ML25268A072
| ML25268A072 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 09/24/2025 |
| From: | Sharbaugh D Public Service Enterprise Group |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| Shared Package | |
| ML25268A071 | List: |
| References | |
| LR-N25-0074, LAR S25-01 | |
| Download: ML25268A072 (1) | |
Text
Attachments 2 and 3 contain Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachments 2 and 3 this document is decontrolled.
David Sharbaugh Site Vice President - Salem Generating Station - PSEG Nuclear PO Box 236 Hancocks Bridge, New Jersey 08038-0221 david.sharbaugh@pseg.com Proprietary Information - Withhold from Public Disclosure Under 10 CFR 2.390 10 CFR 50.90 LR-N25-0074 LAR S25-01 September 24, 2025 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311
Subject:
License Amendment Request to Modify Salem Generating Station Renewed Facility Operating Licenses in Support of Cobalt-60 Production In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to the Renewed Facility Operating Licenses for Salem Generating Station (Salem) Unit 1 and Unit 2.
The proposed change modifies License Condition 2.B.(6) and creates new License Conditions 1.K (Unit 1) and 1.L (Unit 2), 2.B.(7), and 2.C.(23) (Unit 1) and 2.C.(38) (Unit 2) as part of a program to irradiate Cobalt-59 (Co-59) targets to produce Cobalt-60 (Co-60).
PSEG is collaborating with Westinghouse Electric Company LLC (Westinghouse) and Nordion (Canada) Inc. to develop and implement a pilot program for producing Co-60 in the Salem reactors during power operation. The Co-60 is intended for industrial and medical purposes.
These Co-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. PSEG plans to initially load eight Cobalt Burnable Absorbers (COBAs) in host fuel assemblies during the Salem Unit-1, Fall 2026 refueling outage.
The Enclosure to this letter provides an evaluation supporting the proposed changes. The marked-up Facility Operating Licenses, with the proposed changes indicated, are provided in. Attachment 2 to this letter provides Westinghouse technical document Cobalt-60 Supplemental Information to Support the Use of Cobalt Burnable Absorber (COBA) Assemblies in Pressurized Water Reactors. Attachment 3 to this letter provides additional technical information Site Specific Information in Support of Cobalt Burnable Absorber (COBA)
Assemblies in Salem Generating Station Units 1 and 2. Attachments 2 and 3 contain Westinghouse and Nordion proprietary information, which is identified by bracketed and denoted text. Westinghouse and Nordion request that the proprietary information in Attachments 2 and 3 be withheld from public disclosure, in accordance with the requirements of 10 CFR 2.390.
o PSEG I NUCLEAR
LR-N25-0074 Page 2 10 CFR 50.90 Affidavits supporting this request are included as Attachment 4. Attachments 5 and 6 to this letter provide nonpropnetary versions of Attachments 2 and 3, respectively PSEG concludes that the proposed change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50 92(c), and, accordingly, a finding of "no significant hazards consideration" Is Justified PSEG requests approval of this LAR by September 30, 2026 in order to support loading of COBA assemblies during the Urnt-1 Fall 2026 refueling outage Once approved, the amendment will be implemented w1th1n 60 days from the date of issuance There are no regulatory commitments contained in this letter In accordance with 10 CFR 50.91, a copy of this application, with attachments, Is being provided to the designated State of New Jersey Off1c1al If there are any questions or if additional information is needed, please contact Mr. Shane Jurek at shane.Jurek@pseg.com.
