ML25016A326
| ML25016A326 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/19/2025 |
| From: | V Sreenivas NRC/NRR/DORL/LPL1 |
| To: | Blair B Energy Harbor Nuclear Corp |
| Sreenivas, V | |
| References | |
| EPID L 2024 LLA 0059 TSTF-569 | |
| Download: ML25016A326 (1) | |
Text
March 19, 2025 Barry N. Blair Vistra Operations Company LLC Beaver Valley Power Station Mail Stop P-BV-SSB P.O. Box 4, Route 168 Shippingport, PA 15077-0004
SUBJECT:
BEAVER VALLEY POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS 324 AND 214, RE: ADOPTION OF TECHNICAL SPECIFICATIONS TASK FORCE (TSTF) TRAVELER TSTF-569, "REVISE RESPONSE TIME TESTING DEFINITION," (EPID L-2024-LLA-0059)
Dear Barry Blair:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment Nos. 324 and 214 to Renewed Facility Operating License Nos. DPR-66 and NPF-73 for the Beaver Valley Power Station, Unit Nos. 1 and 2, respectively. These amendments are in response to your application dated May 7, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24129A016), requested revisions to the technical specifications (TS) to adopt Technical Specification Task Force (TSTF) Traveler TSTF-569, "Revise Response Time Testing Definition," which is an approved change to the Improved Standard Technical Specifications (ISTS).
The amendments revised the respective TSs to adopt TSTF-569, Revise Response Time Testing Definition, to revise the TS definitions for the engineered safety feature response time and reactor trip system response time.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions Federal Register monthly notice.
Sincerely,
/RA/
V. Sreenivas, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-334 and 50-412
Enclosures:
- 1. Amendment No. 324 to DPR-66
- 2. Amendment No. 214 to NPF-73
- 3. Safety Evaluation cc: Listserv
Amendment No. 324 Renewed Operating License DPR-66 VISTRA OPERATIONS COMPANY LLC ENERGY HARBOR NUCLEAR GENERATION LLC DOCKET NO. 50-334 BEAVER VALLEY POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 324 Renewed License No. DPR-66
- 1.
The Nuclear Regulatory Commission (the Commission) having found that:
A.
The application for amendment by Vistra Operations Company LLC, acting on its own behalf and as agent for Energy Harbor Nuclear Generation LLC (the licensees), dated May 7, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24129A016), 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 renewed operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; 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 TS as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-66 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 324, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license 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 Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 19, 2025 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2025.03.19 14:40:17 -04'00'
Amendment No. 214 Renewed Operating License NPF-73 VISTRA OPERATIONS COMPANY LLC ENERGY HARBOR NUCLEAR GENERATION LLC DOCKET NO. 50-412 BEAVER VALLEY POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 214 Renewed License No. NPF-73 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment by Vistra Operations Company LLC, acting on its own behalf and as agent for Energy Harbor Nuclear Generation LLC (the licensees), dated May 7, 2024, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24129A016) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 regulations of the Commission (except as exempted from compliance in Section 2.D.
below);
C.
There is reasonable assurance: (i) that the activities authorized by this renewed operating license can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.D. below);
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-73 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 214, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto and hereby incorporated in the license. Vistra Operations Company LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license 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 Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 19, 2025 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2025.03.19 14:40:53 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NOS. 324 AND 214 BEAVER VALLEY POWER STATION, UNITS 1 AND 2 RENEWED FACILITY OPERATING LICENSE NOS. DPR-66 AND NPF-73 DOCKET NOS. 50-334 AND 50-412 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Renewed Facility Operating License No. DPR-66 Remove Insert Page 3 Page 3 Renewed Facility Operating License No. NPF-73 Remove Insert Page 4 Page 4 Replace the following pages of the 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.
Appendix A, Technical Specifications Remove Insert 1.1 - 3 1.1 - 3 1.1 - 5 1.1 - 5 Amendment No. 324 Beaver Valley Unit 1 Renewed Operating License DPR-66 (3)
Vistra Operations Company LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Vistra Operations Company LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5)
Vistra Operations Company LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level Vistra Operations Company LLC is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 324, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Auxiliary River Water System (Deleted by Amendment No. 8)
Amendment No. 214 Beaver Valley Unit 2 Renewed Operating License NPF73 C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations set forth in 10 CFR Chapter 1 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 Energy Harbor Nuclear Corp. is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 214, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license.
Energy Harbor Nuclear Corp. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Definitions 1.1 1.1 Definitions ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its actuation setpoint TIME at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.
INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).
LEAKAGE LEAKAGE shall be:
a.
Identified LEAKAGE 1.
LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; 2.
LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or 3.
Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
b.
Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; and c.
Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.
