LR-N25-0074, Attachment 6: Site Specific Information in Support of Cobalt Absorber (Coba) Assemblies in Salem Generating Station, Units 1 and 2

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Attachment 6: Site Specific Information in Support of Cobalt Absorber (Coba) Assemblies in Salem Generating Station, Units 1 and 2
ML25268A078
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Site: Salem  PSEG icon.png
Issue date: 09/24/2025
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Westinghouse
To:
Office of Nuclear Reactor Regulation
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ML25268A071 List:
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LR-N25-0074, LAR S25-01
Download: ML25268A078 (1)
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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 Site Specific Information In Support of Cobalt Burnable Absorber (COBA) Assemblies In Salem Generating Station Units 1 and 2 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

This document provides material specific to Salem 1&2 to compliment the information provided in the Cobalt-60 Supplemental Information (Attachment 2) for designated topics. The section numbering used in this document is consistent with the section numbering of the Supplemental Information attachment. The designated topics are as shown below:

Table of Contents 4.10 Impact of COBA Capsule Breach on Reactor Coolant Chemistry 8.1 Design Parameters and Best Estimate Flows 8.3 Non-LOCA Transients 8.4 Loss of RHR at Midloop / Natural Circulation Cooldown 8.5 FLEX Operation, RHR Cooldown, SFP Cooling and Boron Dilution 8.6.2 Small Break and Large Break LOCA 8.6.3 LOCA Containment Integrity 8.6.4 Post-LOCA Long Term Cooling 8.7 Steam Generator Tube Rupture 8.8 Steamline Break M&E and Steam Release for Dose 8.9 Radiological Consequences (Doses) of Accidents 9.1 Spent Fuel Pool Criticality 9.3 Gamma Heating of the SFP Concrete Structure 9.4 Spent Fuel Pool Cooling 13 First Cycle Post-Irradiation Examination

      • 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

Acronyms AST Alternative Source Term ANS American Nuclear Society BAPC Boric Acid Precipitation Control CFD Criticality Fuel Design COBA Cobalt-60 Burnable Absorber CR Control Room CRSB COBA Rodlet Storage Basket DHR Decay Heat Removal DNBR Departure from Nucleate Boiling Ratio EAB Exclusion Area Boundary ELAP Extended Loss of AC Power EQ Equipment Qualification FLEX Flexible Coping Strategies HLSO Hot Leg Switchover IDHM Integrated Decay Heat Management LBLOCA Large Break Loss of Coolant Accident LAR License Amendment Request LOCA Loss of Coolant Accident LRA Locked Rotor Accident LPZ Low Population Zone MSLB Main Steam Line Break M&E Mass and Energy NCC Natural Circulation Cooldown NFSV New Fuel Storage Vault PCT Peak Clad Temperature RHR Residual Heat Removal RHRS Residual Heat Removal System SFP Spent Fuel Pool SSE Safe Shutdown Earthquake TEDE Total Effective Dose Equivalent

      • 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

4.10 Impact of COBA Capsule Breach on Reactor Coolant Chemistry A comparison has been made between the radioactive release (in Curies) from a COBA capsule and reference concentrations of radioactivity in the RCS. [

]a,c Given the low likelihood of the failure of an individual SS 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. If a failure were to occur, cobalt release would be gradual [

]a,c The plant chemistry program routinely checks for Co-60 in the RCS. Cobalt-60 is monitored periodically as part of routine RCS gamma isotopic analysis. Historical trends of Co-60 activity in the RCS have been evaluated. For both units, there have been no known events that could have had a significant impact on Co-60 RCS concentration. Thus, cobalt levels are considered to have been stable over recent years. This is consistent with not having events (such as a

      • 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

recent introduction of zinc or major component replacement) that could have had a significant impact on Co-60 concentration. Over the past 10 years, the Co-60 concentration has ranged from ~5.0E-5 to ~2.0E-4 Ci/mL (factor of 4) at Unit 1 and ~8.0E-5 to ~2.0E-4 Ci/mL (factor of 2.5) at Unit 2.

The level of RCS Co-60 concentration as presented in UFSAR Table 11.1-8 is 1.9E3 Ci/mL.

Based on the monitoring history of Co-60 levels, an equilibrium RCS Co-60 threshold value that would warrant further investigation is recommended to be [

]a,c the historic maximum equilibrium levels. Implementing the proposed threshold would not invalidate existing analyses that use the UFSAR value as the basis. Also, this threshold would allow for variation in Co-60 that might be expected due to other causes, without unnecessarily disrupting the COBA program.

A significant release of Co-60, although unexpected, would be detected, proceed slowly, and allow for actions to be taken including the discharge of the COBA inserts in the subsequent refueling outage.

Reference:

1. Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Oak Ridge National Laboratory, 1988.
      • 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

8.1 Design Parameters and Best Estimate Flows The impact on design core bypass and best estimate core bypass has been assessed based on the evaluation of the design and best estimate core bypass flows discussed in Section 5.1.

Although the best estimate flows increase slightly with the implementation of COBAs, the revised design and best estimate core bypass flows remain bounded by the previously reported core bypass flow limits. Accordingly, the implementation of COBA assemblies would not impact the current Design Parameters and Best Estimate Flows.

      • 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

8.3 Non-LOCA Transients For the planned use of COBA assemblies to produce commercial quantities of Cobalt-60 at Salem Units 1 and 2, the potential impacts of the COBA assemblies on the safety analyses of the non-LOCA transient events and the impacts of those events on the COBA assemblies were examined. Relative to the potential impacts of the COBA assemblies on the non-LOCA safety analyses (Table 8.3-1 lists the non-LOCA safety analysis events), the following parameters were evaluated:

Reload Safety Analysis Checklist (RSAC) parameters (inputs potentially impacted by a core reload design) such as core axial power shapes, peaking factors, axial offset limits, decay heat, etc.