I declare under penalty of perjury that the foregoing Is true and correct Executed on __ 1-'-+/2_:2-_~-1--/2=1:,=-~'-------
(Dfej Respectfully, D~
David Sharbaugh Site Vice President Salem Generating Station
LR-N25-0074 10 CFR 50.90 Page 3
Enclosure:
Evaluation of the Proposed Changes : Markups of Facility Operating License : Westinghouse Supplemental Information Document, "Cobalt-60 Supplemental Information to Support the Use of Cobalt Burnable Absorber (COBA) Assemblies in Pressurized Water Reactors, - Proprietary : Site Specific Information in Support of Cobalt Burnable Absorber (COBA)
Assemblies in Salem Generating Station Units 1 and 2 - Proprietary : Affidavits Supporting Withholding from Public Disclosure, in Accordance with 10 CFR 2.390 : Westinghouse Supplemental Information Document, "Cobalt-60 Supplemental Information to Support the Use of Cobalt Burnable Absorber (COBA) Assemblies in Pressurized Water Reactors, - Non-Proprietary : Site Specific Information in Support of Cobalt Burnable Absorber (COBA)
Assemblies in Salem Generating Station Units 1 and 2 - Non-Proprietary cc:
Administrator, Region 1, NRC NRC Project Manager, Salem NRC Senior Resident Inspector, Salem Manager, NJ Bureau of Nuclear Engineering PSEG Commitment Tracking Coordinator
LR-N25-0074 LAR S25-01 1
Enclosure License Amendment Request to Modify Salem Generating Station Renewed Facility Operating Licenses in Support of Cobalt-60 Production Table of Contents 1
SUMMARY
DESCRIPTION................................................................................................ 2 2
PROPOSED CHANGE....................................................................................................... 2 3
TECHNICAL EVALUATION................................................................................................ 3 3.1 Background and General Approach............................................................................ 3 3.2 Plant Implementation.................................................................................................. 4 3.3 Detailed Analytical Evaluations................................................................................... 4 4
REGULATORY EVALUATION............................................................................................ 6 4.1 Applicable Regulatory Requirements/Criteria.............................................................. 6 4.2 Precedent................................................................................................................... 7 4.3 No Significant Hazards Consideration........................................................................ 8 4.4 Conclusions...............................................................................................................11 5
ENVIRONMENTAL CONSIDERATION............................................................................ 111
- Markups of Facility Operating License : Westinghouse Supplemental Information Document, "Cobalt-60 Supplemental Information to Support the Use of Cobalt Burnable Absorber (COBA) Assemblies in Pressurized Water Reactors, - Proprietary : Site Specific Information in Support of Cobalt Burnable Absorber (COBA)
Assemblies in Salem Generating Station Units 1 and 2 - Proprietary : Affidavits Supporting Withholding from Public Disclosure, in Accordance with 10CFR 2.390 : Westinghouse Supplemental Information Document, "Cobalt-60 Supplemental Information to Support the Use of Cobalt Burnable Absorber (COBA) Assemblies in Pressurized Water Reactors, - Non-Proprietary : Site Specific Information in Support of Cobalt Burnable Absorber (COBA)
Assemblies in Salem Generating Station Units 1 and 2 - Non-Proprietary
LR-N25-0074 LAR S25-01 2
1
SUMMARY
DESCRIPTION In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to Renewed Facility Operating License for Salem Generating Station Unit 1 and Unit 2 (Salem).
The proposed change modifies License Condition 2.B.(6) and creates new License Conditions 1.K (Unit 1) and 1.L (Unit 2), 2.B.(7) (Units 1 and 2), and 2.C.(23) (Unit 1) and 2.C.(38) (Unit 2) as part of a program to irradiate Cobalt-59 (Co-59) targets to produce Cobalt-60 (Co-60). End-use customers primarily use the Co-60 to sterilize a wide range of 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. Recent activities affecting the supply of Co-60 have resulted in increased demand and constrained availability of Co-60 isotope 2
PROPOSED CHANGE PSEG proposes to add a new License Condition 1.K (Unit 1) and License Condition 1.L (Unit-2) which states, "The receipt, production, possession, transfer, and use of Cobalt-60 as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Part 30."
This new license condition allows for the production and transfer of Co-60 in accordance with 10 CFR Part 30.
Condition 2.B.(6) of the current Operating License for Salem is revised to state, "PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
Mechanical disassembly of the Westinghouse Cobalt Burnable Absorber (COBA) Assemblies containing Cobalt-60 is not considered separation." This change is intended to provide clarification of the term "separation" relative to the removal from the COBAs of enclosed capsules containing Co-60 after discharge from a Salem reactor core.
PSEG proposes to add a new License Condition 2.B.(7) which states, "PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60." This new License Condition supports the project at Salem by allowing intentional production of Co-60 during operation of the Salem facility.
PSEG proposes to add a new License Condition 2.C.(23) (Unit 1) and 2.C.(38) (Unit 2) which states, Prior to off-load of fuel assemblies containing irradiated COBA fuel inserts, verify that the incident energy flux on the SFP walls is less than 1010 MeV/cm2-sec per NUREG/CR-6927 with the associated fuel assemblies placed at the targeted SFP rack locations. This new License Condition will address concerns regarding the potential effect on Spent Fuel Pool wall integrity due to gamma heating effects from the irradiated COBAs.
