ML23286A260
| ML23286A260 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 11/08/2023 |
| From: | Robert Kuntz Plant Licensing Branch III |
| To: | Rhoades D Constellation Energy Generation |
| Purnell, B A | |
| References | |
| EPID L-2022-LLA-0173 | |
| Download: ML23286A260 (43) | |
Text
November 8, 2023 Mr. David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
LASALLE COUNTY STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 260 AND 245 TO RENEWED FACILITY OPERATING LICENSES RE: RELOCATION OF PRESSURE AND TEMPERATURE LIMIT CURVES TO THE PRESSURE TEMPERATURE REPORT (EPID L-2022-LLA-0173)
Dear Mr. Rhoades:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 260 and 245 to Renewed Facility Operating License Nos. NPF-11 and NPF-18, for the LaSalle County Station, Units 1 and 2, respectively. The amendments consist of changes to the technical specifications in response to your application dated November 10, 2022, as supplemented by letters dated January 10, 2023, and June 30, 2023.
The amendments replace the existing reactor vessel heatup and cooldown rate limits and the pressure and temperature limit curves with references to the Pressure and Temperature Limits Report.
A copy of our related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Robert F. Kuntz, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-374
Enclosure:
- 1. Amendment No. 260 to NPF-11
- 2. Amendment No. 245 to NPF-18
- 3. Safety Evaluation cc: Listserv CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-373 LASALLE COUNTY STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 260 Renewed License No. NPF-11
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Constellation Energy Generation, LLC (the licensee) dated November 10, 2023, as supplemented by letters dated January 10, 2023, and June 30, 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. NPF-11 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 260, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Jeffrey A. Whited, 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: November 8, 2023 Jeffrey A.
Whited Digitally signed by Jeffrey A. Whited Date: 2023.11.08 15:35:26 -05'00'
CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-374 LASALLE COUNTY STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 245 Renewed License No. NPF-18
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Constellation Energy Generation, LLC (the licensee) dated November 10, 2023, as supplemented by letters dated January 10, 2023, and June 30, 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. NPF-18 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 245, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Jeffrey A. Whited, 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: November 8, 2023 Jeffrey A.
Whited Digitally signed by Jeffrey A. Whited Date: 2023.11.08 15:35:46 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NOS. 260 AND 245 RENEWED FACILITY OPERATING LICENSE NOS. NPF-11 AND NPF-18 LASALLE COUNTY STATION, UNITS 1 AND 2 DOCKET NOS. 50-373 AND 50-374 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Renewed Facility Operating License No. NPF-11 REMOVE INSERT Page 3 Page 3 Renewed Facility Operating License No. NPF-18 REMOVE INSERT Page 3 Page 3 Technical Specifications REMOVE INSERT 1.1-8 1.1-8 1.1-9 1.1-9 3.4.11-3 3.4.11-3 3.4.11-4 3.4.11-4 3.4.11-5 3.4.11-5 3.4.11-6 3.4.11-7 3.4.11-8 3.4.11-9 3.4.11-10 3.4.11-11 5.6-4 5.6-4 5.6-5 5.6-5
Renewed License No. NPF-11 Amendment No. 260 (3)
Constellation Energy Generation, 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)
Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
Constellation Energy Generation, 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 LaSalle County Station, Units 1 and 2, and such Class B and Class C low-level radioactive waste as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, and Clinton Power Station, Unit 1.
C.
This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3546 megawatts thermal).
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 260, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
DELETED (4)
DELETED (5)
DELETED Am. 198 09/16/10 Am. 194 08/28/09 Am. 194 08/28/09 Am. 194 08/28/09 Am. 260 11/08/23
Renewed License No. NPF-18 Amendment No. 245 (2)
Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3)
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)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
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 LaSalle County Station, Units 1 and 2, and such Class B and Class C low-level radioactive waste as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, and Clinton Power Station, Unit 1.
C.
This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3546 megawatts thermal). Items in shall be completed as specified. Attachment 1 is hereby incorporated into this license.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 245, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan Am. 185 09/16/10 Am. 245 11/08/23
Definitions 1.1 LaSalle 1 and 2 1.1-8 Amendment No. 252/238 1.1 Definitions (continued)
MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.
MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
OPERABLEOPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
PRESSURE AND TEMPERATURE The PTLR is the document that provides the reactor LIMITS REPORT (PTLR) vessel pressure and temperature limits, including heatup and cooldown rates, for the current vessel fluence period. The pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.
RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 3546 MWt.
REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is (continued) 260/245
Definitions 1.1 LaSalle 1 and 2 1.1-9 Amendment No. 242/228 1.1 Definitions REACTOR PROTECTION measured. In lieu of measurement, response time SYSTEM (RPS) RESPONSE may be verified for selected components provided TIME that the components and method for verification (continued) have been previously reviewed and approved by the NRC.
SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:
a.
The reactor is xenon free; b.
The moderator temperature is > 68°F, corresponding to the most reactive state; and c.
All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.
STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TURBINE BYPASS SYSTEM The TURBINE BYPASS SYSTEM RESPONSE TIME shall be RESPONSE TIME that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions. 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.
260/245
RCS P/T Limits 3.4.11 LaSalle 1 and 2 3.4.11-3 Amendment No.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1
NOTE--------------------
Only required to be performed during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing.
Verify:
a.
RCS pressure and RCS temperature are within the applicable limits specified in the PTLR; b.
RCS heatup and cooldown rates are within the limits specified in the PTLR; and c.
RCS temperature change during system leakage and hydrostatic testing is within the limits specified in the PTLR.
In accordance with the Surveillance Frequency Control Program SR 3.4.11.2 Verify RCS pressure and RCS temperature are within the criticality limits specified in the PTLR.
Once within 15 minutes prior to control rod withdrawal for the purpose of achieving criticality (continued) 260/245
RCS P/T Limits 3.4.11 LaSalle 1 and 2 3.4.11-4 Amendment No. 200/1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.3
NOTE--------------------
Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.
Verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature is within the limits specified in the PTLR.
Once within 15 minutes prior to each startup of a recirculation pump SR 3.4.11.4
NOTE--------------------
Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.
Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is within the limits specified in the PTLR.
Once within 15 minutes prior to each startup of a recirculation pump SR 3.4.11.5
NOTE--------------------
Only required to be performed when tensioning the reactor vessel head bolting studs.
Verify reactor vessel flange and head flange temperatures are within the limits specified in the PTLR.
In accordance with the Surveillance Frequency Control Program (continued) 260/245
RCS P/T Limits 3.4.11 LaSalle 1 and 2 3.4.11-5 Amendment No. 200/18 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.6
NOTE--------------------
Not required to be performed until 30 minutes after RCS temperature 77°F for Unit 1 and 91°F for Unit 2 in MODE 4.
Verify reactor vessel flange and head flange temperatures are within the limits specified in the PTLR.
In accordance with the Surveillance Frequency Control Program SR 3.4.11.7
NOTE--------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature 92°F for Unit 1 and 106°F for Unit 2 in MODE 4.
Verify reactor vessel flange and head flange temperatures are within the limits specified in the PTLR.
In accordance with the Surveillance Frequency Control Program 260/245
Reporting Requirements 5.6 LaSalle 1 and 2 5.6-4 Amendment No. 257/243 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 1.
NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel."
The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).
c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
(continued) 260/245
Reporting Requirements 5.6 LaSalle 1 and 2 5.6-5 Amendment No. 177/163 5.6 Reporting Requirements (continued) 5.6.6 Post Accident Monitoring (PAM) Instrumentation 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.
5.6.7 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
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 Limiting Condition for Operation and Surveillance Requirements Section 3.4.11, "RCS Pressure and Temperature (P/T) Limits."
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:
a.
BWROG-TP-11-022-A, Revision 1 (SIR-05-044),"Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated August 2013, (ML13277A557).
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplements thereto.
260/245
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 260 AND 245 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-11 AND NPF-18 CONSTELLATION ENERGY GENERATION, LLC LASALLE COUNTY STATION, UNITS 1 AND 2 DOCKET NOS. 50-373 AND 50-374
1.0 INTRODUCTION
By letter dated November 10, 2022, (Agencywide Documents and Access Management System (ADAMS) Accession No. ML22332A448), as supplemented by letters dated January 10, 2023, (ML23010A227) and June 30, 2023, (ML23181A149) Constellation Energy Generation Company, LLC (CEG or the licensee) submitted a license amendment request (LAR) to modify the technical specifications (TSs) for LaSalle County Station (LSCS), Units 1 and 2.
The LAR states that the format and contents of TS changes proposed are consistent with those defined Technical Specifictions Task Force (TSTF) Traveler TSTF-419, Revise PTLR [Pressure and Temperature Limits Report] Definition and References in ISTS [Improved Standard Technical Specification] 5.6.6, RCS [Reactor Collant System] PTLR (ML012690234), and the criteria for relocating pressure and temperature (P-T) limit curves and other RCS limits into a PTLR, as established in the U.S. Nuclear Regulatory Commission (NRC or Commission)
Generic Letter (GL) 96-03, Relocation of Pressure and Temperature Limit Curves and Low Temperature Overpressure Protection System Limits (ML031110004). The licensees proposed revisions to the LSCS, Units 1 and 2, TSs are given in Attachment 2 (ML22332A450) of the November 22, 2022, letter.
The LAR included the first PTLR, Revision 0, in Attachment 6 (ML22332A454) which was developed in accordance with newly proposed TS, section 5.6.7, Revision 1, of the PTLR was submitted in Attachment 2 of the January 10, 2023, LAR supplement, and Revision 3 of the PTLR in Attachment 2 of June 30, 2023, LAR supplement (ML23181A152). In this regard, the June 30, 2023, version of the PTLR replaces the November 10, 2022, version of the PTLR that was enclosed as Attachment 6 in the November 22, 2023, letter. Therefore, further reference to the PTLR in this safety evaluation (SE) refers to the June 30, 2023, version unless otherwise noted.
The NRC staff performed an audit of the LAR from April 12, 2023, through May 5, 2023. A summary of the staffs audit is provided in the Audit Report of August 3, 2023 (ML23212A901).
The staff held a public meeting with the licensee on August 18, 2023. A summary is provided in the staffs public meeting summary of August 29, 2023 (ML23234A234).
The supplemental letters dated January 10, 2023, and June 30, 2023, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on March 7, 2023 (88 FR 14181).
2.0 REGULATORY EVALUATION
2.1 Proposed TS Changes to Adopt TSTF-419 The licensee proposed changes that would relocate the P-T limits for the reactor pressure vessel to the licensee-controlled PTLR in order to adopt TSTF-419. The LAR proposed to change TS Section 1.0, Definitions to add a new definition for PTLR as follows:
Pressure and Temperature Limits Report (PTLR)
The PTLR is the document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current vessel fluence period. The pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.
