estimate Flows, core bypass flow fraction, or core pressure drop. Item 6 is therefore met.

The impact on the non-LOCA safety analyses of an expected maximum decay heat contribution from cobalt-60 production (Section 8.2) was evaluated to address Item 5 above. The decay heat curves used in the non-LOCA analyses are based on the American Nuclear Society (ANS) 5.1-1979 decay heat standard plus 2. The evaluation concluded there is sufficient margin in the standard ANS 5.11979-based decay heat models used in the non-LOCA safety analyses to offset the additional heat from cobalt-60 production. The additional heat due to cobalt-60 production does not adversely impact the non-LOCA analyses because the revised core compositions forming the basis for the margin used to offset the decay heat from cobalt-60 production will be confirmed on a reload basis.

In summary, the impact on the non-LOCA safety analyses from cobalt-60 production has been evaluated. Cobalt-60 production has a negligible to no impact on the non-LOCA safety analyses based on the expectation that the margin available in the decay heat curves will be confirmed during the reload process to be applicable and the expectation that any deviations from the current limits are expected to be small and in line with typical violations seen during typical reload evaluations.

Based on the above discussion, the bulk transient conditions in the core (i.e., power, flow, inlet temperature) will not be adversely affected by the COBAs for events where DNBR is a concern. The local effects of having COBA rods adjacent to fuel rods relative to DNBR were also considered and were found to have no adverse effects, as discussed in Section 5.1.

• • • Additional details are provided in the Site-Specific attachment to this LAR.

8.4 Loss of RHR at Midloop / Natural Circulation Cooldown The most limiting condition for a loss of RHR cooling occurs when the water level in the RCS has been drained to the middle of the hot leg (mid-loop). Having a lower inventory of water reduces the total enthalpy change needed to raise the bulk temperature of the reactor coolant to the boiling point.

For the Loss of RHR event, an increase in decay heat results in a shorter time for the core to heat up to saturation and boiling conditions.

The Natural Circulation Cooling (NCC) analysis determines the capability of the RCS to remove heat from the reactor core following the loss of forced flow from the reactor coolant pumps. Following a loss of AC power, forced flow in the Reactor Coolant System is lost. The only remaining method

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 55 available for heat transfer in the RCS is via natural circulation, which results from density differences between the hot and cold reactor coolant. For the NCC event, a higher decay heat will require more heat transfer and may slow the cooldown rate.

The decay heat values for the Loss of RHR at Midloop and NCC analyses need to be evaluated to determine whether there is adequate margin available to bound the additional heat from fully-irradiated COBAs. More refined calculational methods exist to calculate decay heat than those used in ANS-5.1-1979, and one possible method is to compare the analyzed decay heat to a calculation of the best-estimate decay heat for the analyzed conditions. The results of this evaluation will determine whether the existing analysis has sufficient margin to account for both the heat from COBAs and a 2-sigma uncertainty of 2%.

The heat produced by COBAs is generally small enough to have no adverse impact on the analyses of record. Additional proposed changes to plant operation may also need to be considered. The decay heat margin in the Loss of RHR and NCC analyses may be adequate to accommodate only the heat from COBAs. Other changes that affect decay heat production in the fuel, such as a transition to 24-month cycles or a power uprate may result in a total impact to decay heat beyond the available margin.

• • • Additional details are provided in the Site-Specific attachment to this LAR.

8.5 FLEX Operation, Shutdown Cooling, and SFP Cooling and Boron Dilution Cobalt-60 production using cobalt burnable absorbers (COBA) results in an additional decay heat source that will increase the total reactor core decay heat transient. This can affect plant operations that require decay heat removal, including Flexible Coping Strategies (FLEX) that have been implemented to address an extended loss of AC power (ELAP) (Reference 8-1), spent fuel pool (SFP) cooling and heat-up time, SFP boron dilution, and plant cooldown.

FLEX plant functions that could be affected include:

1. Core Cooling & Heat Removal

2. RCS Inventory Control

3. Containment Integrity

4. Spent Fuel Pool Cooling A generic, comparative evaluation of various decay heat models, both with and without COBA decay heat, was completed to examine conservatism of the decay heat models used in these analyses and to establish screening criteria to determine the need for possible further evaluation of analyses of record. Note that, the term conservative is used to mean that a particular decay heat model yields higher decay heat at any given time relative to another model.

Decay Heat Model Comparison and Impact Various decay heat models are available for comparison for the scope of FLEX, shutdown cooling (SDC) or Residual Heat Removal (RHR) cooldown, and SFP analyses. The models that were evaluated include U.S. Nuclear Regulatory Commission (NRC) Branch Technical Position (BTP)

ASB 9-2 (Reference 8-2), Westinghouse BOP-FR-8 (Reference 8-3), and ANSI/ANS-5.1-1979 at standard deviations () of = 0, = 1, and = 2 (Reference 8-4). The decay heat output from these models for a specific core full power level were compared to determine which model produces the

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 56 minimum decay heat and thus could be considered as a baseline reference model. Comparing the ANS-5.1-1979 = 0, ASB 9-2, and BOP-FR-8 models shows that ANS-5.1-1979 = 0 yields the minimum decay heat for times after shutdown (ASD) up to 58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br />, while the ASB 9-2 model yields minimum decay heat for times after 58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br />. These models are therefore used as baselines for comparing the impact of COBA to the other models because all other models would yield greater cooling requirements. The Westinghouse RHR cooldown program, RHRCOOL, includes the ANS-5.1-1979 = 0, 1, & 2 and BOP-FR-8 models as analysis options. Of the models included in the RHRCOOL program, the ANS-5.1-1979 = 0 model yields the minimum decay heat for times ASD up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, while the BOP-FR-8 model yields the minimum decay heat for times after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

A conservative COBA decay heat has been established for 24 fully irradiated COBAs (Section 8.2) and was added to the baseline models to establish minimum COBA decay heat models, which were then compared to the other models to develop screening criteria to determine if an analysis is impacted by COBA or may need to be revised. This additional decay heat is a small percentage of the core decay heat, ranging from roughly 1.0% at 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ASD to 2.5% at 7 days (168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />) ASD, which is a FLEX and SFP relevant time range. It is also generally smaller than the difference between the historically accepted models ANS-5.1-1979 = 0 and BOP-FR-8; therefore, the addition of COBA decay heat could be judged to be within general modeling uncertainties; however, the deterministic approach of establishing minimum core plus COBA decay heat models was used to ensure conservatism for analyses with low margins. To this end, a qualitative evaluation is recommended at minimum for COBA plants based on the aforementioned screening criteria, and given below, followed by quantitative evaluation for analyses with low margin or that do not meet the screening criteria.