Markups of the affected Salem Facility Operating Licenses showing the proposed changes are provided in Attachment 1. The proposed OL changes are also described below.
LR-N25-0074 LAR S25-01 3
No. Change Justification 1
New License Condition is proposed, 1.K (Unit 1) and 1.L (Unit 2) which states, "The receipt, production, possession, transfer, and use of Cobalt-60 as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Part 30."
This new license condition allows for the production and transfer of Cobalt-60 in accordance with 10 CFR Part 30.
2 Modification of Condition 2.B.(6) is proposed to state, "PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. Mechanical disassembly of the Westinghouse Cobalt Burnable Absorber (COBA) Assemblies containing Cobalt-60 is not considered separation."
This change is intended to provide clarification of the term "separation" relative to the removal from the COBAs of enclosed capsules containing Co-60 after discharge from a Salem reactor core.
3 New License Condition is proposed, License Condition 2.B.(7), which states, " PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60.
This new License Condition supports the project at Salem by allowing intentional production of Co-60 during operation of the Salem facility.
4 New License Condition is proposed, 2.C.(23)
(Unit 1) and 2.C.(38) (Unit 2), which states, Prior to off-load of fuel assemblies containing irradiated COBA fuel inserts, verify that the incident energy flux on the SFP walls is less than 1010 MeV/cm2-sec per NUREG/CR-6927 with the associated fuel assemblies placed at the targeted SFP rack locations.
This new License Condition will address concerns regarding the potential effect on Spent Fuel Pool wall integrity due to gamma heating effects from the irradiated COBAs.
3 TECHNICAL EVALUATION 3.1 Background and General Approach Westinghouse and Nordion (Canada) have developed an innovative isotope production technology, and with their partner utilities, plan to produce Co-60 in Westinghouse-designed Pressurized Water Reactors (PWRs). Nordion is a major supplier of Co-60 sources for industrial and medical purposes. These Co-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.
Co-60 is produced by subjecting Co-59 material to neutron irradiation in a nuclear reactor core.
Nickel-plated slugs of Co-59 target material would be provided to Westinghouse by Nordion.
The target material is placed in metal capsules and loaded into fuel component assemblies, called Cobalt Burnable Absorber (COBA) assemblies, at the Westinghouse Columbia Fuel Fabrication Facility.
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
LR-N25-0074 LAR S25-01 4
components such as wet annular burnable absorbers (WABAs)) where they are irradiated for multiple fuel cycles. When subjected to the neutron flux in the core, the Co-59 is converted to Co-60.
After removal from the core, the now-activated capsules are harvested (removed from the COBA rodlets), loaded into a Nordion-supplied, Canadian Nuclear Safety Commission (CNSC)-
certified, and U.S. Department of Transportation-approved transport package, and transferred to Nordions Canada facility. At Nordion, the Co-60 slugs are removed from the Westinghouse irradiated capsules, and are blended with Co-60 slugs from other suppliers so that Nordion can create finished sources of specific activity levels to meet various end-use specifications.
Nordion then uses the Co-60 slugs to manufacture sealed sources for distribution to their customers for sterilization and medical end-use applications.
3.2 Plant Implementation PSEG plans to load eight COBA assemblies into the Salem Unit 1 reactor core as part of Reload 31 during the Fall 2026 refueling outage. Each COBA is inserted into a fresh fuel assembly and loaded into the reactor core with its host fuel assembly. The COBAs are inserted into the fuel assemblies much like any other burnable absorber, such as a WABA. The fuel assembly itself is not modified. This process will continue at each subsequent refueling outage with irradiated COBAs shuffled into fresh fuel assemblies and additional unirradiated COBAs loaded into fresh fuel assemblies. After the requisite number of activation cycles in the core, the COBAs will be moved to the cask pit for harvesting of the Co-60-containing capsules.
During harvesting, each rodlet will be removed from the COBA baseplate and each capsule will have its activity measured. 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.
3.3 Detailed Analytical Evaluations (nonproprietary Attachment 5) documents the evaluations and analyses on a common basis that are applicable to Westinghouse designed PWRs. As such, there are additional details of these evaluations and analyses that are specific to each individual plant site. These details are provided in Attachment 3 (nonproprietary Attachment 6). Details of the evaluations and analyses are provided in the following sections of Attachments 2 and 3:
Topic Discussion / Conclusion Supplemental Information --
Site Specific Information --
Co-60 Production Process An overview of the design of the COBA assembly and the process for producing Co-60 is provided.
