ML25063A119

MD 8.3 Evaluation for Dresden Ccsw MOV Fto 2-27-2025

ML25063A119
Person / Time
Site:	Dresden
Issue date:	02/27/2025
From:	Billy Dickson NRC/RGN-III/DORS/RPB1
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ML25063A140	List: ML25063A119
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MD 8.3 Evaluation Decision Documentation for Reactive Inspection (Deterministic and Risk Criteria Analyzed)

PLANT:

Dresden Unit 2 EVENT DATE:

06/20/2024 DETERMINISTIC CRITERIA EVALUATION DATE:

02/27/2025 Brief Description of the Significant Operational Event or Degraded Condition:

Note: Baseline inspection and evaluation of the subject issue/valve have been continuous since December 2024. Upon discovery in November 2024, a team of regional decisionmakers met initially to evaluate entry criteria for the MD 8.3 process and it was determined that entry was not warranted at the time. As inspectors continued their review and new information became available regarding plausible failure mechanisms, subsequent meetings were reconvened to determine if entry into the MD 8.3 process was warranted. Such new information was discovered by inspectors on February 25, 2025, regarding the potential for a valve hammering phenomenon to potentially be causal, for which an evaluation team reconvened and entered the MD 8.3 process on February 26, 2025.

On June 20, 2024, the licensee completed a past operability review to assess a failure of motor operated valve 2-1501-3A, Unit 2 containment cooling heat exchanger division I tube side discharge (containment cooling service water). The past operability review concluded the valve, which failed to open on April 30, 2024, had been inoperable from April 2, 2024, through May 1, 2024. The licensee further reviewed the past operability assessment on November 27, 2024, and concluded the event was reportable under 10 CFR 50.73. The licensee documented the issue in Licensee Event Report (LER) 237/2024-001-00 on January 27, 2025. The LER also mentioned a similar failure on the same valve, which occurred in November of 2023, and is being investigated separately.

The licensee concluded the failure-to-open resulted from low margin on a valve position feedback potentiometer within the valves control system. The licensees long-term corrective actions included reducing the valves stroke length to improve the potentiometer margin. The inspectors review of the event (which began in December 2024) found the licensee performed its lowest level of causal evaluation for the event and the licensee failed to evaluate other credible failure mechanisms associated with the potentiometer feedback assembly. The licensee, after discussions with inspectors, planned to investigate the additional failure mechanisms in late March, when the valve would be out of service for planned maintenance.

On February 18, 2025, the licensee identified a weld crack between the actuator mounting plate and the top of the valve yoke. During the maintenance to repair the weld, the licensee also identified the valve had a bent stem adapter. The licensee replaced the bent stem adapter and repaired the valve yoke and actuator mounting surface. The licensee reviewed past valve testing data and did not identify an overthrust condition. The licensee has not yet definitively determined the cause of the failed weld or bent stem adapter but preliminarily believes it to be due to a latent weld failure.

On February 25, 2025, inspectors identified Dresden-specific operating experience that showed the current torque switch circuit design created the potential for a valve hammering phenomenon to occur and inspectors believe that it could potentially be causal. Given this new information, the Region III evaluation team reconvened and entered the MD 8.3 process on February 26, 2025.

2 The licensee informed the Region that a Root Cause Investigation would be conducted and installed condition monitoring instrumentation (high speed recorders connected to monitor motor current for the valve actuators of both Unit 2 CCSW heat exchanger outlet valves to collect 90 days of data) to provide objective evidence of the existence/non-existence of the valve hammering phenomena.

This issue is being followed by the resident inspectors with assistance from DORS engineering specialists and the NRR subject matter expert for valves/actuators.

Y/N DETERMINISTIC CRITERIA

1. Involved operations that exceeded, or were not included in, the design bases of the facility N

Remarks:

2. Involved a major deficiency in design, construction, or operation having potential generic safety implications N

Remarks: None were identified during the review of this issue.

3. Led to a significant loss of integrity of the fuel, primary coolant pressure boundary, or primary containment boundary of a nuclear reactor N

Remarks:

4. Led to the loss of a safety function or multiple failures in systems used to mitigate an actual event N

Remarks: Although this issue involved the failure of one train (Division 1) of a multi-train system, the other train (Division 2) remained functional, thereby maintaining safety-function.

5. Involved possible adverse generic implications N

Remarks: None identified during the inspection.

6. Involved significant unexpected system interactions N

Remarks: None observed during the event or identified during the inspection.

