ML24270A155
ML24270A155 | |
Person / Time | |
---|---|
Site: | Monticello ![]() |
Issue date: | 12/19/2024 |
From: | Ballard B Plant Licensing Branch III |
To: | Hafen S Northern States Power Company, Minnesota |
Ballard, B | |
References | |
EPID L-2024-LLA-0000 | |
Download: ML24270A155 (1) | |
Text
December 19, 2024 Shawn Hafen Site Vice President Northern States Power Company
- Minnesota Monticello Nuclear Generating Plant 2807 West County Road 75 Monticello, MN 55362
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT - ISSUANCE OF AMENDMENT NO. 213 RE: REVISE TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR), AND UPDATE NEUTRON FLUENCE METHODOLOGY (EPID L-2024-LLA-0000)
Dear Shawn Hafen:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 213 to Renewed Facility Operating License No. DPR-22, for the Monticello Nuclear Generating Plant (Monticello). The amendment consists of changes to the technical specifications (TSs) in response to your application dated December 29, 2023.
The amendment revises TS 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), to update the PTLR methodology to the Structural Integrity Associates methodology report SIR-05-044-A, Revision 1, Pressure Temperature Limits Report Methodology for Boiling Water Reactors. In addition, this amendment approves a new neutron fluence methodology for use at Monticello.
A copy of the Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Brent T. Ballard, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263
Enclosures:
- 1. Amendment No. 213 to DPR-22
- 2. Safety Evaluation cc: Listserv
NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 213 Renewed License No. DPR-22
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company dated December 29, 2023, 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, 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. DPR-22 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 213, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.
Additionally, the use of the TransWare Enterprises, Inc. Radiation Analysis Modeling Application (RAMA) neutron fluence methodology is approved for use with the Pressure Temperature Limits Report.
- 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 Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: December 19, 2024 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2024.12.19 10:55:24 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 213 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE DOCKET NO. 50-263 Renewed Facility Operating License No. DPR-22 Replace the following page of the Renewed Facility Operating License No. DPR-22 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.
REMOVE INSERT Page 3 Page 3 Technical Specifications Replace the following pages of 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.
REMOVE INSERT 5.6-5 5.6-5 Amendment No. 213
- 2.
Pursuant to the Act and 10 CFR Part 70, NSPM to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operations, as described in the Final Safety Analysis Report, as supplemented and amended, and the licensees filings dated August 16, 1974 (those portions dealing with handling of reactor fuel);
- 3.
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM 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.
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and
- 5.
Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess, but not separate, such byproduct and special nuclear material as may be produced by operation of the facility.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations 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 NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 2004 megawatts (thermal).
- 2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 213, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.
- 3.
Physical Protection NSPM shall implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 5.6.5 Monticello Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) a.
RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
1.
Limiting Conditions for Operation Section 3.4.9, "RCS Pressure and Temperature (PIT) Limits" 2.
Surveillance Requirements Section 3.4.9, "RCS Pressure and Temperature (PIT) Limits" b.
The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1.
SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," Revision 1, dated August 2013.
c.
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplements thereto.
5.6-5 Amendment No. 213
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 213 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263
1.0 INTRODUCTION
By application dated December 29, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23363A174), Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (NSPM, the licensee) requested changes to the technical specifications (TSs) for Monticello Nuclear Generating Plant (Monticello).
The proposed license amendment request (LAR) would revise TS 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), to update the PTLR methodology to the Structural Integrity Associates (SIA) methodology report SIR-05-044-A, Revision 1, Pressure Temperature Limits Report Methodology for Boiling Water Reactors.
The amendment requested the NRC staff approve the TransWare Enterprises, Inc. (TransWare)
Radiation Analysis Modeling Application (RAMA) neutron fluence methodology for use at Monticello.
2.0 REGULATORY EVALUATION
The U.S. Nuclear Regulatory Commission (NRC or Commission) established requirements in Title 10 of the Code of Federal Regulations (10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities), to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. The NRC staff evaluates the acceptability of a facilitys proposed pressure and temperature (P-T) limits based on the following NRC regulations and guidance.
2.1 Applicable Regulations Per Title 10 of the Code of Federal Regulations (10 CFR) Part 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 10 CFR Part 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 or construction permits to the extent applicable and appropriate. Both the common standards for licenses and construction permits in 10 CFR Part 50.40(a), and those specifically for issuance of operating licenses in 10 CFR Part 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.
The regulations in 10 CFR Part 50.36, Technical specifications, paragraph (a)(1), require that each operating license application for a production or utilization facility include proposed TSs and a summary statement of the bases for such specifications. Paragraph (b) requires the TS be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto.
The regulations in 10 CFR Part 50.36(c) require, in part, that TSs include the following categories related to facility operation: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls.
The regulations in 10 CFR Part 50.60, Acceptance criteria for fracture prevention measures for light-water nuclear power reactors for normal operation, require that all light-water nuclear power reactors meet the fracture toughness and material surveillance program requirements for the RCPB set forth in 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, and Appendix H, Reactor Vessel Material Surveillance Program Requirements.