Beaver Valley Units 1 and 2 1.1 - 3 Amendments 324 / 214
Definitions 1.1 1.1 Definitions QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore detector RATIO (QPTR) calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant as specified in the Licensing Requirements Manual, and shall not exceed 2900 MWt.
REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage.
The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
a.
All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM, and b.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
Beaver Valley Units 1 and 2 1.1 - 5 Amendments 324 / 214
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 324 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-66 AND RELATED TO AMENDMENT NO. 214 TO RENEWED FACILITY OPERATING LICENSE NPF-73 VISTRA OPERATIONS COMPANY, LLC BEAVER VALLEY POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-334 AND 50-412
1.0 INTRODUCTION
By application dated May 7, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24129A016), Vistra Operations Company, LLC (Vistra) the licensee) submitted a license amendment request (LAR) for the Beaver Valley Power Station (BVPS), Units 1 and 2. The amendments would revise technical specification (TS) definitions for engineered safety feature (ESF) response time and reactor trip system (RTS) response time that are referenced in surveillance requirements (SRs), hereafter, referred to as response time testing (RTT).
The proposed changes are based on Technical Specifications Task Force (TSTF) traveler TSTF-569, Revision 2, Revise Response Time Testing Definition, dated June 25, 2019 (ML19176A034). The U.S. Nuclear Regulatory Commission (NRC or the Commission) issued a final safety evaluation (SE) approving TSTF-569, Revision 2, on August 14, 2019 (ML19176A191). The description of the generic changes and their justification are contained in these two documents.
2.0 REGULATORY EVALUTATION 2.1 Description of Response Time Testing The RTS for BVPS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and the reactor coolant system (RCS) pressure boundary during anticipated operational occurrences and to assist the engineering safety feature actuation system (ESFAS) in mitigating accidents. The ESFAS initiates necessary safety systems, based on the values of selected unit parameters, to protect against violating core design limits and the RCS pressure boundary and to mitigate accidents.
The RTT verifies that the individual channel or train actuation response times are less than or equal to the maximum values assumed in the accident analyses. The RTT acceptance criteria are under licensee control. Individual component response times are not modeled in the accident analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor to the point at which the equipment reaches the required functional state (e.g., control and shutdown rods fully inserted in the reactor core).
2.2 Proposed Changes to the Technical Specifications BVPS Limiting Condition for Operation (LCO) 3.3.2, requires the ESFAS instrumentation for each Function in TS Table 3.3.2-1 "Engineered Safety Feature Actuation System Instrumentation," to be OPERABLE. To assure the LCO is met, surveillance requirement (SR) 3.3.1.14 requires the licensee to verify that ESF RESPONSE TIMES are within limits.
Similarly, BVPS LCO 3.3.1 requires the RTS instrumentation for each Function in TS Table 3.3.1-1 "Reactor Trip System Instrumentation" to be OPERABLE, and SR 3.3.1.16 requires the licensee to verify that RTS RESPONSE TIMES are within limits. Section 1.1 of the TS definitions for ESF RESPONSE TIME and RTS RESPONSE TIME, which states acceptable means to measure each response time, and provide an alternative that may be used "[i]n lieu of measurement."
In its application, the licensee stated that it requests adoption of NRC-approved TSTF-569, Revise Response Time Testing Definition. The only revision of TSTF-569 that is NRC approved is Revision 2. As described in Section 1, Summary Description, of Revision 2 of TSTF-569:
The proposed change revises the definitions to eliminate the requirement for prior NRC review and approval of the response time verification of similar components, while retaining the requirement for the verification to be performed using the methodology contained in Attachment 1, titled, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing. The proposed change will permit licensees to verify the response time of similar component types using the methodology contained in Attachment 1, without obtaining prior NRC approval for each component.
Accordingly, as shown in the LAR, the request would add an additional "in lieu of measurement" alternative to measuring ESF RESPONSE TIME and RTS RESPONSE TIME. The additional alternative for ESF RESPONSE TIME would be "[i]n lieu of measurement, response time may be verified for selected components provided... the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology." Similarly, for RTS RESPONSE TIME, "[i]n lieu of measurement, response time may be verified for selected components provided that... the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology."
The application stated that the licensee concluded that the justifications presented in TSTF-569 and the safety evaluation prepared by the NRC staff are applicable to BVPS and provide the justification for the amendment request.
2.2.1 Variations
- 1) The BVPS TS Bases number the SR differently in comparison to TSTF-569. TSTF-569 states SR 3.3.1.16 is for RTS Instrumentation; however, BVPS use SR 3.3.1.14.
Additionally, TSTF-569 states SR 3.3.2.10 is for Engineered Safety Feature Actuation System (ESFAS) Instrumentation; however, BVPS use SR 3.3.2.9.
- 2) The BVPS TS Bases do not itemize references for the documents referred to in TSTF-569 change. The change includes the documents instead of listing a reference.