7. Involved repetitive failures or events involving safety-related equipment or deficiencies in operations Y

Remarks: As noted in the event description above, the valve failed to open on April 30, 2024, and had a similar failure in November of 2023. Additionally, given the site-specific history of valve hammering being causal to MOV failures in the 1980s, and instances of hammering being observed by operators in the field as recently as 2021 during a surveillance test window, inspectors continue to question whether hammering occurs while the system is in a standby state. The licensees recent installation of condition monitoring instrumentation will provide insight/objective evidence of the existence/non-existence of the valve hammering phenomena over a 90-day observation

3 period. Additionally, the licensees performance of a root cause evaluation will provide greater insight into the event.

8. Involved questions or concerns pertaining to licensee operational performance N

Remarks:

4 CONDITIONAL RISK ASSESSMENT RISK ANALYSIS BY: Josh Havertape DATE: 02/27/2025 Brief Description of the Basis for the Assessment (may include assumptions, calculations, references, peer review, or comparison with licensees results):

A regional senior reactor analyst (SRA), using SAPHIRE version 8.2.12 and the Dresden SPAR model version 8.82, completed a condition assessment for the degraded conditions associated with the division I containment cooling service water (CCSW) heat exchanger outlet valve, MO-2-1501-3A. The following key assumptions and factors were considered in the quantification:

A degraded condition involving MO-2-1501-3A resulted in the overrange of the position indication potentiometer. Therefore, there was a loss of function of the valve because it wouldve been unable to open on demand for an exposure period of 43 days. The exposure period was based on a latent degraded condition that revealed itself on two separate occasions after successful operation of the valve. Using a t + repair time model both unavailability periods were combined and evaluated by setting RHR-MOV-CC-F03A, CCSW HX DISCHARGE VALVE 1501-3A FAILS TO OPEN, to TRUE.

A second degraded condition involving MO-2-1501-3A resulted in damage to the valve yoke. This was assumed to decrease the reliability of MO-2-1501-3A and resulted in an increased failure rate for the remainder of the year (322 days) by setting RHR-MOV-CC-F03A to a failure rate of 1E-2. Additionally, 0.1 was used as a sensitivity.

The degraded condition impacted the torus cooling function of the low pressure coolant injection system, which supports the containment pressure control safety function.

Therefore, the analyst reviewed defense-in-depth associated with this safety function and determined that operator actions to initiate shutdown cooling were important to mitigate the degraded condition. To represent uncertainties regarding time available to perform manual actions to make the shutdown cooling system available (e.g., install valve / pump fuses, fill / vent piping), several human error probabilities were used for failure to initiate shutdown cooling: 0.15, 1E-2, and 5E-4 (nominal).

Common cause failure was considered.

Credit was given for diverse and flexible coping strategies (FLEX).

The estimated incremental conditional probability of core damage (ICCDP) range was determined to be 2.7E-7 to 1.5E-6 and places the ICCDP in the range of baseline inspection follow-up and overlap between baseline inspection follow-up and a special inspection. This assessment did not include contributions from fire, which was judged to be important for this degraded condition.

The most significant risk contributors to the evaluation were loss of turbine building cooling water events involving loss of instrument air, main steam isolation valve closure, failure of a safety relief valve to close, failure of torus cooling, and failure to recover instrument air leading to core damage.

5 RESPONSE DECISION USING THE ABOVE INFORMATION AND OTHER KEY ELEMENTS OF CONSIDERATION AS APPROPRIATE, DOCUMENT THE RESPONSE DECISION TO THE EVENT OR CONDITION, AND THE BASIS FOR THAT DECISION DECISION AND DETAILS OF THE BASIS FOR THE DECISION:

Given the depth and breadth of the baseline inspection effort conducted to date with engineering and subject matter expert support throughout December 2024, conducting a special inspection is not actionable at this time given the absence of additional information to review by an immediate special inspection, and the ongoing root cause evaluation by the licensee. The licensees activities, including review of data recorder results monitoring for occurrences of valve hammering, will be reviewed through focused baseline inspection. Reevaluation of this decision will be considered as new information becomes available.

BRANCH CHIEF: Rob Ruiz DATE: 03/04/2025 SRA: Josh Havertape DATE: 03/04/2025 DEPUTY DIRECTOR: Billy C. Dickson, Jr.

DATE: 03/04/2025 DIVISION DIRECTOR:

DATE:

RA (if reactive inspection is initiated): N/A DATE:

ADAMS ACCESSION NUMBER: ML25063A119 ADAMS PACKAGE ACCESSION NUMBER: ML25063A140 EVENT NOTIFICATION REPORT NUMBER (as applicable): N/A Internal Distribution List is at the end of this document.

