ML25268A175
| ML25268A175 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 09/26/2025 |
| From: | Markley M Plant Licensing Branch II |
| To: | Coleman J Southern Nuclear Operating Co |
| Kalathiveettil, D | |
| References | |
| EPID L-2025-LLR-0017 | |
| Download: ML25268A175 (1) | |
Text
September 26, 2025 Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Co., Inc.
3535 Colonnade Parkway Birmingham, AL 35243
SUBJECT:
EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 - AUTHORIZATION OF ALTERNATIVE REQUEST RR-V SIXTH INSERVICE TESTING INTERVAL (EPID L-2025-LLR-0017)
Dear Jamie Coleman:
By letter dated January 30, 2025, Southern Nuclear Operating Company (SNC, the licensee) submitted Alternative Request RR-V-4 for the Sixth Inservice Testing (IST) interval at the Edwin I. Hatch Nuclear Plant (Hatch), Units 1 and 2, which is scheduled to begin on January 1, 2026.
The licensee requested Alternative Request RR-V-4 for Hatch, Units 1 and 2, on the basis that the proposed alternative provides an acceptable level of quality and safety for the IST activities for the American Society of Mechanical Engineers (ASME) valves affected. The U.S. Nuclear Regulatory Commission (NRC) staff has completed its review and concludes that SNC has adequately addressed the regulatory requirements set forth in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a, Codes and standards, paragraph (z)(1) for this alternative. Therefore, pursuant to 10 CFR 50.55a(z)(1), the NRC staff authorizes the use of this alternative to the 2022 Edition of the ASME Operation and Maintenance of Nuclear Power Plants, Division 1, OM Code: Section IST (OM Code), as incorporated by reference in 10 CFR 50.55a, for the Code of Record interval, as defined in 10 CFR 50.55a(y), Definitions, that implements the 2022 Edition of the ASME OM Code, for Hatch, Units 1 and 2. Use of this alternative with other Codes of Record is not authorized.
All other ASME OM Code requirements as incorporated by reference in 10 CFR 50.55a for which relief or an alternative was not specifically requested, and granted or authorized (as appropriate), in the subject request remain applicable.
If you have any questions, please contact the Project Manager, Dawnmathews Kalathiveettil at Dawnmathews.Kalathiveettil@nrc.gov or 301-415-5905.
Sincerely, Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-321 and 50-366
Enclosure:
Safety Evaluation cc: Listserv MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.09.26 13:31:10 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AUTHORIZATION OF VALVE ALTERNATIVE REQUEST RR-V-4 SIXTH INSERVICE TESTING INTERVAL SOUTHERN NUCLEAR OPERATING COMPANY EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-321 AND 50-366
1.0 INTRODUCTION
By letter dated January 30, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25030A363), and supplemented by letter dated May 15, 2025 (ML25135A409), Southern Nuclear Operating Company (SNC, the licensee) submitted RR-V-4 to the U.S. Nuclear Regulatory Commission (NRC) requesting authorization for a proposed alternative for specific affected valves in lieu of certain inservice testing (IST) requirements of the 2022 Edition of the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants, Division 1, OM Code: Section IST (OM Code), as incorporated by reference in Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.55a, Codes and standards, for the Sixth Inservice Testing (IST) interval at the Edwin I. Hatch Nuclear Plant (Hatch), Units 1 and 2.
The licensee submitted Alternative Request RR-V-4 pursuant to 10 CFR 50.55a(z)(1),
Acceptable level of quality and safety, on the basis that the proposed alternative will provide an acceptable level of quality and safety for IST activities applied to the specific affected valves within the scope of this request.
2.0 REGULATORY EVALUATION
The NRC regulations in 10 CFR 50.55a(f)(4), Inservice testing standards requirement for operating plants, state, in part, that:
Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves that are within the scope of the ASME OM Code must meet the inservice test requirements (except design and access provisions) set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in paragraphs (f)(2) and (3) of this section
[10 CFR 50.55a] and that are incorporated by reference in paragraph (a)(1)(iv) of this section [10 CFR 50.55a], to the extent practical within the limitations of design, geometry, and materials of construction of the components.
The regulations in 10 CFR 50.55a(z), Alternative to codes and standards requirements, state, in part, that alternatives to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation. A proposed alternative must be submitted and authorized prior to implementation.
