ML23093A095
| ML23093A095 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 05/10/2023 |
| From: | Mahesh Chawla Plant Licensing Branch IV |
| To: | Diya F Ameren Missouri, Union Electric Co |
| Chawla M | |
| References | |
| EPID L-2022-LLA-0132 | |
| Download: ML23093A095 (1) | |
Text
May 10, 2023 Mr. Fadi Diya Senior Vice President and Chief Nuclear Officer Ameren Missouri Callaway Energy Center 8315 County Road 459 Steedman, MO 65077
SUBJECT:
CALLAWAY PLANT, UNIT NO. 1 - ISSUANCE OF AMENDMENT NO. 232 REGARDING TECHNICAL SPECIFICATION CHANGES FOR SPENT FUEL STORAGE (EPID L-2022-LLA-0132)
Dear Mr. Diya:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 232 to Renewed Facility Operating License No. NPF-30 for the Callaway Plant, Unit No. 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated August 29, 2022, as supplemented by letters dated October 26, 2022, and February 21, 2023.
The amendment revises TS 3.7.16, Fuel Storage Pool Boron Concentration; TS 3.7.17, Spent Fuel Assembly Storage; and TS 4.3.1, Criticality, to accommodate a simplified storage configuration that establishes two regions in the spent fuel pool and provide a bounding nuclear criticality safety analysis for additional fuel assembly designs.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Mahesh L. Chawla, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483
Enclosures:
- 1. Amendment No. 232 to NPF-30
- 2. Safety Evaluation cc: Listserv
UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 232 License No. NPF-30
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Union Electric Company (UE, the licensee),
dated August 29, 2022, as supplemented by letters dated October 26, 2022, and February 21, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-30 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan*
The Technical Specifications contained in Appendix A, as revised through Amendment No. 232 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This amendment is effective as of its date of issuance, and shall be implemented within 90 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-30 and the Technical Specifications Date of Issuance: May 10, 2023 Jennifer L.
Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2023.05.10 10:43:47 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 232 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-30 CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483 Replace the following pages of the Renewed Facility Operating License No. NPF-30 and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Renewed Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 3.7-41 3.7-41 3.7-43 3.7-43 3.7-44 3.7-44 4.0-1 4.0-1 4.0-2 4.0-2 4.0-3 4.0-3
Renewed License No. NPF-30 Amendment No. 232 (3)
UE, 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; (4)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
UE, 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 UE is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan*
The Technical Specifications contained in Appendix A, as revised through Amendment No. 232 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Environmental Qualification (Section 3.11, SSER #3)**
Deleted per Amendment No. 169.
Amendments 133, 134, & 135 were effective as of April 30, 2000 however these amendments were implemented on April 1, 2000.
The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
CALLAWAY PLANT 3.7-41 Amendment No. 232 Fuel Storage Pool Boron Concentration 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Fuel Storage Pool Boron Concentration LCO 3.7.16 The fuel storage pool boron concentration shall be 2165 ppm.
APPLICABILITY:
When fuel assemblies are stored in the fuel storage pool.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Fuel storage pool boron concentration not within limit.
NOTE -------------------
LCO 3.0.3 is not applicable.
A.1 Suspend movement of fuel assemblies in the fuel storage pool.
AND Immediately A.2 Initiate action to restore fuel storage pool boron concentration to within limit.
Immediately
CALLAWAY PLANT 3.7-43 Amendment No. 232 Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 The combination of initial enrichment and burnup of each spent fuel assembly stored in Region 2 shall be within the Acceptable Domain of Figure 3.7.17-1.
APPLICABILITY:
Whenever any fuel assembly is stored in the fuel storage pool.
ACTIONS SURVEILLANCE REQUIREMENTS CONDITION REQUIRED ACTION COMPLETION TIME A.
Requirements of the LCO not met.
A.1
NOTE -----------
LCO 3.0.3 is not applicable.
Initiate action to move the noncomplying fuel assembly to Region 1.
Immediately SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial enrichment and burnup of the fuel assembly is in accordance with Figure 3.7.17-1.
Prior to storing the fuel assembly in Region 2
CALLAWAY PLANT 3.7-44 Amendment No. 232 Spent Fuel Assembly Storage 3.7.17 Figure 3.7.17-1 (page 1 of 1)
MINIMUM REQUIRED FUEL ASSEMBLY BURNUP AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2
CALLAWAY PLANT 4.0-1 Amendment No. 232 Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Callaway Plant site consists of approximately 2,767 acres of rural land 10 miles southeast of the city of Fulton in Callaway County, Missouri, and 80 miles west of the St. Louis metropolitan area.
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitution of fuel rods by zirconium alloy or stainless steel filler rods may be used in accordance with approved applications of fuel rod configurations. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies. The control rod material shall be silver indium cadmium, hafnium metal, or a mixture of both types, as approved by the NRC.
4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum nominal U-235 enrichment of 5.0 weight percent; (continued)
Design Features 4.2 4.0 DESIGN FEATURES CALLAWAY PLANT 4.0-2 Amendment No. 232 4.3 Fuel Storage (continued) b.
keff < 1.0 if fully flooded with unborated water and keff 0.95 if flooded with borated water, which includes an allowance for uncertainties as described in Section 9.1 of the FSAR; c.
A nominal 8.99 inch center to center distance between fuel assemblies placed in the fuel storage racks; d.