The regulations in 10 CFR Part 50.61, Fracture toughness requirements for protection against pressurized thermal shock (PTS) events, requires, for pressurized water reactors, that the PTS reference temperature of the reactor vessel shell material be limited at the end of plant life to less than the PTS screening criterion of 270°F for plates, forgings, and axial weld materials, and 300°F for circumferential weld materials.1 The regulations in 10 CFR Part 50, Appendix G, establish fracture toughness requirements to maintain the integrity of the RCPB in nuclear power plants. P-T limit requirements for the reactor pressure vessel (RPV) are established in paragraph IV.A.2 and Table 1 of Appendix G.
Paragraph IV.A.2 and Table 1 specify that P-T limit curves and minimum temperature requirements for the RPV are defined by the operating condition (i.e., pressure testing or normal operation, including anticipated operational occurrences), the reactor vessel pressure, whether or not fuel is in the vessel, and whether the core is critical. In Table 1, the reactor vessel pressure is defined as a percentage of the preservice system hydrostatic test pressure. The requirements for both the P-T limit curves and the minimum temperature must be met for all normal operating and pressure test conditions. Additionally, 10 CFR Part 50, Appendix G, requires that applicable surveillance data from RPV material surveillance programs be incorporated into the calculations of the P-T limits, and that the P-T limits be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials.
The regulations in 10 CFR Part 50, Appendix H, require a material surveillance program to monitor fracture toughness properties of ferritic materials in the reactor vessel beltline region which result from exposure of these materials to neutron irradiation and the thermal environment.
1 If this criterion cannot be satisfied, 10 CFR 50.61 may allow continued operation if other specified criteria are satisfied.
2.2 Design Criteria Monticello was designed prior to the publishing of the Atomic Energy Commission (AEC)
General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, draft for comment, 32 FR 10214 (July 1967) (AEC Proposed GDC). Therefore, Monticello was not licensed to the current 10 CFR Part 50, Appendix A, GDC. The Monticello principal design criteria for the design, construction and operation of the plant are listed in the Monticello Updated Safety Analysis Report (USAR), Section 1.2, Principal Design Criteria. The licensees evaluation comparing the Monticello design basis to the AEC proposed GDCs of 1967 is presented in the Monticello USAR, Appendix E, Plant Comparative Evaluation with the Proposed AEC 70 Design Criteria.
The NRC staff reviewed the LAR to confirm that the intent of the requirements in the equivalent 10 CFR, Part 50, Appendix A, GDCs given below are met:
Criterion 14 - Reactor coolant pressure boundary. The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.
Criterion 15 - Reactor coolant system design. The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
The AEC Proposed GDC Criterion corresponding to 10 CFR Part 50, Appendix A, GDCs 14 and 15 can be found in AEC 70 GDC, Criterion 33 - Reactor Coolant Pressure Boundary Capability.
Criterion 31 - Fracture prevention of the reactor coolant pressure boundary. The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.
The AEC Proposed GDC Criterion corresponding to 10 CFR Part 50, Appendix A, GDC 31 can be found in AEC 70 GDC, Criterion 34 - Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention, and AEC 70 GDC, Criterion 35 - Reactor Coolant Pressure Boundary Brittle Fracture Prevention.
Criterion 32 - Inspection of reactor coolant pressure boundary. Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to
assess their structural and leak tight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.
The AEC Proposed GDC Criterion corresponding to 10 CFR Part 50, Appendix A, GDC 32 can be found in AEC 70 GDC, Criterion 36 - Reactor Coolant Pressure Boundary Surveillance.
2.3 Applicable Guidance Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, (ML003740284) describes the calculative procedures used to estimate and project neutron radiation embrittlement of low-alloy steels used in light-water-cooled reactor vessels, such as the nil-ductility transition RTNDT and adjusted reference temperature (ART).
The guidance in RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (ML010890301), provides guidance on methods for determining reactor pressure vessel fluence that are acceptable to the NRC staff. The guidance in RG 1.190 states that an acceptable neutron fluence calculation has the following attributes:
Fluence estimation using an appropriate calculational methodology, and Analytic uncertainty analysis identifying possible sources of uncertainty, and Comparisons with benchmark measurements and calculations from applicable test facilities including:
o Plant-specific operating reactor measurements, o Pressure vessel simulator measurements, o Calculational benchmarks.
The NRC Generic Letter (GL) 96-03, Relocation of Pressure and Temperature Limit Curves and Low Temperature Overpressure Protection System Limits (ML031110004), permits relocation of the P-T limits from the TS to a PTLR. GL 96-03 calls for licensees to (1) generate their P-T limits in accordance with an NRC-approved methodology, (2) comply with 10 CFR Part 50, Appendices G and H, (3) reference NRC-approved methodologies in the TS, (4) define the PTLR in TS Section 1.0, (5) develop a PTLR to contain the P-T limit curves, and (6) modify applicable sections of the TS accordingly.
Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, (ML14149A165), provides guidance on the scope and detail of information that should be provided in RPV fracture toughness and associated licensing applications in order to facilitate NRC staff review. The discussion includes P-T limits and P-T limit reports, as well as consideration of neutron fluence and structural discontinuities in the development of P-T curves.
The guidance in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP), Section 5.3.2, Revision 2, Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock, (ML070380185), provides an acceptable method for determining the P-T curves based on 10 CFR Part 50, Appendix G, and the methodology set forth in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Appendix G.