The LAR proposed the following changes to TS Section 3.4.11, RCS Pressure and Temperature (P/T) Limits:
The text referring to the P-T curves and the associated TS wording in Surveillance Requirements (SR) 3.4.11.1 thru 3.4.11.7 would be replaced with reference to the within the limits specified in the PTLR, and P-T curves specified in TS Figures 3.4.11-1 thru 3.4.11-6 would be deleted.
The LAR would add a new Section 5.6.7 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) as follows:
5.6.7 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
RCS pressure and temperature limits for heat up, 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 Limiting Conditions for Operation and Surveillance Requirement Section 3.4.11, "RCS Pressure and Temperature (P/T) Limits."
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 document:
- a. BWROG [Boiling Water Reactor Owners Group]-TP-11-022-A, Revision 1 (SIR-05-044), "Pressure Temperature Limits Report Methodology for Boiling Water Reactors," dated August 2013, (ML13277A557).
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
2.2 Variation from TSTF-419 The LAR proposed one variation to TSTF-419, as follows:
There is one variation between the content in TSTF-419-A and this application.
Specifically, the PTLR definition in the TSTF and versions of the Standard Technical Specifications use the phrase "unit specific" to describe the report.
However, the vendor-supplied report for LSCS has both units contained within one report. To prevent confusion and maintain alignment with how this definition is applied in the precedent relied on in this application, LSCS is removing the phrase "unit specific" from the PTLR definition in the LSCS TS. Even if future PTLRs are unit specific, this would not necessitate a change to the definition to remain compliant with this new definition.
2.3 Applicable Regulation and Guidance The regulation in Title 10 of the Code of Federal Regulations (10 CFR), section 50.36, Technical specifications, paragraph (a), requires that each operating license application for a production or utilization facility include proposed TS and a summary statement of the bases for such specifications. Paragraph (c) of 10 CFR 50.36 requires, in part, that TS include the following categories related to facility operation: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.
The regulations in 10 CFR 50.60, Acceptance Criteria for Fracture Prevention Measures for Light Water Nuclear Power Reactors for Normal Operation, requires that all operating light-water nuclear power reactors meet the fracture toughness requirements for the reactor coolant pressure boundary (RCPB) set forth in 10 CFR, part 50, appendix G, Fracture Toughness Requirements.
The regulations in 10 CFR part 50, appendix G, Fracture Toughness Requirements, require:
(1) sufficient fracture toughness for reactor pressure vessel (RPV) ferritic materials to provide adequate safety margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests; (2) P-T limits that satisfy the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Appendix G, and the minimum temperature requirements during normal heatup, cooldown, and pressure test operations; and (3) applicable surveillance data from RPV material surveillance programs developed in accordance with 10 CFR, part 50, appendix H, Reactor Vessel Material Surveillance Program Requirements, be incorporated into the calculations of P-T limits.
The regulations in 10 CFR, part 50, appendix H, Reactor Vessel Material Surveillance Program Requirements, requires licensees to implement a material surveillance program to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of light-water nuclear power reactors which result from exposure of these materials to neutron irradiation and the thermal environment.
GL 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection Systems Limits (ML031110004), dated January 31, 1996, permits
relocation of the P-T limits from the TS to a PTLR. GL 96-03 establishes a position that licensees seeking a license amendment for relocation of the applicable P-T limits into a PTLR should: (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 Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (ML010890301), describes methods acceptable to the NRC staff for determining the RPV neutron fluence with respect to meeting the regulatory requirements discussed above.
RG 1.99, Revision 2 (RG 1.99), Radiation Embrittlement of Reactor Vessel Materials, dated May 1988 (ML003740284), describes procedures for calculating the adjusted nil-ductility transition reference temperature (ART) due to neutron irradiation.
GL 92-01, Revision 1, Reactor Vessel Structural Integrity, 10 CFR 50.54(f), dated March 6, 1992 (GL 92-01, ML031070438), requested that licensees submit their plant specific RPV data to the NRC staff for review. GL 92-01, supplement 1, dated May 19, 1995 (ML031070449),
requested that licensees provide and assess data from other licensees that could affect their RPV integrity.
The NRC Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014 (ML14149A165), provides evaluation guidance for P-T limit curves and PTLRs, including the consideration of neutron fluence and structural discontinuities in the development of P-T limit curves.
The NRC 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 limits based on the requirements in 10 CFR, part 50, appendix G, and 10 CFR 50.61.
By letter dated September 4, 2013, BWROG issued BWROG-TP-11-022-A, Revision 1, Pressure-Temperature Limits Report Methodology for Boiling Water Reactors (BWROG-TP-11-022-A, ML13277A557).
3.0 TECHNICAL EVALUATION
The NRC staff reviewed the LAR and the proposed PTLR implementation in accordance with the guidance in GL 96-03 and TSTF-419. As such, the NRC staff evaluated the updated adjusted reference temperature (ART) and P-T limit calculations included in the PTLR, as assessed in accordance with methods or guidance in BWROG-TP-11-022-A, RG 1.99, and SRP 5.3.2, and the applicable requirements for P-T limit assessments in 10 CFR, part 50, appendix G, and RPV material surveillance programs in 10 CFR, part 50, appendix H.
3.1 Evaluation of the PTLR Information with the Criteria in GL 96-03 The NRC staff evaluated the proposed PTLR in accordance with the criteria in Attachment 1 to GL 96-03 as discussed below.
3.1.1 GL 96-03, Criterion 1 - Neutron Fluence Values GL 96-03, Criterion 1, establishes that the PTLR should provide the values of neutron fluences that are used in the calculations of ART.
3.1.1.1 Conditions and Limitations The NRC safety evaluation report for the BWROG-TP-11-022-A, contained one condition for future potential applicants to address in their application of this licensing technical review (LTR) to their plant-specific P-T limits or PTLR submittal:
Each applicant referencing this LTR shall confirm that, in addition to the requirements in the ASME Code, section XI, appendix G, the lowest service temperatures for all ferritic RCPB components that are not part of the RPV, are below the lowest operating temperature in the proposed P-T limits.
LSCS, Units 1 and 2, have confirmed the lowest service temperatures for all ferritic reactor RCPB components that are not part of the RPV, are below the lowest operating temperature in the proposed P-T limits. This confirmation has been included in section 4.0, Operating Limits, of the LSCS, Units 1 and 2, PTLR. The NRC staff confirms the changes are, therefore, acceptable.
3.1.1.2 Pressure and Temperature Limits Report Three regions of the RPV were evaluated to develop the revised P-T curves: the beltline region, the bottom head region, and the non-beltline region. These regions bound all other regions with respect to brittle fracture. The methodology used to generate the P-T curves in this submittal is approved by the NRC and uses ART values determined in accordance with RG 1.99, Revision
- 2. The revised P-T curves and outputs from the integrated surveillance program (ISP) ensure that adequate RPV safety margins against non-ductile failure will continue to be maintained during normal operations, anticipated operational occurrences, and inservice leak and hydrostatic testing. Together, these measures ensure that the integrity of the RCPB will be maintained for the life of the plant. These proposed changes are consistent with the guidance provided in GL 96-03, as supplemented by TSTF-419. The NRC staff finds the changes are, therefore, acceptable.
3.1.1.3 Neutron Fluence Calculations During its review of the licensees license renewal application, the NRC staff previously determined fluence calculational method acceptability by confirming that the methodology used to determine 54 effective full power years (EFPY) fluence values was the same as the NRC-approved methodology used to produce 32 EFPY fluence values (see NUREG-2055, ML16271A039). The licensee stated that the LAR was prepared in accordance with the guidelines of GL 96-03 and TSTF-419. The regulations in 10 CFR, part 50, appendix G, require reactor vessel beltline materials to be tested in accordance with the surveillance program requirements of 10 CFR, part 50, appendix H.
The LAR stated that neutron fluence is calculated in accordance with RG 1.190, using the Radiation Analysis Modeling Application (RAMA) computer code. The ART values for the limiting beltline materials are calculated in accordance with NRC RG 1.99. The LAR stated that GL 96-03 allows plants to relocate their P-T curves and associated numerical limits (such as heatup and cooldown rates) from the plant TS to a PTLR, which is a licensee-controlled document. As stated in GL 96-03, during the development of the improved STS, a change was proposed to relocate the P-T limits currently contained in the plant TS to a PTLR. As one of the improvements to the STS, the NRC staff agreed with the industry that the curves may be relocated outside the plant TS to a PTLR so that the licensee could maintain these limits efficiently. One of the prerequisites for having the PTLR option is that the P-T curves and limits be derived using methodologies approved by the NRC, and that the associated licensing topical reports describing the approved methodologies be referenced in the plant TS.
The NRC staff evaluated the proposed LSCS, Unit 1, PTLR and LSCS, Unit 2, PTLR in accordance with the criteria in Attachment 1 to GL 96-03 as discussed below.
Criterion 1 provides that the PTLR methodology should describe the transport calculation methods including computer codes and formula used to calculate neutron fluence values.
As stated in the LAR, the licensee utilized the BWROG-TP-11-022-A, report methodology to generate the P-T limits. The methodology is an NRC-approved method for use in generating PTLRs. The PTLR methodology describes the transport calculation methods, including computer codes and formula used to calculate neutron fluences. The NRC staff finds that LSCS, Units 1 and 2, PTLRs have satisfied Criterion 1 of Attachment 1 to GL 96-03 because both PTLRs provide appropriate transportation calculation methods that satisfy RG 1.190.
LSCS, Units 1 and 2, has replaced the original RPV material surveillance program with the Boiling Water Reactor Vessel and Internals Project (BWRVIP) ISP. LSCS, Units 1 and 2, are committed to using the BWRVIP ISP during the current licensed period. Use of the BWRVIP ISP for LSCS, Units 1 and 2, was approved by the NRC in NUREG-2205. LSCS, Units 1 and 2, have made a license renewal commitment to use the ISP during the period of extended operation. The Reactor Vessel Surveillance program is based on BWRVIP-86, Revision 1-A, BWR Vessel and Internals Project Updated BWR Integrated Surveillance Program (ISP)
Implementation Plan, (BWRVIP-86) dated May 2013 (ML13176A097). The NRC staff finds the use of the BWRVIP ISP acceptable.
Based on the above, the NRC staff finds: (1) the fluence calculational method inputs are representative of past operating conditions, and (2) the licensee has provided confidence that current and future operating conditions will be appropriately accounted for and will result in updated fluence projections when necessary. The fluence calculational method described is acceptable for use with the PTLR methodology based on appropriate 54 EFPY fluence projections made using a RG 1.190 adherent methodology. The licensee stated that the LAR was prepared in accordance with the guidelines of GL 96-03 and TSTF-419. Therefore, the NRC staff has reasonable assurance that the proposed PTLR and subsequent updates will use appropriate fluence calculational method inputs with an acceptable fluence calculational method adherent to RG 1.190.