The applicant or licensee must demonstrate that:
(1) Acceptable level of quality and safety. The proposed alternative would provide an acceptable level of quality and safety; or (2) Hardship without a compensating increase in quality and safety. Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
NUREG/CR-5928, Interfacing System LOCA (ISLOCA) Research Program (ML072430731),
July 1993.
Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program (ML23073A154), June 2023.
3.0 TECHNICAL EVALUATION
The applicable Code of Record for the Hatch, Units 1 and 2, Sixth IST interval is the 2022 Edition of the ASME OM Code. The Hatch, Units 1 and 2, Sixth IST interval is scheduled to begin on January 1, 2026.
3.1 Licensees Alternative Request RR-V-4 3.1.1 Applicable ASME OM Code Requirements The requirements in the ASME OM Code, Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants, as incorporated by reference in 10 CFR 50.55a, related to Alternative Request RR-V-4 are as follows:
Paragraph ISTC-3630, Leakage Rate for Other Than Containment Isolation Valves, states:
Category A valves with a leakage requirement not based on an Owners 10 CFR 50, Appendix J Program, shall be tested to verify their seat leakages within acceptable limits. Valve closure before seat leakage testing shall be by using the valve operator with no additional closing force applied.
Paragraph ISTC-3630(a), Frequency, states:
Tests shall be conducted at least once every two yr. [years].
3.1.2 Components for Which Alternative is Requested The alternative was proposed for the following components:
Equipment ID Description OM Code Category Type LLRT Type Test Media Comments 1E11-F008 RHR SDC Suction Outboard Isol.
Valve A
CIV/PIV C
Air Class 1 1E11-F009 RHR SDC Suction Inboard Isol. Valve A
PIV ASME Water Class 1 1E11-F015A LPCI Inboard Isolation Valve A
CIV/PIV C
Air Class 1 1E11-F015B LPCI Inboard Isolation Valve A
CIV/PIV C
Air Class 1 1E11-F050A LPCI Injection Check Valve AC PIV ASME Water Class 1 -
Tested with 1E11-F122A 1E11-F050B LPCI Injection Check Valve AC PIV ASME Water Class 1 -
Tested with 1E11-F122B 1E11-F122A RHR F050A Bypass Valve A
Tested with 1E11-F050A 1E11-F122B RHR F050A Bypass Valve A
Tested with 1E11-F050B 1E21-F005A CS Injection Inboard Valve A
CIV/PIV C
Air Class 1 1E21-F005B CS Injection Inboard Valve A
CIV/PIV C
Air Class 1 1E21-F006A CS Injection Check Valve AC PIV ASME Water Class 1 -
Tested with 1E21-F037A 1E21-F006B CS Injection Check Valve AC PIV ASME Water Class 1 -
Tested with 1E21-F037B 1E21-F037A CS F006A Bypass Valve A
Tested with 1E21-F006A 1E21-F037B CS F006A Bypass Valve A
Tested with 1E21-F006B 1E41-F005 HPCI Injection Check Valve AC PIV ASME Water Class 2 1E41-F006 HPCI Injection Outboard Isolation Valve A
CIV/PIV C
Water Class 2 -
Not 0.6 La -
Tested with Outboard FWCV 1E51-F013 RCIC Injection Outboard Isolation Valve A
CIV/PIV C
Water Class 2 -
Not 0.6 La -
Tested with Outboard FWCV 1E51-F014 RCIC Injection Check Valve AC PIV ASME Water Class 2 2E11-F008 RHR SDC Suction Outboard Isolation Valve A
CIV/PIV C
Air Class 1 2E11-F009 RHR SDC Suction Inboard Isolation Valve A
PIV ASME Water Class 1 2E11-F015A LPCI Inboard Isolation Valve A
CIV/PIV C
Air Class 1 2E11-F015B LPCI Inboard Isolation Valve A
CIV/PIV C
Air Class 1 2E11-F050A LPCI Injection Check Valve AC PIV ASME Water Class 1 -
Tested with 2E11-F122A 2E11-F050B LPCI Injection Check Valve AC PIV ASME Water Class 1 -
Tested with 2E11-F122B 2E11-F122A RHR F050A Bypass Valve A
Tested with 2E11-F050A 2E11-F122B RHR F050B Bypass Valve A
Tested with 2E11-F050B 2E21-F005A CS Injection Inboard Valve A
CIV/PIV C
Air Class 1 2E21-F005B CS Injection Inboard Valve A
CIV/PIV C
Air Class 1 2E21-F006A CS Injection Check Valve AC PIV ASME Water Class 1 -
Tested with 2E21-F037A 2E21-F006B CS Injection Check Valve AC PIV ASME Water Class 1 -
Tested with 2E21-F037B 2E21-F037A CS F006A Bypass Valve A
Tested with 2E21-F006A 2E21-F037B CS F006A Bypass Valve A