Signed by Ruiz, Robert on 03/04/25 Signed by Havertape, Joshua on 03/04/25 Signed by Dickson, Billy on 03/04/25

6 Decision Documentation for Reactive Inspection (Deterministic-only Criteria Analyzed)

PLANT:

Dresden Unit 2 EVENT DATE:

06/20/2024 EVALUATION DATE:

02/27/2025 Brief Description of the Significant Event or Degraded Condition:

See above section.

REACTOR SAFETY Y/N IIT Deterministic Criteria 1.

Led to a Site Area Emergency N

Remarks:

2.

Exceeded a safety limit of the licensee's technical specifications N

Remarks:

3.

Involved circumstances sufficiently complex, unique, or not well enough understood, or involved safeguards concerns, or involved characteristics the investigation of which would best serve the needs and interests of the Commission N

Remarks:

Y/N SI Deterministic Criteria N

4.

Significant failure to implement the emergency preparedness program during an actual event, including the failure to classify, notify, or augment onsite personnel N

Remarks:

N 5.

Involved significant deficiencies in operational performance which resulted in degrading, challenging, or disabling a safety system function or resulted in placing the plant in an unanalyzed condition for which available risk assessment methods do not provide an adequate or reasonable estimate of risk.

Remarks:

7 RADIATION SAFETY Y/N IIT Deterministic Criteria 1.

Led to a significant radiological release (levels of radiation or concentrations of radioactive material in excess of 10 times any applicable limit in the license or 10 times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, when averaged over a year) of byproduct, source, or special nuclear material to unrestricted areas N

Remarks:

2.

Led to a significant occupational exposure or significant exposure to a member of the public. In both cases, significant is defined as five times the applicable regulatory limit (except for shallow-dose equivalent to the skin or extremities from discrete radioactive particles)

N Remarks:

3.

Involved the deliberate misuse of byproduct, source, or special nuclear material from its intended or authorized use, which resulted in the exposure of a significant number of individuals N

Remarks:

4.

Involved byproduct, source, or special nuclear material, which may have resulted in a fatality N

Remarks:

5.

Involved circumstances sufficiently complex, unique, or not well enough understood, or involved safeguards concerns, or involved characteristics the investigation of which would best serve the needs and interests of the Commission N

Remarks:

Y/N AIT Deterministic Criteria 6.

Led to a radiological release of byproduct, source, or special nuclear material to unrestricted areas that resulted in occupational exposure or exposure to a member of the public in excess of the applicable regulatory limit (except for shallow-dose equivalent to the skin or extremities from discrete radioactive particles)

N Remarks:

8 7.

Involved the deliberate misuse of byproduct, source, or special nuclear material from its intended or authorized use and had the potential to cause an exposure of greater than 5 rem to an individual or 500 mrem to an embryo or fetus N

Remarks:

8.

Involved the failure of radioactive material packaging that resulted in external radiation levels exceeding 10 rads/hr or contamination of the packaging exceeding 1000 times the applicable limits specified in 10 CFR 71.87 N

Remarks:

9.

Involved the failure of the dam for mill tailings with substantial release of tailings material and solution off site N

Remarks:

Y/N SI Deterministic Criteria

10. May have led to an exposure in excess of the applicable regulatory limits, other than via the radiological release of byproduct, source, or special nuclear material to the unrestricted area; specifically occupational exposure in excess of the regulatory limits in 10 CFR 20.1201 exposure to an embryo/fetus in excess of the regulatory limits in 10 CFR 20.1208 exposure to a member of the public in excess of the regulatory limits in 10 CFR 20.1301 N

Remarks:

11. May have led to an unplanned occupational exposure in excess of 40 percent of the applicable regulatory limit (excluding shallow-dose equivalent to the skin or extremities from discrete radioactive particles)

N Remarks:

12. Led to unplanned changes in restricted area dose rates in excess of 20 rem per hour in an area where personnel were present, or which is accessible to personnel N

Remarks:

13. Led to unplanned changes in restricted area airborne radioactivity levels in excess of 500 DAC in an area where personnel were present, or which is accessible to personnel and where the airborne radioactivity level was not promptly recognized and/or appropriate actions were not taken in a timely manner N

Remarks:

9

14. Led to an uncontrolled, unplanned, or abnormal release of radioactive material to the unrestricted area for which the extent of the offsite contamination is unknown; or, that may have resulted in a dose to a member of the public from loss of radioactive material control in excess of 25 mrem (10 CFR 20.1301(e)); or, that may have resulted in an exposure to a member of the public from effluents in excess of the ALARA guidelines contained in Appendix I to 10 CFR Part 50 N

Remarks:

15. Led to a large (typically greater than 100,000 gallons), unplanned release of radioactive liquid inside the restricted area that has the potential for ground-water, or offsite, contamination N

Remarks:

16. Involved the failure of radioactive material packaging that resulted in external radiation levels exceeding 5 times the accessible area dose rate limits specified in 10 CFR Part 71, or 50 times the contamination limits specified in 49 CFR Part 173 N

Remarks:

17. Involved an emergency or non-emergency event or situation, related to the health and safety of the public or on-site personnel or protection of the environment, for which a 10 CFR 50.72 report has been submitted that is expected to cause significant, heightened public or government concern N

Remarks:

SAFEGUARDS/SECURITY Y/N IIT Deterministic Criteria 1.

