ML25128A279
| ML25128A279 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/12/2025 |
| From: | Ilka Berrios Plant Licensing Branch III |
| To: | Rhoades D Constellation Energy Generation |
| Wall, S | |
| References | |
| EPID L-2025-LLR-0046 | |
| Download: ML25128A279 (1) | |
Text
May 12, 2025 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:
BRAIDWOOD STATION, UNITS 1 AND 2 - ISSUANCE OF ALTERNATIVE REQUESTS RR-3 ASSOCIATED WITH POST-MAINTENANCE TESTING OF AIR-OPERATED CONTROL VALVE FOURTH INSERVICE TESTING PROGRAM INTERVAL (EPID L-2025-LLR-0046)
Dear Mr. Rhoades:
By letter dated April 24, 2025 (Agencywide Documents Access and Management System Accession No. ML25114A247), Constellation Energy Generation, LLC (CEG, the licensee) submitted alternative request RR-3 regarding certain inservice testing (IST) requirements of the 2012 Edition of the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants, Division 1, OM Code: Section IST (OM Code) for the IST Program at Braidwood Station (Braidwood), Unit 1, for the Fourth IST Program Interval.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee requested to use the proposed alternatives in request RR-3 on the basis that complying with the requirements of the ASME OM Code would result in hardship without a compensating increase in the level of quality and safety.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that CEG has adequately addressed the requirements in 10 CFR 50.55a(z)(2). Therefore, the NRC staff approves the use of RR-3 until the end of the next Braidwood, Unit 1 refueling outage scheduled in the fall of 2025, and the end of the next Braidwood, Unit 2 refueling outage scheduled in the spring of 2026.
All other requirements in the ASME OM Code for which relief was not specifically requested and approved in this request remains applicable.
If you have any questions, please contact the Senior Project Manager, Scott Wall, at 301-415-2855 or e-mail at Scott.Wall@nrc.gov.
Sincerely, Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456 and STN 50-457
Enclosure:
Safety Evaluation cc: Listserv ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.05.12 15:18:05 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE REQUEST RR-4 POST-MAINTENANCE TESTING OF AIR-OPERATED CONTROL VALVE FOURTH INSERVICE TESTING PROGRAM INTERVAL CONSTELLATION ENERGY GENERATION, LLC BRAIDWOOD STATION, UNITS 1 AND 2 DOCKET NOS. STN 50-456 AND STN 50-457
1.0 INTRODUCTION
By letter dated April 24, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25114A247), Constellation Energy, LLC (CEG, the licensee) submitted Alternative Request RR-3 to the U.S. Nuclear Regulatory Commission (NRC) regarding certain inservice testing (IST) requirements of the 2012 Edition of the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants, Division 1, OM Code: Section IST (OM Code) for the IST Program at Braidwood Station (Braidwood), Units 1 and 2, for the Fourth IST Interval Program.
Specifically, the licensee requested an alternative to the manual test requirements in the ASME OM Code as incorporated by reference in Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.55a, Codes and standards, for certain Auxiliary Feedwater (AF) air-operated flow control valves 1(2)AF005 in Braidwood, Units 1 and 2, on the basis that performance of those tests at this time would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety under 10 CFR 50.55a(z)(2). The licensees submittal requested a one-time IST Program manual testing interval extension for the 1AF005G-H valves until the end of the next Braidwood, Unit 1 refueling outage scheduled in the fall of 2025, and a one-time IST Program manual testing interval extension for the 2AF005A-H valves until the end of the next Braidwood, Unit 2 refueling outage scheduled in the spring of 2026.
2.0 REGULATORY EVALUATION
The NRC regulations in 10 CFR 50.55a(f)(4), Inservice testing standards requirement for operating plants, state in part:
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 [10 CFR 50.55a(f)(2) and (3)] and that are incorporated by reference in [10 CFR 50.55a(a)(1)(iv)], to the extent practical within the limitations of design, geometry, and materials of construction of the components.
The NRC regulations in 10 CFR 50.55a(z), Alternatives to codes and standards requirements, state:
Alternatives to the requirements of [10 CFR 50.55a(b) through (h)] 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.
3.0 TECHNICAL EVALUATION
3.1 Licensees Alternative Request RR-3 Applicable ASME OM Code Edition The applicable Code of record for the Fourth IST Interval Program at Braidwood is the 2012 Edition of ASME OM Code as incorporated by reference in 10 CFR 50.55a.