Tested with 2E21-F006B 2E41-F005 HPCI Injection Check Valve AC PIV ASME Water Class 2 2E41-F006 HPCI Injection Outboard Isolation Valve A
CIV/PIV C
Water Class 2 -
Not 0.6 La -
Tested with Outboard FWCV 2E51-F013 RCIC Injection Outboard Isolation Valve A
CIV/PIV C
Water Class 2-Not 0.6 La -
Tested with Outboard FWCV 2E51-F014 RCIC Injection Check Valve AC PIV ASME Water Class 2 Notes:
RHR = residual heat removal SDC = shutdown cooling LPCI = low pressure coolant injection CS = core spray HPCI = high pressure coolant injection RCIC = reactor core isolation cooling CIV = containment isolation valve PIV = pressure isolation valve FWCV = feedwater check valve 3.1.3 Licensees Proposed Alternative and Basis In its submittal dated January 30, 2025, as supplemented by letter dated May 15, 2025, the licensee states that:
Hatch proposes to perform PIV testing at intervals ranging from every refuel to every third refuel. Hatch also proposes to use Air Tests for valves that are CIVs and PIVs. Some CIVs will be tested with Water and all PIVs that only perform a PIV function will be tested with Water. The specific interval for each valve would be a function of its performance and would be established in a manner consistent with the Containment Isolation Valve (CIV) process under 10 CFR 50 Appendix J Option B. 12 of the 36 valves listed are classified as CIVs and are currently leak tested with air according to 10 CFR 50 Appendix J methodology every 2 years to satisfy their PIV leakage test requirement (with acceptance criteria correlated to water at function maximum pressure differential). Whether the valve is a CIV/PIV or PIV only, the valve must have two consecutive leakage tests which meet its acceptance criteria to be considered a good performer. That is, the test interval may be extended to every third refuel outage upon completion of two consecutive periodic PIV tests with results within the prescribed acceptance criteria. The test interval will be extended to a specific value in a range of frequencies from 30 months up to a maximum of 75 months (as described in NEI 94-01 Revision 3-A). Additionally, the test interval may be extended in accordance with, and subject to the limitations of, NEI 94-01 Revision 3-A. Any test failure will require a return to initial (every RFO) interval until good performance can again be established.
The primary basis for this relief request is historically good performance of the PIVs and desire to reduce personnel exposure dose (ALARA). With the testing being performed every refueling outage has resulted in approximately 180 tests with 2 failures, which yields a failure rate of approximately 1 percent.
Additional basis for this relief request is provided below:
Separate functional testing of power-operated PIVs and Condition Monitoring of Check Valve PIVs per ASME OM Code.
Low likelihood of valve mispositioning during power operations (procedures, interlocks).
Air test vs. water test - degrading seat conditions tend to be identified with air testing.
Relief valves in the low pressure (LP) piping - these relief valves may not provide Inner-System Loss of Coolant Accident (ISLOCA) mitigation for inadvertent PIV mispositioning but their relief capacity can accommodate conservative PIV seat leakage rates.
Alarms that identify high pressure (HP) to LP leakage - Operators are highly trained to recognize symptoms of a present ISLOCA and to take appropriate actions.
By letter dated December 30, 2015 (ML15310A406), a similar alternative request was authorized for Hatch, Units 1 and 2.
3.1.4 Licensees Reason for its Request In its submittal dated January 30, 2025, the licensee states, in part, that:
Pursuant to 10 CFR 50.55a(z)(1), relief is requested from the requirements of the ASME OM Code, 2022 Edition, ISTC-3630(a). While the valves are not in Series, they are not individually tested, due to essentially being in the same flow path.
