ML24304A960
| ML24304A960 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 11/05/2024 |
| From: | Markley M Plant Licensing Branch II |
| To: | Flippin N, Pigott E Duke Energy Carolinas, Office of Nuclear Reactor Regulation |
| Stone Z | |
| References | |
| EPID L-2024-LLR-0002, EPID L-2024-LLR-0003 | |
| Download: ML24304A960 (11) | |
Text
November 5, 2024 Nicole Flippin Site Vice President Catawba Nuclear Station Duke Energy Carolinas, LLC 4800 Concord Road York, SC 29745 Edward Pigott Site Vice President McGuire Nuclear Station Duke Energy Carolinas, LLC 12700 Hagers Ferry Road Huntersville, NC 28078
SUBJECT:
CATAWBA NUCLEAR STATION, UNITS 1 AND 2, AND MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 - PROPOSED ALTERNATIVE FOR THE INSPECTION OF REACTOR VESSEL CLOSURE HEAD PENETRATIONS (EPID L-2024-LLR-0002 AND L-2024-LLR-0003)
Dear Nicole Flippin and Edward Pigott:
By letter dated January 10, 2024, as supplemented by letter dated August 28, 2024, Duke Energy Carolinas, LLC (Duke Energy, the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of an alternative to certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code),Section XI, requirements at Catawba Nuclear Station, Units 1 and 2, and McGuire Nuclear Station, Units 1 and 2.
The proposed alternative (RA-23-0242) would allow Duke Energy to use the ASME BPV Code Cases N-729-6 Alternative Examination Requirements for PWR [pressurized water reactor]
Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, and N-770-5 Alternate Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, for the requirements of the follow-up examination of the Reactor Vessel Closure Head (RVCH)
Penetration Nozzles subsequent to the performance of peening, and to the depth of compression in the performance of peening, respectively. The supplement letter dated August 28, 2024, removed the alternative minimum nominal compressive residual depth requirement for the Auxiliary Head Adapter RVCH Penetration Nozzles and supersedes the original proposed alternative with RA-24-0193.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee requested to use an alternative on the basis that complying with the specified ASME
N. Flippin, et al.
BPV Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that Duke Energy has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2).
All other ASME BPV Code,Section XI requirements for which an alternative was not specifically requested and authorized remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions, please contact Jack Minzer Bryant, Catawba Licensing Project Manager, at (301)-415-0610 or via email at Jack.Minzerbryant@nrc.gov.
Sincerely, Michael T. Markley, Branch Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos.
50-413, 50-414 50-369, 50-370
Enclosure:
Safety Evaluation cc: Listserv MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2024.11.05 13:58:12 -05'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PROPOSED ALTERNATIVE NO. RA-24-0193 INSPECTION OF REACTOR VESSEL CLOSURE HEAD PENETRATIONS DUKE ENERGY CAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413, 50-414, 50-369, AND 50-370
1.0 INTRODUCTION
By letter dated January 10, 2024 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML24010A033) as supplemented by letter dated August 28, 2024 (ML24241A062), Duke Energy Carolinas, LLC (the licensee) requested the use of an alternative to certain requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code),Section XI, for the requirements of the follow-up examination of the Reactor Vessel Closure Head (RVCH) Penetration Nozzles subsequent to the performance of peening, and to the depth of compression in the performance of peening for Catawba Nuclear Station (CNS), Units 1 and 2, and McGuire Nuclear Station (MNS), Units 1 and 2.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee requested to use the proposed alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 REGULATORY EVALUATION
2.1 Background
The licensee implemented a peening mitigation process on various RVCH penetrations including control rod drive mechanism (CRDM) penetrations, auxiliary head adapter (AHA) penetrations and the head vent penetration at CNS, Unit 1 in spring 2023 (C1R27), MNS, Unit 2 in spring 2023 (M2R28), CNS, Unit 2 in fall 2022 (C2R25), and at MNS Unit 1 in fall 2023 (M1R29). In accordance with 10 CFR 50.55a(g)(6)(ii)(D)(5), the licensee is required to implement MRP-335, Revision 3-A Materials Reliability Program: Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement, (Electric Power Research Institute (EPRI) submitted MRP-335, Revision 3, for U.S. Nuclear Regulatory Commission (NRC) review on February 19, 2016 (ML16055A216). The NRC staff approved MRP-335, Revision 3, on August 24, 2016 (ML16208A485), and requested an approved version of the topical report be published with an -A designation. MRP-335, Revision 3-A, dated November 7, 2016, is publicly available at EPRI.com under Product ID 3002009241) for the follow-up volumetric examination of these penetrations in order to fully credit the peening mitigation for the subsequent, long-term, volumetric examination frequency. All four RVCHs had accumulated fewer than 8 effective degradation years (EDYs) at the time of peening. So, the authorized peening follow-up volumetric examinations for the CRDM and RVCH vent penetrations, in accordance with 10 CFR 50.55a(g)(6)(ii)(D)(5), are to be performed in the second refueling outage subsequent to peening application (N+2 outage).