ASME OM Code Components Affected In its submittal, the licensee proposed alternative testing for the following valves:
Table 1 Valve Number Description ASME Code Class OM Code Category 1AF005G Unit 1 Aux Feedwater Air Operated Flow Control Valve G 3
B 1AF005H Unit 1 Aux Feedwater Air Operated Flow Control Valve H 3
B 2AF005A Unit 2 Aux Feedwater Air Operated Flow Control Valve A 3
B 2AF005B Unit 2 Aux Feedwater Air Operated Flow Control Valve B 3
B 2AF005C Unit 2 Aux Feedwater Air Operated 3
B Valve Number Description ASME Code Class OM Code Category Flow Control Valve C 2AF005D Unit 2 Aux Feedwater Air Operated Flow Control Valve D 3
B 2AF005E Unit 2 Aux Feedwater Air Operated Flow Control Valve E 3
B 2AF005F Unit 2 Aux Feedwater Air Operated Flow Control Valve F 3
B 2AF005G Unit 2 Aux Feedwater Air Operated Flow Control Valve G 3
B 2AF005H Unit 2 Aux Feedwater Air Operated Flow Control Valve H 3
B Applicable ASME OM Code Requirements ASME OM Code, Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants, paragraph ISTC-3540, Manual Valves, states:
Manual valves shall be full-stroke exercised at least once every 2 yr [years],
except where adverse conditions may require the valve to be tested more frequently to ensure operational readiness. Any increased testing frequency shall be specified by the Owner. The valve shall exhibit the required change of obturator position.
Licensees Proposed Alternative The licensees proposal in Alternative Request RR-3 is as follows:
CEG is proposing an extension associated with performing the manual exercise of 1AF005G-H, and 2AF005A-H in accordance with 10 CFR 50.55a(z)(2) on the basis that compliance results in hardship or unusual difficulty without a compensating increase in level of quality or safety. The requested extension allows performing the manual stroke testing prior to the end of the units next refueling outage to mitigate the risk associated with a loss of the AF system design basis function. LCO [Limiting Condition for Operation] 3.7.5 for the AF system is only applicable in Modes 1, 2, and 3. Performing the flow control valve testing prior to the end of each units next refueling outage mitigates the risk of having to shutdown the associated unit due to an inoperable AF train, since the unit would already be shutdown and out of the AF system LCOs Mode of Applicability.
Licensees Basis for Use The licensees basis for use of the proposal in Alternative Request RR-3 is as follows:
The AF system provides cooling water flow to the steam generators [SGs] to allow for decay heat removal during emergency conditions. Performing valve strokes requires closing the associated AF flow-control valve and entry in LCO 3.7.5 Required Action for the associated train.
Closing of the 1(2)AF005s via the handwheel is not required for the AF system to meet its design safety function to provide flow to the SG in accordance with LCO 3.7.5, but further closure of the valve via the handwheel may be necessary to ensure margin to overfill during a SGTR [steam generator tube rupture].
A history of IST performance and manual valve strokes using air over the previous 7 years was reviewed. The 1(2)AF005 valves have been manually stroked successfully during the previous test interval in 2022 (Unit 1 valves) and in 2023 (Unit 2 valves). The manual stroke was added to the IST program in 2022 following a violation received during the Design Basis Assurance Inspection (ML22164A905). The 1(2)AF005 valves undergo quarterly stroke time testing via the air operator. Since 2018, none of the valves have failed these tests due to exceeding their operating limit.
This testing history provides reasonable assurance the AOVs [air-operated valves] are currently operable and ready to perform their safety function to provide feedwater to the SGs to remove decay heat from the Reactor Coolant System upon the loss of normal feedwater supply. In the event of a SGTR, the 1(2)AF013 valves function as the primary isolation of feedwater flow to the ruptured steam generator. In the event the 1(2)AF013 fails to close, the associated 1(2)AF005 valve is closed with air and has its manual handwheel closed to ensure isolation. Based on this testing history and the back-up nature of the manual handwheel operation, CEG concludes that the proposed extension is acceptable and the extension of the AF AOV manual, handwheel stroke will not result in an adverse consequence to safety.
