ML24145A177
| ML24145A177 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 07/15/2024 |
| From: | James Kim Plant Licensing Branch 1 |
| To: | Mcfeaters C Public Service Enterprise Group |
| Kim J | |
| References | |
| EPID L-2023-LLA-0125 | |
| Download: ML24145A177 (1) | |
Text
July 15, 2024 Charles V. McFeaters President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038
SUBJECT:
HOPE CREEK GENERATING STATION AND SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 ISSUANCE OF AMENDMENT NOS. 236, 349, AND 331 RE: MODIFY EXCLUSION AREA BOUNDARY (EPID L-2023-LLA-0125)
Dear Charles McFeaters:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 236, 349, and 331, to Renewed Facility Operating License Nos. NPF-57, DPR-70, and DPR-75 for the Hope Creek Generating Station (Hope Creek), and Salem Nuclear Generating Station, Unit Nos. 1 and 2 (Salem), respectively, in response to your application dated September 6, 2023, as supplemented by letters dated October 30, 2023, December 21, 2023, April 5, 2024, and April 26, 2024.
The amendments changed the licensing basis as described in the Salem, Units 1 and 2, and Hope Creek Updated Final Safety Analysis Reports to account for modifications to the Exclusion Area Boundary for Salem and Hope Creek.
C. McFeaters 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/
James S. Kim, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-354, 50-272, and 50-311
Enclosures:
- 1. Amendment No. 236 to NPF-57
- 2. Amendment No. 349 to DPR-70
- 3. Amendment No. 331 to DPR-75
- 4. Safety Evaluation cc: Listserv
PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 236 Renewed License No. NPF-57
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by PSEG Nuclear LLC, dated September 6, 2023, as supplemented by letters dated October 30, 2023, December 21, 2023, April 5, 2024, and April 26, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 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, by Amendment No. 236, Renewed Facility Operating License No. NPF-57 is hereby amended to authorize the change to the Updated Final Safety Analysis Report (UFSAR) as requested by letter dated September 6, 2023, as supplemented by letters dated October 30, 2023, December 21, 2023, April 5, 2024, and April 26, 2024, and evaluated in the NRC staffs safety evaluation enclosed with this amendment.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 180 days of the date of issuance. The licensee shall submit the update of the UFSAR authorized by this amendment in accordance with 10 CFR 50.71(e).
FOR THE NUCLEAR REGULATORY COMMISSION Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: July 15, 2024 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2024.07.15 15:30:30 -04'00'
PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 349 Renewed License No. DPR-70
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Constellation Energy Generation, LLC (the licensees), dated September 6, 2023, as supplemented by letters dated October 30, 2023, December 21, 2023, April 5, 2024, and April 26, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 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, by Amendment No. 349, Renewed Facility Operating License No. DPR-70 is hereby amended to authorize the change to the Updated Final Safety Analysis Report (UFSAR) as requested by letter dated September 6, 2023, as supplemented by letters dated October 30, 2023, December 21, 2023, April 5, 2024, and April 26, 2024, and evaluated in the NRC staffs safety evaluation enclosed with this amendment.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 180 days of the date of issuance. The licensee shall submit the update of the UFSAR authorized by this amendment in accordance with 10 CFR 50.71(e).
FOR THE NUCLEAR REGULATORY COMMISSION Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: July 15, 2024 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2024.07.15 15:30:51 -04'00'
PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 331 Renewed License No. DPR-75
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Constellation Energy Generation, LLC (the licensees), dated September 6, 2023, as supplemented by letters dated October 30, 2023, December 21, 2023, April 5, 2024, and April 26, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 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, by Amendment No. 331, Renewed Facility Operating License No. DPR-75 is hereby amended to authorize the change to the Updated Final Safety Analysis Report (UFSAR) as requested by letter dated September 6, 2023, as supplemented by letters dated October 30, 2023, December 21, 2023, April 5, 2024, and April 26, 2024, and evaluated in the NRC staffs safety evaluation enclosed with this amendment.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 180 days of the date of issuance. The licensee shall submit the update of the UFSAR authorized by this amendment in accordance with 10 CFR 50.71(e).
FOR THE NUCLEAR REGULATORY COMMISSION Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: July 15, 2024 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2024.07.15 15:31:11 -04'00'
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 236, 349, AND 331 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-57, DPR-70, AND DPR-75 PSEG NUCLEAR LLC HOPE CREEK GENERATING STATION AND PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-354, 50-272, AND 50-311
1.0 INTRODUCTION
By letter dated September 6, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23249A260), as supplemented by letters dated October 30, 2023, December 21, 2023, April 5, 2024, and April 26, 2024 (ML23303A144, ML23355A273, ML24096A184, and ML24117A128, respectively), PSEG Nuclear LLC (PSEG or the licensee) requested a license amendment for Hope Creek Generation Station (Hope Creek) and Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2. The license amendment request (LAR) would revise the licensing basis as described in the Hope Creek and Salem Updated Final Safety Analysis Reports (UFSARs) to account for modifications to the Exclusion Area Boundary (EAB) for Hope Creek and Salem. The purpose of this LAR is to allow for land parcels designated for use by the New Jersey Wind Port (NJWP) project to be removed from the exclusion area.
The supplements listed above 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, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register (FR) on October 31, 2023 (88 FR 74532).
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations (10 CFR) 50.67, Accident source term, establishes acceptable radiation dose limits resulting from design basis accidents for an individual located at the exclusion area boundary or low population zone, and for occupants of the control room. In its review of this LAR, the staff evaluated whether the revised design basis radiological analyses incorporating the proposed changes to the EAB demonstrate that the regulatory limits in 10 CFR 50.67 continue to be met.