However, the licensee was not able to fully qualify the peening mitigation to meet the performance criteria for the AHA penetrations with alloy 82/182 piping butt welds per Section 4.2.8 APPENDIX: Performance Criteria and Measurement or Qualification Criteria for Mitigation by Surface Stress Improvement (Peening) of Alloy 82/182 Piping Butt Welds in PWR
[pressurized water reactor] Primary System Piping, of MRP-335, Revision 3-A. In its submittal dated January 10, 2024, the licensee proposed an alternative qualification criterion for the peening mitigation of the AHA penetration butt welds. In its supplement dated August 28, 2024, the licensee removed this alternative qualification to allow more time to develop its technical basis. This defaults the unmitigated AHA penetration butt welds to the inspection requirements of 10 CFR 50.55a(g)(6)(ii)(F), which mandates use of ASME Code Case N-770-5. Under Examination Item B-1 of Code Case N-770-5, the volumetric examination requirement is every second inspection period not to exceed 7 years.
The licensee still intends to meet the follow-up examination requirements of a peened AHA penetration butt weld as they work to justify the qualification process of the peened AHA penetration butt welds, which would require a volumetric and surface examination no sooner than the third (N+3) refueling outage. Therefore, in order to satisfy the requirements for both unmitigated and peened AHA penetrations, the next examination of the AHA penetrations at each unit are scheduled for the N+3 refueling outage subsequent to peening for each unit.
To eliminate performance of multiple volumetric examinations in a locked high radiation area (LHRA) in two sequential refueling outages (RFO), the licensee is requesting deferral of the CNS, Units 1 and 2, and MNS, Units 1 and 2, post-peening follow-up volumetric examinations for the CRDM and head vent penetrations until the N+3 refueling outage. The licensee supports the request noting previous NRC approvals, the unmitigated volumetric inspection frequency requirements of RVCH penetrations, and a flaw analysis. Further, the licensee confirms there will be no deviation in visual examination requirements, as bare metal visual examinations will be performed each RFO in accordance with MRP-335, Revision 3-A, Section 4.3.4 Subsequent ISI Program, and that each unit follows enhanced leakage rate monitoring to identify leakage rates as low as 0.1 gallon per minute in a timely fashion. As such, due to hardship, and with reasonable assurance of structural integrity, the licensee proposes a one-time alternative examination timing for each units follow-up peening volumetric examinations for the subject penetrations to the N+3 refueling outage following peening application.
2.2 Regulations Regulations in 10 CFR 50.55a(g)(6)(ii)(D)(5) state, in part, that in order for a Reactor Pressure Vessel upper head with nozzles and associated J-groove welds mitigated by peening to obtain examination relief from the requirements of Table 1 for unmitigated heads, peening must meet the performance criteria, qualification, and examination requirements stated in MRP-335, Revision 3-A, with the exception that a plant-specific alternative request is not required and NRC condition 5.4 of MRP-335, Revision 3-A does not apply.
Regulations in 10 CFR 50.55a(g)(6)(ii)(F) state, in part, that holders of operating licenses or combined licenses for pressurized-water reactors as of or after June 3, 2020, shall implement the requirements of ASME BPV Code Case N-770-5 instead of ASME BPV Code Case N-770-2.
Regulations in 10 CFR 50.55a(z) state, in part, that alternatives to the requirements of paragraph (b) through (h) of 50.55a may authorized by the NRC if the licensee demonstrates that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3.0 TECHNICAL EVALUATION
3.1 Licensees Proposed Alternative RA-24-1093 3.1.1 Applicable Code Edition and Addenda Unit ISI Interval ASME ISI Code Edition Interval Start Date Interval End Date License End Date CNS-1 Fourth 2007 Edition, 2008 Addenda 8/19/2015 6/28/2026 12/05/2043 CNS-2 Fourth 2007 Edition, 2008 Addenda 8/19/2015 6/28/2026 12/05/2043 MNS-1 Fifth 2007 Edition, 2008 Addenda 12/01/2021 11/30/2031 06/12/2041 MNS-2 Fifth 2019 Edition 3/01/2024 2/28/2034 3/3/2043 Notes-
- 1. These interval end dates are current estimates, with flexibilities due to transition in fuel cycle period or implementation of ASME Code Case N-921.