The guidance in NUREG 0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition (SRP), Chapter 16.0,
Technical Specifications, Revision 3, (ML100351425) also provides NRC staff with guidance for the review of TSs.
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the P-T limit methodology, LAR, licensees RPV material surveillance capsule program, the chemistry factor, ART calculations, the proposed P-T limit curves, neutron fluence calculations, and proposed changes to the TSs.
On January 9, 2023, the licensee submitted the Monticello subsequent license renewal application (SLRA) (ML23009A353). The licensee performed neutron fluence projections for the Monticello RPV and reactor vessel internal components and plant structures using the RAMA fluence methodology for the projected 20-year period of subsequent extended operation (i.e.,
through 72 EFPY (or 80 calendar years).
By letter dated July 11, 2023 (ML23193B026), the licensee submitted a third supplement to the Monticello SLRA containing a TransWare topical report discussing the fluence methodology and qualification of that model for Monticello. In the proposed LAR, the licensee stated the analysis performed during the Monticello SLRA was applicable to update the PTLR under the current licensing basis as the analysis for the period of subsequent extended operation bounds the analysis through the end of the current operating license. Therefore, the licensee submitted this LAR to seek approval for the use of the TransWare topical report and update the methodology to develop a new PTLR for use at Monticello through the end of the current operating license.
The current PTLR, Revision 1, submitted by letter dated August 28, 2014 (ML14246A206) is applicable for Monticello through the end of the initial license renewal period (i.e., up to 54 effective full power years (EFPY), or 60 calendar years).
3.1 Pressure-Temperature Limit Methodology The regulations of 10 CFR Part 50, Appendix G require the use of the ASME Code,Section XI, Appendix G, which specifies a procedure for calculating P-T limit curves that is based on linear elastic fracture mechanics. The key material property used in the P-T limit curve calculation is the fracture toughness (KIC). As specified in paragraph G-2210 of the ASME Code,Section XI, KIC is an exponential function of the difference in metal temperature at the postulated crack tip and the RTNDT for the ferritic RPV material.Section IV.A of 10 CFR Part 50, Appendix G, requires that the values of RTNDT for RPV beltline materials used in the P-T limit calculations account for the effects of neutron irradiation, and incorporate RPV surveillance material test data that are reported as part of the RPV materials surveillance program, which is required by 10 CFR Part 50, Appendix H.
The RTNDT value of RPV beltline materials will increase with operating time because of neutron irradiation, thereby, causing a shift in the RTNDT which leads to a corresponding shift in the P-T limit curves. This shift in the P-T limit curves will be more restrictive for the RPV operation to maintain the required safety margins to protect the RPV shell material against brittle fracture per 10 CFR Part 50, Appendix G.
The ASME Code,Section XI, Appendix G, specifies that the P-T limits be developed by postulating a flaw at the inside and outside diameter surface with a depth equal to 1/4 of the wall thickness of the RPV shell, and a length equal to 1.5 times the wall thickness of the RPV shell.
The critical locations in the RPV shell thickness (T) for calculating P-T limit curves are the 1/4T
and 3/4T locations, which correspond to the maximum depth of the postulated inside surface and outside surface defects, respectively.
The P-T limit curve calculations are based, in part, on the RTNDT of the RPV shell material, as specified in the ASME Code,Section XI, Appendix G. The RTNDT is the parameter for determining the critical stress intensity factor (fracture toughness, KIC) for the RPV shell material. The regulations of 10 CFR Part 50, Appendix G, require that RTNDT values for materials in the RPV beltline region must be adjusted to account for the effects of neutron irradiation.
The effects of neutron fluence on the RPV shell material can be determined by the ART, which is calculated using the guidance of RG 1.99, Revision 2. The ART is defined as the sum of the initial (unirradiated) RTNDT, the mean value of the shift in RTNDT caused by irradiation (RTNDT),
and a margin term. The RTNDT is a product of a chemistry factor and a fluence factor. The chemistry factor is dependent upon the amount of copper (Cu) and nickel (Ni) in the RPV shell material. The Cu and Ni values are used with Tables 1 and 2 of RG 1.99 to determine a chemistry factor per paragraph 1.1 of RG 1.99 for welds, plates and forgings. Paragraph 2.1 of RG 1.99 provides guidance on calculating the chemistry factor using surveillance capsule data.
The Cu and Ni values of the RPV shell material are obtained from vessel purchase order records, certified material test reports, or from values previously submitted to and approved by the NRC staff. The fluence factor is dependent upon the neutron fluence at the postulated flaw depths at the 1/4T and 3/4T locations. The margin term is used to account for uncertainties in the values of the initial RTNDT, the Cu and Ni contents, the neutron fluence and the calculational procedures.
3.2 License Amendment Request of the LAR describes the reason for the proposed changes in neutron fluence methodology, P-T limit curve development, and relevant TS pages. Enclosure 2 of the LAR discusses the PTLR, including the proposed P-T limits curves and reactor vessel material surveillance program. Enclosure 3 of the LAR discusses dated ART calculations. Enclosure 4 of the LAR provides the updated P-T limit curve and ART calculations for the 72 EFPY.
By letter dated February 27, 2013 (ML13025A155), the NRC staff approved the licensees previous LAR to revise and relocate the P-T limit curves to a PTLR from the Monticello TSs. The approved PTLR uses the General Electric methodology in NEDO-32983-A, Revision 2, Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations, dated January 2006 (ML072480121) to perform neutron fluence calculations.