Partially spent fuel assemblies with a discharge burnup in the Acceptable Domain for Region 2 storage of Figure 3.7.17-1 may be allowed unrestricted storage in the fuel storage racks, except for the empty cells associated with a Region 1 assembly; e.
Rules governing the storage configurations in Regions 1 and 2 are the following:
For cells containing a Region 1 assembly:
1.1 None of the face-adjacent cells may contain a Region 1 assembly; 1.2 A minimum of two of the face-adjacent cells must be empty; 1.3 A maximum of two of the remaining face-adjacent cells may contain Region 2 assemblies. See also Rule 2.3; 1.4 If both of the remaining face-adjacent cells are Region 2 assemblies, then Rule 2.1 is restricted to one Region 1 assembly for those cells.
For cells containing a Region 2 assembly:
2.1 A maximum of two of the face-adjacent cells may contain Region 1 assemblies. See also Rule 1.4; 2.2 The remaining face-adjacent cells may contain Region 2 assemblies or be empty; 2.3 If two face-adjacent cells contain Region 1 assemblies, then Rule 1.3 is restricted to one Region 2 assembly for those cells.
f.
New or partially spent fuel assemblies with a discharging burnup in the "Unacceptable Domain for Region 2 Storage of Figure 3.7.17-1 will be stored in Region 1.
(continued)
Design Features 4.2 4.0 DESIGN FEATURES CALLAWAY PLANT 4.0-3 Amendment No. 232 l 4.3 Fuel Storage (continued) 4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum nominal U-235 enrichment of 5.0 weight percent; b.
keff 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the FSAR; c.
keff 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 9.1 of the FSAR; and d.
A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks.
4.3.2 Drainage The fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 2040 ft.
4.3.3 Capacity The fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2363 fuel assemblies in the spent fuel pool and no more than 279 assemblies in the cask loading pool.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 232 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483
1.0 INTRODUCTION
By letter dated August 29, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22242A122), as supplemented by letters dated October 26, 2022 (ML22299A232), and February 21, 2023 (ML23052A041), Union Electric Company, doing business as Ameren Missouri (the licensee), submitted a license amendment request (LAR) that proposed changes to the Technical Specifications (TSs) for Callaway Plant, Unit No. 1 (Callaway).
The proposed changes would modify TS 3.7.16, Fuel Storage Pool Boron Concentration; TS 3.7.17, Spent Fuel Assembly Storage; and TS 4.3.1, Criticality, to accommodate a simplified storage configuration that establishes two regions in the spent fuel pool (SFP) and provide a bounding nuclear criticality safety (NCS) analysis for additional fuel assembly designs.
The supplemental letter dated February 21, 2023, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff proposed no significant hazards consideration determination as published in the Federal Register on February 7, 2023 (88 FR 8003).
2.0 REGULATORY EVALUATION
2.1 Proposed Changes The updated SFP NCS analysis evaluates SFP storage racks for placement of fuel within storage arrays defined in the TSs. Proposed revisions to the TSs include:
TS 3.7.16 The proposed revision to TS 3.7.16 would include:
Delete and a fuel storage pool verification has not been performed since the last movement of fuel assemblies in the fuel storage pool from the applicability and Delete Required Action A.2.2 and the associated completion time and renumber Required Action A.2.1 to A.2.
TS 3.7.17 The proposed revision to TS 3.7.17 would include:
Delete or 3 and or in accordance with Specification 4.3.1.1 in Limiting Condition for Operation (LCO) 3.7.17; Delete or Specification 4.3.1.1 and or 3 in Surveillance Requirement 3.7.17.1; Replace figure 3.7.17-1 in its entirety with a new figure describing storage restrictions related to initial enrichment and assembly exposure; and Delete and 3 from the label for figure 3.7.17-1.
TS 4.3.1 The proposed revision to TS 4.3.1 would include:
Delete For fuel with enrichments greater than 4.6 nominal weight percent of U-235
[Uranium-235], the combination of enrichment and integral fuel burnable absorbers shall be sufficient so that the requirements of 4.3.1.1.b are met from TS 4.3.1.1.a; Delete 0.95 from TS 4.3.1.1.b; Add < 1.0 and and keff [k-effective] 0.95 if flooded with borated water to TS 4.3.1.1.b; Delete Burnup, and 3, and in the checkerboarding configuration from TS 4.3.1.1.d; Add associated with a and Region 1 assembly to TS 4.3.1.1.d;
TS 4.3.1.1.e is deleted in its entirety and replaced with a set of logical rules that govern acceptable storage configurations of fuel assemblies in the SFP; and Delete Burnup and or 3 in TS 4.3.1.1.f.
2.2 Regulatory Requirements and Guidance The regulatory requirements and guidance documents that the NRC staff used in the review of the LAR, as supplemented, are listed below.
Title 10 to the Code of Federal Regulations (10 CFR), Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, General Design Criteria for Nuclear Power Plants, General Design Criterion (GDC) 62, Prevention of criticality in fuel storage and handling, requires that, Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
The regulations in 10 CFR 50.68, Criticality accident requirements, requires, in part, under 10 CFR 50.68(a), that each holder of an operating license shall comply with either 10 CFR 70.24 or the requirements in 10 CFR 50.68(b). The licensee has elected to comply with 10 CFR 50.68(b), which describes the keff limits for the NCS analysis as well as limits on the maximum nominal U-235 enrichment of fresh fuel assemblies to 5 percent by weight.