Plant operating conditions such as temperatures, pressures, and flows (including core bypass flow and core pressure drop)

Reactor trip characteristics (rod drop time, rod position versus time curve, reactivity versus rod position)

Departure from nucleate boiling ratio (DNBR) results Further examination of these potential impacts concluded that only the decay heat input could be significantly impacted (see Section 8.2 in the Supplemental Information). However, it was determined that there is sufficient margin in the decay models used in the non-LOCA safety analyses to offset the added decay heat, and this will be reconfirmed for future core reloads with COBA assemblies. Based on the decay heat margin that exists, the use of COBA assemblies will have no significant impacts on the non-LOCA safety analyses.

Relative to the impacts of non-LOCA events on the COBA assemblies, a mostly bounding set of reactor core thermal power and coolant temperature and pressure conditions was first determined based on the results of the non-LOCA safety analyses. A single bounding case was developed which combined transient conditions, including:

A maximum overpower condition (from the limiting cool-down event)

Bulk coolant in the core at saturated conditions (indicative of a heat-up event)

Stagnant flow in the water gap between the COBA capsule and the outer tube (indicative of a loss-of-flow event)

With corresponding COBA rod heating rates, these conditions were then applied as boundary conditions in COBA rod temperature distribution calculations. Those calculations showed significant margins to the various material melting temperatures, which demonstrates the integrity of the COBA rods will be maintained. The only non-LOCA safety analysis described in Chapter 15 of the UFSAR that had results outside the bounds of reactor core conditions discussed above is that of the RCCA ejection event (UFSAR Section 15.4.7). A conservative assessment of the RCCA ejection event was performed to determine if the event could cause the cobalt slugs in the COBA rods to melt. It was determined that because the majority of the power increase that results from an RCCA ejection event occurs within the UO fuel pellets, the cobalt slug temperatures in an adjacent COBA rod would not increase sufficiently to reach the cobalt melting temperature. Therefore, this demonstrates the integrity of the COBA rods will be maintained during an RCCA ejection event.

      • 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

In summary, the COBA assemblies that are planned for insertion in the reactor cores will have no significant impacts on the non-LOCA safety analyses and the COBA rods are capable of withstanding the transient conditions associated with the non-LOCA safety analyses described in Chapter 15 of the UFSAR. Therefore, it is concluded that the use of COBA assemblies is acceptable with respect to the non-LOCA safety analyses.

      • 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

Table 8.3-1: Salem Units 1 and 2 Non-LOCA Safety Analysis Events UFSAR Section Event 15.2.1 Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal from a Subcritical Condition 15.2.2 Uncontrolled RCCA Bank Withdrawal at Power 15.2.3 RCCA Misalignment - Dropped RCCA(s)

RCCA Misalignment - Statically Misaligned RCCA 15.2.4 Uncontrolled Boron Dilution 15.2.5 Partial Loss of Forced Reactor Coolant Flow 15.2.6 Startup of an Inactive Reactor Coolant Loop 15.2.7 Loss of External Electrical Load and/or Turbine Trip 15.2.8 Loss of Normal Feedwater 15.2.9 Loss of Offsite Power to the Station Auxiliaries 15.2.10 Excessive Heat Removal Due to Feedwater System Malfunctions 15.2.11 Excessive Load Increase Incident 15.2.12 Accidental Depressurization of the Reactor Coolant System 15.2.13 Accidental Depressurization of the Main Steam System 15.2.14 Spurious Operation of the Safety Injection System at Power 15.3.2 Minor Secondary System Pipe Breaks 15.3.3 Inadvertent Loading of a Fuel Assembly into an Improper Position 15.3.4 Complete Loss of Forced Reactor Coolant Flow 15.3.5 Single RCCA Withdrawal at Full Power 15.4.2 Major Secondary System Pipe Rupture 15.4.3 Major Rupture of a Main Feedwater Line 15.4.5 Single Reactor Coolant Pump (RCP) Locked Rotor and RCP Shaft Break 15.4.7 Rupture of a Control Rod Drive Mechanism Housing (RCCA Ejection)

      • 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 8.4 Loss of RHR at Midloop / Natural Circulation Cooldown The most limiting condition for a loss of Residual Heat Removal (RHR) cooling occurs when the water level in the Reactor Coolant System (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.

Decay heat is an important input to the analysis, because decay heat directly affects how fast the reactor coolant will heat up. The implementation of COBAs will result in an increase in the decay heat produced by the reactor core. An increase in decay heat results in a shorter time for the core to heat up to saturation and boiling conditions. The current Loss of RHR at Midloop analysis has sufficient margin to account for the additional decay heat from COBAs; therefore, there is no impact to the Loss of RHR at Midloop.

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 RCPs. Following a loss of AC power, forced flow in the Reactor Coolant System is lost. The only method available for heat transfer in the RCS is via natural circulation, which results from density differences between the hot and cold reactor coolant. Decay heat is an important input to the analysis, because a higher decay heat will require more heat transfer and may slow the cooldown rate.

The current analysis has sufficient margin to account for the additional decay heat from COBAs; therefore, there is no impact to NCC.

Since the analyses of record for Loss of RHR at Midloop and NCC bounds the additional decay heat from COBAs, there is no impact to the licensing basis for these functional areas.

      • 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 8.5 FLEX Operation, RHR Cooldown, SFP Cooling and Boron Dilution The added decay heat from COBAs can impact a range of analyses including post-Fukushima Flexible Coping Strategies (FLEX) analyses that address an extended loss of AC power (ELAP),

shutdown cooling (RHR cooldown), spent fuel pool (SFP) cooling and heatup, and SFP boron dilution analyses that are input to SFP criticality. A generic impact evaluation was performed to determine the conservativism of various decay heat models relative to the addition of COBA decay heat. The result of the generic impact evaluation was a set of screening criteria to determine that the results of the related analyses bound the use of COBAs. The screening criteria are as follows:

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 COBA decay heat is recommended to account for model uncertainties.