Specifically, the NRC staff evaluated the radiological consequences of affected design-basis-accidents (DBAs) against the dose criteria specified in 50.67(b)(2) and the guidance in Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.
10 CFR part 50, appendix A, General Design Criteria for Nuclear Power Plants: GDC 19, Control room, states, in part:
A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.
The staff also evaluated the meteorological data, inputs to the atmospheric dispersion modeling analyses, and related descriptions that are integral to the proposed changes to the EAB. The staff used the following regulatory guidance and standards in its review:
NRC RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 (ML003716792),
NUREG-0800, Standard Review Plan (SRP) Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, Revision 0, July 2000 (ML003734190).
NUREG-0800, SRP Section 2.3.3, Onsite Meteorological Measurements Program, Revision 3, March 2007 (ML063600394).
NUREG-0800, SRP Section 2.3.4, Short-Term Atmospheric Dispersion Estimates for Accident Releases, Revision 3, March 2007 (ML070730398).
RG 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Revision 1, November 1982 (Reissued February 1983) (ML003740205).
RG 1.23, Meteorological Monitoring Programs for Nuclear Power Plants, Revision 1, March 2007 (ML070350028).
NUREG/CR-2858, PAVAN - An Atmospheric-Dispersion Program for Evaluating Design-Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations, November 1982 (ML12045A149).
NUREG/CR-2260, Technical Basis for Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, October 1981 (ML12045A197).
3.0 TECHNICAL EVALUATION
As part of implementing the reduction in size of the EAB the licensee made few changes to the design basis radiological dose calculations contained in their current licensing basis. Upon review of the initial application and its supplement, the staff found that most of the initial conditions, inputs, and assumptions used in calculating the dose consequences contained in the LAR are variables which were either part of previously approved LARs or incorporated into the current licensing basis for Hope Creek and Salem Units 1 and 2 via the licensees 10 CFR 50.59, "Changes, tests and experiments" (50.59 process). Therefore, the staff did not review these variables as part of its review of this LAR. In the staffs review, the updated initial conditions, inputs and assumptions were reviewed, and confirmed to be appropriately utilized in the licensees dose calculations. The staff also conducted confirmatory calculations. The significant changes since the last docketed calculations are summarized in the supplement provided on April 3, 2024.
3.1 DBAs Affected by the LAR To support the implementation of a reduced size of EAB, the licensee analyzed the radiological consequences for all DBAs and determined that four of the DBAs meet the criteria for requiring a license amendment request:
Hope Creek Generating Station Loss-of-coolant-accident (LOCA)
Main steam line break (MSLB)
Salem Nuclear Generating Station, Units 1 and 2 LOCA Steam generator tube rupture (SGTR)
The staffs review of these DBAs are described in Sections 3.2.1, 3.2.2, 3.2.3 and 3.2.4 of this SE.
3.2 Confirmatory Calculations and Sensitivity Studies To perform the dose analyses, the licensee used the computer code, RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation, Version 3.03, as described in NUREG/CR-6604. This code estimates transport and removal of radionuclides and radiological consequences at selected receptors. The NRC staff reviewed all of the initial conditions, inputs, and assumptions used in the licensees RADTRAD files for each of the four DBAs. The staff compared the RADTRAD inputs to the changes proposed in the license amendment request and the licensing basis of each operating unit. The staff performed confirmatory calculations and various sensitivity studies of each of the modified DBAs using RADTRAD for all updated DBAs.
One aspect of the sensitivity studies performed was focused on incorporating all of the changes made to the DBA analyses since last docketed calculations, including all 50.59 process changes. The sensitivity studies provided an understanding of the magnitude of change to the final calculated dose values at the receptor locations, while using the previously approved atmospheric dispersion factors. In addition, the sensitivity studies provided insight as to how the calculated dose values change with different input parameters when implementing the changes since the last docketed calculations, including all changes conducted under the 50.59 process. These sensitivity studies note that the values obtained by NRC staff show less than a minimal increase in dose in each of the analyzed DBAs with all of the changes are incorporated into the calculations. These sensitivity studies do not evaluate any of the individual 50.59 process changes and do not supplement nor substitute the NRC inspection procedures associated with licensee controlled 50.59 process and evaluations. These sensitivity studies were performed to gain an in-depth understanding of the RADTRAD inputs and for the staff to further understand the impact of the changes performed under the licensee controlled 50.59 process.
3.2.1 Salem LOCA The current LOCA radiological consequence analysis, provided in Salem UFSAR Section 15.4.1, Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident),
(ML24130A164) is based on the accident source term described under 10 CFR 50.67.
Consistent with the alternative source term (AST) application and implementation as described in RG 1.183, the licensee assumed that the inventory of fission products is based on the maximum full power operation of 3,632 megawatts thermal (MWt), which is 1.05 times the current licensed thermal power level of 3,459 MWt.
Salem LOCA Analysis:
The staff reviewed the initial conditions, inputs and assumption in the dose consequence analysis and found that the majority of the inputs to the calculations discussed in the LAR are part of the licensing basis for both Salem, Units 1 and 2 through previously docketed calculations supporting earlier license amendment requests, or through use of the licensee controlled 50.59 process and subsequent incorporation into the current version of the licensees UFSAR. In the supplement dated April 5, 2024, the license identified the input variables that significantly impact dose calculations that have been changed since the last docketed calculation (LDC).