ISTC-3630(a) requires that leakage rate testing for pressure isolation valves be performed at least once every 2 years. Pressure Isolation Valves (PIVs) are not specifically included in the scope of performance-based testing as provided for in 10 CFR 50 Appendix J Option B. However, OM Code Case OMN-23, Alternative Rules for Testing Pressure Isolation Valves, does allow for performance-based testing of Pressure Isolation Valves. Additionally, NEI 94-01 describes the risk-informed basis for extended test intervals under Option B. That justification shows that for valves which have demonstrated good performance by passing their leak rate tests for two consecutive cycles, further failures appear to be governed by the random failure rate of the component. NEI 94-01 also presents the results of a comprehensive risk analysis, including the statement that risk impact associated with the increasing [leakrate] test intervals is negligible (less than 0.1% of total risk). The valves identified in this relief request are all in water applications, CIV valves are tested in accordance with Appendix J Requirements using air. PIV testing is typically performed at lower pressures, such as for Appendix J Requirements, are acceptable provided the results are extrapolated to system functional differential pressure. Plant Hatch applies the extrapolated values to both PIV and CIV values. This relief request is intended to provide for a performance-based scheduling of PIV tests at Hatch. The reason for requesting this relief is dose reduction / ALARA. Recent historical data was used to identify that PIV testing alone each refuel outage incurs a total dose of approximately 400 millirem.
Assuming all of the PIVs remain classified as good performers the extended test intervals would provide for a savings of 800 mR over a 6 year period per unit.
NUREG 0933 Issue 105 (Interfacing Systems LOCA at LWRs [light water reactors])
discussed the need for PIV leakage rate testing primarily based on the 3 pre-1980 historical failures of applicable valves industry-wide. These failures all involved human errors in either operations or maintenance. None of these failures involved inservice equipment degradation. The performance of PIV leak rate testing provides assurance of acceptable seat leakage with the valve in the closed condition. Typical PIV testing does not identify functional problems which may inhibit the valves ability to reposition from open to close. For check valves, such functional testing is accomplished per ASME OM Code ISTC-3522. Power-operated valves are routinely full stroke tested per ASME OM Code to ensure their functional capabilities. At Plant Hatch, these functional tests for PIVs are performed only at Cold Shutdown or Refuel Outage frequencies.
Such testing is not performed online in order to prevent any possibility of an inadvertent ISLOCA condition. The 24 month functional testing of the PIVs is adequate to identify any abnormal condition that might affect closure capability. Performance of the separate 24 month PIV leak rate testing does not contribute any additional assurance of functional capability, it only determines if seat tightness of the closed valves.
This alternative request is a re-submittal of NRC approved 4th and 5th Interval(s) Relief Request RR-V-10, previously submitted and approved for testing of CIVs and PIVs at Plant Hatch per Technical Specifications. There have been no substantive changes to this alternative or to the basis for use, which would alter the previous NRC Safety Evaluation conclusions for previous IST Intervals for Plant Hatch. (See Precedents for SERs [safety evaluation reports]).
3.2
NRC Staff Evaluation
At nuclear power plants, the PIVs are defined as two valves in a series within the reactor coolant pressure boundary which separate the HP reactor coolant system from an attached LP system. Failure of a PIV could result in an over-pressurization event which could lead to a system rupture and possible release of fission products to the environment. This type of failure event was analyzed under NUREG/CR-5928, Interfacing System LOCA (ISLOCA) Research Program, (ML072430731). The purpose of NUREG/CR-5928 was to quantify the risk associated with an ISLOCA event. NUREG/CR-5928 analyzed boiling water reactor (BWR) and pressurized water reactor (PWR) designs.
10 CFR 50, Appendix J, Option B, references specific guidance concerning acceptable leakage rate test methods, procedures, and analyses that may be used to implement a performance-based leakage test program. The guidance and acceptance criteria are provided in Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program (ML23073A154).
RG 1.163 endorsed, with appropriate regulatory positions, NEI Topical Report 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 2012 (ML12221A202).
The licensee has proposed an alternative test in lieu of the requirements in the ASME OM Code, Subsection ISTC, paragraph ISTC-3630(a) for all of the PIVs listed in the request.
Specifically, the licensee proposed to verify the leakage rate of PIVs using the 10 CFR Part 50, Appendix J, Option B performance-based schedule. Valves would initially be tested at the required interval schedule that ranges from every refueling outage (RFO) to every third RFO.