FLEX analyses used the ANS-5.1-1979 = 2 model for all FLEX analyses; therefore, there is no impact to the FLEX analyses due to the addition of COBA.

Single train RHR cooldown represents the limiting analysis relative to added COBA decay heat.

When the analysis is reviewed against the screening criteria, it is concluded that the results are conservative due to the use of the BOP-FR-8 model and relevant cooldown times that are less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. A sensitivity study has been performed to assess the impact of COBA insertion.

The sensitivity of cooldown times to the added Cobalt 60 decay heat results in an increase of nominally 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the single-train cooldown cases examined, which represents an increase in cooldown duration of approximately 4-5%. Per UFSAR Section 5.5.7.3.4, in compliance with Branch Technical Position RSB 5-1, the safe shutdown design basis of the Salem units is Hot Standby, Mode 3. Consequently, there is no requirement to achieve single-train cooldown to cold shutdown in a specific time after shutdown based on compliance with the RSB 5-1 requirements. Assuming loss of all Nonseismic Category I equipment, the station can be maintained in a safe, hot standby condition while manual actions are taken to permit achievement of cold shutdown conditions following a safe shutdown earthquake (SSE) with loss of offsite power. Therefore, because it has been shown that the plant can be cooled to Mode 5 with existing equipment, the addition of COBA is acceptable but may prolong the time to reach a specific RCS temperature.

SFP cooling is assessed by the SFP outage management program for each outage to ensure that the SFP temperature can be maintained below specified limits (Section 9.4). SFP Time to Boil has been addressed generically for a bounding case which showed minimal impact (Attachment 2, Section 9.2).

      • 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 The results of the SFP boron dilution analysis show that dilution of the SFP below minimum boron requirements is not credible as any dilution event would be mitigated by operator action (Reference 8.5-1).

The results of the evaluations for FLEX and RHR cooldown demonstrate there is no effect on the FLEX mitigative actions with the addition of Cobalt 60 burnable absorbers. While single train RHR cooldown times determined for this effort are longer, the results show that the RCS can be cooled to 200°F (Mode 5 entry) over a period of time that is extended by approximately 4-5% due to the presence of COBA. The post-Fukushima FLEX analyses are conservative for the added COBA decay heat due to the use of a conservative decay heat model. Therefore, all design and licensing requirements for FLEX Operation, RHR Cooldown, SFP Cooling and Boron Dilution will continue to be met.

Reference 8.5-1 PSEG Nuclear LR-N25-0050, License Amendment Request to Update the Spent Fuel Pool and New Fuel Vault Criticality Safety Analysis, July 21, 2025.

      • 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 8.6.2 Small Break and Large Break LOCA Section 8.1 concludes that implementation of the COBA assemblies would not impact the current Design Parameters and Best Estimate Flows. Furthermore, there is negligible to no effect on the core bypass flow fraction, core pressure drop and the rod drop time. Therefore, no further assessment of design parameter data, pressure losses in the reactor and core bypass flow is needed.

COBA Decay Heat and Gamma Energy The core decay heat addition from COBAs is conservatively treated as a constant value based on the maximum decay heat addition for a single COBA assembly and the maximum number of COBA assemblies allowed in the core. The COBA decay heat does not originate in the fuel rods and therefore it does not directly contribute to the cladding heat up analyzed in the loss-of-coolant accident (LOCA) analysis. The additional energy produced in the core can impact the local fluid conditions and thus the ability to transfer heat from the fuel rods. However, the amount of decay heat produced by the COBAs is significantly lower than the decay heat produced by the fuel rods and would have negligible effect on the thermal hydraulic response calculations contained in the SBLOCA and LBLOCA analyses.

The decay heat modeled in the small break LOCA (SBLOCA) and large break LOCA (LBLOCA) analyses is prescribed by 10 CFR 50 Appendix K. Appendix K decay heat is based on the ANS 1971 standard and conservatively assumes infinite operating time and includes 20%

uncertainty. The COBA decay heat remains less than 0.5% of the fuel decay heat at the end of the LOCA transients and will be even less at the respective limiting peak cladding temperature (PCT) time. This relatively small heat addition to the core will have a negligible impact on the core average thermal hydraulic response and the ability to remove heat from the fuel, and no direct impact to fuel rod heat-up.

The local effects of a COBA insert located in the hot assembly are also considered at the limiting PCT time for the respective transients. Using a relative power in the hot assembly (PHA) consistent with the Reload Safety Analysis Checklist current limit, the maximum COBA decay heat in one assembly is still relatively small, less than 2% of the hot assembly decay heat for SBLOCA and less than 1.5% of the hot assembly decay heat for LBLOCA at the respective PCT times. As noted before, this additional decay heat does not directly heat the cladding, and is small enough that it would negligibly impact the ability to transfer heat from the cladding to the surrounding coolant. For radiation heat transfer, since the COBA inserts produce much less decay heat compared to fuel rods, the direction of heat transfer would be from the fuel rods to the COBAs, thus precluding additional fuel rod cladding heatup due to the COBA decay heat.

A SBLOCA is a slower, more stratified event compared to LBLOCA and the core average core mixture level is used to represent the mixture level for fuel rod heatup calculations. The SBLOCA analysis conservatively assumes steam only cooling in the uncovered portion of the core and rod to rod radiation is neglected. As discussed earlier, the core wide increase in decay heat is negligible and further examination of the local effects is not necessary. A LBLOCA event is more sensitive to local effects due to the shorter, more severe core uncovery, and higher decay heat and fuel rod heatup. The hot assembly quench front progression is critical to the fuel rod heatup transient. Considering the coarse resolution of translating the core inlet flooding rate to the fuel rod heatup calculation and the minimal potential contribution to overall hot

      • 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 assembly heat addition, the impact to the local fluid conditions will be negligible. Based on this, the COBA assemblies will have a negligible impact on both the core wide and local behavior modeled in the LBLOCA and SBLOCA analyses.