Involved circumstances sufficiently complex, unique, or not well enough understood, or involved safeguards concerns, or involved characteristics the investigation of which would best serve the needs and interests of the Commission N

Remarks:

N 2.

Failure of licensee significant safety equipment or adverse impact on licensee operations as a result of a safeguards initiated event (e.g., tampering).

10 Remarks:

3.

Actual intrusion into the protected area N

Remarks:

Y/N AIT Deterministic Criteria 4.

Involved a significant infraction or repeated instances of safeguards infractions that demonstrate the ineffectiveness of facility security provisions N

Remarks:

5.

Involved repeated instances of inadequate nuclear material control and accounting provisions to protect against theft or diversions of nuclear material N

Remarks:

6.

Confirmed tampering event involving significant safety or security equipment N

Remarks:

7.

Substantial failure in the licensees intrusion detection or package/personnel search procedures which results in a significant vulnerability or compromise of plant safety or security N

Remarks:

Y/N SI Deterministic Criteria 8.

Involved inadequate nuclear material control and accounting provisions to protect against theft or diversion, as evidenced by inability to locate an item containing special nuclear material (such as an irradiated rod, rod piece, pellet, or instrument)

N Remarks:

9.

Involved a significant safeguards infraction that demonstrates the ineffectiveness of facility security provisions N

Remarks:

11

10. Confirmation of lost or stolen weapon N

Remarks:

11. Unauthorized, actual non-accidental discharge of a weapon within the protected area N

Remarks:

12. Substantial failure of the intrusion detection system (not weather related)

N Remarks:

13. Failure to the licensees package/personnel search procedures which results in contraband, or an unauthorized individual being introduced into the protected area N

Remarks:

14. Potential tampering or vandalism event involving significant safety or security equipment where questions remain regarding licensee performance/response, or a need exists to independently assess the licensees conclusion that tampering or vandalism was not a factor in the condition(s) identified N

Remarks:

12 RESPONSE DECISION USING THE ABOVE INFORMATION AND OTHER KEY ELEMENTS OF CONSIDERATION AS APPROPRIATE, DOCUMENT THE RESPONSE DECISION TO THE EVENT OR CONDITION, AND THE BASIS FOR THAT DECISION.

DECISION AND DETAILS OF THE BASIS FOR THE DECISION:

BRANCH CHIEF: Rob Ruiz DATE: 03/04/2025 SRA: Josh Havertape DATE: 03/04/2025 DEPUTY DIRECTOR: Billy C. Dickson, Jr.

DATE: 03/04/2025 DIVISION DIRECTOR:

DATE:

ADAMS ACCESSION NUMBER: ML25063A119 ADAMS PACKAGE ACCESSION NUMBER:ML25063A140 EVENT NOTIFICATION REPORT NUMBER (as applicable): N/A Distribution: Alejandro.Alen@nrc.gov; Scott.Morris@nrc.gov; Jason.Carneal@nrc.gov; John.Giessner@nrc.gov; Mohammed.Shuaibi@nrc.gov; Blake.Welling@nrc.gov; Ray.McKinley@nrc.gov; Mark.Franke@nrc.gov; Gregory.Suber@nrc.gov; Laura.Pearson@nrc.gov; LaDonna.Suggs@nrc.gov; Ravi.Penmetsa@nrc.gov; Jason.Kozal@nrc.gov; Billy.Dickson@nrc.gov; David.Curtis@nrc.gov; Jared.Heck@nrc.gov;Geoffrey.Miller@nrc.gov; Nick.Taylor@nrc.gov; Karla.Stoedter@nrc.gov; Doris.Chyu@nrc.gov; Joshua.Havertape@nrc.gov; Lionel.Rodriguez@nrc.gov; Matthew.Leech@nrc.gov; NRR_Reactive_Inspection.Resource@nrc.gov; Robert.Ruiz@nrc.gov Signed by Ruiz, Robert on 03/04/25 Signed by Havertape, Joshua on 03/04/25 Signed by Dickson, Billy on 03/04/25