3.1.2 GL 96-03 Criterion 2 - RPV Surveillance Capsule Withdrawal Schedule The reviewers notes included with the model TSs in GL 96-03 states that The Reactor Vessel Material Surveillance Program shall comply with Appendix H to 10 CFR, part 50. The reactor
vessel material irradiation surveillance specimen removal schedule shall be provided, along with how the specimen examinations shall be used to update the PTLR curves.
GL 96-03, Criterion 2, establishes that the PTLR shall provide the RPV surveillance capsule withdrawal schedule, or reference (by number and title) the record(s) that provide the RPV surveillance withdrawal schedule. GL 96-03, Criterion 2, also establishes that the PTLR should provide or reference applicable RPV surveillance capsule reports if the ART values are calculated using surveillance data.
Consistent with the NRC staffs approval of BWRVIP-86, the NRC staff noted that the licensee defined how the BWRVIP ISP is applied to the LCSC, Unit 1 and 2, current licensing basis (CLB) in chapter 5, Reactor Coolant System and Connected Systems (ML22111A234), of the LSCS Updated Final Safety Analysis Report (UFSAR) with UFSAR, section 5.3.1.6 and UFSAR table 5.2-12, providing the times of the past RPV surveillance capsule withdrawals and future capsule withdrawal schedules of LSCS, Unit 1, and LSCS, Unit 2, RPV surveillance capsules implemented under the BWRVIP ISP. The following SE subsections (3.1.2.1 and 3.1.2.2) address how the licensees implementation of the BWRVIP ISP for LSCS, Unit 1, differs from that implemented for LSCS, Unit 2.
3.1.2.1 Application of the BWRVIP ISP for the LSCS, Unit 1 Licensing Basis The NRC staff confirmed that the PTLR identifies that the licensee has removed two RPV surveillance capsules from the LSCS, Unit 1, RPV, one at 6.5 EPFY and a second capsule at 18.9 EPFY. The licensee indicates that one additional (supplemental) capsule is available in the LSCS, Unit 1, RPV for further use under the BWRVIP ISP, and that the capsule will be removed (for testing of test specimens in the capsule) at a time that is consistent with the proprietary schedule for the capsules removal in BWRVIP-86. The NRC staff finds this acceptable for implementation because it is based on the time specified for the LSCS, Unit 1, supplemental capsule withdrawal in BWRVIP-86, and the licensees withdrawal schedule for LSCS, Unit 1, surveillance capsules in UFSAR, table 5.2-12.
The NRC staff noted that the PTLR references the following documents containing information regarding the BWRVIP ISP surveillance plate and weld materials that are identified for LSCS, Unit 1, in BWRVIP-86:
Electric Power Research Institute (EPRI) Report No. BWRVIP-250-NP (ML11326A290),
which covers the prior removal and testing of the LSCS, Unit 1, 300º Capsule at 6.5 EPFY and the 120º Capsule at 18.9 EFPY EPRI Proprietary Report No. BWRVIP-135, Revision 4 (BWRVIP-135, non-proprietary version of the report was submitted as Attachment 7 in the November 10, 2022, letter (ML22332A455)), which evaluates the applicable capsule data for LSCS, Unit 1, surveillance plate and weld materials in the 120º and 300º capsules.
Thus, based on its review, the NRC staff concludes that the PTLR conforms to GL 96-03, Criterion 2, because: (1) the PTLR adequately describes the application implementation of BWRVIP ISP withdrawal schedule (as referenced to BWRVIP-86) for the LSCS, Unit 1, CLB, (2) the PTLR appropriately references the applicable reports for BWRVIP ISP surveillance plate and weld materials applying to LSCS, Unit 1, including the BWRVIP-250-NP and BWRVIP-135, reports, and (3) the PTLR addresses the future LSCS, Unit 1, supplemental capsule withdrawal
and testing needs consistent with the staffs October 20, 2011, SE for BWRVIP-86, and the NRC staffs approval for LSCS, Units 1 and 2, implementation of BWRVIP-86, in NUREG-2205.
3.1.2.2 Application of the BWRVIP ISP for the LSCS, Unit 2 Licensing Basis The NRC staff confirmed that the licensee removed one RPV surveillance capsule at 6.98 EPFY, and that two additional (supplemental) surveillance capsules remain available for potential use by the licensee, as supported by capsule withdrawal schedule information for LSCS, Unit 2, in UFSAR, table 5.2-12.
The NRC staff noted that LSCS, Unit 2, is not designated as a host plant for the BWRVIP ISP capsule withdrawal schedule bases that apply to LSCS, Unit 2, in BWRVIP-86. Instead, the staff confirmed that BWRVIP-86, identifies that the RPV beltline plate and weld materials selected as the RPV target materials for LSCS, Unit 2, are represented by surveillance materials in the U.S. [United States] BWR [boiling-water reactor] RPV surveillance capsules other than the LSCS, Unit 2, RPV capsules.1 Thus, the NRC staff confirmed that no further capsule removals of LSCS, Unit 2, surveillance capsules are required or proposed under the BWRVIP ISP capsule withdrawal schedule that applies to LSCS, Unit 2. Based on its review, the staff concludes that the PTLR conforms to GL 96-03, Criterion 2, for the LSCS, Unit 2, CLB because: (1) the PTLR adequately describes the application and implementation of the BWRVIP ISP withdrawal schedule as it references BWRVIP-86, for the LSCS, Unit 2, CLB, (2) the NRC staff has confirmed that the CLB for LSCS, Unit 2, does not require any removals of LSCS, Unit 2, RPV surveillance capsules for the BWRVIP ISP surveillance material monitoring objectives that apply to LSCS, Unit 2, and (3) the PTLR addresses that the two remaining LSCS, Unit 2, surveillance capsules are available as potential sources of RPV surveillance data (if needed), as confirmed in UFSAR, table 5.2-12.
3.1.3 GL 96-03 Criterion 3 - Description of the Low-Temperature Overpressure Protection System Limits GL 96-03, Criterion 3, establishes that, for pressurized-water reactor (PWR) designs, the PTLR should provide the low-temperature overpressure protection (LTOP) system setpoint curves or setpoint values. LTOP does not apply to LSCS, Units, 1 and 2, because these are BWR designs and the criterion is only applicable to PWR designs. Thus, the NRC staff finds that the LSCS, Units 1 and 2, PTLR did not need to address conformance with GL 96-03, Criterion 3.
3.1.4 GL 96-03 Criterion 4 - Methods for Calculating ART Values and Identification of Limiting ART Values GL 96-03, Criterion 4, establishes that the PTLR methodology shall describe the method for calculating the ART values using the methods of analysis in RG 1.99. GL 96-03, Criterion 4, establishes that the PTLR identify the limiting materials and limiting values of ART at the one-quarter thickness (1/4T) and three-quarter thickness (3/4T) locations of the RPV.
Section 3.0 of the PTLR states that the calculations of RPV beltline material ART values are performed in accordance RG 1.99. The NRC staff also confirmed that appendix A of the 1
For the BWRVIP ISP defined in BWRVIP-86, EPRI BWRVIP defines a BWR RPV target plate or weld material as the specific vessel material to which the ISP test matrix assigns a representative surveillance material.
BWROG-TP-11-022-A, methodology defines how the ART values will be calculated consistent with methods contained in RG 1.99.
The NRC staff verified that the PTLR did not need to include any 54 EFPY 3/4T ART values because this is consistent with the NRC-approved methodology in BWROG-TP-11-022-A.
The NRC staff also reviewed the 54 EPFY 1/4T ART values for the LSCS, Unit 1 and Unit 2 RPV beltline materials. The June 30, 2023, LAR supplement includes ART information and calculations for LSCS, Unit 1 RPV middle shell axial weld 3-308 BG in response and resolution of request for additional information (RAI) No. LSCS-PTLR-LAR-2. For LSCS, Unit 1, the NRC staff confirmed a limiting 54 EPFY 1/4T ART value of 136.3 ºF, which was calculated using the chemistry factor (CF) ratio procedure defined in Position 2.1 of RG 1.99, for the LSCS, Unit 1 surveillance weld material.
For LSCS, Unit 2, the NRC staff confirmed that the licensee used RG 1.99, Position 1.1 (and the CF tables in RG 1.99) to derive the CF and 54 EPFY 1/4T ART values reported for the LSCS, Unit 2, RPV beltline plates, nozzles and welds in PTLR, table 8. The NRC staff confirmed a limiting 54 EPFY 1/4T ART value of 85.7 ºF (degree Fahrenheit) for the LSCS, Unit 2, beltline materials for lower shell plate 21-2 (plate heat No. C9425-1) in PTLR, table 8.
Thus, based on the confirmations provided in the previous paragraphs, the NRC staff finds that the PTLR appropriately conforms to GL 96-03, Criterion 4, because the NRC staff has confirmed the PTLR:
states that the calculation of RPV beltline material CF and 54 EFPY 1/4T ART values are performed in accordance with RG 1.99.
includes the 54 EFPY 1/4T ART data and values for all LSCS, Unit 1, RPV beltline materials and surveillance materials in PTLR, table 7 and all LSCS, Unit 2, RPV beltline materials and surveillance materials in PTLR, table 8, and identifies the limiting 54 EFPY 1/4T ART values for the LSCS, Unit 1 and 2 RPV beltline materials (i.e., limiting 54 EFPY 1/4T ART of 136.3 ºF for LSCS, Unit 1 and 85.7 ºF for LSCS, Unit 2) in section 5.0 of the PTLR.
3.1.5 GL 96-03, Criterion 5 - Description of the Fracture Toughness Methods and Inclusion of P-T Limit Curves in the PTLR GL 96-03, Criterion 5, establishes that the PTLR methodology should describe the application of fracture mechanic methods used in the construction of P-T limit curves based on the methods of analysis in ASME Code, section XI, Appendix G, and in SRP, section 5.3.2. For the contents of the PTLR, GL 96-03, Criterion 5, establishes that the PTLR should provide the P-T limit curves for plant heatups, cooldowns, critical operations, and RCS leakrate or hydrostatic pressure testing.
The NRC staff confirmed that the methodology criteria in chapter 3 of BWROG-TP-11-022-A, includes the applicable fracture mechanics basis for calculating P-T limit heatup and cooldown curves associated with RCS leakrate or hydrostatic pressure test conditions, normal operations
- core non-critical conditions, and normal operations with the unit in the core critical condition.