Sections 1 & 2 COBA Assembly Design and Fabrication A detailed description of the design and COBA fabrication process is provided demonstrating the application of the Co-60 production process.
Section 3
LR-N25-0074 LAR S25-01 5
Topic Discussion / Conclusion Supplemental Information --
Site Specific Information --
Mechanical Design Evaluation including stress analysis, testing, operational risk, and coolant monitoring The structural acceptability of the COBA assembly and the capsules containing the cobalt targets is demonstrated by stress and creep analyses, production leak testing, prototype wear testing and ISO 2919 testing. Assessment of operational risk, and RCS chemistry impact also provides additional confirmation of capsule integrity.
Section 4 Section 4 Thermal and Hydraulic Evaluation Thermal hydraulic analyses and evaluations conclude that the core components consisting of the COBA assemblies and capsules meet the thermal-hydraulic design criteria.
Thermal hydraulic analysis and benchmarking demonstrates that the Computational Fluid Dynamics methodology for the COBA rodlet thermal hydraulic analyses is appropriate.
Section 5 Materials Materials of construction of the COBA components are described and demonstrated to be appropriate for the Co-60 process.
Section 6 Nuclear Design Nuclear design analysis methodology demonstrates applicability of current design methods for use of COBAs.
Section 7 Safety Analyses Evaluation of design parameters and results of safety analyses demonstrate acceptability of COBAs for normal operation and accident conditions.
Section 8 Section 8 Effects on Spent Fuel Pool Impact of irradiated COBA assemblies on SFP parameters is shown to be acceptable.
Section 9 Section 9 Harvesting Harvesting process is described including the tooling, software, and waste disposal necessary for extraction and transfer of the Co-60 material.
Section 10
LR-N25-0074 LAR S25-01 6
Topic Discussion / Conclusion Supplemental Information --
Site Specific Information --
Co-60 Activity Measurement Cobalt activity measurement hardware and process activities are described that assure adherence to the licensed shipping requirements.
Section 11 Byproduct Material Tracking and Reporting Process of byproduct material control is presented, demonstrating that the applicable regulatory requirements are satisfied.
Section 12 First Cycle Post-Irradiation Examination Examinations may be performed to provide confirmatory data relative to COBA performance.
NA Section 13
- No additional discussion is provided in the Site-Specific Attachment 4
REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine that applicable regulations and requirements continue to be met.
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.
Therefore, the requirement for a Part 30 byproduct material license is addressed by amendment to the existing Part 50 Facility Operating License.
To implement Part 30 into the Salem Units 1 and 2 Operating License, the following OL changes will be made.
Proposed License Conditions 1.K (Unit 1) and 1.L (Unit 2) allow for the production and transfer of Co-60, Modification of Condition 2.B.(6) provides clarification of the term "separation" relative to the removal from the COBAs of enclosed capsules containing Co-60, Proposed License Condition 2.B.(7) allows intentional production of Co-60 during operation, and Proposed License Conditions 2.C.(23) (Unit 1) and 2.C.(38) (Unit 2) address concerns regarding the potential effect on Spent Fuel Pool wall integrity due to gamma heating.
LR-N25-0074 LAR S25-01 7
Salem was designed in accordance with Atomic Energy Commission (AEC) proposed General Design Criteria published in July 1967. The applicable AEC proposed criteria, as documented in Salem Updated Final Safety Analysis Report (UFSAR) Section 3.1, were compared to 10 CFR 50 Appendix A General Design Criteria (GDC) deemed applicable to the use of COBA assemblies. The proposed amendment does not affect Salems conformance with any General Design Criteria or Regulatory Guide.
The Co-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.
Although the byproduct material sources (Co-60) are produced in sealed metal capsules, 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, 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, and are also not 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. PSEG has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the facility operating license, and do not affect conformance with any GDC differently than described in the UFSAR.
4.2 Precedent This license amendment request is similar in nature to the following license amendments previously approved by the NRC to allow production of byproduct material at the Clinton Power Station, Hope Creek Generating Station, Watts Bar Nuclear Plant, and Sequoyah Nuclear Plant.
- a.
Letter from C. S. Goodwin (U.S. Nuclear Regulatory Commission) to C. G. Pardee (Exelon Nuclear), "Clinton Power Station, Unit No. 1 - Issuance of Amendment Re: Request to Modify Facility Operating License In Support of the Use of Isotope Test Assemblies (TAC No. ME1643)," dated January 15, 2010 (ADAMS Accession No. ML100200005)
- b.