In the LAR dated December 29, 2023, the licensee requested the NRC staff approve the use of the TransWare RAMA neutron fluence methodology to develop the PTLR at Monticello. After the NRC staff approves the TransWare RAMA methodology for use, the licensee will replace the prior neutron fluence methodology with the TransWare RAMA as the licensing basis methodology to estimate RPV neutron fluence at Monticello.
The December 29, 2023, LAR also request that the methodology listed under TS item 5.6.5.b.1 be revised from the 2007 (Revision 0) to the 2013 version (Revision 1) of Boiling Water Reactor Owners Group (BWROG) topical report, BWROG TP-11-022-A, Revision 1/SIA LTR SIR 044-A, Revision 1 (ML13277A557), Pressure Temperature Limits Report Methodology for Boiling Water Reactors.
In addition, the LAR included a revised PTLR that incorporated the results from the evaluation of the Monticello 120-degree surveillance capsule removed during the spring of 2021 refueling outage into the supporting analyses for the proposed P-T limits in the PTLR.
3.3 Reactor Vessel Material Surveillance Capsule Program The regulation of 10 CFR Part 50, Appendix H, allows licensees to use an integrated surveillance program (ISP) in which representative materials for the reactor are irradiated in one or more reactors of sufficiently similar design and operating features to permit accurate comparisons of the predicted amount of radiation damage.
The licensee stated that the Monticello is a member of the Boiling Water Reactor (BWR) Vessel and Internals Project, Integrated Surveillance Program (BWRVIP ISP), which is administered by the Electric Power Research Institute and the BWROG. The ISP combines the BWR surveillance programs into a single integrated program so that similar heats of materials in the surveillance programs of various BWR nuclear plants (i.e., host plants) can be used to represent the limiting materials in other BWR RPVs. The scope of the program is described in the BWRVIP ISP guidance, and the technical basis of the program is described in BWRVIP-78, BWR Vessel and Internals Project, BWR Integrated Surveillance Program Plan (ML003704011). The ISP capsule removal schedule is included in BWRVIP-86, Revision 1-A, BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program Implementation Plan (ML023190487).
The licensee stated that Monticello is currently operating under, and is licensed to use, the BWRVIP ISP during the initial renewed license period of extended operation up to 2030. The licensee further stated that it plans to adopt, for the subsequent license renewal period, BWRVIP-321-A, Boiling Water Reactor Vessel and Internals Project, Plan for Extension of the BWR Integrated Surveillance (ISP) Through the Second License Renewal (SLR)
(ML21152A130). The NRC staff noted that BWRVIP-321, Rev 1-A (ML23143A345) is available for generic use, which incorporates changes requested by Electric Power Research Institute (EPRI) letter dated March 29, 2022 (ML22091A218).
The licensee removed the surveillance capsule at the 120-degree azimuthal location of the RPV during the spring of 2021 refueling outage. The 120-degree capsule was irradiated from initial startup through 30 fuel cycles of operation. The licensee stated that this was the last of the three surveillance capsules installed in the Monticello reactor. In addition to the 120-degree capsule, the licensee previously removed the surveillance capsules at the 30-degree and 300-degree azimuthal locations of the RPV.
By letter dated October 28, 2022 (ML22304A092), the licensee submitted to the NRC, for its information, the test results of the 120-degree capsule as shown in BWRVIP-347, BWR Vessel and Internals Project, Testing and Evaluation of the Monticello 120° ISP(E) Surveillance Capsule (ML22304A093). The licensee stated that the test results for the 120-degree capsule will be added to the next revision of BWRVIP-135, BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations (ML17018A153).
Based on information provided in the subject LAR, the NRC staff noted that the Monticello reactor material surveillance program is in compliance with the regulations of 10 CFR Part 50, Appendix H, in terms of the integrated surveillance program, reporting and testing.
3.4 Chemistry Factor The chemistry factor is a key parameter in deriving ART values and is related to the chemical elements in the RPV shell material that affect the fracture toughness, notably copper (Cu) and nickel (Ni). The vessel beltline Cu and Ni values were obtained from the evaluation of the Monticello vessel plate, weld, and forging materials, as shown in Enclosure 3 of the submittal.
The NRC staff noted that although Enclosure 3 of the submittal calculates an outdated ART, the Cu and Ni values of the Monticello RPV shell material are still valid and consistent with the Monticello current licensing basis because Cu and Ni are material property values that are inherent in the vessel material and do not change as a result of neutron irradiation. The licensee stated that the updated chemistry factor evaluation included the results of three surveillance capsules as shown in Table 2 of Enclosure 4. Table 2 of Enclosure 4 shows that the chemistry factor for the bounding ART is 174°F which occurs at the lower intermediate shell plate, I-14, heat number C2220-1.
The NRC staff performed an independent calculation using paragraph 2.1 of RG 1.99, Revision 2 to evaluate the chemistry factor for the limiting material shell plate I-14, using the test data from the three surveillance capsules at the 30-degree, 120-degree, and 300-degree azimuthal locations of the RPV. The NRC staff verified that the chemistry factor of 174°F for the limiting material, shell plate I-14, is acceptable based on the data from the three surveillance capsules.