The applicable requirements of 10 CFR 50.68 to this LAR are:
10 CFR 50.68(b)(4), which states, in part, If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.
The regulations in 10 CFR 50.36, Technical specifications, contain the requirements for the content of TSs. The regulations in 10 CFR 50.36(b) require, in part, TSs to be derived from the analyses and evaluations included in the safety analysis report, and amendments thereto. In accordance with 10 CFR 50.36(c)(2), Limiting conditions for operations are the lowest functional capability or performance levels of equipment required for safe operation of the facility. In accordance with 10 CFR 50.36(c)(3), Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. The regulation in 10 CFR 50.36(c)(4), Design features, requires that TSs include design features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3) of
[10 CFR 50.36]. The regulation in 10 CFR 50.36(c)(5), Administrative controls, requires that the TSs include, provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
Nuclear Energy Institute (NEI) 12-16, Revision 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants, dated September 2019 (ML19269E069),
provides guidance for acceptable approaches that may be used by industry to perform criticality analyses for the storage of new and spent fuel at LWR power plants in compliance with
10 CFR Part 50. NEI 12-16, Revision 4, was endorsed by the NRC in Regulatory Guide 1.240, Fresh and Spent Fuel Pool Criticality Analyses (ML20356A127), with some clarifications and exceptions.
NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology (ML050250061), describes the procedures by which licensees may validate codes for criticality safe analyses including determination of calculational bias, bias uncertainty, and upper safety limits.
3.0 TECHNICAL EVALUATION
3.1 Background
The licensees NCS analyses describes the methodology and analytical models to show that the requirements of 10 CFR 50.68 are met.
The licensee is proposing to change the number of storage regions from three regions to two regions, with Region 1 being a checkerboard pattern for high reactivity assemblies and Region 2 with uniform loading for low reactivity assemblies. The terms checkerboard and uniform, which are described in TS 4.3.1.1.e in a set of logical rules for permissible loading, are not exact descriptions of the Region 1 and Region 2 storage configurations. TS figure 3.7.17-1 is revised to only include two regions for permissible loading and also includes credit for assembly cooling time from 0 to 20 years.
3.1.1 Applicable Fuel Assembly Designs The NCS analysis must consider all potential fuel assembly designs that may be loaded into, or are currently within, the SFP. The fuel assemblies considered in the licensees NCS analysis include:
Westinghouse 17x17 Standard (STD)
Westinghouse 17x17 Optimized (OFA)
Westinghouse 17x17 Vantage+ (V+)
Framatome GAIA 17x17 (GAIA)
The above list of fuel designs, described in section 3.3.1, Design Basis Fuel Assembly Design, of the LAR, as supplemented by letter dated October 26, 2022, includes all assembly designs currently stored or intended to be stored in the Callaway SFP.
3.2 Method of Analysis There is no generic or standard NRC-approved methodology for performing NCS analyses for fuel storage and handling. The plant-specific methods used for the NCS analyses for fuel in the Callaway SFP are described in section 3.0, Methodology, of enclosures 2 (publicly available) and 3 (not publicly available) of the letter dated October 26, 2022. The licensee used the guidance described in NEI 12-16, Revision 4, to demonstrate compliance with the requirements of 10 CFR 50.68(b).
3.2.1 Computer Code Validation The licensees NCS analyses consider the decrease in fuel reactivity typically seen in pressurized water reactors (PWRs) as the fuel is depleted during reactor operation. This approach is frequently used in PWR NCS analyses and is sometimes referred to as burnup credit. Burnup credit NCS analyses involves a two-step process. The first step relates to depletion where a computer code simulates the reactor operation to calculate the changes in the fuel composition of the fuel assembly. The second step is a modeling of the depleted fuel assembly in the SFP storage racks and the determination of the system keff. The validation of the computer codes in each step is a significant portion of the analysis. Since the licensees NCS analyses credit fuel burnup, in its review, the NRC staff considered validation of the computer codes and data used to calculate burned fuel compositions, and the computer code and data that utilize the burned fuel compositions to calculate keff for systems with burned fuel.
3.2.1.1 Criticality Code Validation The licensees NCS analysis utilizes MCNP5 for criticality calculations. MCNP5 is a three-dimensional continuous energy Monte Carlo code. The MCNP5 requires the user to input values for the following parameters that may have a significant effect on the statistics of the final results: number of neutrons per generation, number of generations skipped prior to averaging, total number of generations, and initial source distribution. The results of the criticality benchmark analysis adequately demonstrate convergence; therefore, the input parameters described above are acceptable.
Based on its review, the NRC staff finds that the validation was performed in a manner consistent with NUREG/CR-6698. Further, the selection of critical benchmark experiments appropriately bound the design application range for the Callaway SFP.
Therefore, the NRC staff concludes that the MCNP5 validation is acceptable because the inputs are appropriate and result in adequate convergence checks, which ensures that an accurate keff is determined.
3.2.1.2 Depletion Code Validation The licensee uses the CASMO5 code to model irradiation of the fuel in the core. The CASMO5 code, used to determine the isotopic composition of irradiated fuel, was reviewed and approved by the NRC in a letter dated September 15, 2017 (ML17236A419). The licensees analyses calculated the depletion uncertainty as 5 percent of the reactivity difference between fresh unpoisoned fuel and the burnup of interest. The NRC staff finds that this method is acceptable because the method is consistent with the guidance in NEI 12-16, Revision 4.