- 2. After the first period of the fifth ISI interval, MNs Unit-1 is scheduled to transition to the 2019 Edition of ASME Code Section XI.
3.1.1 ASME BPV Code Requirements ASME Code Case N-729-6 contains requirements for the inspection of J-groove welded RVCH penetration nozzles, with or without flaws, as conditioned by Code of Federal Regulations (CFR) 10 CFR 50.55a(g)(6)(ii)(D).
3.1.1 Proposed Alternative In its letter dated August 28, 2024, the licensee stated, in part, that:
Pursuant to 10 CFR 50.55a(z)(2), Duke Energy is requesting relief from the requirements of 10 CFR 50.55a(g)(6)(ii)(D) for the timing of the follow-up examination of the RVCH penetration nozzles subsequent to the performance of peening (CRDM penetrations #1-78 and head vent). Specifically, Duke Energy is requesting a one-time alternative to the examination frequency requirements of 10 CFR 50.55a(g)(6)(ii)(D)(5), in which a single post-peening follow-up examination of all the penetrations in each RVCH is performed in the third (N+3) refueling outage after peening.
In accordance with 10 CFR 50.55a(g)(6)(ii)(D)(5), Duke Energy shall confirm that the performance criteria, qualification, and examination requirements stated in MRP-335 R3-A are satisfied, with the exception that a plant-specific alternative request is not required and NRC condition 5.4 of MRP-335 R3-A does not apply, prior to obtaining examination relief from the requirements of 10 CFR 50.55a(g)(6)(ii)(D) for unmitigated RVCH penetrations.
3.2
NRC Staff Evaluation
The NRC staff reviewed and evaluated the licensees request on the basis of 10 CFR 50.55a(z)(2), such that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The specified requirement under 10 CFR 50.55a(g)(6)(ii)(D) is performance of a follow-up volumetric examination after two operating cycles following the application of peening on the RVCH penetrations with partial penetration J-groove welds. These include the CRDM and head vent penetrations. The licensee used the same peening mitigation process on the AHA penetrations with alloy 82/182 butt welds located just outboard of the CRDM penetrations on the RVCH. All of these penetrations are located in a LHRA when the plant is shut down and the RVCH is removed. The purpose of the licensees proposed alternative is to defer the regulatory required volumetric examination of the CRDM and head vent penetrations to be in conjunction with the AHA penetration volumetric examination requirements.
3.2.1 Hardship The licensee noted the hardship in performing the examination in two sequential RFOs would add unnecessary radiological dose due to a second entry to the same area. The nozzles to be examined are in a LHRA and estimated personnel radiological exposure to perform a volumetric examination of the AHA nozzles would be approximately 250 to 300 mRem but could be higher if tool breakdowns or issues occurred requiring additional personnel entry. Additional hardship was identified of increased risk of an industrial accident and the potential for contamination exposure when additional entries into a LHRA are required. The NRC staff acknowledges the licensees radiological and industrial safety concerns and finds that the licensee has justified sufficient hardship to meet the requirement of 10 CFR 50.55a(z)(2).
3.2.2 Quality and Safety The NRC staff reviewed the level of quality and safety of the licensees proposed alternative that the volumetric examinations of the subject RVCH penetrations be delayed for one cycle of operation. The licensee provided supporting basis noting (1) previous NRC approvals, (2) the unmitigated volumetric inspection frequency requirements of RVCH penetrations at similar head ages and temperatures, (3) a further flaw analysis to address a previously repaired condition with its impact on examination requirements and (4) defense-in-depth actions. The NRC staff reviewed each of these factors in evaluating the level of quality and safety in the licensees proposed alternative.
The NRC staff notes that the degradation mechanism of concern is leakage of primary coolant containing boric acid from the RVCH penetrations and/or associated J-groove welds. This leakage can cause two issues to challenge the structural integrity of the reactor coolant pressure boundary of the RVCH or penetration nozzles. The first challenge is circumferential cracking, and thereby potential ejection, of a penetration nozzle from the RVCH. This could cause a small break loss-of-coolant accident (LOCA) or control rod misalignment. The second challenge is that the leakage could cause boric acid corrosion of the low alloy steel material that comprises the bulk thickness of the RVCH. Boric acid corrosion rates of low alloy steel could be up to 6 inches/year under very severe conditions as discussed in NRC report, NUREG/CR-6875, Boric Acid Corrosion of Light Water Reactor Pressure Vessel Materials, J.H. Park, O. K. Chopra, K. Natesan, and W. J. Shack, July 2005 (ML052360563). After sufficient corrosion, a small or medium break LOCA could occur. To prevent significant degradation in RVCH and penetration nozzles, 10 CFR 50.55a(g)(6)(ii)(D) requires an inspection program for these components, including volumetric examinations and bare metal visual examinations. The NRC staff further notes that the licensee applied peening on the subject nozzles and associated J-groove weld surfaces, in accordance with MRP-335, Revision 3-A, to mitigate against primary water stress corrosion cracking (PWSCC) initiation in the components.