The COBA rodlets release gamma energy that will be absorbed by the fuel rods. The gamma energy deposited in the fuel from the decay of cobalt is at least an order of magnitude less than the fuel heat generation for SBLOCA and LBLOCA and will have a negligible impact on the SBLOCA and LBLOCA fuel rod cladding heat up.

COBA Coolability Examination of the temperature distribution across the COBA rodlet under LOCA conditions showed that the COBA rodlet outer tube temperature remains below the LOCA acceptance criterion for peak cladding temperature of 2200°F and the materials in the COBA rodlet are below the respective melting temperatures. Given these considerations, the demonstration of survivability of the [

]a,c capsule under LOCA conditions, and the relatively small decay heat from the COBA compared to the fuel rods, coolability of both the COBA rodlets and core will be maintained as demonstrated from the LOCA licensing basis analyses.

Conclusions The effects of the COBA assemblies were assessed for the large break and small break LOCA analyses. Based on the evaluation, the COBA assemblies will have a negligible impact on the LBLOCA and SBLOCA analyses, leading to a 0°F peak cladding temperature (PCT) impact for 10 CFR 50.46 reporting purposes. Coolability of both the COBA rodlets and core will be maintained.

      • 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 8.6.3 LOCA Containment Integrity In addition to the overall peak containment pressure and peak containment temperature, the long-term containment temperature response used to evaluate the equipment qualification (EQ) program has also been evaluated.

The current analyses support plant operation in the reactor vessel upflow configuration as described in WCAP-18707-P (Reference 8.6.3-1) and WCAP-18659-P (Reference 8.6.3-2).

The long-term LOCA mass and energy (M&E) releases and containment response analyses that are documented in Section 9.4 of WCAP-18659-P bound both units.

The inclusion of Co-60 burnable absorber rod assemblies will not increase the RCS initial temperatures, pressure, primary side fluid volume, secondary side fluid volume, secondary side metal mass, or the initial core stored energy. There would also be only a negligible increase in metal mass in the reactor vessel. There would be a slight increase in the decay heat energy release rate for a [

]a,c The additional decay heat energy release is accounted for by applying a conservative constant decay heat energy rate to the LOCA mass and energy releases used as the forcing function for the limiting containment response cases.

The limiting long-term LOCA containment transient is a double-ended pump suction (DEPS) break assuming minimum safeguards (Min SI). This break scenario considers two cases: one case with recirculation sprays operating, and a second case without recirculation sprays operating. These cases have been reanalyzed to address the impact for the additional energy associated with Co-60 burnable absorber rod assemblies in the core.

Table 8.6.3-1 provides a comparison of the limiting containment results with the Co-60 energy included to the analysis of record results from Table 9-26 of WCAP-18659-P. Based on these results, the peak containment pressure and temperature results remain unchanged from those reported in Table 9-26 of WCAP-18659-P. However, the additional energy will result in a slight increase in the containment pressure and steam temperature at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after accident initiation and an increase in the long-term containment temperature results used in the equipment qualification (EQ) program.

The long-term containment response results are provided in Table 9-36 of WCAP-18659-P for the DEPS Min SI case with recirculation sprays operable and in Table 9-37 without recirculation sprays in operation. A comparison of the AOR transient results to the Co-60 case with recirculation sprays operational and the Co-60 case without recirculation sprays was performed.

Due to the minor increase in the transient pressure and temperature, the transient figures provided in Section 9.4.2.5 of WCAP-18659-P remain valid.

The addition of the Co-60 decay heat energy results in an increase in the containment pressure and vapor temperature at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of 0.1 psi & 0.7°F and 0.4 psi & 1.2°F with recirculation sprays and without recirculation sprays, and a vapor temperature increase of 1.0°F and 2.1°F at approximately 120 days for EQ for the case with recirculation sprays and case without recirculation sprays, respectively.

      • 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 References 8.6.3-1 WCAP-18707-P, Revision 0, Reactor Internals Upflow Conversion Program Engineering Report For Salem Nuclear Generating Station Unit 1, May 2022.

8.6.3-2 WCAP-18659-P, Revision 0, Reactor Internals Upflow Conversion Program Engineering Report For Salem Nuclear Generating Station Unit 2, August 2021.

      • 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 8.6.3-1 Salem Units 1 and 2 LOCA Containment Response Results (Loss-of-Offsite Power Assumed)

Case Peak Pressure (psig)

Peak Temperature

(°F)

Pressure (psig) @

24 hrs Vapor Temperature

(°F) @ 24 hrs Vapor Temperature

(°F) @ ~120 days DEPS Min SI with Recirc Sprays (AOR) 45.8 @

799 sec 269.3 @ 799 sec 7.3 165.0 105.2 DEPS Min SI with Recirc Sprays (Co-

60) 45.8 @

799 sec 269.3 @ 799 sec 7.4 165.7 106.2 DEPS Min SI without Recirc Sprays (AOR) 45.8 @

799 sec 269.3 @ 799 sec 17.0 209.3 115.5 DEPS Min SI without Recirc Sprays (Co-

60) 45.8 @

799 sec 269.3 @ 799 sec 17.4 210.5 117.6 Containment Design Pressure = 47 psig Containment Atmosphere Design Temperature = 271°F

      • 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 8.6.4 Post-LOCA Long Term Cooling Section 8.1 concludes that the implementation of the COBA assemblies would not impact the current Design Parameters and Best Estimate Flows. The inclusion of COBAs does not impact the post-LOCA analysis inputs which are used to define the RCS fluid volume. Therefore, no further assessment of design parameter data and other post-LOCA LTC analysis inputs is needed.

The COBA rodlets release gamma energy that will be absorbed by the fuel rods. Section 8.6.2 concludes that the gamma energy deposited in the fuel from the COBA assemblies will have a negligible effect on the fuel rod cladding heating analyzed for the short term LOCA response. In comparison to the short term response, the post-LOCA LTC analyses consider core wide conditions, and the additional gamma energy deposited in the fuel from the decay of cobalt is negligible and not considered further.