3.3 Depletion Analysis 3.3.1 Core Moderator and Fuel Temperature The licensees NCS analysis considered moderator and fuel temperatures varied from the nominal case by 100 degrees Kelvin (°K) for moderator temperature and 300 °K for fuel temperature. The nominal case is the upper bound of moderator and fuel temperatures as the upper bound is expected to provide the most reactive configuration. Based on its review, the NRC staff finds that the temperature ranges appropriately demonstrate that the upper limit of each temperature is bounding. The results of this temperature analysis are found in table 7-6,
Reactivity Effect of Core Operating Parameters, of the LAR, as supplemented by letter dated October 26, 2022. The analysis demonstrates that the maximum temperature is limiting for both moderator and fuel temperature. Therefore, the NRC staff finds that licensees NCS analysis, with respect to core moderator and fuel temperature, is acceptable because the most limiting conditions are used in determining the maximum keff.
3.3.2 Core Soluble Boron Concentration The licensees NCS analysis considered core soluble boron concentration varied from the nominal case by 300 parts per million (ppm). The nominal case is the upper bound of soluble boron concentration as the maximum soluble boron concentration is expected to provide the most reactive configuration. The result of this soluble boron concentration analysis is found in table 7-6 of the LAR, as supplemented. The analysis demonstrates that the highest soluble boron concentration is limiting. Therefore, the NRC staff finds that the licensees NCS analysis with respect to core soluble boron concentration is acceptable because the most limiting condition is used in determining the maximum keff.
3.3.3 Fuel Power Density The licensees NCS analysis considered fuel power density varied from the nominal case by 5 megawatt per metric ton of uranium (MWD/MTU). The nominal case is the highest allowable power density. The precise effect of power density on fuel reactivity is unclear. The result of this power density analysis is found in table 7-6 of the LAR, as supplemented. The NRC staff finds that licensees NCS analysis demonstrates that power density has an insignificant effect on the reactivity. Therefore, the NRC staff concludes that the licensees NCS analysis with respect to fuel power density is acceptable because the most limiting condition is used in determining the maximum keff.
3.3.4 Fixed, Removable, and Integral Burnable Absorbers The use of burnable absorbers can have a significant effect on the reactivity of a spent fuel assembly due to changes in the local neutron flux spectrum. The burnable absorbers that should be considered in NCS analyses include poison rods and integral absorbers, such as gadolinium and control rods.
The licensees NCS analysis contains a sensitivity study to determine the bounding configuration of burnable absorbers. Because multiple absorbers are being used simultaneously, the absorbers are all modeled explicitly in the depletion analysis, in accordance with the guidance in NEI 12-16, Revision 4. The result of this analysis is documented in table 7-8, Reactivity Effect of Irradiation with the IBA [Integral Burnable Absorber] and Fuel Inserts, of the LAR, as supplemented by letter dated October 26, 2022, and table 3, Alternative IFBA [Integral Fuel Burnable Absorber] loading of enclosures 7 (publicly available) and 8 (not publicly available) of the supplement dated February 21, 2023. The NRC staff finds that this approach to modeling fuel depletion with absorbers is acceptable because the most limiting condition is used in determining the maximum keff.
3.3.5 Control Rod Usage If rod cluster control assemblies (RCCAs) are present in assemblies for significant amounts of time in the reactor, the associated spectral hardening can increase plutonium generation, leading to higher fuel reactivity for the same burnup. The Callaway core frequently operates with
RCCAs, at most, inserted 8 inches for silver-indium-cadmium (Ag-In-Cd) RCCAs and 12 inches for hafnium-zirconium (Hf-Zr) RCCAs. The licensees NCS analysis considered partial insertion of RCCAs in the depletion analysis. The NRC staff finds this approach to modeling fuel depletion with RCCAs is acceptable because modeling the maximum use of RCCAs during normal operation will result in highest contribution to the maximum keff.
3.3.6 Axial Burnup Profile Fuel depletion follows a near-cosine shape along the axial height of the fuel assembly with the highest burnups occurring in the middle of the assembly. The result is that the top and bottom ends of a fuel assembly have a lower burnup and that assembly average burnup takes on a cosine shape; therefore, the top and bottom ends are more reactive. As burnup increases, the cosine shape flattens. The licensees NCS analysis considers multiple bounding burnup profiles and selects the most limiting case for use in design-basis criticality calculations. The results of this analysis are documented in table 7-9, Reactivity Effect of Axial Burnup Profile, of the LAR, as supplemented by letter dated October 26, 2022. The analysis determined that the Westinghouse 17x17 axial burnup profiles are most limiting. The NRC staff finds the selection of axial burnup profiles to be acceptable because the most limiting case has been determined and will be used in design-basis criticality calculations.
3.3.7 Depletion Uncertainty The licensees analyses calculated the depletion uncertainty as 5 percent of the reactivity difference between fresh unpoisoned fuel and the burnup of interest. The NRC staff finds that the method used by the licensee is consistent with the guidance in NEI 12-16, Revision 4, and is, therefore, acceptable.
The licensees NCS analysis also considers uncertainties related to burnup, minor actinides, and fission products. The licensee applies a conservative 5 percent uncertainty to burnup. The licensee applies a conservative 1.5 percent bias to bound the reactivity effect of minor actinides and fission products. The NRC staff finds that this depletion uncertainty and additional bias is acceptable because it is consistent with the guidance in NEI 12-16, Revision 4, and an additional 1.5 percent bias is sufficient to bound the uncertainty associated with the reactivity of any minor actinides not explicitly modeled.