Consistent with the fracture mechanics methods defined in BWROG-TP-11-022-A, Revision 1, the staff confirmed that the licensee provides its P-T limit heatup/cooldown curves for these
types of operating conditions in PTLR, figures 1 - 3 (i.e., the Curve A, B, and C figures) for LSCS, Unit 1, and in PTLR, figures 4 - 6 (i.e., the Curve A, B, and C figures) for LSCS, Unit 2, with each P-T limit curve figure containing individual P-T limit curves for the beltline, non-beltline, and bottom head regions of the RPV and a composite P-T limit curve that was derived from the most bounding P-T points.
Therefore, based on this review, the NRC staff finds that the PTLR conforms to GL 96-03, Criterion 5, because the PTLR contains the appropriate P-T limit curves for the applicable pressure test, normal - core not critical, and normal - core critical operating conditions.
3.1.6 GL 96-03, Criterion 6 - Incorporation of 10 CFR, Part 50, Appendix G, Minimum Temperature Requirements into Derivation of the P-T Limit Curves GL 96-03, Criterion 6, establishes that the PTLR methodology should describe how the minimum temperature requirements in table 1 of 10 CFR, part 50, appendix G, for the RCS during leakrate or hydrostatic pressure test operating conditions, normal operations of the unit with the unit in the non-critical condition, and normal operations of the unit with the unit in the core critical condition, will be factored into the development of the P-T limit curves. For the contents of the PTLR, GL 96-03, Criterion 6, establishes that the PTLR should identify the minimum temperatures that are accounted for and included in the P-T limit calculations.
Consistent with the minimum temperature criteria defined and discussed in section 2.7 of BWROG-TP-11-022-A, the NRC staff confirmed that the licensee accounted for all appropriate 10 CFR, part 50, appendix G, minimum temperature requirements in the licensees development of the P-T limit curves of the beltline, non-beltline and bottom head regions and the composite P-T limit curves that were included in the Curve A, B, and C P-T limit figures (i.e.,
PTLR, figures 1 - 3 for LSCS, Unit 1, and figures 4 - 6 for LSCS, Unit 2, respectively) of the PTLR.
Therefore, based on this review, the NRC staff finds that the PTLR conforms to GL 96-03, Criterion 6, for inclusion of minimum service temperature requirements in the P-T limit curves because the NRC staff has confirmed that the P-T limit curves appropriately account for the minimum temperature requirements specified in table 1 of 10 CFR, part 50, appendix G, for the applicable operating condition categories.
3.1.7 GL 96-03, Criterion 7 - Incorporation of RPV Surveillance Data into ART Calculations GL 96-03, Criterion 7, establishes that the PTLR methodology should describe how the data from multiple surveillance capsules are used in the ART calculations, and describe the procedure that is used if the measured ART value exceeds the predicted ART value (i.e.,
describe the procedure for assessing data credibility). For the contents of the PTLR, GL 96-03, Criterion 7, establishes that the PTLR should provide the supplemental data and calculations of the CF in the PTLR if the surveillance data are used in the ART calculations, evaluate the surveillance data to determine if they meet the credibility criteria in RG 1.99, and provide the results of the ART calculations.
The NRC staff confirmed that the licensee provided the BWRVIP ISP surveillance data and calculations of CF using the surveillance data by enclosing the BWRVIP-135 report as (non-proprietary version of the report), and Attachment 10 (proprietary version of the report) in the November 22, 2022, LAR submittal. The NRC staff also confirmed that the
licensee provided the 54 EPFY 1/4T ART data and results for the BWRVIP ISP surveillance plate materials and surveillance weld materials representing LSCS, Unit 1, in table 7 of the PTLR.
Similarly, for LSCS, Unit 2, the NRC staff has confirmed that the licensee included applicable 54 EFPY 1/4T ART data and results for the BWRVIP ISP surveillance plate materials and surveillance weld materials representing LSCS, Unit 2, in table 8 of the PTLR.
The NRC staff also confirmed that the licensee included the applicable surveillance data credibility assessments for the designated LSCS, Units 1 and 2, surveillance materials in the BWRVIP-135 report that was submitted with the LAR. Thus, the staff finds that the PTLR conforms to provisions for contents of the PTLR as defined in GL 96-03, Criterion 7, because the licensee appropriately includes the calculations of ART using the applicable BWRVIP ISP surveillance data for LSCS, Units 1 and 2, in PTLR, tables 7 and 8 (respectively,) and has included the applicable surveillance data credibility assessments as part of the BWRVIP-135 provided in the LAR Attachment.
3.2 NRC Staff Evaluation of the PTLR 3.2.1 Evaluation of the Adjusted Nil-Ductility Transition Reference Temperature Values Used in the Derivation of Critical Stress Intensity Values ART values for an evaluated RPV beltline material are calculated for both the assumed crack tips at both the 1/4T and 3/4T wall thickness locations of the evaluated RPV material, as specified in RG 1.99. However, the approved P-T limit methodology in BWROG-TP-11-022-A, has sufficient conservatisms built into the methodology such that the licensee only needed to calculate the 54 EFPY ART values of the LSCS, Units 1 and 2, RPV beltline materials at the 1/4T location of the assessed RPV locations (for evaluated low-pressure coolant injection (LPCI) nozzles and feedwater (FW) nozzles at the 1/4T location of the nozzle inside radius dimension).
For the assessment of ART, the NRC staff performed independent calculations and verified that the licensees 54 EPFY 1/4T ART values for LSCS, Unit 1 and 2, RPV beltline materials were calculated in accordance with BWROG-TP-11-022-A and RG 1.99. The following subsections address the NRC staffs evaluations of the PTLR unit-specific ART results.
3.2.1.1 ART Evaluation for LSCS Units 1 and 2 - Evaluated Beltline Shell Locations LSCS, Unit 1 The NRC staff confirmed that all initial RTNDT values reported for the LSCS, Unit 1, RPV beltline materials in PTLR, table 7, were consistent with those that were provided for the materials in General Electric Report GE-NE-0000-0003-5526-02R1a, Revision 1 (ML041950197), and approved in License Amendment No. 210 for LSCS, Unit 1 (ML14220A517).
In its response to RAI No. LSCS-PTLR-LAR-2, the licensee made an administrative correction of PTLR, table 7, to identify that all three of the LSCS, Unit 1, RPV middle shell axials welds (ID Nos. 3-308 BG, BH and BJ) were fabricated from weld heat number 1P3571 tandem. This change corrected the prior Heat 305424 designation that was provided for RPV middle axial shell weld 3-308 BH in the prior version of table 7 in the November 10, 2022, version of the PTLR, and resulted in changes to the reported Cu-Ni (copper-nickel) chemistry values and CF value for weld 3-308 BH, as amended from previously reported Wt-% (weight percentage) for
Cu-Ni values of 0.273 and 0.629, and a CF of 189 ºF, for the weld in table 7 of the November 22, 2023, version of the PTLR to reported Wt-% Cu-Ni values of 0.287 and 0.756, and a CF of 214 ºF for the same weld material in table 7 of PTLR. The NRC staff confirmed that these chemistry data changes and the change in the material heat number designation for LSCS, Unit 1, middle shell axial weld 3-308 BH are supported by the licensees referencing of the WCAP-15074, Revision 0, report (PTLR Reference 31; ML111861650) for RPV weld chemistries applying to all LSCS, Unit 1, RPV middle shell axial welds with ID No. 3-308 designations (i.e.,
welds 3-308 BG, BH, and BJ) and linked to weld heat No. 1P3571 in the PTLR.
In the response to RAI No. LSCS-PTLR-LAR-2, the licensee also edited the following data or results associated with the ART row at the end of PTLR, table 7, that applies to the designated surveillance weld material for LSCS, Unit 1:
a minor increase in the reported chemistry factor CF value to reflect the proper proprietary surveillance data-based CF value reported for the designated LSCS, Unit 1 surveillance weld material in the BWRVIP-135 report, an increase in the reported 54 EFPY 1/4T neutron fluence from a value of 3.25x1017 n/cm2 (E > 1.0 MeV) to a value of 5.70x1017 n/cm2 (E > 1.0 MeV),
an increase in the reported 54 EFPY 1/4T fluence factor value (f value) in table 7 from a 0.230 to a value of 0.314, and an increase in the reported 54 EFPY 1/4T ART in table 7 from a value of 98.5 ºF to a value of 136.3 ºF; the licensee also amended section 5.0 of the PTLR to identify 136.3 ºF value as the updated, limiting 54 EFPY 1/4T ART for the RPV beltline shell plate and weld materials in LSCS, Unit 1.
To assess the licensees ART calculations, NRC staff performed independent 54 EFPY 1/4T ART calculations for all LSCS, Unit 1 RPV beltline plate, nozzle, and weld materials listed in table 7 of the PTLR. The NRC staff verified that, for LSCS, Unit 1, the beltline is limited by the ART value reported for the LSCS, Unit 1, RPV surveillance weld material at the end of PTLR, table 7, as referenced in the evaluation for the material in the BWRVIP-135 report. The NRC staff calculated a surveillance data-based CF value for the LSCS, Unit 1 surveillance weld material that is consistent with the proprietary CF value reported for this surveillance weld material in the BWRVIP-135 report and a 54 EFPY 1/4T ART value of 136.3 ºF reported for the referenced LSCS, Unit 1, surveillance weld material (as calculated by the NRC staff using the CF ratio procedure defined in Position 2.1 of RG 1.99). The staffs 54 EFPY 1/4T ART value of 136.3 ºF is consistent with the value reported for the LSCS, Unit 1 surveillance weld material in PTLR, table 7.