Screening Criteria for initial evaluation of conservatism of existing analyses that model decay heat (FLEX, RHR cooldown, or SFP):

1. FLEX analyses that follow the recommendation of WCAP-17601-P to use the ANS-5.1-1979

= 2 are conservative and do not require re-analysis. All others should be reviewed, although ANS-5.1-1979 = 1 is also shown to be conservative for all times up to 414 hours0.00479 days <br />0.115 hours <br />6.845238e-4 weeks <br />1.57527e-4 months <br />, after which, the negative margin relative to the ANS = 0 + COBA model is negligible.

2. Any analyses for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or less using ANS-5.1-1979 = 1 or = 2 or BOP-FR-8 are considered conservative and do not require re-analysis.

3. Any analysis using ANS-5.1-1979 = 2 will be conservative for the addition of COBA at all times after shutdown.

4. Any analysis using ANS-5.1-1979 = 0 after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> can be considered conservative for the addition of COBA; however, addition of the best-estimate realistic COBA decay heat value of 225 kW is recommended to account for model uncertainties.

Recommendations:

1. Any analysis using ANS-5.1-1979 = 0 for times less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> should be reviewed for sufficient margin relative to the impact of decay heat on key results. If margin is low, the analysis should be revised to account for the added COBA decay heat.

2. Analyses based on ASB 9-2 or BOP-FR-8 and starting 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ASD or later should be revised to include the additional COBA decay heat. Either model can be used.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 57 Conclusions For those plants considering the addition of COBAs, the decay heat models used in the AOR for those plants should be evaluated against the screening criteria to determine whether the addition of COBA may impact plant operations that require decay heat removal, including FLEX strategies that address an extended loss of AC power; shutdown cooling analyses; and spent fuel pool analyses. The screening criteria were developed using a deterministic approach to establish minimum core decay heat plus COBA decay heat models to ensure conservatism in the case of analyses with low margins for decay heat removal

• • • Additional details are provided in the Site-Specific attachment to this LAR.

8.6 Loss of Coolant Accident 8.6.1 COBA Capsule Survivability The survivability of the COBA rodlet [

]a,c capsule during a loss-of-coolant accident (LOCA) has been assessed based on an evaluation of [

]a,c as determined by a single rod burst test (SRBT), evaluation of material properties at elevated temperatures, and dimensional analysis of the capsule-to-rodlet and rodlet-to-thimble gaps.

Material Rupture Properties

[

]a,c were evaluated in a Westinghouse single rod burst test (SRBT) (Reference 8-5) and in two Oak Ridge National Laboratory studies (References 8-6 and 8-7). The relevant dimensional values for each study cited as compared to the nominal and most conservative COBA capsule dimensions is given in Table 8-1.

Isobaric Ramp Testing Westinghouse SRBT specimens (References 8-5 and 8-8) were subjected to a series of internal pressures up to 2000 psia. Specimens were pressurized at temperature marginally above Hot Full Power (HFP) hot leg conditions (630F) and ramped to 2300F, then held for 60 seconds. Alumina pellets were used to create void fractions representative of loaded cladding, with two diametrical gaps examined. An 8-mil gap was examined as a prototypical specimen, while a 24-mil gap was tested as an oversized gap condition. Burst temperature for each specimen was recorded.

ORNL testing performed in 2016 (Reference 8-6) investigated time-temperature histories approximating LOCA condition. Zirconia pellets were used to create void fractions representative of loaded cladding, and the specimens were pressurized at temperature marginally above Hot Full Power (HFP) cold leg conditions (572F) and ramped at (9°F/sec) to 2200F, then held for 3 minutes, ramp-down at (9°F/sec) to 1472°F, and finally, quenching with room temperature water. Burst temperature for each specimen was recorded. Table 8-2 provides the average burst temperature vs pressure results from Reference 8-5 and the individual burst temperature vs pressure results from Reference 8-6.

The volume of gas inside the COBA capsule uses the same methodology used to estimate the gas volume in the other SRBT articles. It is assumed that the slug stack takes up the entire length of the tube. The gas volumes are given in Table 8-1. From the calculated gas volumes, the Westinghouse 8-mil SRBT and ORNL SRBT specimens are more comparable to the COBA capsules than the Westinghouse 24-mil SRBT specimens. The gas volume is calculated for the SRBT specimens using the following form:

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 58

4 2

4 2 Eq. 1 The burst temperature vs pressure for References 8-5 and 8-6 is plotted in Figure 8-1 in addition to digitally extracted data from a previous burst study performed by Coffman (Reference 8-9). The equation derived from the Coffman data indicates an exponential relationship between hoop stress and temperature that takes the following form:

T(F) = 2432.9e-2*10^-5x Eq.2 where x is the hoop stress in psia. At 2200F, the hoop stress required to produce a rupture is calculated to be 5031psia. From this, using the wall thickness and outer diameter of the COBA capsule, the internal pressure at HFP required to burst at LOCA conditions is calculated using the following form (Reference 8-10):

P=2t/d Eq. 3 where P is the internal pressure (psi), is the hoop stress (psi), t is the wall thickness (inch), and d is the outer diameter (inch).

For the nominal dimensional values given in Table 8-1, the internal pressure required to burst at 2200F is 520 psia. A calculation of the internal pressure required to burst with the most conservative dimensions in tolerance (0.3015 OD and 0.0148 wall thickness) is 494 psia. A conservative calculation of the internal pressure generated by gas production in the COBA capsule during residence in reactor estimates the internal pressure at HFP will be 154 psia and will not exceed 334 psi at LOCA temperatures, which provides sufficient margin to indicate rupture will not occur in the event of a LOCA within 3 minutes.