Evaluation of Decay Heat The core decay heat addition from COBAs is conservatively treated as a constant value based on the maximum decay heat addition for a single COBA assembly and the maximum number of COBA assemblies allowed in the core. The additional decay heat from COBAs is addressed in the following subsections.

a. Post-LOCA Subcriticality Calculations The post-LOCA subcriticality calculations determine a mixed mean boron concentration in the sump. These calculations do not consider the impact of decay heat and therefore post-LOCA subcriticality is not impacted by the COBA inserts.
b. Boric Acid Precipitation Control (BAPC)

The BAPC analysis calculates the boric acid concentration in the core and determines an appropriate hot leg switchover (HLSO) time that will prevent the precipitation of boric acid in the core. Decay heat is an important input to this analysis because increased boiling of core fluid will increase the boric acid concentration in the core. The hot leg switchover (HLSO) time is conservatively reported as 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> (Unit 1) and 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (Unit 2) after the initiation of the LOCA. As time goes on, the relatively small amount of heat generated by the COBA inserts becomes more significant relative to the decay heat produced by the fuel rods.

The BAPC analyses assessed or explicitly modeled Appendix K decay heat. The decay heat from COBAs increases to just over 1% of the Appendix K decay heat by 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> after the LOCA. An increase in decay heat will result in a quicker build-up of boric acid in the core and therefore, a potentially earlier HLSO time. Assuming the increase in decay heat of 1% results in a 1% decrease in the calculated HLSO time, there is sufficient margin in the reported HLSO times such that the reported times remain conservative.

c. Decay Heat Removal The DHR calculations verify the post-LOCA minimum flow requirements that ensure that there is sufficient vessel inventory during cold leg and hot leg recirculation and preclude the potential for
      • 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 boron precipitation. The COBA decay heat represents only a fractional increase in the core boiloff rate relative to the analyzed boiloff rates, such that the minimum flow requirements continue to be met.

Evaluation of COBA Coolability Section 8.6.2 concludes that coolability of the COBA rodlets will be maintained as demonstrated from the large break and small break LOCA licensing basis analyses, based on the survivability of the [

]a,c capsule under LOCA conditions, consideration of the relatively small decay heat from COBA compared to the fuel rods, and the demonstration of the temperature distribution across the COBA rodlet under LOCA conditions. As discussed in Section 8.6.2 for the short term LOCA response, and applicable for the long term LOCA response, the COBA rodlet outer temperature remains below the LOCA acceptance criterion for peak cladding temperature of 2200°F and the COBA rodlet materials are below the respective melting temperatures under LOCA conditions. Based on this, coolability of the COBA rodlets and the core will be maintained as demonstrated from the post-LOCA LTC licensing basis analyses.

Conclusions The effects of the COBA assemblies were evaluated for the post-LOCA analyses, concluding that:

The COBA inserts will have no impact on the long term core cooling subcriticality calculations, The increased decay heat from the COBA inserts can be accommodated within the analysis margin for the BAPC and DHR calculations, and Long term coolability of the COBA rodlets and the core will be maintained.

      • 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 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. The plant response considered in the SGTR analyses is dominated by parameters such as break size, safety injection flow rates, decay heat, latent heat, and operator action timing. 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. The SGTR input to dose calculations provide mass release data considering removal of decay heat via steam releases from the intact steam generators. The SGTR dose analyses are based on intact steam generator steam releases continuing until the residual heat removal system (RHRS) cut-in time of 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />. The decay heat added by the COBA assemblies would not change the analysis decay heat modeling or the RHRS cut-in time. The decay heat added by the COBA assemblies is not significant for the 32-hour duration of the SGTR analysis and would not change the analysis modeling.

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 21 8.8 Steamline Break and Steam Release for Dose Steam Release for Dose Analyses The impact on the steam release for dose analysis of an expected maximum decay heat contribution from Co-60 production is evaluated (see Section 8.2 in the Supplemental Information). The evaluation calculates the additional steam releases as an energy balance using the specific enthalpy of the auxiliary feedwater (AFW) assuming the steam generators (SGs) are in equilibrium with the reactor coolant system (RCS) at the residual heat removal system (RHR) operating temperature.

From the Salem Unit 2 steam release analysis, the maximum AFW enthalpy is 91 Btu/lbm (corresponding to 120°F) and the RHR cut-in temperature is 350°F. The AFW enthalpy and RHR temperature in the Salem Unit 2 analysis are conservative and applicable for Salem Unit 1 as well. Therefore, this approach will be used to evaluate the impact on both units.

The steam released to remove fuel, RCS, and SG stored energy; core decay heat; and to account for changes in RCS and SG fluid masses and associated conditions would already be accounted for in the existing analyses. Therefore, the total energy to be removed in this evaluation is the additional decay heat from Co-60 production.

An energy balance for the intact steam generators gives:

[

]a,c Conservatively estimating the steam release at [

]a,c the release needed to remove the decay heat generated from Co-60 production can be compared to the overall steam release currently considered in the dose analyses. Since the RHR temperature is conservatively low compared to the RCS temperature at other times in the cooldown and the AFW enthalpy is conservatively high over the entire time range, the steam release due to decay heat is conservatively high over the duration of the releases.

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 22 Tables 8.8-1 and 8.8-2 provide the analysis of record (AOR) releases for the steam line break and locked rotor events and present the steam release due to decay heat from Co-60 production as both a value for each interval and a percentage increase of steam release.

Rod ejection releases are not provided because as shown in Table 15.4-12A of the UFSAR, the rod ejection releases are provided for only slightly more than 100 seconds. Therefore, the releases over this short duration would not be impacted by the small increase in decay heat from Co-60 production.