However, the NRC staff acknowledges that there is additional surveillance data for surveillance welds made from the same weld heat number as that for the designated LSCS, Unit 1, surveillance weld material in BWRVIP-86 from the Kewaunee and Maine Yankee (nuclear power plants) RPV material surveillance programs. The NRC staff discussed the subject of the Kewaunee and Maine Yankee sister plant data for the specified LSCS, Unit 1 surveillance weld material during a public meeting with the licensee on August 18, 2023. As has been documented in the staffs public meeting summary of August 29, 2023 (ML23234A234), during the August 18, 2023, public meeting, the licensee informed the staff that it reviewed all relevant data for surveillance weld materials made from the specified (but proprietary) weld heat number
for the LSCS, Unit 1, in BWRVIP-86, Revision 1-A. The staff noted that, as is stated on page 9 of the PTLR, the licensee used the surveillance data evaluated in the BWRVIP-135 to derive the CF value and 54 EFPY 1/4T ART value for the LSCS, Unit 1, surveillance weld material made from the weld heat number. The NRC staff verified that use of the data reported in BWRVIP-135 yields an acceptable 54 EFPY 1/4T ART value of 136.3 ºF for the LSCS, Unit 1, surveillance weld material, as calculated using the surveillance data-based methods of ART analysis (i.e., using procedure 1) in appendix A of BWROG-TP-11-022-A, The NRC staff also confirmed that the use of all available surveillance data for the designated surveillance weld material would yield a reported 54 EFPY 1/4T ART value for the designated LSCS, Unit 1, surveillance weld material that is less than that calculated by the licensee (i.e.,
136.3 ºF) using the LSCS, Unit 1, specific data alone. Thus, based on this review, the NRC staff finds that the PTLR provides an acceptable limiting ART value for the evaluated LSCS, Unit 1, RPV beltline materials because:
the licensee has demonstrated that it has performed the applicable 54 EFPY 1/4T ART calculations of the ferritic RPV beltline materials in the LSCS, Unit 1, RPV and all BWRVIP ISP RPV surveillance materials applying to LSCS, Unit 1, in accordance with BWROG-TP-11-022-A and RG 1.99; the licensee has provided the applicable 54 EFPY 1/4T ART value inputs in PTLR, table 7, for all LSCS, Unit 1, RPV beltline materials and for the proprietary RPV surveillance plate and weld materials identified and evaluated for LSCS, Unit 1, in BWRVIP-135, and that were included as referenced LSCS, Unit 1, surveillance materials in PTLR, table 7; the licensee is applying a value of 136.3 ºF as the limiting 54 EFPY 1/4T ART value for the P-T limit analyses of the LSCS, Unit 1, RPV beltline materials, and cites this value as the limiting ART value for LSCS, Unit 1, in section 5.0 of the PTLR; and the NRC staff has performed independent calculations of the 54 EFPY 1/4T ART values for the LSCS, Unit 1, RPV beltline materials, and for the LSCS, Unit 1, surveillance plate and weld materials, and has confirmed that the licensees ART assessments and reported 54 EFPY 1/4T ART values have been performed in accordance with methods of analysis in RG 1.99, and in BWROG-TP-11-022-A and are valid for implementation.
LSCS, Unit 2 The NRC staff confirmed that all initial RTNDT values reported for the LSCS, Unit 2, RPV beltline materials in PTLR, table 8, were consistent with those that were provided for the materials in General Electric Report GE-NE-0000-0003-5526-01R1a, Revision 1 (ML101130372), and approved in License Amendment No. 188 for LSCS, Unit 2 (ML110890368).
For LSCS, Unit 2, ART evaluation, the NRC staff noted that licensees implementation of the BWRVIP ISP for LSCS, Unit 2,and the criteria for performing 1/4T ART calculations in appendix A of BWROG-TP-11-022-A, does not credit or apply any of the surveillance data from the past LSCS, Unit 2 RPV surveillance capsule withdrawal (or surveillance data from past non-LSCS unit capsule withdrawals) for derivation of limiting 54 1/4T ART value (i.e., 85.7 ºF) that was reported for the LSCS, Unit 2, RPV beltline materials in section 5.0 of the PTLR. As has
been explained in SE, sections 3.1.2 and 3.1.4, this is based on the licensees determination and the NRC staffs confirmation that the LSCS, Unit 2, RPV beltline plate and weld materials selected as the RPV target materials for LSCS, Unit 2, in BWRVIP-86 are made from different plate or weld heat numbers from those used in the fabrication of the corresponding non-LSCS host plant surveillance materials for the selected target materials in BWRVIP-86.
Thus, the NRC staff confirmed that the licensee used the non-surveillance methods of CF and ART analysis in RG 1.99, position 1.1 (and the CF tables in RG 1.99) to derive the CF and 54 EPFY 1/4T ART values reported for LSCS, Unit 2, RPV beltline plates and welds in table 8 of the PTLR. The NRC staff also confirmed that the 54 EFPY 1/4T ART values provided at the end of PTLR, table 8, for the BWRVIP ISP surveillance plate materials and surveillance weld materials representing LSCS, Unit 2 (as referenced to BWRVIP-135), were not used by the licensee to derive the limiting 54 EPFY 1/4T ART value of 85.7 ºF reported for LSCS, Unit 2, in section 5.0 of the PTLR which is due to the lack of a match in the heat numbers between the designated RPV representative surveillance weld and plate materials for LSCS, Unit 2, and the selected RPV target materials for LSCS, Unit 2, as defined in BWRVIP-86.
Thus, based on this review, the staff finds that the PTLR provides an acceptable ART value for the evaluated LSCS, Unit 2, RPV beltline materials because:
the licensee has demonstrated that it has performed the applicable 54 EFPY 1/4T ART calculations for all LSCS, Unit 2, RPV beltline materials in accordance with the methods of analysis in RG 1.99; the licensee has provided the applicable 54 EFPY 1/4T ART inputs and values in PTLR, table 8, for all LSCS, Unit 2, RPV beltline materials and for the proprietary RPV surveillance plate and weld materials that were identified and evaluated for LSCS, Unit 2, in the BWRVIP-135 report and included as referenced surveillance materials for LSCS, Unit 2, in PTLR table 8; the licensee is applying a value of 85.7 ºF as the limiting 54 EFPY 1/4T ART value for the P-T limit analyses of the LSCS, Unit 2, RPV beltline materials and cites this value as the limiting ART for LSCS, Unit 2, in section 5.0 of the PTLR ; and the NRC staff has performed independent calculations of the 54 EFPY 1/4T ART values for the LSCS, Unit 2 RPV beltline materials, and for the LSCS, Unit 2, surveillance plate and weld materials, and has confirmed that the licensees ART assessments and reported 54 EFPY 1/4T ART values have been performed in accordance with methods of analysis in RG 1.99 and in BWROG-TP-11-022-A and are valid for implementation.
3.2.1.2 ART Evaluation for LSCS, Units 1 and 2 - Evaluated Beltline Nozzle Discontinuity Locations LSCS, Unit 1 For the evaluation of the LSCS, Unit 1, RPV beltline nozzle materials, the NRC staff performed independent 54 EFPY 1/4T ART calculations of the LSCS, Unit 1, N6 LPCI nozzle materials and the RPV N12 water level instrumentation nozzle materials (henceforth, instrument nozzles or instrument nozzle materials when referring to the materials). The NRC staff verified that PTLR table 7 provided 54 EFPY 1/4T ART values for all ferritic N6 LPCI nozzle materials and all ferritic
materials associated with the N12 instrument nozzles, with a limiting 54 EFPY 1/4T ART value of 24.5 °F reported for the LPCI nozzles and 25.7 °F reported for the instrument nozzles. These ART evaluations are based on the CF tables and methods of CF analysis in position 1.1 of RG 1.99. For the instrument nozzles, the NRC staff verified that the licensee used the most limiting 54 EFPY 1/4T ART value (i.e., 25.7 ºF) reported for the ferritic RPV middle shell plates adjoined to the instrument nozzles for the ART assessment because instrument nozzles are made from stainless steel materials and the associated nozzle-to-plate welds are made from Nickel-alloy weld filler materials, which do not require ART evaluations under the requirements of 10 CFR, part 50, appendix G. However, the NRC staff noted that the licensees application of a limiting ART of 25.7 ºF for the instrument nozzles (using the limiting value from the adjacent plates) is acceptable because the licensee appropriately indicates in PTLR, section 5.0, that it performed P-T limit stress analyses of the instrument nozzles to account for potential stress increases that may be caused by configurations of the instrument nozzles in the RPV. Also, the specified ART value is needed to derive the critical stress intensity (KIc) values of the instrument nozzles, used as part of the inputs for the P-T limit calculations for RPV beltline regions in the PTLR.
Thus, based on this review, the NRC staff finds that the PTLR provides an acceptable ART basis for the evaluated LSCS, Unit 1, RPV beltline nozzle materials because:
the licensee is applying the limiting 54 EFPY 1/4T ART values of 24.5 ºF, and 25.7 ºF as the ART inputs for the KIc assessments of LSCS, Unit 1, RPV LPCI nozzles and instrument nozzles and the basis for the ART values is consistent with the staff-approved methodology in BWROG-TP-11-022-A and the methods for ART analysis in RG 1.99.
LSCS, Unit 2 Similar to the LSCS, Unit 1, assessment, the NRC staff performed independent 54 EFPY 1/4T ART calculations of the LSCS, Unit 2, N6 LPCI nozzles materials and N12 instrumentation nozzle materials. The NRC staff verified that PTLR, table 8, provided 54 EFPY 1/4T ART values for all ferritic LSCS, Unit 2, N6 LPCI nozzle materials and all ferritic materials associated with the N12 instrument nozzle materials, with a limiting 54 EFPY 1/4T ART value of 34.0 °F reported for the LPCI nozzles and 68.3 °F reported for the instrument nozzles, as calculated using the CF tables and methods of CF analysis in position 1.1 of RG 1.99. For the limiting 54 EFPY 1/4T ART value reported for the LSCS, Unit 2 instrument nozzles, the value is appropriately established and based on the most limiting 54 EFPY 1/4T ART value reported for the ferritic RPV beltline shell plates (i.e., RPV lower-intermediate shell plates) adjoined to the instrument nozzles, as the instrument nozzles themselves are made from stainless steel materials and the associated nozzle-to-plate welds are made from nickel-based alloy weld filler materials.
Thus, based on this review, the NRC staff finds that the PTLR provides acceptable ART values for the evaluated LSCS, Unit,2 RPV beltline nozzle materials because:
the licensee is applying the limiting 54 EFPY 1/4T ART values of 34.0 ºF, and 68.3 ºF as the ART inputs for the KIc assessments of LSCS, Unit 2, RPV LPCI nozzles and instrument nozzles, and
the basis for the ART values is consistent with the staff-approved methodology in BWROG-TP-11-022-A and the methods for ART analysis in RG 1.99.
3.2.1.3 ART Evaluation for LSCS, Units 1 and 2 - Evaluated Non-Beltline Locations LSCS, Unit 1 The NRC staff confirmed that the PTLR applies the following initial ART values (initial RTNDT values) for the limiting RPV non-beltline locations in LSCS, Unit 1:
12 ºF for the LSCS, Unit 1, RPV closure flange region, 40 ºF for the LSCS, Unit 1, RPV feedwater nozzle (as the limiting upper vessel non-beltline component), and 47 ºF for the LSCS, Unit 1, RPV bottom head.
The NRC staff confirmed the initial RTNDT values for these non-beltline locations were consistent with those previously reported in General Electric Report No. GE-NE-0000-0003-5526-02R1a and approved in License Amendment No. 210 for LSCS, Unit 1.