The NRC staff reviewed the LAR and documents associated with the plants licensing basis and found that with two exceptions, all initial conditions, inputs, and assumptions used in the licensees accident analysis are contained in both Salem Unit 1 and Unit 2s licensing basis. The first change is the change to the atmospheric dispersion factor for the EAB due to the change in the EAB boundary, which is discussed in section 3.3.4 of the LAR. The second change is comprised of changes to refueling water storage tank (RWST) contribution to EAB dose calculations.
RWST Back-leakage Parameters:
LDC-Post-LOCA EAB, Low Population Zone (LPZ), and Control Room (CR) Doses - AST Calculation RWST Back-leakage = 100 cc/hr (0.5 gpm in USFAR)
RWST Minimum Air Volume = 35,806 ft3 LAR-Table 3.2-4 Data for LOCA Model RWST Back-leakage = 2.9 gpm RWST Minimum Air Volume = 51,696 ft3 The RWST back-leakage parameter changes from the LDC to the values used in the LAR are conservative and would result in a higher contribution to dose at the receptor locations.
Specifically, the licensee calculated the RWST back-leakage contribution to the LOCA dose is 1.7 x 10-2 roentgen equivalent man (rem) (0.017 rem) total effective does equivalent (TEDE).
The RWST back-leakage contribution to the total LOCA dose of 6.86 rem TEDE (with an allowable limit of 25 rem TEDE), is not significant. Based on its review, the NRC staff finds that this change is acceptable in that it is based on a more conservative set of assumptions, which would lead to higher calculated dose. The below dose summary table provides the licensees breakdown of contributors to the updated EAB dose calculation.
Dose Summary for Salem LOCA Post-LOCA Activity Release Path EAB Post-LOCA TEDE (rem)
Containment Leakage Sprayed Region 3.02 Unsprayed Region 1.53 Total 4.55 ESF Leakage 2.06 Containment Relief Line Release 3.9795E-04 RWST Back-Leakage (2.9 gpm back-leakage) 0.017 Containment Shine 0.223 Total 6.86 Current Licensing Basis 4.1 Allowable TEDE Limit 25 The NRC staff performed independent review of all inputs, assumptions, and initial conditions in the Salem LOCA accident analysis provided in LAR dated September 6, 2023, and the supplement dated April 3, 2024. The staff performed independent confirmatory calculations using RADTRAD Version 5.0.3, as necessary, to ensure a thorough understanding of the licensee's methods and results. Based on the above, the NRC staff finds that the EAB radiological doses for the Salem LOCA accident analysis meet the applicable regulatory dose criteria in 10 CFR 50.67 and are, therefore, acceptable.
3.2.2 Salem Steam Generator Tube Rupture The current LOCA radiological consequence analysis, provided in Salem UFSAR Section 15.4.4, Steam Generator Tube Rupture is based upon an accident in which a complete severance of a single steam generator tube occurs. The accident is assumed to take place at power with the reactor coolant contaminated with fission products corresponding to continuous operation with a limited amount of defective fuel rods. The licensee assumes a loss of off-site power, since this renders the main condenser unavailable, and the operators cool the plant by releasing steam to the environment through the atmospheric release valves. The licensee assumes that primary-to-secondary flow continues for 30 minutes, at which time the effected SG is isolated and the reactor coolant system is depressurized. Appendix F of RG 1.183 identifies acceptable radiological analysis assumptions for a SGTR.
Salem SGTR Analysis The licensees accident evaluation analyzes the accident with two different initial conditions of the reactor coolant activity concentrations: the first initial condition is corresponding to a pre-accident iodine spike, and the second analysis assumes a concurrent iodine spike. There is no fuel failure during the SGTR event. The NRC staff reviewed the licensees initial conditions, inputs and assumption in the dose consequence analysis and found that the majority of the inputs to the calculations discussed in the LAR are currently part of the Salem Unit 1 and Salem Unit 2s licensing basis through previously docketed calculations supporting earlier license amendment requests, or through use of the licensee controlled 50.59 process and subsequent incorporation into the current version of the licensees UFSAR.
In the LDC, the SGTR dose consequence analysis of the EAB of Salem, Units 1 and 2, are combined. The licensee has revised the calculation under the 50.59 process and separated the units into individual RADTRAD calculations. This provided a more accurate calculation of the dose consequence analysis. The separation of calculations is now part of the licensing basis for the SGTR dose calculation for Salem, Units 1 and 2.
There are three changes to the Salem, Units 1 and 2 SGTR accident analysis. The first change is the change in EAB atmospheric dispersion factor which is discussed in section 3.3.4 of the LAR.
The second change is a change in the liquid iodine partition coefficient. The licensee currently uses values for liquid iodine partition coefficient which are more conservative than the values contained in RG 1.183. The change in the liquid iodine partition coefficient in the LAR is made to conform with Appendix F of RG 1.183, Assumptions for Evaluating the Radiological Consequences of a PWR Steam Generator Tube Rupture Accident. Regulatory Position 5.6 of Appendix F directs you to Regulatory Positions 5.5 and 5.6 of Appendix E to address transport of radionuclides. As stated in Regulatory Position 5.5.4 of RG 1.183:
The radioactivity in the bulk water is assumed to become vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for iodine of 100 may be assumed. The retention of particulate radionuclides in the steam generators is limited by the moisture carryover from the steam generators.
The NRC staff finds that the licensees proposed change to less conservative values for iodine partition coefficients is acceptable in that it conforms with Appendix F of RG 1.183. The staff notes that these updated partition coefficients are used in the dose consequence analysis.