Isothermal Testing ORNL testing performed in 1963 (Reference 8-7) investigated isothermal time-to-rupture for a single heat of [

]a,c at temperature ranging from 1100°F to 2200°F and time up to 4000 hours0.0463 days <br />1.111 hours <br />0.00661 weeks <br />0.00152 months <br /> in air; and in 90 vol % He, 8 vol % N2, 2 vol % O2; and in vacuum. Two specimen lengths, 2.5 and 6 inches, were investigated, and no discernable difference in stress-rupture behavior was observed until 1800F. It is noted at 1800F that end effects on the shorter tubes became sufficient to lower the effective stress on the tube wall, resulting in a greater apparent strength in the shorter specimens. All specimens tested above 1800F were the shorter tube length.

Additionally, no pellets were placed inside the test specimens to reduce the void fraction. Table 8-3 provides the results of this testing at relevant temperatures (1800F, 2000F, and 2200F). Figure 8-2 plots the time-to-rupture at 1800F for 6-inch and 2.5-inch test specimens, along with fitted power-law curves. Figure 8-3 plots the time-to-rupture for 2000F and 2200F. The capsule maximum internal pressure can be calculated from the hoop stress given in Table 8-3 again using Eq. 3:

The maximum internal pressure calculated at LOCA conditions of 334 psia is lower than the predicted 387.4 psia, based on COBA dimensions given in Table 8-1 and hoop stress of 3750 psia at 2200°F from Table 8-3, with time to rupture estimated to be 101 seconds.

Additionally, prior to burst, the capsule is expected to diametrically strain. If the tangential strain is greater than the diametrical gap with the outer rodlet, the capsule outer diameter will contact the outer rodlet inner diameter and reduce the applied hoop stress, preventing capsule rupture.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 59 Material Mechanical And Physical Properties Analysis of the LOCA/Non-LOCA temperature conditions establishes upper bounding temperatures for LOCA conditions for the capsule and slug centerline, as well as upper bounding temperatures for non-LOCA overpower conditions for the capsule and slug centerline. These bounding temperatures remain below the melting ranges for [

]a,c and provide satisfactory margin to indicate neither the capsule nor the slugs will melt in a LOCA or in a non-LOCA overpower scenario. As the capsules are not under load, the impact on mechanical properties will be negligible to the survivability of the capsules.

Dimensional Analysis Table 8-4 summarizes the OD, ID and wall thickness data for the capsule, rodlet, and the thimble.

Nominal dimension and gap calculations are presented. All dimensions are in inches.

Based on the dimensions above its predicted that a rupture in the [

]a,c under pressure will not result in escape of cobalt slugs from the ruptured capsule. The [

]a,c out of the capsule. Similarly, if the capsule to rodlet interaction results in a break in the rodlet, cobalt slugs will still be contained inside the rodlet. The [

]a,c out of the rodlet. The rodlet tube is [

]a,c Any release of cobalt slugs that may result from a rupture in the capsule or a break in the rodlet tube would be contained within the thimble tube.

Material Interaction The material interaction between [

]a,c has been evaluated. During normal operation the interaction is limited to the differences in irradiation growth rates, coefficients of thermal expansion, and slight friction wear (Section 4.6 discusses the wear interaction between the capsule and rodlet). Since RCS water is passing through [

]a,c

[

]a,c However, since this type of reaction takes longer than the time at peak temperature during LOCA, the impact of eutectic reaction will not result in significant damage to the rodlet. Its also noted that at a temperature of about 100°C lower than peak temperature the reaction time increases by a factor of 60 which significantly minimizes any impact on the capsule and the rodlet. Any damage that may occur to the rodlet or the capsule will not result in the release of cobalt slugs to the fuel assembly.

Conclusion The COBA material rupture, mechanical, physical, and interaction properties in addition to the COBA assembly dimensions have been evaluated for survivability of the capsule in LOCA conditions. The following conclusions have been made:

The build-up of internal pressure due to gas generation during residence will not be sufficient to induce rupture in the capsule during the peak LOCA temperature duration.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 60 Peak temperatures during LOCA and transient non-LOCA conditions will not result in melting of the slugs or capsule.

In the event of a capsule rupture, the defense in depth of the capsule, outer cladding, and thimble tube will prevent cobalt slugs from escaping the COBA assembly.

Oxide formation on the outer tube and capsule will mitigate the risk of eutectic formation, and in the event of hard contact resulting in eutectic formation will result in localized melting in the zirconium tube rather than bulk melting.

8.6.2 Small Break and Large Break LOCA The impact to Small Break and Large Break LOCA Analyses is concerned with the behavior of the COBA inserts during irradiation and under LOCA conditions. COBA inserts that behave passively with no negative consequences during a high temperature, low reactor pressure transient are evaluated with respect to LOCA based on the impact of nuclear fuel peaking factors, axial power distributions, rod burnups, any changes to fuel rod design parameters, and the additional thermal mass. Additionally, the impact on design parameters and pressure losses in the reactor and core bypass flow are considered. Lastly, coolability of the COBA assemblies under LOCA conditions is addressed.

The previous section discusses the survivability of the stainless steel capsule during a LOCA. Based on the conclusions in Section 8.6.1, the COBA inserts are assumed to behave passively under LOCA conditions, except for the thermal heat addition from decay heat and gamma energy.

Based on a review of expected impacts on the RSAC (key safety parameters) of including COBA in the core design, the power shape distributions assumed in the LOCA analyses and the peaking factor vs burnup behavior are not expected to be impacted by the inclusion of COBA. This will be confirmed as part of the standard plant-and cycle-specific RSAC and reload evaluation process. Additionally, a review of the impacts on the fuel generalized energy deposition model (GEDM) concludes the overall gamma energy deposition in the fuel remains appropriate. Finally, no changes to the fuel rod design parameters were identified.

• • • Additional details are provided in the Site-Specific attachment to this LAR.