3.5 Adjusted Reference Temperature Section 4.2.3 of the Monticello SLRA discusses ART for RPV materials due to neutron embrittlement. Table 4.2.3-2 of the SLRA provides the limiting ART of 182.7°F at the 1/4T location at 72 EFPY for the plate I-14, heat number C2220-1. The licensee stated in the letter dated December 29, 2023, that the ART of 182.7°F that appears in Enclosure 3 of the current submittal does not include the latest test data from the 120-degree surveillance capsule. of the current submittal contains the updated ART and P-T curves calculations.
The NRC staff noted that P-T curves are developed based on a postulated flaw at the 1/4T and 3/4T locations of the reactor vessel. When combining pressure and thermal stresses, it is necessary to consider total stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw) because the tensile stress that causes crack propagation occurs in the interior and exterior reactor wall surfaces depending on the heat-up and cool down operation. The thermal gradient causes tensile stress in the inner wall during cool-down and in the outer wall during heat-up. However, as a conservative simplification, the licensee assumes the tensile stress occurring at the inside surface for both heat-up and cool-down. This results in applying the maximum tensile stresses at the 1/4T location flaw. The licensee stated that this approach is conservative because irradiation effects cause the allowable fracture toughness at the 1/4T location to be less than that at 3/4T for a given metal temperature. The licensee stated that this conservative approach does not cause operational difficulties because the BWR is at steam saturation conditions during normal operation. The NRC staff noted that for a given pressure, the coolant saturation temperature is well above the P-T curve limiting temperature.
As such, the material fracture toughness at a given pressure would exceed the allowable fracture toughness which is desirable. Based on the ASME Code,Section XI, Appendix G, the NRC staff verified that using the limiting ART at the 1/4T location to develop the P-T curves in lieu of using the ART at the 3/4T location is conservative, and therefore is acceptable.
Table 2 of Enclosure 4 of the current LAR shows the chemistry factor, initial RTNDT, neutron fluence, fluence factor, shift of RTNDT, margin, and the ART of all RPV shell materials at the 1/4 T location for 72 EFPY. Table 2 of Enclosure 4 shows that the lower intermediate shell plate, I-14, heat number C2220-1 has the highest, and thus limiting ART of 178.1°F at the 1/4T location for 72 EFPY based on a neutron fluence of 4.38E+18 neutron/cm2.
The NRC staff determined that the Monticello surveillance plate heat number C2220 is the ISP representative material for the Monticello target vessel, and the ISP data can be used directly in the evaluation of the ART for the vessel material because plate heat number C2220 is an exact match for the Monticello vessel. As such, the NRC staff performed an independent calculation using the methodology of RG 1.99, Revision 2, and verified that 178.1°F is the limiting ART at 72 EFPY and is acceptable because it is derived based on the chemistry factor of 174°F, the initial RTNDT of 27°F, the RTNDT of 134.1°F, and a margin of 8.5. The NRC staff determined that the licensees initial RTNDT of 27°F, the RTNDT of 134.1°F, and a margin of 8.5 are derived based on RG 1.99, Revision 2 and are, therefore, acceptable. The NRC staff finds that the ART at 72 EFPY bounds the ART at the 54 EFPY.
3.6 Proposed Pressure-Temperature Curves of the submittal provides the calculation for the P-T limit curves for the beltline, bottom head, and non-beltline regions of the RPV for 72 EFPYs in accordance with the guidance of the topical report BWROG TP-11-022-A, Revision 1/SIA LTR SIR-05-044-A, Revision 1. Because the NRC staff has approved the generic use of the BWROG topical report, the NRC staff evaluated whether the licensee followed the guidance in the BWROG topical report to develop the proposed P-T limit curves in the Monticello PTLR.
The NRC staff noted that the Monticello P-T curves consist of three curves representing three operating conditions: Pressure Test (Curve A), Normal Operation - Core Not Critical (Curve B),
and Normal Operation - Core Critical (Curve C). Each of these curves contains the pressure and temperature limits for each of the following three regions of the RPV; i.e., the beltline region, bottom head region, and non-beltline region, including the top head flange. The curves for each region are combined to generate a composite, bounding curve for all RPV regions. The operator is required to heat-up, cool down, and perform pressure tests within the composite curve in Curves A, B and C.
3.6.1 Extended Beltline Region NRC RIS 2014-11 clarifies that the development of P-T limit curves should consider the stresses in RPV shell regions that are outside of the beltline region and outside of the structural discontinuities of the RPV shell. The licensee considered feedwater nozzle (non-beltline) and recirculation inlet nozzle (beltline) as part of the extended beltline region because these nozzles may incur stresses that could be significant and should be considered in the development of the P-T curves. The licensee used the ANSYS finite element computer program to develop the stress distributions based on pressure and thermal loadings through the feedwater nozzle and recirculation inlet nozzle as well as the vessel shell. The licensee used these stress distributions to develop the stress intensity factors to construct the P-T curves for the feedwater and recirculation inlet nozzles as shown in Enclosure 4 of the submittal. The NRC staff verified that the licensee considered the stress intensity factors from the feedwater nozzle and recirculation inlet nozzle as part of the P-T curve development and, therefore, the licensee has satisfied the guidance in NRC RIS 2014-11.