Letter from R. B. Ennis (U.S. Nuclear Regulatory Commission) to T. Joyce (PSEG Nuclear LLC), Hope Creek Generating Station - Issuance Of Amendment RE: Use of Isotope Test Assemblies For Cobalt-60 Production (TAC No. ME2949), dated October 7, 2010 (ADAMS Accession No. ML102700271)
The Co-60 production process utilized in the Clinton and Hope Creek plants was similar to that being proposed for Salem. Co-59 targets were inserted into the reactor core, exposed to neutron flux during reactor operation which converts the Co-59 into Co-60, and then are removed from the core to be used for medical and industrial purposes. However, there are several differences between the previous process used in the boiling water reactors (BWRs) and the process being proposed for the Salem PWR plant.
LR-N25-0074 LAR S25-01 8
In the BWR application, several fuel rods in selected fuel bundles were replaced with isotope rods containing cobalt targets, which necessitates a change to the fuel design.
For the Salem application, the cobalt is inserted into the core as a fuel assembly component. There is no change to the fuel assembly design.
In a BWR, there is a direct path from the reactor core outside containment during normal operation. In the PWR environment, the reactor coolant loop is a closed system inside the containment.
With the BWR approach, the isotope rods are removed from the fuel bundle following irradiation and replaced with a mechanically equivalent rod. In the PWR process, fuel reconstitution is not required. The COBA is removed from the fuel assembly like any other fuel assembly component.
- c. Letter from R. E. Martin, (U.S. Nuclear Regulatory Commission) to O. D. Kingsley (Tennessee Valley Authority), Issuance of Amendment On Tritium Producing Burnable Absorber Rod Lead Test Assemblies (TAC No. M98615), dated September 15, 1997 (ADAMS Accession No. ML020780128).
- d. Letter from Mr. R. W. Hernan (NRC) to Mr. J. A. Scalice (TVA), "Issuance of Amendments Regarding Technical Specification Change No. 00-06 (TAC No. MB2972 and MB2973),"
dated September 30,2002 (ADAMS Accession No. ML022770566)
Both the Co-60 process and tritium production process utilize fuel assembly inserts that are inserted into the fuel assembly like a discrete burnable absorber or other core component. The primary difference between the Co-60 production process that will be used at Salem and the tritium production at Watts Bar and Sequoyah is that tritium is a volatile substance that can result in both liquid and gaseous effluent releases. By contrast, Co-60 is grouped with noble metals that would be treated as nonvolatile particulates which would not be released from the water surrounding the fuel and would not become airborne.
4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," PSEG Nuclear LLC (PSEG) requests an amendment to Renewed Facility Operating License No. DPR-70 for Salem Generating Station, Unit 1, and Renewed Facility Operating License No. DPR-75 for Salem Generating Station, Unit 2 (Salem). Specifically, the proposed changes modify the Salem Facility Operating Licenses to permit the production of Cobalt-60 (Co-60) by the neutron irradiation of Cobalt-59 (Co-59) in Cobalt Burnable Absorber (COBA) assemblies in the Salem reactors during power operation. The Co-60 would ultimately be used to produce sealed sources for sterilization of medical products and other applications.
PSEG has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
LR-N25-0074 LAR S25-01 9
- 1) Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the license conditions do not impact plant operation in any manner. The handling of byproduct material (i.e., Co-60) will be done in accordance with the requirements of 10 CFR 30 and the requirements of the Salem Facility Operating Licenses. The use of the cobalt burnable absorbers (COBAs) has been evaluated for impact on the previously evaluated transients and design basis accidents for Salem. (nonproprietary Attachment 5), "Cobalt-60 Supplemental Information to Support the Use of Cobalt Burnable Absorber (COBA) Assemblies in Pressurized Water Reactors and Attachment 3 (nonproprietary Attachment 6), Site Specific Information in Support of Cobalt Burnable Absorber (COBA) Assemblies in Salem Generating Station Units 1 and 2, document the results of the analyses and evaluations completed to demonstrate the impact on operation following introduction of the COBAs in the Salem cores. During power operation, any impact on core thermal limits would be minor and within the variations typically seen on a cycle-to-cycle basis. The use of the COBAs does not adversely affect accident initiators, design assumptions, or the manner in which the plant is operated and maintained. The COBAs do not adversely affect the ability of any structures, systems or components (SSCs) to perform their intended safety function to mitigate the consequences of an accident within the applicable regulatory limits.