8.6.3 LOCA Containment Integrity The effect of COBA assemblies in the core on the long-term LOCA containment integrity analysis has been evaluated. The long-term LOCA mass and energy (M&E) releases are most sensitive to changes in the following on the primary side:

Increases in the reactor coolant system (RCS) temperatures, pressures, fluid volumes Increases in the core stored energy Core Power increases Pumped emergency core cooling system (ECCS) flow and to changes in the following on the secondary side:

Fluid mass Metal mass

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 61 The addition of the Co-60 burnable absorber (COBA) assemblies will not increase the licensed core power for a plant. The initial conditions of the RCS will not be impacted. For plants that operate with wet annular burnable absorbers (WABAs) some of the WABAs would be replaced by the COBAs. In that case, the RCS fluid volume would be unchanged. There would not be any change in the primary fluid mass. For plants that currently do not have WABAs, installing the COBAs would occupy volume so the initial primary side fluid volume would decrease, which would be less limiting for the initial primary side blowdown into the containment from the postulated ruptured primary side piping. In general, the primary side fluid volume will not increase with the addition of COBA assemblies so the current licensing basis initial condition would remain valid.

The core region in the LOCA mass and energy approved methodology presented in WCAP-10325-P-A and its supporting documentation (References 8-11, 8-12, and 8-13) does not model any miscellaneous metal within the core. Metal is modeled in the reactor vessel and the remainder of the RCS. Based on the total mass of metal in the reactor vessel and the mass of the fuel rods in a representative Westinghouse PWR, the maximum increase in the metal mass due to adding [

]a,c is insignificant. Since the additional mass from [

]a,c is insignificant, the surface area of the COBAs does not need to be considered. Also, since the additional metal mass from the proposed [

]a,c has been shown to be insignificant, any sensible energy from the metal is considered to be negligible.

The inclusion of Co-60 burnable absorber rod assemblies will not increase the initial core stored energy of a fuel product. However, there will be an increase to the decay heat energy release rate contribution from the COBAs. The additional decay heat energy release must be accounted for in the long-term LOCA containment analysis to meet the requirements of Section 6.2.1.3 of NUREG-0800.

This additional decay heat energy can be addressed by applying a conservative, constant decay heat energy rate to the mass and energy releases generated using the methodology from WCAP-10325-P-A for the duration of the limiting long-term LOCA containment response cases.

The addition of COBAs will not have any impact on the pumped safety injection flow or on the secondary side metal or fluid mass of the steam generators for a PWR.

The only impact on the long term LOCA mass & energy releases for containment integrity would be the consideration of the increase in decay heat energy. This consideration would likely have a minor impact on the post-24-hour containment pressure and temperature response and would need to be evaluated for any impact on the equipment qualification (EQ) program.

• • • Additional details are provided in the Site-Specific attachment to this LAR.

8.6.4 Post-LOCA Long Term Cooling The typical post-LOCA long term cooling (LTC) analysis of record (AOR) consists of three elements:

subcriticality, boric acid precipitation control (BAPC), and decay heat removal (DHR). The impact to post-LOCA LTC analyses is concerned with the behavior of the COBA inserts during irradiation and under LOCA conditions. Additionally, the impact on design parameters and other post-LOCA LTC analysis inputs is considered. Lastly, coolability of the COBA assemblies under post-LOCA conditions is addressed.

As noted in Section 8.6.2, the COBA inserts are assumed to behave passively under LOCA conditions, except for the thermal heat addition from decay heat and gamma energy. Furthermore, there is no expected impact on the RSAC parameters when including COBA in the core design, and

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 62 this will be confirmed as part of the standard plant-specific and cycle-specific reload evaluation process.

• • • Additional discussion is provided in the Site-Specific attachment to this LAR.

8.7 Steam Generator Tube Rupture The purpose of the Steam Generator Tube Rupture (SGTR) analysis is to produce mass release data for input to the dose calculations. In addition, many plants also analyze to demonstrate margin to steam generator overfill, which confirms liquid releases do not need to be considered in the dose analysis. The plant response to an SGTR is dominated by parameters such as break size, steam generator relief valve size, auxiliary feedwater flow rates, safety injection flow rates, decay heat, latent heat, and operator action timing (if applicable). The SGTR event is not sensitive to local conditions in the reactor vessel, and the mass of fuel in the core is not a significant input. The additional mass of the COBA assemblies is not significant and would not change the analysis modeling. SGTR input to dose calculations have historically provided mass release data considering removal of decay heat via steam releases from the intact steam generators until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after shutdown. The decay heat added by the COBA assemblies is not significant up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after shutdown and would not change the analysis decay heat modeling. However, many plants SGTR dose analyses model steam releases extending beyond 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after shutdown to consider a plant-specific time for the transition to cooling by the residual heat removal system (RHRS). The impact of COBA assemblies on the decay heat increases with time and thus, the effects of COBA on the decay heat and steam releases past 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after shutdown must be examined on a plant-specific basis. The effects of the COBA on the plant-specific time to transition to cooling by the RHRS to terminate steam releases must also be examined.

Thus, the addition of the COBA assemblies must be examined on a plant-specific basis.

• • • Additional discussion is provided in the Site-Specific attachment to this LAR.

8.8 Steamline Break M&E and Steam Release for Dose Steamline Break (SLB) M&E Release Analyses The cobalt-60 production process utilizes the fuel assembly (FA) thimbles. Cobalt-59 is positioned inside the FA thimbles. The analyses of SLB M&E releases for containment and compartment pressure and temperature response inside and outside containment do not model the FA thimbles.

As such, it is concluded that no impact on the SLB M&E release analyses would occur from the cobalt-60 production as long as the six assumptions presented in Section 8.3 are confirmed. The impact of all six items is dispositioned as discussed in Section 8.3.

Therefore, cobalt-60 production has a negligible to no impact on the SLB M&E release inside and outside containment analyses.

• • • Additional discussion is provided in the Site-Specific attachment to this LAR.