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 23 Table 8.8-1 Salem Steamline Break Releases Time (hour)

Steamline Break Release (lbm)

(UFSAR Table 15.4-7 Co-60 Decay Heat Release (lbm) 1 Increase In Steam Release due to Co-60 Production (% of UFSAR value)

From To Unit 1 Unit 2 Unit 1 Unit 2 0

2 500,000 431,800 2

8 452,000 2,648,000 8

32 2,008,000 Note 1:

The increased steam release from 2 to 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> for Unit 2 is [

]a,c Table 8.8-2 Salem Locked Rotor Releases Time (hour)

Locked Rotor Release (lbm) (UFSAR Table 15.4-6A)

Co-60 Decay Heat Release (lbm) 1 Increase In Steam Release due to Co-60 Production (% of UFSAR value)

From To Unit 1 Unit 2 Unit 1 Unit 2 0

2 655,000 715,700 2

8 540,000 3,148,900 8

32 2,400,000 Note 1:

The increased steam release from 2 to 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> for Unit 2 is [

]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 24 8.9 Radiological Consequences (Doses) of Accidents A post-accident release of Cobalt-60 from COBA fuel inserts is not expected based on the design details and analyses described in Section 8.6.1 of the Cobalt-60 Supplemental Information Attachment. As such, the contribution to the overall post-accident source term from COBA fuel inserts is driven by compliance with the recommendations of Regulatory Guide 1.183. 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 using a mechanistic approach based on the COBA capsule and cobalt slug design rather than applying the noble metal release fraction. A study was performed using both high-and low-pressure severe accident sequences to inform the resultant Cobalt-60 release fraction.

Potentially impacted dose consequence calculations were identified based on the bounding Cobalt-60 content of COBA fuel inserts as well as the small addition of decay heat [

]a,c resulting from the use of COBA fuel inserts. Dose consequence calculations were reassessed for the following UFSAR Chapter 15 events:

8.9.1 Post-LOCA EAB, LPZ And CR Doses - Alternative Source Term (AST)

Calculation S-C-ZZ-MDC-1945 determines post-Loss of Coolant Accident (LOCA) doses at the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR). Doses at these locations minimally increase due to higher reactor core Co-60 activity. The results are shown in Table 8.9-1 below.

8.9.2 EAB, LPZ, & CR Doses - Main Steam Line Break (MSLB) Accident - AST Calculation S-C-ZZ-MDC-1950 determines EAB, LPZ, and CR doses due to a Main Steam Line Break (MSLB) accident. Doses are minimally increased due to higher mass and energy release to the environment from the intact steam generators resulting from increased decay heat attributed to the activated COBA assemblies. The results are shown in Table 8.9-2 for Unit-1 and Table 8.9-3 for Unit-2 below.

8.9.3 EAB, LPZ, & CR Doses - RCP Locked Rotor Accident (LRA) - AST Calculation S-C-ZZ-MDC-1951 determines EAB, LPZ, and CR doses due to a Locked Rotor Accident (LRA). Doses are minimally increased due to higher mass and energy release to the environment from the intact steam generators resulting from increased decay heat attributed to the activated COBA assemblies. The results are shown in Table 8.9-4 below.

The increase in the above dose consequences as a result of Co-60 result in a less than 10%

decrease in available dose margin and, therefore, the change is acceptable.

      • 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 Table 8.9-1 Post-LOCA EAB, LPZ and Control Room Dose Consequence Results Location Regulatory Limit (Rem)

Original TEDE Dose Value (Rem)

New TEDE Dose Value (Rem)

Percent Change in Margin EAB 25 6.86 6.91 0.3%

LPZ 25 1.21 1.21 0%

Control Room 5

4.33 4.35 3%

Table 8.9-2 Unit-1 Main Steam Line Break EAB, LPZ and CR Dose Consequence Results Location Regulatory Limit (Rem)

Original Pre-Accident Iodine Spike -

TEDE Dose Value (Rem)

New Pre-Accident Iodine Spike TEDE Dose Value (Rem)

Percent Change in Margin Original Concurrent Iodine Spike TEDE Dose Value (Rem)

New Concurrent Iodine Spike TEDE Dose Value (Rem)

Percent Change in Margin EAB 25 / 2.5 0.127 0.127 0%

0.724 0.725 0%

LPZ 25 / 2.5 0.0224 0.0224 0%

0.151 0.151 0%

Control Room 5

0.204 0.204 0%

1.58 1.58 0%

Table 8.9-3 Unit-2 Main Steam Line Break EAB, LPZ and CR Dose Consequence Results Location Regulatory Limit (Rem)

Original Pre-Accident Iodine Spike -

TEDE Dose Value (Rem)

New Pre-Accident Iodine Spike TEDE Dose Value (Rem)

Percent Change in Margin Original Concurrent Iodine Spike TEDE Dose Value (Rem)

New Concurrent Iodine Spike TEDE Dose Value (Rem)

Percent Change in Margin EAB 25 / 2.5 0.127 0.127 0%

0.682 0.682 0%

LPZ 25 / 2.5 0.0225 0.0225 0%

0.142 0.142 0%

Control Room 5

0.215 0.215 0%

1.6 1.6 0%

      • 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 Table 8.9-4 Locked Rotor Accident Dose Consequence Results Location Regulatory Limit (Rem)

Original Unit-1 TEDE Dose Value (Rem)

New Unit-1 TEDE Dose Value (Rem)

Percent Change in Margin Original Unit-2 TEDE Dose Value (Rem)

New Unit-2 TEDE Dose Value (Rem)

Percent Change in Margin EAB 2.5 0.216 0.218 0.1%

0.201 0.202 0%

LPZ 2.5 0.0241 0.0241 0%

0.0227 0.0230 0%

Control Room 5

0.0845 0.0852 0%

0.0825 0.0831 0%

      • 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 9.1 Spent Fuel Pool Criticality The criticality safety analysis for the storage of fuel assemblies in the SFP and the new fuel storage vault (NFSV) has been updated to address the implementation of COBA fuel assemblies. The updated analysis is in agreement with NEI 12-16 Revision 4. It addresses the implementation of COBA fuel assemblies as well as several future station initiatives including a power uprate. Given the broad scope of changes covered in the updated analysis, it was submitted as a separate License Amendment Request (LAR) in July 2025 (Reference 9.1-1).