The NRC staff acknowledges that similar proposed alternatives have been approved to extend the volumetric inspection frequency of the follow-up examinations from the N+2 to N+3 refueling outages following peening mitigation at plants with similar EDYs of operation and RVCH operating temperatures. The five subject plants submitted alternatives under 10 CFR 50.55a(z)(2) with similar radiological hardships for the follow-up examinations to be performed for various different head penetrations in different RFOs. While the previous NRC authorizations were not for precisely the same plant issues, the NRC staff recognizes the previous precedents in plant-specific reviews below.
The NRC staff notes that, if peening was not performed on the subject RVCH penetrations, the next volumetric examination could be performed for the CNS-1, MNS-1 and MNS-2 RVCHs during the N+3 refueling outage following the last volumetric examination. This examination frequency would be less than the required 2.25 Reinspection Years or 8 calendar years of ASME Code Case N-729-6 for cold leg temperature operating RVCHs. The NRC staff notes that because peening was performed, the stress state of the penetration nozzles was changed. The purpose of the follow-up examinations is to verify that no small flaws go undetected prior to extension of the volumetric examination frequency to the long-term alternative of once per interval.
The NRC staff notes for CNS-2 that an indication of primary water stress corrosion cracking (PWSCC) was identified in the J-groove weld surface of RVCH Penetration #74 with an apparent volumetric leak path. Therefore, an embedded flaw repair (EFR) was performed to isolate the defect from the environment which allows PWSCC. The outer surfaces of this penetration were overlaid with alloy 52/52M weld metal, which is highly PWSCC-resistant. As CNS-2 has a cold leg operating temperature RVCH with a penetration that has identified previous PWSCC, per Note 8 of Table 1, CNS-2s RVCH penetration nozzles would require volumetric examination at least once every 36 months, within the N+2 follow-up volumetric examination requirement for a peened RVCH. The NRC staff finds that the EFR of the cracked penetration nozzle and the peening mitigation of the remaining nozzles provides a basis for further evaluation of the follow-up examination frequency requirement, which the licensee performed utilizing the previous NRC-approved plant evaluations and a plant-specific evaluation.
The licensee utilized flaw analyses performed by Dominion Engineering, Inc. Technical Notes TN-4069-00-01, Revision 0 (ML18248A060) and TN-4069-00-02, Revision 0 (ML18270A066) which were submitted previously to support a proposed alternative for Byron Station, Unit 2, and Braidwood, Unit 1, to defer some volumetric examinations. The licensee reviewed the flaw analysis and concluded that it was also applicable to CNS-1, CNS-2, MNP-1, and MNP-2 which all use a nominal 18-month fuel cycle and operate their RVCH at similar cold loop operating temperature conditions. Additionally, to address the issue of one penetration found to have PWSCC at CNS-2, a plant-specific flaw evaluation was provided to evaluate the shortest time to leakage from a hypothetical PWSCC flaw in the penetration nozzles at CNS-2. The time to potential leakage of the hypothetical flaw was beyond the operating time of the N+3 proposed alternative follow-up volumetric examination.
The NRC staff reviewed each of the flaw analyses and finds that the crack growth analyses were based on conservative assumptions and industry-wide crack size measurement data applicable for CNS-1, CNS-2, MNS-1, and MNS-2. The TN-4069-00-01 and TN-4069-00-02 analyses included a matrix of deterministic PWSCC crack growth calculations. The matrix considered various crack growth cases that involve different initial crack sizes, crack aspect ratios, operating temperatures, and severity levels of stress profiles. The crack growth analysis discussed the effectiveness of follow-up volumetric examination to monitor pressure boundary leakage of the nozzles. The analysis further estimates the growth of hypothetical, shallow PWSCC cracks that may have been missed in previous examinations. The reports indicated that extending the currently approved examination schedule by one cycle of operation would result in a very low fraction of cases that could cause nozzle leakage. Further, the report concluded the possibility of leaking nozzles after the follow-up volumetric examination would be the same for examinations performed after two or three cycles of operation. These analyses were developed to bound the fleet of cold head operating temperature plants applicable to CNS-1, CNS-2, MNS-1, and MNS-2. The licensees plant-specific analysis for CNS-2 utilized inputs, assumptions for flaw locations, crack growth rates and methodologies the NRC staff found consistent with previously approved plant-specific PWSCC analyses and ASME Code requirements. Therefore, the NRC staff found the licensees plant-specific analyses and results acceptable.