3.4 Spent Fuel Pool Criticality Safety Analysis 3.4.1 Spent Fuel Pool Water Temperature The guidance in NEI 12-16, Revision 4, states that the NCS analysis should use a water temperature and density that results in the maximum reactivity. Typically, the maximum reactivity will occur at either at the highest or lowest temperatures allowed. The licensees NCS analysis considers SFP water temperatures near 4 degrees Celsius (°C), 20 °C, and 120 °C.
The results of NCS analysis related to SFP water temperature are documented in table 7-4, Reactivity Effect of SFP Water Temperature, of the LAR, as supplemented by letter dated October 26, 2022. The results show that the highest reactivity configuration occurs at the lowest temperature water at maximum density. The NRC staff finds the water temperature analysis acceptable because the most reactive conditions will be used to generate a maximum keff.
3.4.2 Spent Fuel Pool Storage Racks The Callaway SFP contains 15 high density storage racks for spent fuel. The racks are all of the same design and materials; therefore, their treatment is the same throughout the entire NCS analysis. The racks have fixed BORAL neutron absorbers between each storage cell. The length of the absorbers is such that the entire length of the active fuel is shadowed by the BORAL absorber with some conservatisms associated with poison panel length. The storage racks are modeled using a 2x2 array with periodic boundary conditions. Lastly the BORAL panels will be modeled as having a minimum Boron-10 (10B) areal density. The NRC staff finds that these modeling considerations are acceptable because they will result in a conservative determination of the maximum keff.
3.4.3 Bounding Spent Fuel Assembly The fuel assemblies considered in this analysis are all 17x17 assemblies manufactured by either Westinghouse or Framatome. The SFP NCS analysis did not determine a single bounding assembly design. Instead, the licensee performed an analysis, the results of which are located in table 7-1, Bounding Fuel Assembly Design, of the LAR, as supplemented by letter dated October 26, 2022, that demonstrates which specific assembly designs are most reactive.
The results show that the V+ assemblies are the most reactive fresh assemblies and will therefore be used in the NCS analysis for Region 1 storage. The GAIA fuel assemblies are the most reactive spent fuel and will be used in the NCS analysis for Region 2 storage. The NRC staff finds that this selection of a bounding fuel assembly design is acceptable because the bounding design will ensure that the maximum keff is determined for each region.
Additionally, the bounding fuel assemblies are modeled with uniform enrichment for fresh fuel assemblies, no credit for IFBA is taken, and bounding depletion characteristics are used. The NRC staff finds that these modeling considerations are acceptable because they will result in a conservative contribution toward determining the maximum keff.
3.4.4 Manufacturing Tolerances and Uncertainties NEI 12-16, Revision 4, allows for the determination of the keff uncertainty due to manufacturing tolerances and uncertainties to be determined by: (1) a root sum square of the individual keff uncertainty values, (2) all tolerance values selected to maximize keff, or (3) a combination of (1) and (2). The licensee used method (2).
Both the fuel assembly and storage rack have manufacturing tolerances and uncertainties that can affect the reactivity of the system. The parameters related to fuel assembly and storage rack reactivity are described in sections 3.3.2, Fuel Assembly Parameters, and 3.3.3, Spent Fuel Rack Parameters, of the LAR, as supplemented by letter dated October 26, 2022. The results of these reactivity analyses are documented in tables 7-2, Reactivity Effect of Fuel Assembly Parameters, and 7-3, Reactivity Effect of SFR [Spent Fuel Rack] Parameters, of the LAR, as supplemented by letter dated October 26, 2022.
The BORAL poison panels are composed of finely ground Boron Carbide (B4C) particles in an aluminum matrix. There is a potential positive reactivity phenomenon caused by self-shielding of large B4C particles. The licensee performed a reactivity analysis of B4C particle size to determine if there is any significant positive reactivity effect. The results of that analysis are documented in table 7-26, Reactivity Effect of B4C Particle Size, of the LAR, as supplemented by letter dated October 26, 2022. The analysis considers particles sized between 45
and 180 micrometers (µm), which make up between 81.2 and 93.5 percent of the particles. The analysis demonstrates that variations in particle size is not a significant contributor to changes in keff. The analysis shows a small increase in keff compared to the nominal values. This small increase is due to the heterogenous distribution model used, which results in increased neutron streaming, thus causing a small increase in calculated keff. This is not representative of an actual poison panel. Therefore, the NRC staff determined that variations in B4C particle size do not significantly impact reactivity for the particle sizes present in the Callaway SFP. Neglecting B4C particle size effects will still allow the licensee to determine an acceptable maximum keff.
The NRC staff finds that the licensees calculation of reactivity effects related to manufacturing tolerances and uncertainties is acceptable because all significant contributors to reactivity are analyzed and their associated biases to the maximum keff are treated appropriately.
3.4.5 Eccentricity of Fuel within the Storage Cell The nominal keff calculation models all fuel assemblies in the center of their respective storage cells. However, the fuel assemblies can be anywhere within their respective storage cells. The eccentricity portion of the analysis is intended to determine the reactivity effect of the fuel assemblies being in positions other than the center of their storage cell. The analysis should include the reactivity effect of the most limiting eccentric position, if any, as either a bias from the nominal centrally positioned assembly or as part of the design-basis calculation.