Steam Release for Dose Analyses The purpose of the steam release for dose analysis is to produce mass release data for input to the non-LOCA dose calculations. The steam release for dose analysis is a conservative mass and energy balance that accounts for the removal of decay heat and removal of the latent heat in the

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 63 primary and secondary systems. It is necessary to evaluate the steam release for dose analyses for the effect of the decay heat generated in the core following a reactor trip.

The impact on the steam release for dose analysis of an expected maximum decay heat contribution from cobalt-60 production has been evaluated.

The decay heat from cobalt-60 production results in a small increase in steam releases used for dose consequences.

• • • Additional discussion is provided in the Site-Specific attachment to this LAR.

8.9 Radiological Consequences (Doses) of Accidents A typical suite of dose analyses would include those covered in Regulatory Guide 1.195, Revision 0 and Regulatory Guide 1.183, Revision 0. Those events include: Loss of Coolant Accident (LOCA),

Main Steamline Break (MSLB), Steam Generator Tube Rupture (SGTR), Control Rod Ejection (CRE)

- Containment Leakage, CRE - Steam Generator Leakage, Locked Rotor, and Fuel Handling Accident (FHA). Some plants may have additional analyses such as Loss of Offsite Power (LOOP),

failure of small lines (e.g. instrument line, letdown line) carrying primary coolant outside containment (referred to as a small line break outside containment, SLBOC), and tank ruptures.

Plants with analyses based on Regulatory Guide 1.195, Revision 0 (or similar methods) are only required to consider accident doses based upon iodine and noble gases. Cobalt is neither an iodine nor a noble gas, and therefore, dose analyses based on Regulatory Guide 1.195, Revision 0 (or similar methods) would not be impacted by increased cobalt-60.

Plants with analyses based on Regulatory Guide 1.183, Revision 0 (or similar methods) include consideration of cobalt as a noble metal. Following Regulatory Guide 1.183, Revision 0, noble metals are only considered as part of the core melt accident, i.e. a design basis LOCA. Other events only consider the release of halogens (Iodine, Bromine), noble gases (Krypton, Xenon), and alkali metals (Cesium, Rubidium).

A key input to accident analyses is the radiological source term. NUREG-1465 developed representative severe accident (beyond design basis) source terms for light water reactors, and was subsequently endorsed in Reg Guide 1.183, Revision 0. In NUREG-1465, cobalt is included in the same fission product group as noble metals. NUREG-1465 states that cobalt was added to the noble metals based on relative dose consequence studies and assuming it is released in large quantities.

However, there is no physical/chemical basis for treating cobalt as a noble metal. Thus, it is reasonable to determine a release fraction for cobalt, given the increase in cobalt inventory from the COBAs.

The LOCA dose impact was assessed using a cobalt-60 release fraction of [

]a,c, as compared to the Regulatory Guide 1.183, Revision 0 noble metal release fraction of 0.25%. The cobalt-60 release fraction was developed from a study of high and low pressure severe accident sequences. The dose impact of the added cobalt-60 in the COBA assemblies on the LOCA dose analysis is expected to be small, on the order of less than 2% of the current calculated dose for plants using Regulatory Guide 1.183, Revision 0 (or similar methods).

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 64 The control rod ejection (CRE) event was examined to determine if the power increase during the event could potentially result in melting of the cobalt pellets within the COBAs. It was found that because the majority of the power increase occurs within the UO fuel, the cobalt temperatures in an adjacent COBA would not increase sufficiently to reach the melting point. Therefore, the cobalt pellets within the COBAs are not expected to melt during a CRE event. Thus, while a CRE dose analysis may model a limited amount of centerline fuel melt, COBA rods would not experience melt as a result of a CRE, and thus, there would be no Co-60 released from COBA following a CRE. This must be confirmed on a plant-specific basis (See section 8.3 in the Site-Specific attachment).

In summary, the dose impact of the added cobalt-60 in the COBA assemblies on the LOCA dose analysis is expected to be small, on the order of less than 2% of the current calculated dose for plants using Regulatory Guide 1.183, Revision 0 (or similar methods). However, the dose impact to a given analysis is highly dependent on plant-specific assumptions such as atmospheric dispersion factors, activity removal mechanisms, etc. Thus, any plant-specific implementation of COBA for a plant that follows Regulatory Guide 1.183, Revision 0 (or similar methods) will require a plant-specific LOCA dose assessment. A plant-specific assessment of the potential for COBA melting during a CRE event will also be required. A plant-specific assessment related to SGTR steam releases and steam release for dose used in non-LOCA dose analyses and the time the releases are terminated will also be required. The addition of COBA assemblies is not expected to challenge the dose analysis limits, but this must be confirmed on a plant-specific basis.

• • • Additional discussion is provided in the Site-Specific attachment to this LAR.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 65 References 8-1 NEI 12-06, Revision 4 Diverse and Flexible Coping Strategies (FLEX) implementation guide, December 2016.

8-2 NUREG-0800 9.2.5, Revision 2 Standard Review Plan Ultimate Heat Sink, July 1981.

8-3 BOP-FR-8 Decay Heat Standard Document, Revision 1, Functional Requirements and Design Criteria Residual Decay Heat Standard, March 1973.

8-4 ANSI/ANS-5.1-1979, Decay Heat Power in Light Water Reactors, © 1979 by the American Nuclear Society.

8-5 WCAP-8611, Stainless Steel Single Rod Burst Test, P.J. Kuchirka, September 1975.

8-6 Massey et al., Cladding Burst Behavior of Fe-Based Alloys Under LOCA, JNM 470, pp128-138 (2016) 8-7 ORNL-TM-535, Stress Rupture Properties of 304 SS Tubing - Oak Ridge National Laboratory, June 7, 1963.

8-8 WCAP-7379, Performance of Zircaloy Clad Fuel Rods During a Simulated Loss-of-Coolant-Accident-Single Rod Tests, September 1969.

8-9 F. D. Coffman, LOCA Temperature Criterion for Stainless Steel Clad Fuel, NUREG-0065, 1976.

8-10 Barlows Formula - Calculate Internal, Allowable and Bursting Pressure, literature (Attached in PRIME).