The third change is the addition of three initial conditions to the licensees dose consequence analysis used to calculate a flashing fraction (RCS nominal operating pressure, RCS outlet temperature, and Design steam pressure). The volume of water that flashes in a ruptured steam generator immediately flashes to vapor and is assumed to be released without decontamination, and therefore, no credit is taken for scrubbing as the vapor rises through the bulk water of the steam generator. The Salem Unit 1 and 2 SGTR design basis radiological analyses do not currently model flashing of the primary coolant released from the ruptured tube in the faulted steam generator, which results in an immediate and unfiltered release. The NRC staff finds that use of a flashing fraction in the SGTR analysis is acceptable because it aligns with the guidance in RG 1.183, and it results in a more conservative analysis.
For this DBA analysis, the licensee updated the dose values for the control room, EAB and LPZ.
All initial conditions, inputs, and assumptions described in section 3.3 Salem Steam Generator Tube Rupture (SGTR) Accident other than those three changes discussed above, are part of the Salem Unit 1 and Salem Unit 2 licensing basis and have been reviewed and approved in previous LARs or have been incorporated in the licensing basis through other approved regulatory processes.
The NRC staff performed independent review of all inputs, assumptions, and initial conditions in the Salem Unit 1 and Unit 2 SGTR accident analysis provided in LAR and the supplement to the LAR dated April 5, 2024. The tables below provide the licensees breakdown of contributors to the updated dose calculations for the Salem Unit 1 and Unit 2 SGTR Accident pre-accident and concurrent iodine spikes. The staff performed independent confirmatory calculations using RADTRAD Version 5.0.3, as necessary, to ensure a thorough understanding of the licensee's methods and results. Based on the above, the NRC staff finds that the CR, EAB, and LPZ radiological doses for the Salem SGTR accident analysis meet the applicable regulatory dose criteria contained in 10 CFR 50.67 and, are therefore, acceptable.
Salem, Unit 1, SGTR Accident-Pre-accident Iodine Spike TEDE (rem)
Receptor Location CR EAB LPZ P-T-S Iodine Release 0.009 0.032 0.004 P-T-S Iodine Release (Flashing) 1.220 2.100 0.198 SC Liquid Iodine Release 0.001 0.001 0.000 Noble Gas Release 0.009 0.051 0.005 Total 1.240 2.184 0.207 Current Licensing Basis 0.61 2.06 0.30 Allowable TEDE Limit 5
25 25 Salem, Unit 1, SGTR Accident-Concurrent Iodine Spike TEDE (rem)
Receptor Location CR EAB LPZ P-T-S Iodine Release 0.025 0.033 0.007 P-T-S Iodine Release (Flashing) 0.368 2.080 0.196 SC Liquid Iodine Release 0.001 0.001 0.000 Noble Gas Release 0.009 0.051 0.005 Total 0.403 2.165 0.208 Current Licensing Basis 0.40 1.49 0.22 Allowable TEDE Limit 5
2.5 2.5 Salem, Unit 2, SGTR Accident-Pre-accident Iodine Spike TEDE (rem)
Receptor Location CR EAB LPZ P-T-S Iodine Release 0.009 0.039 0.004 P-T-S Iodine Release (Flashing) 0.994 2.250 0.172 SC Liquid Iodine Release 0.001 0.001 0.000 Noble Gas Release 0.008 0.058 0.005 Total 1.010 2.350 0.180 Current Licensing Basis 0.59 2.00 0.29 Allowable TEDE Limit 5
25 25 Salem, Unit 2, SGTR Accident-Concurrent Iodine Spike TEDE (rem)
Receptor Location CR EAB LPZ P-T-S Iodine Release 0.023 0.039 0.006 P-T-S Iodine Release (Flashing) 0.296 2.100 0.160 SC Liquid Iodine Release 0.001 0.001 0.000 Noble Gas Release 0.008 0.058 0.005 Total 0.328 2.200 0.171 Current Licensing Basis 0.37 1.37 0.20 Allowable TEDE Limit 5
2.5 2.5 3.2.3 Hope Creek LOCA The current LOCA radiological consequence analysis, provided in the Hope Creek UFSAR Section 15.6.5, Loss-of-Coolant Accident Resulting from the Spectrum of Postulated Piping Breaks Within the Reactor Collant Pressure Boundary Inside Primary Containment, is based on the accident source term described under 10 CFR 50.67. As consistent with the AST application and implementation, the licensee assumed that the inventory of fission products is based on the maximum power level of 3,917 MWt. This reactor power level is the current licensed power level of 3,840 MWt plus 2 percent for instrument uncertainty consistent with RG 1.183.
Hope Creek LOCA Analysis:
The NRC staffs review of the initial conditions, inputs, and assumption in the dose consequence analysis contained LAR and supplement identified one change from the LOCA analysis in the Hope Creek UFSAR. This change is the updated EAB atmospheric dispersion factor which is discussed in Section 3.3.4 of the LAR. The staffs review of the LAR and supplements did not identify any other changes from the Hope Creek LOCA analysis in the UFSAR. The below dose summary table provides the licensees breakdown of contributors to the updated EAB dose calculation.
EAB Dose Summary for a Hope Creek LOCA Post-LOCA Activity Release Path EAB Post-LOCA TEDE (rem)
Containment Leakage 1.58 ESF Leakage 2.25 MSIV Leakage 9.11 Total 12.94 Current Licensing Basis 3.02 Allowable TEDE Limit 25 The NRC staff performed independent review of all inputs, assumptions, and initial conditions in the Hope Creek LOCA accident analysis provided in LAR and the supplement dated April 5, 2024. The staff performed independent confirmatory calculations using RADTRAD Version 5.0.3, as necessary, to ensure a thorough understanding of the licensee's methods and results. Based on the above, the NRC staff finds that the EAB radiological doses for the Hope Creek LOCA accident analysis meet the applicable regulatory dose criteria contained in 10 CFR 50.67 and are, therefore, acceptable.