The major aspects of the updated criticality safety analysis that address the storage of COBA fuel assemblies in the SFP include the following:

  • The explicit modeling of irradiated COBA fuel assemblies. Once irradiated, the presence of the COBA in a fuel assembly changes the isotopic composition and associated axial profiles within the fuel assembly from what would be present in a non-COBA fuel assembly, impacting the criticality safety analysis.
  • The inclusion of COBA fuel assemblies in Criticality Fuel Designs (CFDs) within the analysis. Both pre-uprate fuel assemblies with COBA and post-uprate fuel assemblies with COBA were modeled as part of development of these CFDs.

There is no impact on NFSV criticality from COBA because the new COBA fuel assemblies and COBA fuel inserts are unirradiated. New COBA fuel inserts are received inserted in new fuel assemblies but, without irradiation of the fuel assembly, have no impact on the fuel assembly isotopic compositions and associated axial profiles. No credit was taken for the poison contribution of new COBA fuel inserts. This approach has precedent in that the same treatment was followed for new Wet Annular Burnable Absorbers (WABA), another form of fuel poison insert.

Details surrounding the inclusion of COBA fuel assemblies in the updated criticality safety analysis can be found in the LAR referenced below.

Reference 9.1-1 PSEG Nuclear LR-N25-0050, License Amendment Request to Update the Spent Fuel Pool and New Fuel Vault Criticality Safety Analysis, July 21, 2025 (ADAMS Accession Nos. ML25203A071/ ML25203A075).

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 28 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 spent fuel pool (SFP) at SGS due to the anticipated increase in the incident gamma energy flux on the SFP walls/floor resulting from:

a) Storage of COBA Fuel assemblies in the SFP (Figure 9.3-1 presents a sketch of an Integrated Fuel Assembly with COBAs.)

b) Harvesting activities associated with extracting the irradiated COBA rodlets/capsules from the COBA Fuel assemblies. (Figure 9.3-6 presents a rendition of the Cask Pit Area where harvesting activities are conducted.)

9.3.1 Acceptance Criteria Acceptance criteria outlined in NUREG/CR 6927 (Reference 9.3-1) 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.

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 9.3-2) 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 assumptions were utilized to establish the fuel management protocols in the SFP with the primary intent being to maintain the incremental incident gamma flux on concrete surfaces due to storage of a cluster of COBA Fuel assemblies within the specified criterion in NUREG/CR-6927 of < 1010 Mev/cm2-sec and < 1010 rads, such that nuclear heating and irradiation of concrete structures, respectively, need not be considered. The critical assumptions reflected in this evaluation are as follows:

A minimum initiation time for fuel transfer to the SFP after reactor shutdown of 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> based on prior operating experience, A minimum distance between the SFP racks to the SFP walls of 0.5 inches,

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 29 A minimum distance between the SFP racks to the SFP floor of 12 inches, and COBA rodlet harvesting activities not permitted during refueling / core-offload scenarios, and when the temperature of the SFP water exceeds 120oF.

The following three potential Cobalt-60 harvesting / temporary storage configuration scenarios were evaluated against the specified criterion in NUREG/CR-6927 of < 1010 Mev/cm2-sec, and, where the criterion is exceeded, to confirm that the associated maximum temperature in the concrete walls remained within the temperature limits set by Section A.4.1 of ACI 349-01 Appendix A:

1. Scenario 1 - A fully loaded COBA Rodlet Storage Basket (CRSB) containing [

]a,c COBA rodlets installed adjacent to the wall in the new fuel elevator recon basket located in the cask pit area (Figure 9.3-2)

2. Scenario 2 - A fully loaded source cage and/or a COBA rodlet located in the harvesting workstation in the cask pit (Figures 9.3-3 and 9.3-4)
3. Scenario 3 - 2 fully loaded rectangular source cage storage stands, each containing [

]a,c source cages, stored in the cask pit. (Figures 9.3-4 and 9.3-5) 9.3.3 Results of Assessment

1. Storage of COBA Fuel Assemblies in the SFP The evaluation concluded that the criterion of NUREG/CR-6927 will be met for the SFP wall concrete temperatures due to the introduction of COBA fuel assemblies in the SFP, provided COBA fuel assemblies are stored at a minimum distance from the SFP walls as specified in the evaluation. By meeting the placement requirement, the COBA fuel assemblies could remain in the SFP for the remaining life of the plant without exceeding the concrete temperature limits or irradiation limits of 1010 rads.

The evaluation also determined that the contribution of the COBA rodlets in the COBA fuel assemblies to the total flux at the SFP floor is essentially insignificant when compared to the NUREG/CR-6927 criterion of <1010 MeV/cm2-sec. The relatively low flux value from the COBA rodlets to the floor is attributed to:

the COBA rodlet self-shielding in the axial direction, the distance between the end of the active region of the COBA rodlet to the SFP floor, the shielding provided by the pool water and other structural materials present between the COBA rodlet and the SFP floor, and the fuel rod / COBA rodlet configuration where the active region of the fuel rods extend [

]a,c beyond the active region of the COBA rodlets which serves to reduce the contribution from adjacent COBA rodlets.