The NRC staff performed a series of independent evaluations to verify the licensees assessment. Based on MRP-335, Revision 3-A, the NRC staff determined that there is reasonable assurance that peening of the CNS-1, CNS-2, MNS-1, MNS-2 penetration nozzles will mitigate new crack initiation. The NRC staff also determined that the bare metal visual examination of the RVCH to be performed during each refueling outage is sufficient to ensure there is currently no active indication of nozzle leakage, and that any potential leakage will be identified. The NRC staffs independent evaluations found some cases of crack growth and specific weld residual stress profiles where leakage could result if the examination frequency was increased by one cycle of operation. However, the NRC staff evaluations showed insufficient time for these cases, either in the nozzle or J-groove weld, to allow leakage to challenge the structural integrity of the RVCH. The NRC staff bases this conclusion on the need for additional circumferential crack growth for nozzle ejection or the leaking flaw to grow to allow leakage rates to cause boric acid corrosion rates identified in NUREG/CR-6875. Therefore, the NRC staff finds that the conclusions of the licensees assessment are acceptable.
The NRC staff reviewed the operating experience available for other peened RVCH penetration nozzles. The NRC staff noted that no previous indications of PWSCC have been identified during follow-up volumetric examinations of peened RVCH penetration nozzles of similar operating temperature conditions. Further, no indications of leakage were found at these plants through the J-groove weld by the volumetric leak path examination. The NRC staff has determined that these examination results provide additional assurance that the postulated flaw analyses utilized by the licensee and NRC are conservative.
The NRC staff assessed the adequacy of the defense-in-depth of the licensees examination and monitoring requirements to evaluate the structural integrity of the upper head and nozzles.
The NRC staff confirmed that the licensee plans to perform a bare metal visual examination on each nozzle for evidence of pressure boundary leakage every refueling outage in accordance with MRP-335, Revision 3-A. The NRC staff finds that the visual examination is an effective defense-in-depth inspection. The NRC staff also notes that technical specifications of CNS-1, CNS-2, MNS-1, and MNS-2 require operational leakage monitoring. The NRC notes that the licensee implements a lower 0.1 gallon-per-minute leakage action level on unidentified reactor coolant pressure boundary leakage, consistent with WCAP-16465-NP, Pressurized Water Reactor Owners Group Standard RCS [Reactor Coolant System] Leakage Action Levels and Response Guidelines for Pressurized Water Reactors, (ML070310082). The NRC staff finds the ongoing leakage monitoring program at CNS-1, CNS-2, MNS-1, and MNS-2 during the additional cycle of operation provides sufficient basis to ensure the structural integrity of the RVCH and penetration nozzles for the period of the licensees proposed alternative. The NRC staff also notes that if any leakage is identified, it would be required to be repaired and the examination requirements of 10 CFR 50.55a(g)(6)(ii)(D) would be implemented.
Based on the above, the NRC staff finds that the licensee has provided a sufficient basis for hardship and technical basis to justify extending the follow-up volumetric examination of the subject RVCHs for one operating cycle. The NRC staff also finds that the defense-in-depth actions for bare metal visual examination, along with operational leakage monitoring, provides reasonable assurance that the structural integrity of the RVCH and penetration nozzles is maintained. Therefore, the NRC staff finds that complying with the current volumetric examination requirements in a LHRA under the RVCH in the N+2 refueling outage following peening of the CRDM and head vent penetrations along with the N+3 refueling outage following peening of the AHA penetration butt welds would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
4.0 CONCLUSION
As set forth above, the NRC staff determined that complying with the specified ASME BPV Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) and the proposed alternative provides reasonable assurance of structural integrity of the reactor pressure vessel head.
Therefore, the NRC Staff authorizes the use of proposed alternative RA-24-0193 at CNS, Units 1 and 2, and MNS, Units 1 and 2, until the N+3 scheduled refueling outage following peening of the RVCH penetrations at each plant.
All other ASME BPV Code,Section XI requirements for which an alternative was not specifically requested and authorized remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: J. Collins, NRR Date: November 5, 2024
ML24304A960 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DNRL/NPHP/BC NAME JMinzerBryant KGoldstein MMitchell DATE 10/30/2024 10/31/2024 10/28/2024 OFFICE NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME MMarkley JMinzerBryant DATE 11/5/2024 11/5/2024