PSEG has also evaluated the effects of the COBAs on post-irradiation conditions. The additional heat from the Co-60 decay, although small when compared to the total heat from a normal refueling discharge, has been evaluated for impact on transient analyses and accident conditions, as well as the effect on the spent fuel pool (SFP). The small amount of extra heat added by the Co-60 poses minimal risk for the transient or accident analyses or of SFP local boiling over that previously analyzed. The maximum incident radiation due to a fuel assembly containing an irradiated COBA insert has also been evaluated. Fuel assemblies containing irradiated COBA assemblies will be placed such that the incident energy flux on the SFP walls will not exceed the limit in NUREG/CR-6927, Primer on Durability of Nuclear Power Plant Reinforced Concrete Structures - A Review of Pertinent Factors. However, analysis has demonstrated that the energy deposition rate is well below that required to cause significant concrete heating. Salem procedures exist to guide placement of irradiated fuel assemblies in the SFP to avoid gamma heating of the wall concrete. These procedures will be modified to specify that fuel assemblies containing irradiated COBAs must be stored such that the limits of NUREG/CR-6927 will continue to be met with the placement of COBAs in the SFP, with no limitation on the amount of time a fuel assembly containing an irradiated COBA can be stored in the SFP.
The consequences of previously evaluated accidents are not significantly increased by the proposed change. As documented in Attachments 2 and 3, the proposed change does not affect the performance of any equipment credited to mitigate the radiological consequences of an accident.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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- 2) Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The COBA assembly does not require any modifications to the fuel lattice design with uranium loading enrichment remaining within the plants licensed limit. These modifications will not result in operation of the facility in a different way than currently operated.
The implementation of COBA does not result in a significant perturbation of the core design.
The same feed fuel batch size can be maintained with the uranium enrichment remaining within the plants licensed limit. The modeling of the reactor core uses the same licensed computer codes documented in the Salem licensing basis. Impacts on control rod worth, shutdown margin and trip reactivity are expected to be small and within the typical cycle to cycle variations. Also, use of the COBAs will not alter the design configuration or method of operation of plant equipment beyond its normal functional capabilities. The Co-60 program does not create any new credible failure mechanisms, malfunctions or accident initiators.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3) Do the proposed changes involve a significant reduction in a margin of safety?
Response: No.
These proposed changes will not affect the design or operation of any equipment important to safety. In addition, the proposed changes to the license conditions do not affect the results of any safety calculations.
Analyses and evaluations documented in Attachments 2 and 3 demonstrate that the acceptance criteria for the safety analyses are not affected and will continue to be met.
Cycle specific analyses with the COBAs will be performed using the same approved methodologies to assure that the fuel operating requirements remain in compliance with regulatory limits. COBA implementation does not result in any significant change to the reload core design with the uranium enrichment remaining within the plants licensed limit.
Therefore, the standard reload safety evaluation methodology still applies to core designs implementing COBA. There are no changes being made to safety limits or safety system allowable values that would adversely affect plant safety as a result of the use of COBAs.
The performance of the systems important to safety is not affected. The proposed change does not adversely affect safety analysis assumptions, initial conditions, or acceptance criteria and therefore, the margin of safety in the original safety analyses is maintained.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above evaluation, PSEG concludes that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.
LR-N25-0074 LAR S25-01 11 4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
5 ENVIRONMENTAL CONSIDERATION PSEG has evaluated the proposed amendments for environmental considerations. The review has resulted in the determination that the proposed amendments would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendments do not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendments.
LR-N25-0074 LAR S25-01 Mark-up of Proposed Facility Operating License and Technical Specification Pages The following Technical Specifications page for Renewed Facility Operating License DPR-70 is affected by this change request:
Facility Operating License Page 1.K 2
2.B.(6) 3 2.B.(7) 3 2.C.(23) 9 Technical Specification Page The following Technical Specifications page for Renewed Facility Operating License DPR-75 is affected by this change request:
Facility Operating License Page 1.L 2
2.B.(6) 3 2.B.(7) 3 2.C.(38) 10 Renewed License No. DPR-70 Amendment No. 341 H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of Renewed Facility Operating License No. DPR-70, subject to the conditions for protection of the environment set forth in the Technical Specifications, Appendix B is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied; I.