In RIS 2014-11, the NRC staff summarized a regulatory position that the development of P-T limits for the RPV must consider not only the RPV shell material with the highest ART, but also other RPV materials with structural discontinuities. Specifically, the RIS identifies that for compliance with 10 CFR, part 50, appendix G, all ferritic components within the entire [RPV]
must be considered in the development of P-T limits, and the effects of neutron radiation must be considered for any locations that are predicted to experience a neutron fluence exposure greater than 1x1017 n/cm2 (E > 1 MeV) at the end of the licensed operating period. Thus, based on the guidance provided in the RIS, the NRC staff also confirmed that the initial RTNDT values of these RPV non-beltline materials did not need to be adjusted for neutron irradiation exposure impacts because the PTLR demonstrates that the peak projected inside surface fluences for the non-beltline materials will be less than a fluence of 1x1017 n/cm2 (E > 1 MeV) at 54 EPFY. The NRC staff finds the initial RTNDT values for the specified RPV non-beltline materials to be acceptable because they are consistent with the initial RTNDT values that were approved for these RPV non-beltline materials in License Amendment No. 210 for LSCS, Unit 1.
LSCS, Unit 2 The NRC staff confirmed that the PTLR applies the following initial RTNDT values for the limiting RPV non-beltline locations in LSCS, Unit 2:
26 ºF for the LSCS, Unit 2, RPV closure flange region, 40 ºF for the LSCS, Unit 2, RPV feedwater nozzle (as the limiting upper vessel non-beltline component), and 58.6 ºF for the LSCS, Unit 2, RPV bottom head.
The NRC staff also confirmed that the initial RTNDT values for these non-beltline locations were consistent with the initial RTNDT values previously reported for these non-beltline locations in
General Electric Non-Proprietary Report No. GE-NE-0000-0003-5526-01R1a and approved in License Amendment No. 188 for LSCS, Unit 2.
Consistent with the NRC staffs summary of the RIS 2014-11 guidelines in the corresponding subsection (above) for the LSCS, Unit 1, RPV non-beltline materials, the NRC staff also confirmed that the initial RTNDT values of these LSCS, Unit 2, non-beltline materials did not need to be adjusted for neutron irradiation exposure impacts because the PTLR demonstrates that the peak projected inside surface fluences for the non-beltline materials will be less than a fluence of 1x1017 n/cm2 (E > 1 MeV) at 54 EPFY. The NRC staff finds the initial RTNDT values for the specified LSCS, Unit 2 RPV non-beltline materials to be acceptable because they are consistent with the initial RTNDT values that were approved for these LSCS, Unit 2 RPV non-beltline materials in License Amendment No. 188 for LSCS, Unit 2.
3.2.2 NRC Staff Evaluation of the Contents of the PTLR 3.2.2.1 PTLR and Selected Methodology for Performing the P-T Limit Calculations The newly proposed administrative controls section in TS, section 5.6.7 (as proposed in (ML22332A450), of the November 10, 2022, LAR submittal), indicates that the P-T limits will be calculated in accordance with the P-T limits methodology in BWROG-TP-11-022-A. The NRC staff confirmed that section 1.0 of the PTLR indicates that P-T limits, RCS heatup and cooldown rates, and RPV head closure flange boltup requirements were calculated in accordance with the approved methodology in BWROG-TP-11-022-A.
The NRC staff confirmed that the PTLR identifies that the licensee used the material properties in the ASME Code), section II, part D, Material Properties, 2001 Edition (inclusive of the year 2003 Addenda) as the basis for the unirradiated material properties for evaluated ferritic plate, nozzle, and weld materials. The staff finds the selection of this version of ASME Code, section II, to be acceptable because the licensee is applying an acceptable edition of the ASME Code, section II, for selection of the material properties, as endorsed for use in 10 CFR 50.55a.
3.2.2.2 NRC Staff Evaluation of P-T Limit Curves for the RPV Beltline Regions The PTLR states that the P-T limit curves for the RPV beltline regions in the Curve A, B, and C figures (i.e., figures 1 - 3 for LSCS, Unit 1, and figures 4 - 6, for LSCS, Unit 2, in the PTLR) were developed to bound all ferritic materials with projected neutron fluence exposures exceeding a fluence of 1x1017 n/cm2 (E > 1.0 MeV) at 54 EPFY, including considerations of stress levels from structural discontinuities in the beltline region of the RPVs, such as nozzles.
For LSCS, Unit 1, this includes:
the RPV middle shell plates, lower-intermediate shell plates, and lower shell plates (including the axial seam welds adjoining the plates in middle, lower-intermediate, and lower shell courses, the circumferential seam weld adjoining the middle shell and lower-intermediate shell courses, and the circumferential seam weld adjoining the lower-intermediate shell and lower shell courses),
the N6 LPCI nozzles and associated nozzle-to-vessel welds, and N12 instrument nozzles and associated nozzle-to-vessel welds.
For LSCS, Unit 2, this includes:
the RPV lower-intermediate shell plates and lower shell plates (including the axial seam welds adjoining the plates in the lower-intermediate and lower shell courses and the circumferential seam weld adjoining the lower-intermediate shell and lower shell courses),
the N6 LPCI nozzles and associated nozzle-to-vessel welds, and the N12 instrument nozzles and associated nozzle-to-vessel welds.
In the response to RAI No. LSCS-PTLR-LAR-3, the licensee indicated that it had reconciled the 54 EFPY P-T limit curves provided for the RPV beltline regions in the P-T limit Curve A, B, and C figures (i.e., figures 1 - 3 for LSCS, Unit 1, and figures 4 - 6, for LSCS, Unit 2, in the PTLR) to be based on component-specific curves for the specified RPV beltline shell and nozzle locations evaluated in SI Calculation 2001063.305, Revision 2 (including the LPCI nozzles and instrument nozzles). The NRC staff confirmed based on its review of the response to RAI No.
LSCS-PTLR-LAR-3, that the licensee appropriately amended the P-T limit beltline curves for the RPV beltline regions in Curve A, B, and C figures of the PTLR from those previously provided for the beltline curves in corresponding figures of the November 22, 2023, version of the PTLR.
For the set of P-T limit beltline curves in the PTLR (inclusive of these P-T limit curve changes),
the NRC staff performed independent P-T limit calculations of the curves applying to the beltline regions of the LSCS, Units 1 and 2, RPVs under Curve A hydrostatic/leakrate pressure test conditions and Curve B normal operations - core not critical conditions. The NRC staff applied the methods of calculation for RPV beltline shell and nozzle discontinuity locations in BWROG-TP-11-022-A as part of the basis for the NRC staffs independent P-T limit calculations of the Curve A and B P-T limit curves. For Curve C P-T limit curves (curves for normal operations with the core in the critical condition), the regulations in table 1 of 10 CFR, part 50, appendix G, require the contours of the Curve C P-T limit curves to be at least 40 ºF higher than contours of the Curve B P-T limit curves; therefore, for Curve C P-T limit curves, the NRC staff performed confirmatory activities to verify whether the contours of the Curve C P-T limit curves in the PTLR were at least 40 ºF higher that the contours of the corresponding Curve B P-T limit curves in the PTLR.
Based on the NRC staffs independent calculations, the staff observed that the Curve A and B P-T limit curves provided for the RPV beltline regions in PTLR, figures 2 and 3, for LSCS, Unit 1, and PTLR, figures 4 and 5, for LSCS, Unit 2, were consistent with those calculated by the NRC staff. For Curve C P-T limit curves, the NRC staff verified the contours of the Curve C P-T limit curves (i.e., curves in PTLR, figure 3, for LSCS, Unit 1, and in PTLR, figure 6, for LSCS, Unit 2) were at least 40 ºF higher that the contours of the corresponding Curve B P-T limit curves in the PTLR (i.e., curves in PTLR, figure 2 for LSCS, Unit 1, and in PTLR, figure 5, for LSCS, Unit 2). Thus, the NRC staff verified that the licensees P-T limit curves for the RPV beltline regions in the PTLR addressed the following criteria and incorporated the following inputs defined in the BWROG-TP-11-022-A report:
limiting ART values KIc values for evaluated RPV beltline shell, LPCI nozzle, and instrument nozzle locations;
incorporation of applicable RPV stress inputs on applied pressure stress intensity (KIp) values and thermal stress intensity (KIt) values, including those for RPV LPCI nozzles and instrument nozzles (as evaluated RPV beltline discontinuity regions), as referenced by the guidelines in RIS 2014-11 and performed in accordance with methods for evaluating RPV discontinuity stress inputs in BWROG-TP-11-022-A; for the P-T limit curves of the beltline regions in the Curve A PTLR figures for hydrostatic or leakrate pressure testing conditions (i.e., PTLR, figure 1, for LSCS, Unit 1, and figure 4, for LSCS, Unit 2), inclusion of a maximum heatup/cooldown rate limit of 25 ºF/hr and a safety factor (SF) of 1.5 on applied KIp values; and for the P-T limit curves of the beltline regions in the Curve B P-T limit figures for normal operations - core not critical (i.e., PTLR, figure 2, for LSCS, Unit 1, and figure 5 for LSCS, Unit 2) and in the Curve C P-T limit figures for normal operations - core critical (i.e., PTLR Figure 3 for LSCS, Unit 1, and figure 6 for LSCS, Unit 2), inclusion of a maximum heatup/cooldown rate limit of 100 ºF/hr and a SF of 2.0 on applied KIp values.
Based on its independent calculations, the NRC staff finds that the LAR provides sufficient demonstration that the licensees updated P-T limit curves and associated P-T data for the RPV beltline regions provided in PTLR, figures 1 - 3 for LSCS, Unit 1, and figures 4 - 6, for LSCS, Unit 2, have been performed consistent with the methods in BWROG-TP-11-022-A for RPV beltline region locations. The NRC staffs request in RAI No. LSCS-PTLR-LAR-3 is resolved with respect to the licensees updated P-T limit curves for the RPV beltline regions in PTLR, figures 1 - 6.
3.2.2.3 NRC Staff Evaluation of P-T Limit Curves for the RPV Non-Beltline Regions and Bottom Head Regions Appendix G to 10 CFR, part 50, requires that the P-T limit curves be developed for ferritic reactor RCPB materials that are not part of the RPV beltline region.
In RIS 2014-11, the NRC staff clarified that P-T curve calculations for ferritic RPV materials that are not part of the RPV beltline shell region may define P-T curves that are more limiting than those calculated for the RPV beltline shell materials.