The licensee stated, the ANSYS program was controlled under the vendors 10 CFR Part 50, Appendix B, quality assurance program for nuclear quality-related work.
3.6.2 Minimum Temperature Requirements Table 1 of 10 CFR Part 50, Appendix G, requires that minimum temperature be applied to the P-T curves at 20 percent of the RPV pre-service hydrostatic test pressure in the following operating conditions: the hydrostatic and leakage test, core not critical, and core critical. The NRC staff noted that the RPV pre-service hydrostatic test pressure is 1563 per square inch (psi) for Monticello, as stated in Section 5.3 of Enclosure 4 of the submittal. The 20 percent of the hydrostatic pressure is 313 psi. The NRC staff noted that Curves A, B, and C, have included the condition where, at 313 psi, the minimum temperature requirements are implemented.
Table 1 of 10 CFR Part 50, Appendix G, also requires that (1) core critical P-T limits be 40°F above any Curve A or Curve B limits at all pressures, and (2) above the 20 percent pressure transition point, the Curve C temperatures must be either the reference temperature (RTNDT) of the closure flange region plus 160°F, or the temperature required for the hydrostatic pressure test, whichever is greater.
The licensee stated that for Curves A and B, the initial fluid temperature is typically taken at the bolt-up temperature of the closure flange minus coolant temperature instrument uncertainty. The minimum bolt-up temperature is equal to the limiting material RTNDT of the regions affected by bolt-up stresses. The licensee stated that the minimum bolt-up temperature shall not be lower than 60°F based on guidance in BWROG-TP-11-022-A, Revision 1. Thus, the minimum bolt-up temperature shall be 60°F or the material RTNDT, whichever is higher.
For Curve C, when the reactor is critical, the initial fluid temperature is equal to the calculated minimum criticality temperature in this region. Table 1 of 10 CFR Part 50, Appendix G, indicates that, for a BWR with normal operating water levels, the allowable temperature for initial criticality at the closure flange region is equal to the RTNDT at the flange region plus 60°F.
The NRC staff independently verified that the Curves A, B, and C, in Enclosure 2 of the submittal satisfy the minimum temperature requirements contained in Table 1 of 10 CFR Part 50, Appendix G.
In its safety evaluation (SE), included within BWROG-TP-11-022-A, Revision 1, the NRC staff noted that ferritic RCPB components that are not part of the RPV may have initial RTNDT values, which may define a more restrictive lowest operating temperature in the P-T limits than those for the RPV beltline shell materials. As such, the NRC staff imposed a condition for use of BWROG-TP-11-002-A, Revision 1 that requires the lowest service temperatures for all ferritic RCPB components that are not part of the RPV be below the lowest operating temperature in the proposed P-T limits. The NRC staff identified the following two acceptable ways that a licensee could satisfy the above condition:
- 1.
The lowest operating temperature in the proposed P-T limits is the same as, or higher than, that in the first set of P-T limits approved for initial operation,
[or]
- 2.
The lowest service temperatures for the ferritic components in the RCPB, contained in the component design specifications, are less than the minimum temperature in the proposed P-T limits.
The NRC staff compared and verified that the lowest operating temperature (i.e., 60°F) in the proposed Monticello P-T limits is the same as, or higher than, that in the first set of P-T limits approved for initial operation (60°F). As such, the NRC staff finds that the licensee has satisfied the above condition imposed on the generic use of topical report BWROG-TP-11-022-A, Revision 1.
3.6.3 Pressure Test (Curve A)
The minimum bolt-up temperature of 60°F minus instrument uncertainty (0°F) is applied to all regions as the initial temperature in the iterative calculation process. The static fracture toughness (KIc) is calculated for all regions, (i.e., beltline, bottom head, and non-beltline upper vessel region). The resulting value of KIc, along with a safety factor of 1.5 to calculate the pressure stress intensity factor (KIp). The allowable RPV pressure is calculated for the beltline, bottom head, and upper vessel regions. For the feedwater nozzle/upper vessel region, the licensee applied additional constraints based on the finite element analysis. The licensee stated that because the thermal stress intensity factor is taken as zero for Curve A, the cool-down rate does not affect the results for Curve A.
3.6.4 Normal Operation - Core Not Critical (Curve B)
The licensee calculated static fracture toughness (KIc) for all RPV shell regions. The thermal stress intensity factor (KIt) is calculated for the feedwater nozzle and recirculation Inlet nozzle.
The licensee used the resulting values of KIc and KIt, along with a safety factor of 2.0, to calculate the pressure stress intensity factor (KIp). The allowable RPV pressure is calculated for the beltline, bottom head, and non-beltline regions, as appropriate. The licensee stated that for the non-beltline (feedwater nozzle/upper vessel) region, the additional constraints are applied.
3.6.5 Normal Operation - Core Critical (Curve C)
The licensee stated that the P-T values for Curve C are calculated in a similar manner as Curve B, with several exceptions. Table 1 of 10 CFR Part 50, Appendix G, stipulates that the initial evaluation temperature is calculated as the limiting non-beltline RTNDT that is highly stressed by the bolt preload plus 60°F. The RTNDT of the closure flange region is 10°F as shown in Section 4.0 of Enclosure 4 of the submittal. The resulting minimum criticality temperature is 70°F (10°F + 60°F). Table 1 of 10 CFR Part 50, Appendix G, also requires that when the pressure exceeds 20 percent of the pre-service system hydrostatic test pressure (i.e., 313 pounds per square inch gauge (psig)), the P-T limits are 40°F higher than the Curve B values.