Licensees Reason for Request The licensees reason for Alternative Request RR-3 is as follows:
In accordance with 10 CFR 50.55a, Codes and Standards, paragraph (z)(2),
relief is requested from the requirement of ASME OM Code ISTC-3540. The basis for this relief request is that existing testing requirements result in hardship and unusual difficulty without a compensating increase in the level of quality and safety.
The 1(2)AF005A-H valves control the flow of Auxiliary Feedwater (AF) water to the steam generators (SG). In the event of an accident, the 1(2)AF005 valves can be controlled from the main control room or the remote shutdown panel (RSDP). In the event of a steam generator tube rupture (SGTR), feedwater flow to the ruptured steam generator will be isolated via the motor-operated isolation valves, 1(2)AF013 valves, with the air-operated control valves, 1(2)AF005 valves, as backup. Step 4 of 1/2BwEP-3 (Steam Generator Tube Rupture) directs the operator to isolate AF flow to the affected steam generator by closing the associated 1(2)AF013 valves and setting the demanded flow for the associated 1(2)AF005 valves to 0, closing them. In the event the associated 1(2)AF013 does not close as expected, the operator would go to the response not obtained column where they would be directed to close the associated 1(2)AF005 valve using the handwheel, ensuring the valve does not open if instrument air is lost. In this circumstance, the 1(2)AF005 would be taken closed by air from either the instrument air system or the installed safety-related accumulators prior to operators taking field action.
During normal plant operation the AF system is in standby with the 1(2)AF005A-H valves set to control flow at approximately 170 gallons per minute (gpm) upon actuation. Closing of the 1(2)AF005 valves via the handwheel is not required for the AF system to meet its design safety function to provide flow to the SGs in accordance with Limiting Condition for Operation (LCO) 3.7.5 and the Braidwood Station design basis, but ensuring closure of the valve (following closure via instrument air system or installed accumulators) via the handwheel is credited in the safety analysis to ensure margin to overfill during a SGTR.
The 1(2)AF005 valves have two separate actuation methodologies. The valve can be stroked remotely via air or manually via the handwheel. These two methodologies and the actuation sequences are detailed in Attachment 2. Based on the actuator design, the manual handwheel stroke mechanism and air operation stroke mechanism share a common linkage. Degradation of the actuator that impacts the manual handwheel stroke mechanism can affect the air operated stroke mechanism and can ultimately prevent the valve from being capable of changing position remotely via air.
Failure of any one 1(2)AF005 valve such that it becomes stuck closed as a result of cycling with the handwheel would prevent the associated train from providing feedwater flow to the downstream steam generator during a plant transient resulting in entry into LCO 3.7.5, Two trains of AF remain OPERABLE.
The 1(2)AF005 valves at Braidwood and Byron share a common valve design, and recent operating experience at both stations demonstrates a trend in challenges with manual actuation of the 1(2)AF005 valves from the handwheel.
[The licensee provided specific operating experience examples from Braidwood and Byron nuclear power plants.]
The window for manual stroke testing of the Unit 1 flow control valves expired on 4/20/2025. All the Unit 1 valves except for 1AF005F-H were manually stroke tested during this IST interval. 1AF005A-E were manual full-stroked with no issues noted. Operations commenced stroke testing of 1AF005F, but abandoned the test when increased resistance was noted as described below. The manual stroke testing of the corresponding Unit 2 valves expires on 10/26/2025. This is the late date, including 6 months grace, to maintain compliance with ASME OM code and Code Case OMN-20.
Based on the above information and component specific operating experience, manually exercising the 1(2)AF005 valves utilizing the handwheel has the potential to impact the ability of the valve to be re-positioned and can impact the ability of the Operators to control AF flow remotely. Subsequently, this could challenge the ability of the AF train to provide forward feedwater flow to the downstream steam generator. Full stroking the 1(2)AF005 valve using the handwheel risks failure of the valve in a non-open position. It is the position of the CEG that this risk outweighs the risk of delaying testing and present undue risk without a corresponding increase in safety.
3.2
NRC Staff Evaluation
In Alternative Request RR-3, the licensee proposed an extension of the interval for the performance of the manual exercise test of the 1AF005G-H valves and 2AF005A-H valves at Braidwood, Units 1 and 2, respectively, in accordance with 10 CFR 50.55a(z)(2) on the basis that compliance with the applicable ASME OM Code test requirement would result in a hardship or unusual difficulty without a compensating increase in the level of quality or safety. In particular, the licensee requested a one-time IST Program test interval extension of the manual exercise test for the 1AF005G-H valves until the end of the next Braidwood, Unit 1 refueling outage scheduled for the fall of 2025, and for the 2AF005A-H valves until the end of the next Braidwood, Unit 2 refueling outage scheduled for the spring of 2026.