The licensee stated that P-T limit curves for the non-beltline regions in the Curve A, B, and C P-T Limit Figures (i.e., in PTLR, figures 1, 2, and 3, for LSCS, Unit 1, and figures 4, 5, and 6, for LSCS, Unit 2) are represented by the RPV FW nozzles because they are the limiting RPV non-beltline components. The NRC staff confirmed that the licensee used finite element analyses to develop and calculate the KIp, KIt, and Pallow values used in the construction of the P-T curves for the RPV FW nozzles, as consistent with:
the methods of analysis for BWR-design FW nozzles in BWROG-TP-11-022-A, the finite element diagrams provided for the FW nozzles in the PTLR, and application of the finite element model-derived KIp-applied and KIt values provided for the LSCS Unit 1 and 2 FW nozzles in PTLR Table 9.
In the response to RAI No. LSCS-PTLR-LAR-1, the licensee amended the LAR to provide the configurational design dimensions used as dimensional inputs in the Pallow calculations of the LSCS, Units 1 and 2, FW nozzles, as given in the GE Hitachi Report No. 007N8832, Revision 0, that was submitted and included in Attachment 9 of the June 6, 2023, LAR supplement. The NRC staff confirmed that the configurational design parameters cited in the RAI response as input parameters for the FW nozzle assessments were consistent with those identified as input parameters for FW nozzle Pallow calculations in BWROG-TP-11-022-A for actual corner radius dimensions of the FW nozzles. The NRC staff confirmed that the dimensions provided are given in GE Report No. General Electric Report GE-NE-0000-0003-5526-02-R1a for LSCS, Unit 1, FW nozzles and GE Report No. General Electric Report GE-NE-0000-0003-5526-01R1a for LSCS, Unit 2, FW nozzles. Based on confirmation of the FW nozzle corner radius dimensions, the NRC staff considers RAI No. LSCS-PTLR-LAR-1 to be resolved.
In the response to RAI No. LSCS-PTLR-LAR-3, the licensee indicated (in part) that the basis calculation information and associated PTLR information have been reconciled, as shown in in tables 1 - 6 and figures 1 - 6, of the PTLR with the most recent version of the basis calculation being that in SI Calculation 2001063.305, Revision. 2. The licensee also stated that the calculations abide by the methodology approved in the BWROG-TP-11-022-A.
Based on the response to RAI No. LSCS-PTLR-LAR-3, the NRC staff reviewed the P-T limit curves that were included for the RPV non-beltline regions and bottom head regions in the Curve A, B, and C P-T limit figures of the PTLR (i.e., PTLR, figures 1 - 3, for LSCS, Unit 1, and figures 4 - 6, for LSCS, Unit 2), and the associated P-T limit data that were reported for the non-beltline regions and bottom head regions in tables 1 - 3 (for LSCS, Unit 1) and 4 - 6 (for LSCS, Unit 2) of PTLR. The NRC staff performed sample independent calculations of the P-T limit curves for the RPV bottom head and non-beltline regions under Curve A hydrostatic/leakrate pressure test conditions and Curve B normal operations - core not critical conditions in order to confirm that the P-T limit curves for the non-beltline and bottom head regions in the Curve A and B P-T limit figures (i.e., PTLR, figures 1 and 2 for LSCS, Unit 1, and figures 4 and 5 for LSCS, Unit 2) were calculated in accordance with the approved methodology in BWROG-TP-11-022-A. For the P-T limit curves of the bottom head and non-beltline regions in the Curve C figures (i.e., figure 3 for LSCS, Unit 1, and figure 6 for LSCS, Unit 2), the NRC staff performed confirmatory analyses to verify whether the contours of the Curve C P-T limit curves in the PTLR were at least 40 ºF higher that the contours of the corresponding Curve B P-T limit curves in the PTLR.
Based on its independent calculations, the NRC staff observed that the P-T limit curves for the bottom head and non-beltline regions provided in the Curve A and B P-T limit figures (i.e.,
PTLR, figures 2 and 3 for LSCS, Unit 1, and figures 4 and 5 for LSCS, Unit 2) were at least as conservative as those calculated by the NRC staff. The NRC staff also confirmed that the P-T limit curves for the bottom head and non-beltline regions provided in the Curve C P-T limit figures (i.e., PTLR figure 3 for LSCS, Unit 1, and figure 6 for LSCS, Unit 2) were at least 40 ºF higher that the contours of the corresponding Curve B P-T limit curves for the bottom head and non-beltline regions in the PTLR (i.e., those in PTLR, figure 2 for LSCS, Unit 1, and figure 5 for LSCS, Unit 2). Thus, the NRC staff finds that the licensee has provided sufficient information in the PTLR that the licensees P-T limit curves for the RPV non-beltline and bottom head regions in the PTLR were calculated in accordance with the methods calculation and analysis in the BWROG-TP-11-022-A including methods for addressing RPV FW nozzle and bottom head nozzle stress discontinuity considerations, as referenced in RIS 2014-11. Also, the NRC staff finds that the beltline curves incorporated the minimum SF values on applied KIp values required
by ASME Code, section XI, Appendix G, and the limits on maximum heatup/cooldown rates established in BWROG-TP-11-022-A.
Based on its independent calculations, the NRC staff finds that the LAR provides sufficient demonstration that the updated P-T limit curves of the non-beltline and bottom head regions in Curves A, B, and C P-T Limit figures (i.e., PTLR, figures 1, 2, and 3, for LSCS, Unit 1, and figures 4, 5, and 6, for LSCS, Unit 2, respectively) have been appropriately constructed using the methods of analysis for constructing the P-T limits of the non-beltline and bottom head regions in the BWROG-TP-11-022-A.
The NRC staff also confirmed that the PTLR addresses the RPV closure flange as an additional non-beltline region that must be considered and evaluated for derivation of the P-T limit curves.
The NRC staff evaluation of compliance with minimum temperature requirements (as defined in table 1 of 10 CFR, part 50, appendix G) is provided in SE, subsection 3.2.2.4.
3.2.2.4 Composite P-T Limit Curves, Minimum Temperature Requirements, and Considerations for Ferritic RCPB Components That are Not Part of the RPV Composite Curves and Additional Curve C Requirements Items 2.d and 2.e in table 1 of 10 CFR, part 50, appendix G, require that temperature points of P-T curves applying to the normal operations - core critical conditions (i.e., P-T limit curves in the PTLR Curve C figures) be at least 40 °F greater than those developed for P-T limit curves applying to normal operations - core not critical conditions (i.e., P-T limit curves in the PTLR Curve B figures). The NRC staff verified that the P-T limit subcategory curves for the RPV beltline, non-beltline, and bottom head regions and the resulting composite P-T limit curves in the Curve C P-T limit figures (i.e., PTLR, figure 3 for LSCS, Unit 1, and figure 6 for LSCS, Unit
- 2) are constructed to be 40 °F higher than the corresponding curves in Curve B P-T limit figures (i.e., PTLR, figure 2 for LSCS, Unit 1, and figure 5 for LSCS, Unit 2). Thus, the NRC staff finds that the proposed P-T limit curves in the Curve C P-T limit figures satisfy the Items 2.d and 2.e of table 1, 10 CFR, part 50, appendix G, requirements.
Incorporation of Minimum Temperature Requirements in the P-T Limit Curves The NRC staff verified that the minimum temperature requirements in 10 CFR, part 50, appendix G, into the composite P-T limit curves that were included in the Curve A, B, and C figures of the PTLR (i.e., PTLR, Figures 1 - 3 for LSCS, Unit 1, and figures 4 - 6 for LSCS, Unit
- 2) as discussed below.
Items 1.a and 2.a in table 1 of 10 CFR, part 50, appendix G, as applicable to P-T limit curves for hydrostatic or leakrate pressure test conditions or normal operations - core not critical conditions, require the lowest service temperature at operating pressures less than or equal to
() 20 % of the preservice hydrostatic test pressure to be set to a value equivalent or greater than the highest ART value of the material in the RPV closure flange region that is highly stressed by the bolt preload. Item 2.e in table 1 of 10 CFR, part 50, appendix G (BWRs only), as applicable to P-T limit curves of BWRs during for normal operations - core critical conditions, require the lowest service temperature at operating pressures 20% of the preservice hydrostatic test pressure to be set a value equivalent or greater than the highest ART value of the material in the RPV closure flange region that is highly stressed by the bolt preload plus 60
ºF.
For the portions of the RPV P-T limit curves in the Curve A, B, and C PTLR figures at operating pressures 20% of the preservice hydrostatic test pressure (i.e., 20% of 1563 psig (pounds per square inch gauge) or 312.6 psig), the NRC staff confirmed that the licensee met the applicable minimum temperature requirements through inclusion of vertical line at 72 ºF in the composite P-T limit curves of the Curve A, B, and C figures for LSCS, Unit 1 (i.e., in PTLR, figures 1 - 3) and at 86 ºF in the composite P-T limit curves of the Curve A, B, and C figures for LSCS, Unit 2 (i.e., in PTLR, figures 4 - 6). The NRC staff noted that these values are based on the licensees addition of 60 ºF to the initial RTNDT values of the limiting RPV closure flange materials in LSCS, Units 1 and 2 (i.e., limiting initial RTNDT of 12 ºF for the LSCS, Unit 1, closure flange and 26 ºF for the LSCS, Unit 2, closure flange) and satisfy all Items 1.a, 2.a, and 2.e, minimum temperature requirements specified for BWR type P-T limits in table 1 of 10 CFR part 50, appendix G.
Item 1.b in table 1 of 10 CFR, part 50, appendix G, requires that, when the pressure is > 20% of preservice hydrostatic test pressure during hydrostatic or leakrate pressure testing of the reactor, the minimum temperature must be greater than or equal to a value set by the sum of highest ART of the material in the closure flange region that is highly stressed by the bolt preload plus 90 ºF (i.e., limiting initial RTNDT value of the closure flange + 90 ºF). For the portions of the RPV P-T limit curves in the Curve A figures (P-T curve figures for hydrostatic or leakrate pressure test conditions) at operating pressures > 20% of the preservice hydrostatic test pressure (i.e., > 20% of 1563 psig or > 312.6 psig), the NRC staff confirmed that the licensee met the applicable minimum temperature requirement through inclusion of a vertical minimum temperature line at 102 ºF in the Curve A Figure for LSCS, Unit 1 (i.e., in PTLR, figure
- 1) and at 116 ºF in the Curve A figure for LSCS, Unit 2 (i.e., in PTLR, figure 4). The NRC staff confirmed that these values are based on the licensees addition of 90 ºF to the limiting initial RTNDT of 12 ºF for the LSCS, Unit 1, closure flange and to the limiting initial RTNDT of 26 ºF for the LSCS, Unit 2 closure flange.