The minimum temperature above the 20 percent of the pre-service hydrostatic test pressure is always greater than the RTNDT of the closure region plus 160°F, or is taken as the minimum temperature required for the hydrostatic pressure test.
The NRC staff performed an independent calculation to verify the proposed P-T limit curves (Curves A, B, and C) using the provisions of ASME Code,Section XI, Appendix G, and 10 CFR Part 50, Appendix G. The NRC staff verified that (1) the P-T curves use the limiting ART of 178.1°F; (2) Curve A appropriately uses the safety factor of 1.5 for pressure stress for pressure test; (3) Curves B and C appropriately use the safety factors of 2.0 and 1.5 for pressure stress and thermal stress, respectively; (4) Curves A, B and C, satisfy the minimum required temperatures in accordance with Table 1 of 10 CFR Part 50, Appendix G; and (5) Curves A, B, and C, considered the stress effect of feedwater and recirculation nozzles per RIS 2014-11.
The NRC staff verified that P-T limit Curves A, B, and C, for 72 EFPY at Monticello satisfy 10 CFR Part 50, Appendix G, and the ASME Code,Section XI, Appendix G. The NRC staff also finds that the licensee appropriately used the methodology of BWROG-TP-11-022-A, Revision 1, to develop the updated PTLR method in accordance with RG 1.99, Revision 2.
The NRC staff finds that the proposed Monticello PTLR calculated to 72 EFPY (at a neutron fluence of 4.38E+18 n/cm2 at the 1/4 location of the reactor vessel wall) is acceptable for use up to the end of the current operating license. The PTLR will be limited to operation time in accordance with Monticellos operating license. The NRC staff has determined that the neutron fluence value associated with the development of the P-T limit curves is within the NRC-approved methodology, is bounding, and covers the length of time that is proposed in the submittal and therefore is acceptable for use up to the end of the current operating license. As this approval is bounded by the current analysis performed for the Monticello SLRA, if changes have occurred associated with the pending approval of Monticello SLRA, the licensee will need to assess the impact to the PTLR in the submittal at that time.
The NRC staff notes that (1) if changes have occurred associated with the pending review of the Monticello SLRA, the licensee will need to assess the impact to the PTLR at that time, and (2) approval of the subject PTLR at the neutron fluence value of 4.38E+18 n/cm2 at the 1/4 T location of the RPV wall thickness does not imply that the P-T limits in the SLRA are approved.
3.7 Neutron Fluence Calculations As part of the SLRA, the licensee submitted neutron fluence projections performed for the Monticello RPV and reactor vessel internals components by applying the RAMA fluence methodology. The RAMA fluence methodology used for Monticello is described in EPRI Report No. BWRVIP-114-A (ML092650376). The fluence projections for Monticello are performed for the 20-year period of subsequent extended operation (i.e., through 72 EFPY), which conservatively bounds the 54 EFPY assumed through the end of the current Renewed Facility Operating License. The licensee provided a supplement to the SLRA (ML23193B026) to discuss the Monticello fluence methodology qualification for the RPV and vessel internals. The licensee states in its supplement that the report that was submitted supported the development of the PLTR method.
The NRC staff evaluated the proposed Monticello neutron fluence method for the PTLR in accordance with the criteria in Attachment 1 to GL 96-03, as discussed below:
Criterion 1 requires that the PTLR methodology describes the transport calculation methods including computer codes and formula used to calculate neutron fluence values.
In the BWRVIP-114-A, the NRC staff found the RAMA neutron fluence methodology to be appropriate to perform the transport calculations required to estimate the fluence within the RPV, with no discernable bias in the computed results. The RAMA fluence methodology has been benchmarked against several plant-specific dosimetry measurements from several operating reactors.
The NRC staff SE for the BWRVIP-114-A listed two conditions for application of RAMA fluence methodology for the plants that do not have geometries similar to Susquehanna Steam Electric Station, Unit 2 (Susquehanna), and Hope Creek Generating Station, nuclear plants (BWR-IV),
such as Monticello which is a BWR-III plant. The limitations state that:
To apply the RAMA methodology to plant groups which have geometries that are different than the cited BWR-IV's, at least one plant-specific capsule dosimetry analysis must be provided to quantify the potential presence of a bias and assure that the uncertainty is within the RG 1.190 limits, and justification is necessary for a specific application based on geometrical similarity to an analyzed core, core shroud, and RPV geometry. That is, a licensee who wishes to apply the RAMA methodology for the calculation of RPV neutron fluence must reference, or provide, an analysis of at least one surveillance capsule from a RPV with a similar geometry.
To meet the limitations of the BWRVIP-114-A, since Monticello is a BWR-III plant, the licensee stated that a plant-specific capsules dosimetry analysis was used to quantify the potential presence of any bias and to ensure that the uncertainty is within the RG 1.190 limits. The licensee further stated that the Monticello RPV is modeled with the core, shroud, and the RPV geometry details. The licensee stated, in the supplement to the SLRA, that the computational fluence models constructed for Monticello follow a similar approach that was used to model the Susquehanna, core shroud and top guide benchmarks. The licensee has also incorporated results from evaluation of the Monticello 120-degree surveillance capsule removed during the spring of 2021 refueling outage into the supporting analyses.