The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this renewed license will be in accordance with the Commissions regulations in 10 CFR Parts 30, 40, and 70 including 10 CFR Sections 30.33, 40.32, and 70.23 and 70.31; and J. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1), and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized by this renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facilitys current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commissions regulations.
- 2. Renewed Facility Operating License No. DPR-70 is hereby issued to PSEG Nuclear LLC and Constellation Energy Generation, LLC (the licensees) to read as follows:
A. This renewed license applies to the Salem Nuclear Generating Station, Unit No. 1, a pressurized water nuclear reactor and associated equipment (the facility), owned by PSEG Nuclear LLC and Constellation Energy Generation, LLC and operated by PSEG Nuclear LLC. The facility is located on the applicants site in Salem County, New Jersey, on the southern end of Artificial Island on the east bank of the Delaware River in Lower Alloways Creek Township, and is described in the Final Safety Analysis Report as supplemented and amended and the Environmental Report as supplemented and amended.
B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:
(1)
PSEG Nuclear LLC and Constellation Energy Generation, LLC to possess the facility at the designated location in Salem County, New Jersey, in accordance with the procedures and limitations set forth in this renewed license; (2)
PSEG Nuclear LLC, pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, to possess, use and operate the facility; (3)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; K. The receipt, production, possession, transfer, and use of Cobalt-60 as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Part 30.
Renewed License No. DPR-70 Amendment No. 348 (4)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at a steady state reactor core power level not in excess of 3459 megawatts (one hundred percent of rated core power).
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 348, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications, and the Environmental Protection Plan.
(3)
Deleted Per Amendment 22, 11-20-79 (4)
Less than Four Loop Operation PSEG Nuclear LLC shall not operate the reactor at power levels above P-7 (as defined in Table 3.3-1 of Specification 3.3.1.1 of Appendix A to this renewed license) with less than four (4) reactor coolant loops in operation until safety analyses for less than four loop operation have been submitted by the licensees and approval for less than four loop operation at power levels above P-7 has been granted by the Commission by Amendment of this renewed license.
(5)
PSEG Nuclear LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Mechanical disassembly of the Westinghouse Cobalt Burnable Absorber (COBA) Assemblies containing Cobalt-60 is not considered separation.
(7) PSEG Nuclear LLC, pursuant to the Act and 10 CFR 30 to intentionally produce, possess, receive, transfer and use Cobalt-60
(23) Prior to off-load of fuel assemblies containing irradiated COBA fuel inserts, verify that the incident energy flux on the Spent Fuel Pool walls is less than 1010 MeV/cm2-sec per NUREG/
CR-6927 with the associated fuel assemblies placed at the targeted SFP rack locations. specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC. Changes to the withdrawal schedule or storage requirements shall be submitted to the NRC as a report in accordance with 10 CFR 50.4.
(21) PSEG Nuclear LLC shall take one core sample in the Unit 1 spent fuel pool west wall, by the end of 2013, and one core sample in the east wall where there have been indications of borated water ingress through the concrete, by the end of 2015. The core samples (east and west walls) will expose the rebar, which will be examined for signs of corrosion. Any sample showing signs of concrete degradation and/or rebar corrosion will be entered into the licensee's corrective action program for further evaluation. PSEG Nuclear LLC shall submit a report in accordance with 10 CFR 50.4 no later than three months after each sample is taken on the results, recommendations, and any additional planned actions.
(22) Concurrent with the first use of the chilled water cross-tie as allowed by Technical Specification 3. 7.1 0c, PSEG shall confirm the required performance of the chilled water system cross-tie.
D. Paragraph 2.D. has been combined with paragraph 2.E. per Amendment No. 86, June 27, 1988.
E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 1 0 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, submitted by letter dated May 19, 2006, are entitled: "Salem-Hope Creek Nuclear Generating Station Security Plan," "Salem-Hope Creek Nuclear Generating Station Security Training and Qualification Plan," and "Salem-Hope Creek Nuclear Generating Station Security Contingency Plan." The plans contain Safeguards Information protected under 10 CFR 73.21.
PSEG Nuclear LLC shall fully implement and maintain in effect all provisions of the Commission-approved Cyber Security Plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Salem-Hope Creek CSP was approved by License Amendment No. 300 as supplemented by changes approved by License Amendment Nos. 302, 306, and 318.