Item 2.b in table 1 of 10 CFR, part 50, appendix G, requires that when the reactor operating pressure is > 20% of preservice hydrostatic test pressure during normal operations when the reactor core is not in a critical condition, the minimum temperature must be greater than or equal to a value set by the sum of the highest ART of the material in the closure flange region that is highly stressed by the bolt preload plus a value of 120 ºF (i.e., limiting initial RTNDT value of the closure flange + 120 ºF). For the portions of the P-T limit curves in the Curve B figures (P-T curve figures for normal operations - core not critical conditions) at operating pressures > 20% preservice hydrostatic test pressure (i.e., > 20% of 1563 psig or > 312.6 psig),
the staff confirmed that the licensee met the applicable minimum temperature requirement through inclusion of a vertical minimum temperature line at 132 ºF in the Curve B figure for LSCS, Unit 1 (i.e., in PTLR, figure 2) and at 146 ºF in the Curve B figure for LSCS, Unit 2 (i.e.,
in PTLR, figure 5). The NRC staff confirmed that these values are based on the licensees addition of 120 ºF to the limiting initial RTNDT of 12 ºF for the LSCS, Unit 1, closure flange and to the limiting initial RTNDT of 26 ºF for the LSCS, Unit 2, closure flange.
Item 2.d in table 1 of 10 CFR, part 50, appendix G, requires that when the reactor operating pressure is > 20% of preservice hydrostatic test pressure during normal operations when the reactor core is in a critical condition, the minimum temperature must be greater than or equal to a value set by the sum of the highest ART of the material in the closure flange region that is highly stressed by the bolt preload plus a value of 160 ºF (i.e., limiting initial RTNDT value of the closure flange + 160 ºF). For the portions of the P-T limit curves in the Curve C figures (P-T curve figures for normal operations - core critical conditions) at operating pressures > 20%
preservice hydrostatic test pressure (i.e., > 20% of 1563 psig or > 312.6 psig), the NRC staff
confirmed that the licensee met the applicable minimum temperature requirement through inclusion of a vertical minimum temperature line at 172 ºF in the Curve C Figure for LSCS, Unit 1 (i.e., in PTLR, figure 3) and at 186 ºF in the Curve C figure for LSCS, Unit 2 (i.e., in figure 6).
The NRC staff confirmed that these values are based on the licensees addition of 160 ºF to the limiting initial RTNDT of 12 ºF for the LSCS, Unit 1, closure flange and to the limiting initial RTNDT of 26 ºF for the LSCS, Unit 2, closure flange.
Thus, based on these assessments provided in the previous bulleted paragraphs, the staff has confirmed that the P-T limit curves in the PTLR have met the minimum temperature requirements specified in table 1 of 10 CFR, part 50, appendix G, for the applicable operating condition types and specified operating pressures.
Ferritic RCPB Components Outside of the RPV The SE for the BWROG Report No. BWROG-TP-11-022-A states that licensees should confirm that all ferritic RCPB components that are not part of the RPV will not define a more restrictive operating temperature than the proposed P-T limits. The NRC staff verified that, in section 4.0 of the PTLR, the licensee states that the P-T limit curves bound the lowest service temperature (LST) for ferritic non-RPV components of the reactor coolant pressure boundary... Therefore, the NRC staff finds that the PTLR includes the applicable confirmation (as referenced in the staffs SE for BWROG Report No. BWROG-TP-11-022-A) that the lowest service temperature values included in the PTLR for LSCS, Units 1 and 2 (i.e., 72 ºF for LSCS, Unit 1, and 86 ºF for LSCS, Unit 2) are bounding for all ferritic components of the reactor coolant pressure boundaries that are not part of the RPVs.
3.2.2.5 Confirmation of Other TS, Section 3.4.11, Limits Relocated into the PTLR The NRC staff confirmed that the licensee relocated the following additional RCS limits (previously specified in TS, section 3.4.11) into the section 4.0 of PTLR:
maximum heatup/cooldown rate for Curve A P-T limit curves applying to hydrostatic or leakrate pressure test conditions: heatup/cooldown rate 25 ºF/hr (hour),
maximum heatup/cooldown rate for Curve B P-T limit curves applying to normal operations - core not critical conditions or Curve C P-T limit curves applying to normal operations - core critical conditions: heatup/cooldown rate 100 ºF/hr, RPV bottom head coolant temperature-to-RPV coolant temperature differential (T) limit required during recirculation pump startups: T 145 ºF, and RCS recirculation loop coolant temperature-to-RPV coolant temperature differential (T) limit required during recirculation pump startups: T 50 ºF The NRC staff noted that, for the relocated maximum heatup/cooldown rate limit applying to reactor heatup or cooldown operations during ASME hydrostatic or leakrate pressure testing conditions, the limit is an amended limit criterion (i.e., 25 ºF/hr) from the previous limit value of 20 ºF/hr specified for these types of conditions in TS, section 3.4.11.1.c. The staff finds the amended limit value of 25 ºF/hr to be acceptable for implementation because it is consistent with the maximum heatup/cooldown rate limit of 25 ºF/hr specified for ASME Code hydrostatic
or leakrate pressure testing conditions in the BWROG-TP-11-022-A. For the other relocated RCS limits referenced in this SE section, the staff confirmed that the specified limits are consistent with those previously stated and accepted for the applicable RCS parameters in the subsections of TS, section 3.4.11, and that no revisions to the specified limit values are necessary for relocation and validity of these limits in the contents of the PTLR. Thus, the NRC staff finds the values specified for these relocated limits in the PTLR to be acceptable for implementation because either:
the limits are consistent with those previously accepted for the applicable RCS parameters in subsections of TS, section 3.4.11, and the staff has confirmed that the specified limit values do not require any revision for the objectives of the LAR, or for the amended and relocated limit of 25 ºF/hr for heatups and cooldown operations during ASME hydrostatic or leakrate pressure testing conditions, the maximum heatup/cooldown rate limit value is consistent with that specified for these types of conditions in BWROG-TP-11-022-A.
3.3 NRC Staff Evaluation Conclusion 3.3.1 Relocation of P-T Limits Based on its evaluation as documented above, the NRC staff finds P-T limit curves for the RPV beltline, non-beltline, and bottom head regions, and the final composite P-T limit curves in the Curves A, B, and C figures of the PTLR (i.e., PTLR, figures 1 - 3 for Unit 1 and figures 4 - 6 for Unit 2) are acceptable because the licensee has provided adequate demonstration that:
the P-T limit curves incorporate and satisfy all requirements as defined and specified in section IV.A of 10 CFR, part 50, appendix G, including the requirement in 10 CFR Part 50, Appendix G, that requires the calculations of the ARTs to account for the impacts of neutron irradiation exposures and to incorporate the results of applicable RPV surveillance data from the licensees 10 CFR, part 50, appendix H, RPV materials surveillance program (i.e., the BWRVIP ISP for the LSCS units),
the P-T limit curves for the beltline, non-beltline, and bottom head regions and composite P-T limit curves in the Curve A, B, and C P-T limit figures of PTLR have been calculated in accordance with the approved methods in BWROG-TP-11-022-A, the contents of PTLR are in conformance with all criteria in GL 96-03 for technical contents of PTLRs that apply to BWR designs, and the PTLR does not need to address the criteria for LTOP system setpoints in Criterion 3, GL 96-03, as the criteria only apply to PWR light water reactor designs.
The NRC staff also finds that the licensee has relocated all other RCS limits (previously specified in the subsections of TS, section 3.4.11) into the PTLR, and that the revised maximum heatup/cooldown rate limit established for heatup and cooldown operations during ASME Code hydrostatic or leakrate pressure testing conditions (i.e., a maximum rate change of 25 ºF/hr) is consistent with the maximum heatup/cooldown rate limit set for these types of conditions in BWROG-TP-11-022-A.
3.3.2 Conformance with TSTF-419:
TSTF-419, recommends that reference to NRC-approved topical reports used in the PTLR methodology be cited in the TSs using the full citation, including revision number and date of the topical report. TSTF-419 has since been incorporated into NUREG-1434, Revision 4, Volume 1, (ML12104A195).
The NRC staff noted that while LSCS TSs are formatted according to a precursor to the current format of NUREG-1434, the differences in format and numbering of requirements between LSCS TS and NUREG-1434 do not affect the applicability of TSTF-419 to LSCS TS. The NRC staff reviewed the proposed changes to LSCS TS and determined that the proposed revisions properly reference BWROG-TP-11-022-A, Revision 1, dated August 2013. As such, the NRC staff finds the proposed changes to TS are acceptable and appropriately adopt TSTF-419 in a format commensurate with the LSCS TS format.
TSTF-419 was developed based on changes to Revision 2 of NUREG-1434. The NRC staffs review of this LAR includes consideration of whether the proposed changes are consistent with the latest revision, NUREG-1434, Revision 5, Standard Technical Specifications for General Electric BWR/6 Plants (ML21271A582), which provides example TS LCOs and acceptable remedial actions that meet the requirements in 10 CFR 50.36(c)(2)(i) for a standard plant design.
Regarding the licensees variation described in section 2.2 of this SE, the NRC staff finds that although the vendor-supplied reports for both LSCS units are combined in one report, the Reactor Vessel Pressure and Temperature Limits, including heatup and cooldown rates, are analyzed separately for each of the LSCS units. Plant-specific calculated limits are reported to remain within the applicability of TSTF-419 as determined above, therefore, the NRC staff finds that variation regarding removal of phrase unit specific is editorial and acceptable.
3.
3.3 NRC Staff Conclusion
Based on information submitted, the NRC staff has determined that (1) the proposed P-T curves in the PTLR for LSCS, Units 1 and 2 considered all ferritic RPV materials, (2) the proposed P-T curves are constructed based on the methodology in BWROG-TP-11-022-A, Revision 1; BWROG-TP-11-023-A, Revision 0; SRP 5.3.2; and the ASME Code, section XI, appendix G, and (3) the proposed LSCS, Units 1 and 2 PTLR satisfy 10 CFR, part 50, appendices G and H, 10 CFR 50.36, 10 CFR 50.60, GL 96-03, and TSTF-419. Therefore, the NRC staff concludes that the TSs, as amended by the proposed changes, will continue to meet the requirements of 10 CFR 50.36.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations on, the Illinois State official was notified of the proposed issuance of the amendment on October 13, 2023. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement 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 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 March 7, 2023 (88 FR 14181). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR, section 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's 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: James Medoff, NRR Christopher Jackson, NRR Date of Issuance: November 8, 2023
ML23286A260 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/STSB/BC NRR/DNRL/NVIB/BC NAME RKuntz SRohrer SMehta (A)
ABuford DATE 10/13/2023 10/ 16 /2023 10/20/2023 10/9/2023 OFFICE NRR/DSS/SNSB/BC OGC NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME PSahd ANaber JWhited RKuntz DATE 9/15/2023 11/3/2023 11/8/2023 11/8/2023