8-11 WCAP-10325-P-A, Revision 0, Westinghouse LOCA Mass and Energy Release Model for Containment Design March 1979 Version, May 1983.

8-12 Letter from Hebert N. Berkow, Director (NRC) to James A. Gresham (Westinghouse),

Revision 0, "Acceptance of Clarifications of Topical Report WCAP-10325-P-A, 'Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version' (TAC No. MC7980)," October 18, 2005. (ADAMS Accession No. ML052660242).

8-13 PWROG-17034-P-A, Revision 0, Evaluation of the WCAP-10325-P-A Westinghouse LOCA Mass & Energy Release Methodology, March 2020.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 66 Table 8-1: Relevant burst testing measurements Specimen Dimensional Measurements (inches)

Gas Volume (in3)

Ramp Rate (F/sec)

Length OD Wall Thickness Dia. Gap Westinghouse 1975 SRBT 8-mil Gap 15 0.422 0.0165 0.008 0.07257 20 Westinghouse 1975 SRBT 24-mil Gap 15 0.422 0.0165 0.024 0.21319 20 ORNL 1963 SRBT 6

0.750 0.020 0.710*

2.37551 2.5 0.98980 ORNL 2016 SRBT 12 0.375 0.015 0.022

0. 13731 9

COBA Capsule (Nom &

Conservative) 8.25 0.302 0.0156 0.0258 0.08861 8.253 0.3015 0.0148

0. 0299

0. 10280

• No pellets inside tube to reduce void fraction

• • Isothermal testing Table 8-2: Burst Temperature vs Pressure data from literature 8-mil gap (Reference 8-5)

Internal Pressure (psia)

Hoop Stress (psia)*

Average Burst Temperature (F)

Average Failure Ductility

(%)

550 7033 2212 24 1000 12788 1954 16 2000 25576 1838 16 24-mil gap (Reference 8-5)

Internal Pressure (psia)

Hoop Stress (psia)*

Average Burst Temperature (F)

Average Failure Ductility

(%)

300 3836 2121 14 550 7033 2040 15 1000 12788 1799 18 1500 19182 1757 19 2000 25576 1572 24 22-mil gap (Reference 8-6)

Internal Pressure (psia)

Hoop Stress (psia)

Burst Temperature (F)

Average Failure Ductility

(%)

Not given 6226 2271 Not given Not given 8979 1939 Not given Not given 12930 1822 Not given Not given 15540 1550 Not given

• Calculated from internal pressure using (Reference 8-10).

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 67 Table 8-3: Time-To-Rupture vs Hoop Stress (Reference 8-7)

Temperature (F)

Specimen Length (in)

Hoop Stress (psia)

Time To Rupture (s)

Maximum Tangential Strain at Failure (%)

1800 6

4500 4320 23.0 3600 11160 17.8 3000 17280 17.0 2500 34920 14.7 2000 85320 11.0 1600 146160 10.5 1300 243000 10.9 1000 448560 9.1 2.5 4000 19800 18.8 3000 57960 5.6 2200 136080 10.9 2200 133920 13.2 2000 2.5 5700 72 Not Given 4700 238 Not Given 2200 2.5 4700 72 Not Given 3750 101 Not Given

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 68 Table 8-4: Summary of Diametrical Measurements in COBA Assembly (inches)

Figure 8-1. Burst Temperature versus Internal pressure (References 8-5, 8-6, and 8-9) a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 69 Figure 8-2. Time-To-Rupture vs. Hoop Stress at 1800F (Reference 8-7)

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 70 Figure 8-3. Time-To-Rupture vs. Hoop Stress at 2000F and 2200F (Reference 8-7)

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 71 9

Effects on Spent Fuel Pool 9.1 Spent Fuel Pool Criticality The effect on spent fuel pool criticality is addressed on a site-specific basis.

9.2 Spent Fuel Pool Time to Boil Calculations have been performed to evaluate the impacts of COBAs on the time required for the SFP to heat up following a loss of spent fuel pool cooling. The calculation used the decay heat model from Reference 9-1 with a 10% uncertainty factor included.

A conservative calculation for the decay heat produced by [

]a,c with fully activated COBAs (resident in the core for [

]a,c Section 8.2) assumed that all decay heat in the COBAs is transferred to the SFP water.

The impact of COBAs on the time to boil (TTB) was evaluated for [

]a,c Since this change is small relative to the uncertainty of the decay heat calculations, it is concluded that the COBAs have no significant impact on the SFP heatup time.

9.3 Gamma Heating of the SFP Concrete Structure An assessment was performed to investigate the potential for degradation of the concrete walls and floor of the SFP due to the anticipated increase in the incident gamma energy flux on the SFP walls/floor resulting from:

a) Storage of standard 17x17 Westinghouse fuel assemblies containing cobalt burnable absorbers (COBAs, referred to in this section as COBA fuel assemblies) in the SFP.

b) Harvesting activities associated with extracting the irradiated COBA rodlets/capsules from the COBA Fuel assemblies.

9.3.1 Acceptance Criteria Acceptance criteria outlined in NUREG/CR 6927 (Reference 9-2) and summarized below, were used in this evaluation to assess the impact of a) the incremental incident gamma flux on the surfaces of the SFP structure due to the presence of irradiated COBA rodlets in the COBA fuel assemblies, and b) the incident gamma flux on the surfaces of the SFP structure as a result of the source / receptor configurations associated with harvesting of the irradiated COBA rodlets/capsules from the COBA fuel assemblies. In accordance with Section 4.3.1.1, of NUREG/CR 6927:

a) Nuclear heating of concrete structures need not be considered if the structure is exposed to an incident energy flux < 1010 Mev/cm2-sec, or by extension, if the contribution of the COBA rodlets in the COBA fuel assemblies, to the total incident gamma flux, is insignificant, and b) The degradation of concrete due to irradiation need not be considered if the total integrated dose due to exposure to COBA fuel assemblies remains < 1010 rads, or by extension, if the contribution of the COBA rodlets in the COBA fuel assemblies, to the total integrated dose, is insignificant.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 72 If the incident flux on the SFP wall / floor is estimated to be > 1010 MeV/cm2-sec, acceptance criterion outlined in Section A.4.1 of ACI 349-01 Appendix A (Reference 9-3) is utilized to confirm that there are no adverse effects from introduction of COBA fuel assemblies or COBA rodlet harvesting processes in the SFP. This criterion states that concrete temperatures during normal operation or any other long-term period shall not exceed 150oF except for local areas which are allowed to have increased temperatures not to exceed 200oF.