3.2.4 Hope Creek MSLB The current MSLB radiological consequence analysis, provided in the Hope Creek UFSAR Section 15.6.4, Steam System Piping Break Outside Containment, is based on the accident source term described under 10 CFR 50.67. Consistent with the NRC-approved AST application and implementation as described in RG 1.183, the analysis of the MSLB accident assumes that no fuel rod failures occur because of the transient. With no or minimal fuel damage postulated for the limiting event, the released activity is assumed to be the maximum coolant activity allowed by the technical specification. Two cases are considered: a maximum reactor coolant equilibrium iodine concentration case and a pre-accident iodine spike case.
Hope Creek MSLB Analysis:
The NRC staffs review of the LAR and its supplement identified one change contained in the LAR which is different from the MSLB analysis in the UFSAR. This change is the updated EAB atmospheric dispersion factor which is discussed in section 3.4. The staffs review of the LAR and its supplements did not identify any other changes from the Hope Creek MSLB analysis in the UFSAR. The below table provides the licensing basis MSLB EAB dose values, and the updated values contained in this LAR.
Dose Summary for the MSLB Accident TEDE (rem)
Maximum Equilibrium Iodine Concentration EAB 0.241 Current Licensing Basis 0.055 Allowable TEDE Limit 2.5 TEDE (rem)
Pre-Accident Iodine Spike EAB 4.05 Current Licensing Basis 0.915 Allowable TEDE Limit 25 The NRC staff performed independent review of all inputs, assumptions, and initial conditions in the Hope Creek MSLB accident analysis provided in LAR and the supplement dated April 5, 2024. The staff performed independent confirmatory calculations using RADTRAD Version 5.0.3, as necessary, to ensure a thorough understanding of the licensee's methods and results. Based on the above, the NRC staff finds that the EAB radiological doses for the Hope Creek MSLB accident analysis meet the applicable regulatory dose criteria contained in 10 CFR 50.67 and are, therefore, acceptable.
3.3 Meteorological Measurements and Related Dispersion Model Input Data Considerations A single onsite meteorological monitoring program supports the adjacent Salem and Hope Creek facilities. The monitoring program consists of a 91.4-m (300-ft) primary tower and a 10-m (33-ft) backup tower. Section 3.1.1 of the LAR briefly summarizes the primary towers location, the weather elements measured, and their monitoring elevations. Parameters relevant to the atmospheric dispersion modeling analyses for the Hope Creek and Salem facilities for this LAR include 10-m (33-ft) level wind speeds and wind directions and atmospheric stability class based on temperature differences (or vertical delta-T) between the 46-m minus the 10-m (150-ft minus the 33-ft) measurement levels.
Section 3.1 of the LAR indicates that the meteorological data was obtained and processed using existing station onsite collection methods and guidance in NRC RG 1.23." However, the NRC staff notes that while the LAR references the current version of RG 1.23 (i.e., Revision 1),
which is applicable to both facilities, the Updated Final Safety Analysis Report (UFSAR) for Hope Creek appears to be inconsistent relative to the LAR text including several discrepancies that cite earlier revisions of RG 1.23 and related review guidance. In the supplement dated October 30, 2023, PSEG stated that these discrepancies in the Hope Creek UFSAR will be updated via the PSEG Corrective Action Program.
For this LAR, PSEG used onsite meteorological data for a 3-year period of record (POR) from January 1, 2019, through December 31, 2021. The October 30, 2023, supplemental information indicated that data recoveries were well above the 90 percent criterion in Regulatory Position 5 of RG 1.23. The NRC staff notes that this criterion applies to the individual dispersion modeling-related weather elements (i.e., wind speed, wind direction, and atmospheric stability) as well as to their joint (concurrent) recovery during individual years and to the composite POR.
The licensees LAR Section 3.1.1 indicates that no data substitution was necessary. Further, Subsection 3.1.1.1 of the LAR indicates that wind data for the 3-year POR is representative of long-term conditions based on comparisons with wind roses from two previous monitoring periods. Therefore, the NRC staff finds that the 3-year POR of meteorological data used for this LAR is acceptable.
Nevertheless, the NRC staff observed during its review that the LAR submittal is silent regarding the type of wind speed and wind direction instrumentation used to collect this dispersion model input data. Information in the Salem and Hope Creek UFSARs make mention of mechanical-type instruments regarding wind measurements and their characteristics (e.g., wind vanes and propellors) as part of the onsite meteorological monitoring system. The NRC staff is aware that many facilities in the existing fleet now use ultrasonic anemometer technology for making wind measurements and, as a result, needed to confirm the type(s) of wind instrumentation used during the 2019 to 2021 POR.
In its supplement dated October 30, 2023, PSEG indicated that the wind speed data during this POR was based on ultrasonic anemometer measurements and that it was vector averaged.
PSEG also stated that [u]se of vector wind speed averaging has been assessed to provide conservative results and does not impact the resultant X/Qs [i.e., relative concentrations]. The NRC staff finds that vector-averaged wind speeds can be more conservative as they are either less than or equal to scalar-averaged values over the same time period. This conservatism would be because X/Qs estimated by Gaussian dispersion models are inversely proportional to the wind speed and, as a result, their input to dose calculations would also be the same or conservatively higher. As a result, the NRC staff finds PSEGs input of vector-averaged wind speed data to the dispersion modeling analyses for this LAR acceptable.