Section 3.3.5, Fuel Assembly Radial Positioning and Orientation, of the LAR, as supplemented by letter dated October 26, 2022, describes the analysis the licensee performed to consider the eccentric positioning of fuel assemblies within a storage cell. The results of this analysis are found in table 7-5, Reactivity Effect of Fuel Assembly Radial Positioning, of the LAR, as supplemented by letter dated October 26, 2022. The analysis demonstrates that the most reactive configuration occurs when all assemblies are in the centers of their respective cells.
The licensee also considered radial rotation of a fuel assembly within its storage cell. The fuel assemblies currently stored in the Callaway SFP have uniform enrichments and are rotationally symmetric. Therefore, there is minimal reactivity consequence associated with radial rotation of a fuel assembly within its storage cell. Any reactivity difference associated with an asymmetric burnup will have a minimal impact on the maximum keff.
The NRC staff finds that the licensees analysis of eccentric storage is acceptable because potentially limiting configurations were considered. Additionally, the guidance in NEI 12-16, Revision 4, indicates that a centrally positioned assembly is the most limiting position for storage cells surrounded on all four sides by poison panels with an areal density greater than 0.01 grams per square centimeter (g/cm2.)
3.4.6 Spent Fuel Storage NCS Biases and Uncertainties The Callaway NCS analysis utilizes a total correction factor (TCF) to capture the biases and uncertainties that will combine with the calculated keff to determine the maximum keff. This approach is consistent with the guidance in NEI 12-16, Revision 4. The factors that contribute to TCF are described in figure 3-3, Determination of the Total Correction Factor, of the LAR, as supplemented by letter dated October 26, 2022. The NRC staff reviewed the inputs to the TCF and finds that they are acceptable because the inputs capture all necessary biases and uncertainties.
3.5 Determination of Soluble Boron Requirements Based on its review, the NRC staff determined that the licensees design basis calculations include a soluble boron concentration of 500 ppm and adequately demonstrate compliance with the requirements of 10 CFR 50.68(b)(4) at a boron concentration of 500 ppm under normal conditions. The minimum soluble boron concentration is further increased by 50 ppm due to Assumption 7 in section 4.0, Assumptions, of the LAR, as supplemented by letter dated October 26, 2022. The increase in the soluble boron requirement is sufficiently high to bound any changes in reactivity associated with neglecting minor structural components.
The licensee also performed an analysis of soluble boron requirements under accident conditions. The results of this analysis are documented in table 7-21, Maximum keff Calculations for the Fuel Misload Accident, of the LAR, as supplemented by letter dated October 26, 2022. The maximum soluble boron concentration requirement to meet the regulatory limit is demonstrated to be 1031.2 ppm. The 50 ppm bias from Assumption 7 is also applicable to this concentration.
The licensees current minimum SFP boron concentration, as stipulated in TS 3.7.16 is 2165 ppm. This is significantly higher than the boron concentration requirements above and no change is being proposed to this limit. Therefore, the NRC staff finds that the licensees current soluble boron concentration requirements are bounding and will ensure the licensee meets the requirements of 10 CFR 50.68(b)(4).
3.6 Interface Analysis 3.6.1 Interfaces Between Dissimilar Storage Racks The Callaway SFP only contains a single storage rack design. Therefore, there are no interfaces between dissimilar storage racks. The licensee performed an analysis to measure the reactivity effects between two storage racks. All storage racks in the Callaway SFP have exterior BORAL panels; therefore, there are two panels between each assembly between storage racks. The design basis 2x2 model bounds any reactivity effects between two storage racks. The NRC staff finds this assessment to be acceptable because the presence of multiple BORAL panels between two exterior assemblies is enough to reduce the reactivity such that there is no increase in maximum keff due to this interface.
3.6.2 Storage Configurations The licensees application proposes to transition the Callaway SFP from three storage regions to two. Region 1 is for high reactivity fuel that is either fresh or fuel with low burnup and cooling time. The Region 1 storage pattern is a checkerboard pattern where all assemblies are surrounded by empty cells on face-adjacent sides. Interfaces between Regions 1 and 2 are an exception to this, and the interface analysis is discussed in the next section. Region 2 is for lower reactivity fuel that has either relatively higher burnup or a long cooling time. The Region 2 storage pattern is uniform where all cells may be occupied by a spent fuel assembly. There are no restrictions on the locations of these regions, and they may be adjacent to one another with no additional measures taken.
Instead of relying on terms like uniform and checkboard to describe storage configurations, the licensee introduced a set of logical rules that implicitly apply these storage configurations based on the storage cell of interest and its face-adjacent neighbors. These rules remove any
ambiguity associated with the uniform and checkboard descriptions. The NRC staff finds the storage configurations and associated rules to be acceptable because all possible bounding configurations have been analyzed and the rules prohibit unanalyzed storage configurations.
3.6.3 Interfaces Between Different Storage Configurations The licensee performed a criticality analysis between the two storage regions and varied parameters that could potentially increase reactivity. The results of the analysis are documented in table 7-17, Summary of the Analysis for the SFR Interfaces, of the LAR, as supplemented by letter dated October 26, 2022. The analysis demonstrates that the region interfaces do not result in any increase in reactivity. The reactivity of the system is dominated by the Region 2 assemblies. The licensee considered all potential interface conditions with respect to the rules for permissible loading. Based on its review, the NRC staff finds the licensees assessment of region interfaces to be acceptable because all region interfaces have been analyzed and dispositioned.