The NRC staff found that the fluence projections performed for Monticello in the subsequent license renewal application have a combined uncertainty for the Monticello RPV at 11.6-percent as documented in the SE related to subsequent license renewal of Monticello (ML24077A001),
which is well within the 20-percent criterion established in RG 1.190. The NRC staff noted that the calculations performed in the supplement to the SLRA show that appropriate geometry and material representations were used for the central and upper core shroud shells, top guide plates, fuel structures, upper shroud plenums, and coolant water densities, in addition to the RPV. Hence, the NRC staff finds that the licensee has met the two limitations and conditions for application of RAMA fluence methodology. The NRC staff finds the TransWare RAMA fluence methodology is acceptable for projecting the neutron fluence through the end of the first license renewal period of 54 EFPY, because it is conservatively bounded by the evaluation performed by the licensee and verified by the NRC for the 72 EFPY in the licensees SLRA for Monticello.
3.8 Changes to Technical Specifications The regulations in 10 CFR Part 50.36(b) require that TSs be derived from the analysis. As described in Sections 3.1 through 3.7 of this SE, the proposed methodology will be incorporated into the licensing basis and will provide adequate assurance that controls of RCS operating conditions will be established to ensure that the fracture toughness of the reactor vessel will be maintained such that integrity of the vessel will be maintained. The NRC staff reviewed the licensees methodology to verify that the analysis will acceptably assure vessel integrity. The change to the TS is made to assure that the TS are consistent with the licensees analysis and methodology.
The regulations in 10 CFR Part 50.36(c)(5) requires that administrative controls be established in the TS such that operation in a safe manner is assured. The licensee is revising its PTLR program in TS, Section 5.6.5 (RCS) and (PTLR) to update the methodology used to establish the analytical methods for determining the RCS P-T limits. The licensees methodology is required by TS to be reviewed and approved by the NRC. The technical evaluations described in Sections 3.1 through 3.7 of this SE provide the technical evaluation for NRC acceptance of
the proposed TS changes and the acceptability of the changes to the PTLR. Therefore, the updated methodology proposed to be referenced in the TS program for Section 5.6.5 is acceptable.
The proposal does not request any changes to the other sections of the TS required by 10 CFR Part 50.36. For example, the LCO required actions and surveillance requirements that implement the PTLR program are unchanged. Maintenance of the current TS requirements utilizing new values determined by the updated methodology as approved by the NRC will adequately assure reactor vessel integrity.
3.9 Technical Conclusion The NRC staff has determined that use of the TransWare RAMA fluence methodology for neutron fluence projection in the RPV for the purpose of determining RCS P-T limits for Monticello up to 54 EFPY is acceptable and adherent to RG 1.190. The licensee has addressed the limitations discussed in the NRC staff SE for the methodology listed in BWRVIP-114-A. The NRC staff finds that by meeting the guidance in RG 1.190, the licensee has satisfied the applicable regulations in 10 CFR Part 50.60, 10 CFR Part 50.61, and 10 CFR Part 50, Appendix G, as well as the intent of 10 CFR Part 50, Appendix A, and GDCs 14, 15, 31, and 32 to use the TransWare RAMA fluence Methodology.
The NRC staff has determined that the proposed P-T limit curves in the Monticello PTLR:
(1) considered stresses in the beltline region, bottom head region and non-beltline region; (2) have considered all relevant ferritic RPV shell materials; (3) have used appropriate limiting ART, which was derived based on RG 1.99, Revision 2; (4) are appropriately constructed following the provisions of the ASME Code,Section XI, Appendix G, and SRP Section 5.3.2; (5) have satisfied applicable requirements in 10 CFR Part 50, Appendices G and H; and (6) are acceptable to the end of plant operating license. Therefore, the licensee-proposed changes to the TS are in accordance with the guidance of GL 96-03 to change the methodology used to develop the PTLR to SIR-05-044-A Version 1.
Further, the NRC has determined that the PTLR program implemented by the licensee in Section 5.6.5 of its TSs meets the requirements of 10 CFR Part 50.36 so that adequate assurance of reactor vessel integrity is provided, as described in Section 3.8 above.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendment on August 15, 2024. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20, or changes to the SRs. 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 and there has been no public comment on such finding dated March 19, 2024 (89 FR 19610). Accordingly, the amendment meets the eligibility
criteria for categorical exclusion set forth in 10 CFR Part 51.22(c)(9). Pursuant to 10 CFR Part 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: S. Bhatt, NRR S. Smith, NRR J. Tsao, NRR O. Yee, NRR Date of Issuance: December 19, 2024
ML24270A155 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DNRL/NVIB/BC NRR/DSS/SNSB/BC NAME BBallard SRohrer ABuford PSahd DATE 9/24/2024 9/30/2024 8/1/2024 8/15/2024 OFFICE NRR/DSS/STSB/BC OGC - NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME SMehta ALeatherman IBerrios BBallard DATE 9/25/2024 11/1/2024 12/19/2024 12/19/2024