The NRC staff notes that ultrasonic anemometers, because of the measurement technology, typically determine both wind speed and wind direction as vector averages. However, scalar (or numerical) averages, or their equivalent, can also be derived from the measurement data.
RG 1.23 references the then current version of an American Nuclear Society (ANS) standard (i.e., ANSI/ANS-3.11-2005, Determining Meteorological Information at Nuclear Facilities).
Section 5.3.1 of that standard indicates, in part, that scalar averaging of wind speed and wind direction measurements is to be used when those data are input to straight-line, Gaussian dispersion models such as the PAVAN code (PSEGs use of that model is discussed later).
The October 30, 2023, supplemental information remained silent as to the wind direction data averaging approach. PSEG provided the NRC staff with additional clarifications on December 21, 2023 (ML23355A273), including a statement that [t]he hourly mean wind directions used as input to the PAVAN dispersion models for the new PSEG EAB are unit vector calculations that are determined as sampled wind directions during an averaging period weighted by a 1 m/sec [meter per second] wind speed. The NRC staff recognizes that such unit vector averages are essentially equivalent to scalar-averaged wind directions as opposed to true vector averages where the wind directions are weighted by concurrent wind speeds during sampling over the averaging period. The NRC staff finds that PSEGs input of unit vector averaged wind direction data to the dispersion modeling analyses complies with the intent of the ANS standard referenced in RG 1.23 and is, therefore, considered acceptable.
While outside the scope of this LAR, the NRC staff notes that true vector-averaged wind directions, if input to dispersion modeling analyses for siting and/or adjusting radiation monitoring locations, could influence the accuracy of where such locations are placed. This potential depends on the wind speed characteristics of the area and whether a Gaussian dispersion model would be used but should be considered if the Salem and/or Hope Creek radiation monitoring program(s) is/are revised as a result of the proposed change to the over land portion of the EAB associated with this LAR.
The NRC staff noted that the LAR presented different renderings of the sequence of measurement levels between which vertical delta-T measurements were made on the primary meteorological tower. Section 3.1.1 states that hourly T [delta-T] from 10 m - 46 m (33 ft. - 150 ft.), 10 m - 60 m (33 ft. -
197 ft.), and 10 m - 91 m (33 ft. - 300 ft.) was included and that JFDs [joint frequency distributions of wind speed, wind direction, and atmospheric stability class] were produced using wind speed and direction measured at 10 m (33 ft.) and the Pasquill stability class based on 10 m - 46 m (33 ft. - 150 ft.) T. On the other hand, Subsection 3.1.1.1 states that JFDs were constructed using hourlyT between 46 meters and 10 meters (150 ft.
and 33 ft.).
RG 1.23 defines vertical temperature difference as the upper level temperature measurement minus the lower level temperature measurement. In their request for information (RAI) responses of April 26, 2024, PSEG stated the following:
The T value is an absolute value derived from the difference in sensed temperatures at each of the listed meteorological tower elevations. The order in which the elevations are expressed have no bearing on the numerical value of T since the dash between each elevation is intended to mean to and not a minus sign (i.e., a negative elevation). As such, the resultant value of the Ts determined for each of the listed ranges of elevations are the same regardless of the order in which the elevations are described in each range of Ts.
To clarify, the NRC staff agrees with PSEGs statement that [t]he order in which the elevations are expressed have no bearing on the numerical [i.e., absolute] value of T. However, the NRC staff notes that the order of the elevations used in the calculation does affect the sign associated with the temperature difference, and the sign is directly related to the stability class designations (A through G) defined in Table 1 of RG 1.23.
The NRC staff relied on the statement in Subsection 3.1.1.1 of the LAR as to PSEGs approach for classifying atmospheric stability. Given that, it still was not clear from the hourly meteorological data listings for the POR, provided by PSEG as MS Excel files at NRCs request, if the delta-T values in the files represented values as measured or if they were already adjusted to the 100-m vertical interval in Table 1 of RG 1.23. This aspect of the NRC staffs review was necessary to verify the JFD input before being input to the NRC staffs own confirmatory PAVAN dispersion modeling. Checks of the hourly data as reported consisted of unit conversions (i.e.,
from Fahrenheit to Celsius) and an assumption that the temperature differences (upper minus lower elevations) measured between the 46-m to 10-m (150-ft to 33-ft) levels persisted through 100 m consistent with the RG 1.23 stability classification guidance.
PSEGs stability class counts in the annual and 3-year composite JFDs provided in the supplement dated October 30, 2023, were checked against the NRC staffs review of the hourly meteorological data after the staffs indicated adjustments. The NRC staff found that the results were in reasonable agreement. With regard to PSEGs additional clarifications of December 21, 2023, on the earlier supplemental submittal, the NRC staff finds that PSEGs approach in the LAR is consistent with the approach used by the NRC staff in its confirmatory PAVAN dispersion modeling. The staff also notes that the wind speed classes in these JFDs meet the intent of increased resolution for lower wind speeds as suggested by Table 3 and Regulatory Position 6 of RG 1.23, which provide a suitable format for data compilation of wind speed and wind direction by atmospheric stability classes.
Finally, PSEGs use of the values from the lowest of the three vertical delta-T measurement intervals is consistent with the potential accident releases in the PAVAN modeling analyses as ground-level sources (i.e., with a default release height of 10 m) and is, therefore, considered acceptable.
Based on the above, the NRC staff concludes that the JFD summaries input to the dispersion modeling analyses for this LAR, as provided in PSEGs supplement dated October 30, 2023, and as developed from the onsite hourly meteorological data, are acceptable. However, the NRC staff notes that clarification of discrepancies related to the onsite meteorological monitoring program identified in the review of the supplement dated October 30, 2023, as well as the staffs review of additional clarifications by PSEG in their responses of December 21, 2023, indicate that PSEG will update these aspects relative to the current licensing bases of the respective UFSARs after NRCs approval of this LAR.