9.3.2 Primary Design Inputs The following design inputs applicable to COBA fuel assemblies, and to the proposed COBA rodlet harvesting procedures, are used in this assessment:

A fresh COBA rodlet contains [

]a,c capsules stacked one on top of the other, with each capsule containing [

]a,c cobalt-59 slugs stacked one on top of the other. Thus, each fresh COBA rodlet will contain a total of [

]a,c cobalt-59 slugs.

A maximum of [

]a,c COBA rodlets are attached to the COBA Insert assembly baseplate which is used to insert the COBA rodlets into the COBA fuel assembly guide thimbles (typically reserved for control rods or discrete burnable adsorbers) in a standard 17x17 Westinghouse fuel assembly.

COBA rodlets are always placed in new feed assemblies in the core. [

]a,c However, as a result of [

]a,c After exposure to [

]a,c the COBA rodlets in a host assembly will be harvested within approximately six months of assembly offload to the SFP.

There can be a maximum of [

]a,c COBA Fuel assemblies in SFP at any one time.

After exposure to [

]a,c each cobalt slug is expected to contain a maximum of [

]a,c This activity is a result of activation of the cobalt-59 in each slug while resident in the core in a COBA fuel assembly during power operations.

Based on the above:

o The maximum activity inventory of an irradiated COBA rodlet at the end of this process is

[

]a,c curies of cobalt-60.

o The maximum total additional activity inventory in a COBA Fuel assembly (versus a non-COBA spent fuel assembly) when placed in the SFP for storage could be as high as

[

]a,c of cobalt-60.

Summarized below are the key inputs applicable to the COBA rodlet harvesting procedures.

The COBA rodlet/capsule harvesting process will be initiated:

o After the desired exposure to [

]a,c is complete, and

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 73 o When the plant is not engaged in refueling or core offload operations.

The latter restriction is considered prudent and is intended to prevent procedural conflicts due to two major operations occurring simultaneously, i.e., fuel movement during refueling activities, and COBA rodlet harvesting. [

]a,c The COBA Insert assembly (with its attached inventory of [

]a,c COBA rodlets) is removed from its host COBA fuel assembly in the SFP and [

]a,c The COBA rodlets are then transported, [

]a,c and the enclosed [

]a,c capsules extracted and placed in a source cage. Each source cage can hold a maximum of 52 capsules with the capsules stored vertically around the circumference of the cylindrical source cage. Note that the assumption of [

]a,c curies) being stored in a source cage is conservative for evaluation of the cask pit structure since the curie content for a source cage is administratively restricted to 180,000 curies, and curie content for a transport cask (intended to transport two source cages), is limited to 400,000 curies.

The fully loaded source cage is then removed from the harvesting workstation and temporarily stored on a source cage storage stand until the source cage is transferred into a transport cask.

9.3.3 Results of Assessment

1. Storage of COBA Fuel Assemblies in the SFP Acceptability of storage of COBA fuel assemblies in the SFP is addressed by Judicious fuel management to maintain the incremental incident gamma flux on concrete surfaces due to storage of COBA Fuel assemblies within the specified criterion in NUREG/CR-6927; i.e., 1010 Mev/cm2-sec.

Demonstrating that the corresponding integrated dose to the concrete due to continuous presence of COBA fuel assemblies in the SFP for the remaining life of the plant with an assumed 40-year life extension, will remain within the specified criterion in NUREG/CR-6927; i.e., 1010 rads.

The critical distances between the COBA Fuel assemblies and adjacent concrete surfaces relevant to the temporary storage assessment associated with the storage configuration scenarios of a cluster of COBA assemblies in the SFP are provided in the Site-Specific attachment to this LAR. The results of the assessment of the incident gamma energy flux for the bounding configuration are also provided.

2. COBA rodlet harvesting / temporary storage prior to shipment The critical distances between the COBA sources (i.e., COBA rodlets / capsules) and adjacent concrete surfaces relevant to the assessment associated with harvesting / storage configuration scenarios are provided in the Site-Specific attachment to this LAR. The results of the assessment

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 74 of the incident gamma energy flux for the various configurations, including associated maximum temperatures in the concrete if the incident gamma energy flux acceptance criteria are exceeded, are also included. Acceptability of the harvesting process includes confirmation that the NUREG/CR-6927 acceptance criterion with respect to irradiation is met.

9.3.4 Conclusion It is concluded that there will be no adverse effects on the SFP structure resulting from gamma heating or irradiation due to introduction of COBA fuel assemblies into the SFP or during related COBA rodlet harvesting activities, provided the proposed guidance for storage of the COBA fuel assemblies and during the harvesting process is followed, as discussed in the Site-Specific attachment to this LAR.

• • • Additional discussion is provided in the Site-Specific attachment to this LAR.

9.4 SFP Cooling The added cobalt-60 decay heat noted in Section 8.2 can affect the capability of the SFP cooling system to maintain spent fuel pool temperature following a full core offload. The cobalt-60 assemblies are expected to be either returned to the reactor or harvested post-outage; therefore, the added heat only affects operations during an outage. Since the added heat is a small fraction of the decay heat from the fuel, the impact is expected to be small. However, in cases of low cooling margin and when analyses use decay heat models with minimal conservatism, the impact of the increased decay heat should be evaluated. Plant-specific SFP cooling capability may be evaluated using the screening criteria presented in Section 8.5 or other acceptable methodology.

• • • Additional discussion is provided in the Site-Specific attachment to this LAR.