Renewed License No. DPR-70 Amendment No. 318 Renewed License No. DPR-75 Amendment No. 322 G. The licensees have satisfied the applicable provisions of 10 CFR Part 140, Financial Protection Requirements and Indemnity Agreements, of the Commissions regulations; H. The issuance of this renewed operating license will not be inimical to the common defense and security or to the health and safety of the public; I.
After weighing the environmental, economic, technical and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of Renewed Facility Operating License No. DPR-75 subject to the conditions for protection of the environment set forth herein is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied; J. The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this renewed license will be in accordance with the Commissions regulations in 10 CFR Parts 30, 40 and 70; and K. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1), and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized by this renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facilitys current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commissions regulations.
- 2. Pursuant to approval by the Nuclear Regulatory Commission at meetings on January 14, 1981, April 28, 1981, and May 19, 1981, the License for Fuel-Loading and Low-Power Testing issued on April 18, 1980 is superseded by Renewed Facility Operating License No. DPR-75 hereby issued to PSEG Nuclear LLC and Constellation Energy Generation, LLC (the licensees) to read as follows:
A. This renewed license applies to the Salem Nuclear Generating Station, Unit No. 2, a pressurized water nuclear reactor and associated equipment (the facility), owned by the licensees. The facility is located on the southern end of Artificial Island on the east bank of the Delaware River in Lower Alloways Creek Township in Salem County, New Jersey and is described in the Final Safety Analysis Report as supplemented and amended and the Environmental Report as supplemented and amended.
B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:
(1)
PSEG Nuclear LLC and Constellation Energy Generation, LLC to possess the facility at the designated location in Salem County, New Jersey, in accordance with the procedures and limitations set forth in the renewed license; (2)
PSEG Nuclear LLC, pursuant to Section 104b of the Act and 10 CFR part 50, Domestic Licensing of Production and Utilization Facilities, to possess, use and operate the facility at the designated location in Salem County, New Jersey, in accordance with the limitations set forth in this renewed license; L. The receipt, production, possession, transfer, and use of Cobalt-60 as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Part 30.
Renewed License No. DPR-75 Amendment No. 330 (3)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source or special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration and as fission detectors in amounts as required; (5)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at steady state reactor core power levels not in excess of 3459 megawatts (thermal).
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 330, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Mechanical disassembly of the Westinghouse Cobalt Burnable Absorber (COBA) Assemblies containing Cobalt-60 is not considered separation.
(7) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60.
(38) Prior to off-load of fuel assemblies containing irradiated COBA fuel inserts, verify that the incident energy flux on the Spent Fuel Pool walls is less than 1010 MeV/cm2-sec per NUREG/CR-6927 with the associated fuel assemblies placed at the targeted SFP rack locations. b.
The first performance of the periodic assessment of CRE habitability, Specification 6.17.c(ii), shall be 3 years, plus the 9 month allowance of SR 4.0.2, as measured from June 4, 2003, the date of the most recent successful tracer gas test, as stated in the December 9, 2003 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.
- c.
The first performance of the periodic measurement of CRE pressure, Specification 6.17.d, shall be within 18 months, plus the 138 days allowed by SR 4.0.2, as measured from September 22, 2005, the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.
(34) PSEG Nuclear LLC may make changes to the programs and activities described in the UFSAR supplement, submitted pursuant to 10 CFR 54.21 (d), as revised during the license renewal application review process, provided PSEG Nuclear LLC evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
(35) Appendix A of NUREG-2101, "Safety Evaluation Report Related to the License Renewal of Salem Nuclear Generating Station," dated June 2011, and PSEG Nuclear LLC UFSAR supplement submitted pursuant to 10 CFR 54.21(d), as revised on May 18, 2011, describe certain future programs and activities to be completed before the period of extended operation. PSEG Nuclear LLC shall complete these activities no later than April 18, 2020, and shall notify the NRC in writing when implementation of these activities is complete.
(36) All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation.
All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC. Changes to the withdrawal schedule or storage requirements shall be submitted to the NRC as a report in accordance with 1 0 CFR 50.4.
(37) Concurrent with the first use of the chilled water cross-tie as allowed by Technical Specification 3.7.1 0c, PSEG shall confirm the required performance of the chilled water system cross-tie.
D. An exemption from certain requirements of Appendix J to 10 CFR Part 50 is described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 4.
This exemption was authorized by law and will not endanger life of property or the common defense and security and is otherwise in the public interest. The exemption, therefore, remains in effect. The granting of the exemption was authorized with the
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