3.4 Offsite Atmospheric Dispersion Factors Modeling Analysis (POR)
As part of the LAR, PSEG also performed DBA-related atmospheric dispersion modeling to estimate X/Q values (dispersion factors) applicable to the proposed reduced distances to the Salem and Hope Creek EABs. The X/Qs provide direct input to the accident dose calculations used to demonstrate that the EAB dose limit in 10 CFR 50.67(b)(2)(i) will be met. The applicant used the PAVAN dispersion model to calculate these values. The PAVAN model implements RG 1.145, the associated users guidance in NUREG/CR-2858, and the technical basis document for the regulatory guide in NUREG/CR-2260.
The NRC staff notes that Section 3.1 of the LAR states [the Control Room [CR] and Low Population Zone (LPZ) X/Qs are not revised for this activity, as the distances to the Control Rooms and LPZ are not impacted. In their RAI responses of April 26, 2024, PSEG further clarified that because these distances did not change the original design basis CR and LPZ X/Qs continue to be used for the SGTR [Steam Generator Tube Rupture] analysis.
Although CR and LPZ doses are revised for the SGTR accident due to changes in the associated radiological inventories, the staff finds that the licensees statements and the limited scope of the dispersion modeling analyses with respect to the EAB are reasonable.
PSEG indicated in their supplement dated October 30, 2023, that the PAVAN code was procured from the Radiation Safety Information Computational Center in 2004 as developed by the Pacific Northwest Laboratory for the NRC. PSEG also stated in that supplemental submittal that PAVAN implements the assumptions outlined in Section C of Regulatory Guide 1.145 and further that the program was validated in accordance with their consultants quality assurance program which included commercial grade dedication. Comparable results were obtained as part of the NRC staffs independent confirmatory PAVAN dispersion modeling checks; therefore, the NRC staff finds the applicants use of that version of the PAVAN code is acceptable.
As requested by NRC, PSEG provided separate PAVAN model input and output files for Salem and Hope Creek as part of the supplemental submittal. The 3-year composite JFDs for each of the seven stability classes, as part the December 21, 2023, clarifications discussed above, constituted a large portion of the PAVAN input files.
For Salem, other significant non-meteorological dispersion model inputs and related information provided by PSEG in the LAR submittal include:
Input parameters listed in Table 3.1 that is, no terrain adjustment factors applied, minimum vertical plane cross-sectional area of containment building (2399 square meters), height of that building above plant grade (58.3 m), assumed release height consistent with PAVAN model users guide and corresponding wind speed measurement height (10 m).
Graphical illustration in Figure 2.2-1 of current Salem and Hope Creek site plan, site boundary, and notations of respective minimum EAB distances on-land and over-water (i.e., in the Delaware River).
Graphical illustration in Figure 2.6-1 of proposed Salem minimum EAB distances on-land (including notations for minimum distances applicable to Salem, Unit 1, SGTR accident and for all other DBAs at Salem, Units 1 and 2), minimum EAB distance over-water, and locations of NJWP parcels within the site boundary.
List in Table 2.3-1 of current Salem design-basis accidents affected by LAR, that is -
LOCA, MSLB, SGTR, Locked Rotor Accident (LRA), Fuel Handling Accident (FHA), and Rod Ejection Accident (REA).
List in Table 2.6-1 of proposed Salem minimum EAB distances based on Figure 2.6-1 and the respective distances used for the PAVAN dispersion modeling analyses (i.e.,
equal to or less than the proposed minimum EAB distances) (and as further described in LAR Section 2.6.1 for the Salem, Unit 1, SGTR accident and for all other DBAs at Salem, Units 1 and 2).
Graphical illustration in Figure 3.1-2 of Salem minimum distances to proposed EAB as listed in Table 2.6-1 and as input to the PAVAN dispersion modeling analysis for the Salem, Unit 1, SGTR accident (i.e., the distance from the in-board and out-board Main Steam Safety Valve (MSSV) release points).
Graphical illustration in Figure 3.1-3 of Salem minimum distance to proposed EAB as listed in Table 2.6-1 and as input to the PAVAN dispersion modeling analysis for all other DBAs (i.e., based on the distance between the perimeter of a circle encompassing all other accident sources at Salem, Units 1 and 2, to the minimum distance to the proposed EAB).
For Hope Creek, other significant non-meteorological dispersion model inputs and related information provided by PSEG in the LAR submittal include:
Input parameters listed in Table 3.1 that is, no terrain adjustment factors applied, minimum vertical plane cross-sectional area of the containment building (2940 square meters), height of that building above plant grade (61.7 m), assumed release height and corresponding wind speed measurement height (10 m) consistent with PAVAN model users guide.
Graphical illustration in Figure 2.2-1 of current Salem and Hope Creek site plan, site boundary, and notations of respective minimum EAB distances on-land and over-water (i.e., in the Delaware River).
Graphical illustration in Figure 2.6-2 of proposed Hope Creek minimum EAB distances on-land (applicable to all accidents) and over-water, and locations of NJWP parcels within the site boundary.
List in Table 2.4-1 of current Hope Creek DBAs affected by LAR, that is - LOCA, FHA, Control Rod Drop Accident (CRDA), MSLB, Instrument Line Pipe Break (ILPB), and Feedwater Line Break Outside Containment (FWLB).