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 30

2. COBA rodlet harvesting / temporary storage prior to shipment The critical distances between the COBA sources and adjacent concrete surfaces relevant to the harvesting / temporary storage assessment associated with Scenarios 1, 2 and 3 are summarized below:

In the CRSB - Distance between centerline of the COBA Rodlet closest to the cask pit wall and the cask pit liner 14 In the CRSB - Distance between the bottom of the CRSB to the cask pit floor 30 In the harvesting workstation - Minimum distance between a single COBA rodlet in the harvesting workstation and the cask pit floor 76 In the harvesting workstation - Minimum distance between a single COBA rodlet in the harvesting workstation and the cask pit walls 70 In the harvesting workstation - Minimum distance from a loaded source cage in the harvesting workstation to the cask pit floor 76 In the harvesting workstation - Minimum distance from a loaded source cage in the harvesting workstation and the cask pit walls 70 On the source cage storage stand - Minimum distance between the two [

]a,c source cage storage stands stored side-by-side (each holds 28 source cages) and the cask pit walls The minimum distance of the rectangular source cage stand to the walls parallel to the wider side of the 2 source cage stands The minimum distance of the rectangular source stand to the walls parallel to the narrower sides of the 2 source cage stands 60 60 On the source cage storage stand - Distance between the table-top of the two

[

]a,c source cage storage stands stored side-by-side (each holds

[

]a,c source cages) and the cask pit floor 25.9 Table 9.3-1 presents the results of the assessment of the incident gamma energy flux for the various COBA source / concrete surface configurations described above.

The potential for concrete degradation due to irradiation during the harvesting process over the remaining life of the plant was evaluated and determined to remain within the irradiation limits of 1010 rads.

9.3.4 Conclusion Prior to movement of irradiated COBA fuel assemblies from the core to the SFP,

administrative controls will determine if the above evaluation remains bounding, or whether an outage-specific analysis needs to be developed to maintain the incremental incident gamma flux on concrete surfaces due to storage of a cluster of COBA Fuel assemblies to within the specified criterion in NUREG/CR-6927 of < 1010 Mev/cm2-sec and < 1010 rads,

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 31 such that nuclear heating and irradiation of concrete structures, respectively, need not be considered.

Administrative controls will ensure that COBA rodlet harvesting activities are not permitted during refueling / core-offload scenarios, and when the temperature of the SFP water exceeds 120oF.

References 9.3-1 NUREG/CR-6927, Primer on Durability of Nuclear Power Plant Reinforced Concrete Structures - A Review of Pertinent Factors February 2007.

9.3-2 Code Requirements for Nuclear Safety Related Concrete Structures ACI 349-01, Appendix A, Thermal Considerations, 2001

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 32 Table 9.3-1 Scenarios 1, 2 & 3 - Incident Gamma Flux vs Acceptance Criteria of < 1010 MeV/cm2-sec Source & Receptor location NUREG/CR-6927 Criteria met 1

Flux from the COBA rodlet storage basket (CRSB) to the cask pit walls N1 2

Flux from the CRSB to the cask pit floor Y

3 Flux from a single source cage in the harvesting workstation to the cask pit floor Y

4 Flux from a single source cage in the harvesting workstation to the cask pit walls Y

5 Flux from a single rodlet in the harvesting workstation to the cask pit floor Y

6 Flux from a single rodlet in the harvesting workstation to the cask pit walls Y

7 Flux from [

]a,c source cages on a source cage storage stand to the cask pit walls Y

8 Flux from [

]a,c source cages on a source cage storage stand to the cask pit floor N1

1. Gamma heating rates in the concrete as a function of depth (and width if needed),

were developed for source / receptor configurations 1 and 8. These gamma heating rates were used to establish the temperature profile in the concrete wall or floor to demonstrate that the local peak and bulk average temperatures in the concrete did not exceed the temperature limits set by Section A.4.1 of ACI 349-01 Appendix A, of 200oF and 150oF, respectively.

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 33 Figure 9.3-1 COBA Insert a,c

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 34 Figure 9.3-2 COBA Rodlet Storage Basket (CRSB) in Cask Pit a,c

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 35 Figure 9.3-3 Harvesting Workstation in Cask Pit a,c

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      • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)
      • This record was final approved on 09/24/2025 10:06:12. (This statement was added by the PRIME system upon its validation)
      • 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 9.4 Spent Fuel Pool Cooling The added Cobalt-60 decay heat from COBA fuel inserts is nominally [

]a,c as documented in Section 8.2 of the Cobalt-60 Supplemental Information document. For the most recent Salem Unit 1 and 2 refueling outages, the decay heat input from fuel assemblies at the start of core offload was in the range of 10,000 to 11,000 kW (References 9.4-1 and 9.4-2). The additional [

]a,c from Cobalt-60 decay is approximately [

]a,c of the fuel assembly decay heat.

For each core offload, a decay heat calculation is performed for the specific core being offloaded. PSEG Design Engineering performs SFP heat-up calculations using this decay heat input to verify compliance with the Spent Fuel Pool Integrated Decay Heat Management (IDHM)

Program described in Section 9.1.3.2 of the Salem UFSAR. When COBA is implemented, the decay heat input provided to Design Engineering by Nuclear Fuels will include a [

]a,c fixed term for the nominal Cobalt-60 contribution from COBA. In addition to verifying compliance with the IDHM Program, Design Engineering provides heat-up curves to Operations.

The calculations generating these curves use the same decay heat input provided by Nuclear Fuels for verifying compliance with the IDHM Program and thus include the Cobalt-60 contribution from COBA.

References 9.4-1 DS1.6-0684 - Salem Unit 1 Refueling 30 Decay Heat Calculations - January 2025 9.4-2 DS2.6-0521 - Salem Unit 2 Refueling 27 Decay Heat Calculations - August 2024

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Westinghouse Non-Proprietary Class 3 LR-N25-0074 LAR S25-01 40 13 First Cycle Post-Irradiation Examination A Post-Irradiation Examination (PIE) following the first COBA irradiation cycle may be performed at the discretion of PSEG and Westinghouse. If conducted, the examination would consist of activity measurements for comparison with ANC9 activation model predictions and visual inspections of the rodlets and capsules, performed on a single COBA that has been harvested.

These activities could provide confirmatory data but are not required for validation of the technology.

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