References 9-1 NRC Branch Technical Position ASB 9-2, Residual Decay Energy for Light-Water Reactors for Long-Term Cooling (attached to NRC Standard Review Plan 9.2.5, Ultimate Heat Sink) 9-2 NUREG/CR-6927, Primer on Durability of Nuclear Power Plant Reinforced Concrete Structures

- A Review of Pertinent Factors February 2007.

9-3 Code Requirements for Nuclear Safety Related Concrete Structures ACI 349-01, Appendix A, Thermal Considerations, 2001

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 75 10 Harvesting 10.1 Harvesting Process Overview After the neutron activation of the cobalt in the reactor core, the capsules are harvested, removed from the COBA rodlets and placed in transport casks for shipment to the processing facility.

The cobalt-60 capsule harvesting system and process is used to [

]a,c 10.2 Harvesting Workstation Tooling 10.2.1 Harvesting Workstation The harvesting workstation is the base component for all harvesting operations and provides the attachment locations for all other tooling (Figure 10-1). The workstation structure is [

]a,c 10.2.2 Rodlet Cutter The rodlet cutter [

]a,c from the rodlet for harvesting of the capsules.

The rodlet cutter [

]a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 76

[

]a,c 10.2.3 Capsule Loading Tool The capsule loading tool [

]a,c 10.2.4 Capsule Source Cage The capsule source cage provides controlled storage of harvested capsules. The source cage is designed to hold 52 capsules. [

]a,c 10.3 Electrical and Software The electrical system and software are designed to support the various [

]a,c The software system was developed such that the cobalt-60 harvesting operations can be performed safely. The software functions on a Windows-based system (Figure 10-2) and [

]a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 77 Additional functions include [

]a,c 10.4 Foreign Material Control The harvesting system includes a [

]a,c 10.5 COBA Waste Disposal The COBA [

]a,c The disposition of the remaining parts of the COBA assembly following the removal and harvesting of the COBA capsules from the COBA assembly is addressed in Section 10.1 above.

10.6 Effects Within Spent Fuel Pool The source cage, when filled with capsules, either due to the administrative Ci limit or capsule capacity, will be stored on a temporary storage stand within the SFP to provide distance from the SFP floor. The effects of source cage staging are further discussed in Section 9.3.

10.7 Exposure to Personnel Calculations and evaluations have been performed for personnel working near the water or above the water (SFP bridge). Dose rates have been calculated with 10 feet of water coverage over fuel with COBA inserts. The dose rates are [

]a,c due to the contribution of the COBAs.

Harvesting operations are expected to occur during non-outage time periods (approximately 28-32 days). While the estimated total dose to the harvesting team will depend on actual site conditions, a

[

]a,c is expected for the entire process.

Transfer casks are provided by Nordion for shipping the irradiated capsules to the processing facilities. Casks are prepared and placed in the cask pit. The cobalt-60 filled source cages are loaded while submerged. The loaded source cages are placed into casks, then the casks are removed from the SFP. The expected external dose is <10 mrem/hr. Casks are drained and flushed, purged with argon, then vacuum dried based on total curie content. An additional argon purge is performed and the casks are prepared for shipment.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 78 Figure 10-1 Harvesting Work Station a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 79 Figure 10-2 Computer Software Display Screen a,c

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 80 11 Measurement of Cobalt-60 Activity Capsules containing cobalt-60 sources are loaded into source cages in preparation for shipping. The activity of the cobalt-60 capsules is verified and recorded as the capsules are loaded into a source cage to ensure that the total activity does not exceed the licensed limit of the shipping cask prior to shipment.

11.1 Hardware The Activity Measurement System (AMS) uses [

]a,c is calibrated in the AMS software using known standards prior to the arrival of the equipment to site. The AMS and associated software are non-safety related.

11.2 Process Activities The capsule activity confirmation process involves verifying [

]a,c The SPD output is calibrated against a capsule with a known activity to ensure accuracy of the AMS.

The capsule activity confirmation process is controlled in accordance with the Westinghouse Quality Management System. As operating experience is gained through the use of the AMS, changes to the hardware, software or overall process could be incorporated to optimize or enhance the design.

11.3 Software and Data Capture The AMS software, which is non-safety related, allows the operator to [

]a,c The measurements are logged to provide a report that includes the [

]a,c The software monitors the total activity to ensure the source cage limit is not exceeded. Warning messages are provided when the user approaches this limit and again if it is exceeded.

The AMS software will undergo a Verification and Validation (V&V) process prior to being released for use. This process checks each software feature and calculation to ensure the software functions properly. A V&V will be performed for each software version prior to release.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 81 12 Byproduct Material Tracking and Reporting The Cobalt-60 that is generated in the reactor core is radioactive byproduct source material that, under some circumstances, would be reportable to the National Source Tracking System (NSTS) as defined in 10 CFR 20.1003, Definitions. Such reportability would require certain transaction reports as specified in 10 CFR 20.2207 to document the manufacture, transfer, receipt, disassembly, or disposal of Nationally Tracked Sources.

The radioactive sources that are produced in the licensees reactor are encapsulated in a controlled configuration of sealed metal capsules containing the cobalt slugs. However, with the Westinghouse method of byproduct material production, the sealed sources are not in their final form for end use.

Within the industry, source producers and manufacturers commonly refer to this type of radioactive material as bulk material. Although the term is not defined in the NSTS final rule, it is generally understood to refer to precursor radioactive material, either sealed or unsealed, that is used in the fabrication of sources.

Based on the proposed source manufacturing process described in this document, and because these sources are not in a final form for end use, these sources do not fall within the scope of the nationally tracked source reporting requirements in 10 CFR 20.2207. Consequently, they would also not be subject to the serialization requirements in 10 CFR 32.201. Therefore, reporting the sources to the NSTS and the assignment of unique serial numbers is not required. Based on feedback and observations received from the NRC regarding marking and reporting of the cobalt-60 byproduct material, this position is consistent with NRC guidance.

As discussed in Sections 10 and 11, each capsule will have its activity measured following harvesting from the COBA rodlet as it is loaded into the source cage. These activity levels are documented to assure that the total activity level of the byproduct material in each Nordion transport cask is maintained within the licensed activity limit of the cask certification.

• • • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)