List in Table 2.6-1 of proposed Hope Creek minimum EAB distance based on Figure 2.6-2 and the distance used for the PAVAN dispersion modeling analysis (and as further described in LAR Section 2.6.1).
Graphical illustration in Figure 3.1-1 of Hope Creek minimum distance to proposed EAB as listed in Table 2.6-1 and as input to the PAVAN dispersion modeling analysis for all accidents (i.e., based on the distance between the perimeter of a circle encompassing all accident sources to the minimum distance to the proposed EAB).
The NRC staff reviewed the Hope Creek and Salem dispersion modeling analyses in accordance with SRP Section 2.3.4, RG 1.145, the associated PAVAN model users guidance in NUREG/CR-2858, and the technical basis document for that regulatory guide in NUREG/CR-2260. In general, the staff compared some of the above model inputs against information in the corresponding Salem and Hope Creek UFSARs (e.g., the current EAB and site boundary, containment heights and cross-sectional areas) and finds the model inputs to be acceptable. In particular, the NRC staff finds certain modeling approaches to be conservative, for example:
Methods described for determining the minimum distance to the proposed EAB for the applicable accident scenarios at each facility.
Assignment of the minimum EAB distance (and, therefore, the model-estimated X/Qs) to all direction sectors regardless of the variation in meteorological conditions by sector and the locations of individual release points.
Assumption of a ground-level release for all DBA sources consistent with the existing design basis.
Further, the NRC staff also notes that the PAVAN model input and output files for Salem and Hope Creek, provided by PSEG as part of the supplement dated October 30, 2023, were configured to account for enhanced building wake effects on plume dispersion. The NRC staff considers this approach to be appropriate considering the proximity of the DBA sources to the containment structure.
After review of relevant input information, the NRC staff performed a confirmatory PAVAN model run of the Salem dispersion analyses - that is, for the Salem, Unit 1, SGTR accident at an analytical EAB distance of 790 m and for all other DBA sources at Salem, Units 1 and 2, at the minimum analytical distance of 695 m. The confirmatory X/Q results for all short-term accident averaging intervals, including the controlling 0- to 2-hour average, were found to be essentially the same as the results in the PSEG run file.
Likewise, after review of relevant input information, the NRC staff performed a confirmatory PAVAN model run of the Hope Creek dispersion analysis - that is, for all DBA sources at the minimum analytical distance of 337 m. The NRC staff notes that the PSEG output file also contains an extraneous distance of 362 m that, according to the supplement dated October 30, 2023, was for sensitivity analysis purposes and not included or required as part of the LAR. As above, the confirmatory X/Q results for all short-term accident averaging intervals, including the controlling 0- to 2-hour average, were found to be essentially the same as the results in the PSEG run file.
Consistent with Regulatory Position 4 in RG 1.145 and the guidance in NUREG/CR-2260, the controlling 0- to 2-hour average X/Q is the higher of either the 0.5 percent sector-specific or the 5 percent overall site X/Qs estimated by the model. The PAVAN output from PSEG correctly identifies the controlling X/Q as the 0.5 percent sector-specific value. However, the NRC staff notes that the LAR refers to the controlling value in several places as the 95th percentile value. In their RAI responses of April 26, 2024, though, PSEG acknowledged that the 95th percentile value represents the larger of the 0.5% maximum sector value and the 5% overall site value. The response further indicates that this value is commonly referred to as the 95th percentile X/Q, that the terminology was then carried over to the LAR, and that the intent of that phrase is strictly for nomenclature purposes.
The NRC staff verified that the X/Qs input to the various dose calculations in the LAR submittal and as identified in the supplemental information of April 5, 2024, that summarizes current docketed doses compared to the revised doses in the LAR - that is:
for the Salem, Unit 1, SGTR accident and all other DBAs at Salem, Units 1 and 2, (i.e.,
1.97 E-04 and 2.44 E-04 sec/m3, respectively) and for Hope Creek, all DBAs (i.e., 8.14 E-04 sec/m3),
were the same internally and agreed with those of the staffs confirmatory modeling.
Therefore, based on the NRC staffs review of PSEGs onsite meteorological monitoring program for the POR of that data input to the PAVAN dispersion modeling analyses, and of various other inputs to and assumptions made for those analyses, the staff concludes that Section 3.1 of the LAR is acceptable for providing inputs to the dose consequence analyses of design basis accidents related to the proposed modification of the Exclusion Area Boundary for the Salem and Hope Creek.
3.5 Technical Conclusion The NRC staff reviewed the proposed submittal dated September 6, 2023, and its supplements that would modify the EAB for Salem Units 1 and 2 and Hope Creek. Based on the above, the NRC staff concludes that the proposed changes are acceptable and meet the requirements of 10 CFR 50.67. The NRC staff also concludes that the radiological consequences at the EAB, LPZ, and control room would continue to meet the acceptance criteria provided in 10 CFR 50.67, and 10 CFR Part 50, Appendix A, Criterion 19.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New Jersey State official was notified of the proposed issuance of the amendments on May 29, 2024. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change the format of the license or permit or otherwise make editorial, corrective or other minor revisions. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, as published in the Federal Register on October 31, 2023 (88 FR 74532), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: S. Meigan, NRR M. Mazaika, NRR Date: July 15, 2024
ML24145A177 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DEX/EXHB/BC NRR/DRA/ARCB/BC NAME JKim KEntz BHayes KHsueh DATE 5/24/2024 6/4/2024 5/16/2024 5/20/2024 OFFICE OGC - NLO NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME AGhosh-Naber HGonzález JKim DATE 7/2/2024 7/12/2024 7/15/2024