Valves that have demonstrated good performance for two consecutive cycles may have their test interval extended from every RFO to every third RFO (i.e., 6 years). Any PIV leakage test failure would require the component to be returned to the initial interval of every RFO or 2 years until it can be reclassified as a good performer in accordance with the performance evaluation of Option B. The leakage test interval for these PIVs shall not exceed 75 months with a 3-month grace period based on the performance (i.e., a total of 78 months). The specific interval for each valve will be a function of its performance and will be established in a manner consistent with the CIV process under Option B.
Currently, 12 of the 36 valves in this request are leak tested with air according to 10 CFR Part 50, Appendix J, methodology every 2 years to satisfy PIV leakage test requirements. The NRC staff requested additional information regarding the other 24 valves. By letter dated May 15, 2025 (ML25135A409), the licensee indicated that the remaining 24 valves are being tested every 72 months. The licensee reported that all except four valves (1E11-F050A, 1E11-F050B, 1E21-F005B, and 2E11-F050A) have maintained a history of good performance.
Valves that failed their tests were re-tested each outage until good performance was achieved.
In addition, the licensee routinely functionally tests the full stroke capability of these PIVs in accordance with ASME OM Code requirements, to ensure their close functional capabilities.
Based on the information above regarding licensees reported good valve performance history, coupled with stroking each valve every RFO and the low risk factor, as noted in NUREG/CR-5928, the proposed alternative provides an acceptable level of quality and safety.
In a letter dated May 15, 2025 (ML25135A409), the licensee provided additional information in response to the NRC staffs request for additional information. The licensee referenced ASME OM Code Case OMN-23, Alternative Rules for Testing Pressure Isolation Valves, within the submittal, but the licensee clarified that the code case was not considered as part of the alternative request. Additionally, the licensee clarified language regarding the frequency interval extensions permitted NEI-94-01, Revision 3-A. The staff evaluated Alternative Request RR-V-4 within the context of the additional information provided in the letter dated May 15, 2025.
Based on the above, the NRC staff has determined that Alternative Request RR-V-4 may be authorized pursuant to 10 CFR 50.55a(z)(1) on the basis that the proposed alternative provides an acceptable level of quality and safety, for the specific valves within the scope of this request, in lieu of the applicable valve IST requirements in the ASME OM Code.
The NRC staff notes that the applicable code of record for the Hatch Units 1 and 2, Sixth IST interval, is the 2022 Edition of the ASME OM Code, as incorporated by reference in 10 CFR 50.55a. The Hatch, Units 1 and 2, Sixth IST interval, is scheduled to begin on January 1, 2026.
The ASME OM Code (2022 Edition), Subsection ISTA, General Requirements, paragraph ISTA-3120, Inservice Examination and Test Interval, as incorporated by reference in 10 CFR 50.55a, requires that licensees implement 10-year intervals for their IST programs. Although not requested, the NRC regulations in 10 CFR 50.55a allow licensees to implement the same ASME OM Code as their Code of Record for two consecutive IST program intervals.
4.0 CONCLUSION
As set forth above, the NRC staff determined that proposed alternative numbered RR-V-4 for Hatch, Units 1 and 2, provides an acceptable level of quality and safety for IST activities for the specific affected valves within the scope of the request. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1) for this proposed alternative. Therefore, pursuant to 10 CFR 50.55a(z)(1), the NRC staff authorizes RR-V-4 for the Sixth IST Interval for the specified valves within the scope of the request in lieu of the applicable IST requirements in the 2022 Edition of the ASME OM Code, as incorporated by reference in 10 CFR 50.55a, for Hatch, Units 1 and 2, for the Code of Record interval, as defined in 10 CFR 50.55a(y), Definitions, that implements the 2022 Edition of the ASME OM Code. Use of this alternative with other codes of record is not authorized.
All other ASME OM Code requirements as incorporated by reference in 10 CFR 50.55a for which relief or an alternative was not specifically requested, and granted or authorized (as appropriate), in the subject request remain applicable.
Principal Contributors: N. Hansing, NRR T. Scarbrough, NRR G. Bedi, NRR Date: September 26, 2025
- via eConcurrence NRR-028 OFFICE NRR/DORL/LPL2-1/PM*
NRR/DORL/LPL2-1/LA*
NRR/DEX/EMIB/BC NAME DKalathiveettil KZeleznock SBailey DATE 9/24/2025 9/26/2025 7/16/2025 OFFICE NRR/DORL/LPL2-1/BC*
NRR/DORL/LPL2-1/PM*
NAME MMarkley DKalathiveettil DATE 9/26/2025 9/26/2025