3.6.4 Other Interfaces The licensee also considered an interface between the storage rack and SFP wall. There is no reactivity increase associated with this interface due to exterior BORAL panels and a water gap between the storage rack and SFP wall. The NRC staff finds the licensee assessment of the rack-to-wall interface acceptable because the BORAL panel and water gap neutronically decouple the assembly from the wall, which would otherwise act as a reflector.
3.7 Normal and Accident Conditions 3.7.1 Normal Conditions The licensee considers any regular activities within the SFP that may contribute to changes in reactivity. These activities include:
Fuel Movement Fuel Insertion and Removal Storage of Fuel Rod Storage Rack (FRSR)
Storage of Fuel Assemblies with Missing Rods Storage of Low-Burned Fuel Assemblies Movement of fuel assemblies occurs above the spent fuel storage racks with sufficient distance between the bottom of the moving assembly and the top of stored assemblies to preclude any neutronic interaction.
Reactivity effects of fuel assembly insertion and removal is bounded by the regular storage configuration. That is, inserting and removing assemblies does not present an increase in reactivity relative to a stored assembly.
A FRSR holds fuel rods that have been removed from other assemblies. These racks have a larger pitch and hold fewer rods than the design-basis assembly. Therefore, the FRSR does not present an increase in reactivity.
Fuel assemblies with missing fuel rods can increase reactivity due to an increase in moderation to an extent. Removing enough rods will result in a net decrease in reactivity due to loss of fissile material. The licensees analysis demonstrates a maximum increase in keff of 0.019. Due to this reactivity increase, the licensee has proposed in section 3.5.4 Storage of Fuel Assemblies with Missing Rods, of the LAR, as supplemented by letter dated October 26, 2022, that all fuel assemblies with missing fuel rods or guide tubes be restricted to Region 1 storage where sufficient margin to the regulatory limits exist to accommodate this increase in reactivity.
The NRC staff finds this approach to be acceptable because the Region 1 margin is significantly higher than the keff increase due to missing rods.
Low-burned assemblies may not meet the storage requirements for Region 2 but cannot be moved due to a potential mechanical failure mechanism. All affected assemblies are already stored in a checkerboard pattern, consistent with Region 1 and meet the criteria for storage in Region 1. These assemblies will be subject to the Region 1 storage rules as described in TS 4.3.1.1.e.
Based on its review, the NRC staff finds that the licensees analysis of normal conditions is acceptable because any increase in maximum keff is appropriately dispositioned.
3.7.2 Bounding Accident or Abnormal Condition The licensee considered several accident or abnormal conditions that could potentially result in an increase in keff. This analysis also considers the maximum soluble boron concentration required to mitigate the effects of the accident or abnormal condition such that the maximum keff remains below the regulatory limits. The conditions analyzed are:
Loss of SFP cooling Dropped assembly resting horizontally on the SFR Assembly dropped vertically into a storage cell Mislocated fuel assembly, both within and outside of an SFR Incorrect loading curve Rack movement Boron Dilution (see section 3.7.3 of this safety evaluation)
The potentially bounding non-boron dilution events are mislocated fuel assembly and incorrect loading curve conditions. The licensee performed a keff calculation for both of these events documented in tables 7-21 and 7-22, Maximum keff Calculation for the Incorrect Loading Curve Accident, of the LAR, as supplemented by letter dated October 26, 2022. The goal of the analysis is to determine the minimum soluble boron concentration required to mitigate the accident. The determined minimum soluble boron concentration required to mitigate the above accidents is 1031.2 ppm. This is less than the Callaway TS limit of 2165 ppm.
In some cases, a multiple misload accident can occur with once-burned fuel assemblies, whether by operator error or by an error in documentation. Fresh fuel assembles are easily identifiable when involved in a multiple misload accident due to the shininess of the cladding.
Once-burned and spent fuel assemblies lose this property. Therefore, it is necessary to consider the possibility that once-burned assemblies are involved in a multiple misload accident unless other conditions prevent such an event from happening. In this case, the licensee performed an analysis to determine the minimum assembly burnup required to exceed the regulatory limits during a multiple misload accident. The licensees interpolated minimum burnup was
15.6 gigawatt decay per metric ton uranium (GWD/MTU), which is less than what is expected in a once-burned fuel assembly. Thus, the licensees multiple misload evaluation is bounding, and a multiple misload accident will not result in a condition that exceeds regulatory limits.
Based on its review, the NRC staff finds that the licensees analysis of accident conditions and determination of the minimum soluble boron requirement to mitigate accidents is acceptable because the bounding accidents were considered and there is sufficient margin to the licensees TS limit for minimum soluble boron concentration.
3.7.3 Boron Dilution A boron dilution accident is one in which there is a leak of borated water from the SFP. The water added to the SFP that might compensate for such a leak may not be borated, thus reducing the soluble boron concentration to below the TS minimum concentration of 2165 ppm.
To accommodate a boron dilution accident, the licensee must maintain a minimum boron concentration of 550 ppm, as determined by the minimum boron concentration necessary for normal conditions.