- 3) The BVPS TS Bases contain requirements that differ from the Standard Technical Specifications on which TSTF-569 was based. Specifically, pertaining to RTS Instrumentation SR 3.3.1.14 and ESFAS Instrumentation SR 3.3.2.9, the BVPS TS Bases adds WCAP-15413, Westinghouse 7300A ASIC-Based Replacement Module Licensing Summary Report to provide bounding response times where 7300 cards have been replaced with ASICs cards. WCAP-15413 is referenced in the technical evaluation of TSTF-569 as a topical report that provides additional detailed justification that forms the basis for the methodology in Attachment 1 to TSTF-569, Revision 2. The applicability of this difference is specific to ASIC cards that have replaced 7300 cards.
2.3 Applicable Regulatory Requirements and Guidance Under 10 CFR 50.90, whenever a holder of a license wishes to amend the license, including technical specifications in the license, an application for amendment must be filed, fully describing the changes desired. Under Title 10 of the Code of Federal Regulations (10 CFR) 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commission's regulations.
The licensee's request involves adding an option used to satisfy surveillance requirements. As described in 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.
3.0 TECHNICAL EVALUATION
The NRC staff reviewed the request by comparing the licensees proposal against the changes described in TSTF-569, Revision 2. TSTF-569 is designed to make changes to NUREG-1431, Revision 5.0, Standard Technical Specifications, Westinghouse Plants, September 2021, Volume 1, Specifications (ML21259A155), and Volume 2, Bases (ML21259A159). The staff compared the technical specifications assumed in TSTF-569 with the current technical specifications for BVPS. The staff did not identify any material differences in the relevant technical specifications.
In the NRC SE for TSTF-569, the NRC staff concluded that the proposed change to TS Section 1.1 would eliminate required direct measurement RTT for selected pressure transmitter/sensor and protection channel components but does not eliminate required surveillance testing for the entirety of an instrument channel or the system as a whole (e.g., RTS). Therefore, the NRC staff found that the proposed change is consistent with the surveillance testing requirements of 10 CFR 50.36.
The NRC staff confirmed that the proposed change in this LAR has no effect on the design, fabrication, use, or methods of testing of the instrumentation and will not affect the ability of the instrumentation to perform the functions assumed in the safety analysis. Therefore, compliance with the design criteria General Design Criteria (GDC) 13 and GDC 21 or the equivalent plant-specific criteria is not affected.
The NRC staff finds that the methodology contained in TSTF-569, Rev. 2, Attachment 1, "Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only), provides a consistent, clear, and concise framework for determining that replacement components will operate at a level equivalent to that of the components being replaced. As such, using that methodology will assure that the necessary quality of the components is maintained and that the limiting conditions for operation will be met. Accordingly, approving the incorporation of that methodology into the licensing basis, amending the TS to allow usage of the approved methodology, and approving the LAR to use the methodology in TSTF-569, Rev. 2, results in TS that continue to meet the requirements of 10 CFR 50.36(c)(3). These requirements will be continued to be met by performing SR 3.3.1.16 while using the new "[i]n lieu of" option and will assure that associated aspects of LCO 3.3.1 and 3.3.2 will be met.
3.1 Variations The licensee proposed variations, described in section 2.2.1 of this safety evaluation, from the TS changes described in TSTF-569 and the applicable parts of the NRCs safety evaluation.
The NRC staff reviewed these variations and determined that the first two variations are editorial in nature and continue to meet the intent of TSTF-569. The proposed numbering difference and reference format do not affect the applicability of TSTF-569 to the BVPS TS. The last variation cites Westinghouse Commercial Atomic Power report-15413, Westinghouse 7300A ASIC-Based Replacement Module Licensing Summary Report, which the NRC staff evaluated and confirmed that this can provide additional detailed justification to describe RTS and ESFAS systems of different technologies. Also, this topical report can provide additional detailed justification for the methodology in Attachment to TSTF-569. Therefore, the proposed variations are acceptable.
3.2 Technical Conclusion Based on the preceding evaluation, the NRC staff concludes that the revised TSs proposed by the licensee will continue to meet 10 CFR 50.36(c)(3) and the GDC, as incorporated into the BVPS updated final safety analysis report. The NRC acknowledges the licensee-controlled conforming changes to the TS Bases.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the State of Pennsylvania official was notified of the proposed issuance of the amendments. On January 15, 2025, the State official confirmed that the Commonwealth had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on July 9, 2024 (89 FR 56442).
Accordingly, the amendments meet 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 amendments.
6.0 CONCLUSION
The NRC staff 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 Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: T. Sweat, NRR Dated: March 19, 2025
ML25016A326 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DSS/STSB/BC NAME VSreenivas KEntz SMehta DATE 01/16/2025 1/21/2025 12/23/2024 OFFICE NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME HGonzález VSreenivas DATE 03/19/2025 03/19/2025