At Braidwood, the AF system provides cooling water flow to the SGs to allow for decay heat removal if needed in response to a plant event. During an SGTR event, feedwater flow to the ruptured SG will be isolated using the motor-operated isolation valves 1(2)AF013 with the air-operated control valves 1(2)AF005 as a backup. If the associated 1(2)AF013 valve does not close as expected, the reactor operator would close the associated 1(2)AF005 valve. The reactor operators would close the 1(2)AF005 valves by air from either the instrument air system or the installed safety-related accumulators prior to the reactor operators taking field action. The licensee stated that closing the 1(2)AF005 valves using the handwheel is not required for the AF system to meet its design safety function to provide flow to the SGs in accordance with LCO 3.7.5 and the Braidwood design basis. However, ensuring closure of the 1(2)AF005 valves (following closure via the instrument air system or installed accumulators) using the handwheel is credited in the safety analysis to ensure margin to overfill during an SGTR event.
Based on the actuator design of the 1(2)AF005 valves, the manual handwheel stroke mechanism and air operation stroke mechanism share a common linkage. The licensee reported that degradation of the actuator that impacts the manual handwheel stroke mechanism can affect the air-operated stroke mechanism and can ultimately prevent the valve from being capable of changing position remotely using the air supply. Failure of any 1(2)AF005 valve such that it becomes stuck closed as a result of cycling with the handwheel would prevent the associated train from providing feedwater flow to the downstream SG during a plant transient.
The licensees requested extension of the interval for the manual exercise test of the 1AF005G-H valves and 2AF005A-H valves would allow performing the manual stroke testing prior to the end of the units next refueling outage to mitigate the risk associated with a loss of the AF system design-basis function, and the need to shut down the associated Braidwood unit.
In Alternative Request RR-3, the licensee only requested an extension associated with the 2-year interval for performing the manual exercise of the 1AF005G-H valves and 2AF005A-H valves required by ASME OM Code, Subsection ISTC, paragraph ISTC-3540. Therefore, the quarterly stroke-time testing requirements in ASME OM Code, Subsection ISTC, paragraphs ISTC-3510, Exercising Test Frequency, and ISTC-5113, Valve Stroke Testing, continue to apply to these valves. The licensee did not identify any concerns with the capability of the air-operated closure function of these valves being degraded by the current condition of the valve internals associated with handwheel operation.
Based on the information described above, the NRC staff finds that the licensee has justified in Alternative Request RR-3 that performance of the manual exercise test of the 1AF005G-H valves and 2AF005A-H valves at Braidwood, Units 1 and 2, respectively, for the duration of the alternative request would result in a hardship or unusual difficulty without a compensating increase in the level of quality or safety. The licensee has provided information indicating that handwheel operation at this time might degrade the air-operated capability of the 1AF005 valves. The Braidwood corrective action program under 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, applies to the licensees actions to address degradation of the manual operation of the 1(2)AF005 valves. As a result of its review, the NRC staff finds that Alternative Request RR-3 may be authorized under 10 CFR 50.55a(z)(2) until the end of the next Braidwood, Unit 1 refueling outage scheduled in the fall of 2025, and the end of the next Braidwood, Unit 2 refueling outage scheduled in the spring of 2026.
4.0 CONCLUSION
As set forth above, the NRC staff has determined that the licensee has justified in Alternative Request RR-3 that performance of the manual exercise test of the 1AF005G-H valves and 2AF005A-H valves at Braidwood, Units 1 and 2, respectively, at this time would result in a hardship or unusual difficulty without a compensating increase in the level of quality or safety.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(2).
Therefore, the NRC staff approves the use of RR-39 until the end of the next Braidwood, Unit 1 refueling outage scheduled in the fall of 2025, and the end of the next Braidwood, Unit 2 refueling outage scheduled in the spring of 2026.
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 Contributor: Thomas Scarbrough, NRR Date: May 12, 2025
- via memo NRR-028 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DEX/EMIB/BC*
NAME SWall SLent SBailey DATE 05/08/2025 05/9/2025 05/08/2025 OFFICE NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME IBerrios SWall DATE 05/12/2025 05/12/2025