The licensees analysis of the boron dilution event adequately demonstrates that the event is not credible with respect to exceeding the criticality limits specified in 10 CFR 50.68(b)(4). The licensee has demonstrated that there exists sufficient time for operator response to mitigate the accident before the soluble boron concentration is reduced to below 550 ppm. Therefore, the NRC staff finds this analysis to be acceptable because there exists significant margin between operator response time and time to boron dilution below 550 ppm.
3.7.4 Reduction in BORAL 10B Areal Density There is potential for the 10B areal density to be reduced during the lifetime of the SFR. The NCS analysis uses the minimum allowed areal density, but the licensee considers reductions beyond what is minimally allowed. The results of this analysis are documented in table 7-24, Reactivity Effect of the BORAL' Panel 10B Areal Density, of the LAR, as supplemented by letter dated October 26, 2022. The analysis shows an increase in keff, which was not accommodated for the determination of maximum keff. The licensee performed a margin evaluation to determine modeling conservatisms to demonstrate that modeling conservatisms are sufficient to offset any reactivity increase associated with a reduction in areal density of up to 20 percent. The NRC staff finds that this analysis is acceptable because the increase in reactivity caused by BORAL degradation is offset by the calculated margin resulting from modeling conservatisms and assumptions. Additionally, there is a program in place to take measurements of the areal density to ensure the areal density remains above minimum limit.
Furthermore, there is no significant history of BORAL degradation.
3.8 Cooling Time Credit The licensee proposes to credit cooling time as a function of burnup and initial fuel enrichment.
The reactivity of an assembly decreases over time, potentially allowing a Region 1 assembly to qualify for storage in Region 2 after a certain number of years. The licensee performed an analysis to determine appropriate cooling time curves. The results of that analysis are found in tables 7-12, Summary of the Analysis for Region 2 (Spent Fuel); 7-13, Summary of the Loading Curves for Callaway SFP; and 7-14, Loading Curves Confirmatory Calculations, as well as figure 7-1, Loading Curves for Uniform Loading of Spent Fuel Assemblies (Region 2),
of the LAR, as supplemented by letter dated October 26, 2022. The curves are constructed
such that the target maximum keff is 0.995 when flooded with unborated water and less than 0.95 when flooded with borated water. The licensee credits cooling times of 0, 1, 5, 10, and 20 years. Each of the cooling time curves demonstrate compliance with the requirements of 10 CFR 50.68(b)(4). Therefore, the NRC staff finds that the licensees proposal to credit cooling time in the NCS analysis is acceptable.
3.9 NCS Analysis Results An acceptable SFP NCS analysis must demonstrate compliance with the requirements of 10 CFR 50.68(b). The licensee has demonstrated in the NCS analysis that the maximum keff will remain below the regulatory limits. The results of the NCS analysis demonstrate that the maximum keff for fuel assemblies in Region 1 will remain below 0.8554 when flooded with unborated water. The NRC staff finds that this is well below the regulatory limit and, is therefore, acceptable. The maximum keff for fuel assemblies in Region 2 when flooded with unborated water is 0.9952, and when flooded with 500 ppm borated water is 0.9487. While these values are close to the regulatory limit, there are several conservative assumptions and modeling choices that increase the maximum keff. The licensee determined the maximum keff with a soluble boron concentration of 500 ppm for normal conditions. The licensees TS limit for minimum soluble boron concentration is 2165 ppm, which will result in a significant reduction in maximum keff. The NRC staff finds that the NCS analysis results for Region 2 assemblies is acceptable because the determined maximum keff is below the regulatory limit and there are additional conservatisms to ensure than any increases in reactivity that remain unaccounted for are not sufficient to exceed the regulatory limit.
3.10 NRC Staff Conclusion Based on its review, the NRC staff concludes that there is reasonable assurance that the Callaway SFP meets the applicable regulatory requirements in 10 CFR 50.68 and GDC 62.
Additionally, the NRC staff determined that the proposed TSs would continue to be based on the analyses and evaluations included in the updated final safety analysis report and amendment thereto in accordance with 10 CFR 50.36(b). The NRC staff determined that the TSs, as amended, would continue to require the lowest functional capability or performance levels of equipment required for safe operation of the facility, in accordance with 10 CFR 50.36(c)(2), and contain requirements relating to test calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met, in accordance with 10 CFR 50.36(c)(3).
The NRC staff also determined that the proposed TSs will continue to include required design features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety, in accordance with 10 CFR 50.36(c)(4), and provisions relating to organization and management, procedures, recordkeeping, review and audit and reporting necessary to assure operation of the facility in safe manner, in accordance with 10 CFR 50.36(c)(5). Therefore, the NRC staff concludes that the proposed changes are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Missouri State official was notified of the proposed issuance of the amendment on March 31, 2023. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration published in Federal Register on February 7, 2023 (88 FR 8003), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: Brandon Wise, NRR Kent Wood, NRR Date: May 10, 2023
- concurrence via email OFFICE NRR/DORL/LPL4/PM*
NRR/DORL/LPL4/LA*
NRR/DSS/STSB/BC*
NRR/DSS/SFNB/BC*
NAME MChawla PBlechman VCusumano SKrepel DATE 3/31/2023 4/6/2023 4/7/2023 3/15/2023 OFFICE OGC*
NRR/DORL/LPL4/BC*
NRR/DORL/LPL4/PM*
NAME AGhosh JDixon-Herrity MChawla DATE 4/26/2023 5/9/2023 5/10/2023