ML24344A106

MD 8.3 Evaluation for Donald C Cook, Turbine Control Valve Leads to Increase in Reactor Thermal Power

ML24344A106
Person / Time
Site:	Cook
Issue date:	11/27/2024
From:	NRC/RGN-III/DORS/ERPB
To:
References
MD 8.3
Download: ML24344A106 (1)

Text

MD 8.3 Evaluation Decision Documentation for Reactive Inspection (Deterministic and Risk Criteria Analyzed)

PLANT:

D.C. Cook EVENT DATE:

11/21/2024 DETERMINISTIC CRITERIA EVALUATION DATE:

11/27/2024 Brief Description of the Significant Operational Event or Degraded Condition:

At 1022 on November 21,2024, licensee staff performed testing on Unit 1 control valve CV-2.

During testing CV-2 opened greater than the expected demand of the valve resulting in a transient to 102.8 percent reactor thermal power. Operators took prompt (less than 2 minutes) action to reduce reactor thermal power and power was restored below 100 percent within approximately 8 minutes.

Specifically, the licensee was investigating servo conditions on each of the control valves after previously identifying that CV-4 had automatically swapped servo coils from SPC-A to SPC-B after a load change. It was later discovered that one of the two servos on CV-4 had faulted, presumably due to age. From the concern of failing servos, the licensee wanted to confirm that the other control valves did not have the same issue and began testing the servos by manually swapping between SPC-A and SPC-B. After switching the servo, if the servo control automatically switched back, that would be a sign of degradation in the servo similar to what was identified in CV-4.

While testing on CV-2, immediately after swapping from SPC-A to SPC-B, the Linear Variable Differential Transformer (LVDT) for SPC-B indicated 16 percent valve position when its actual position was 41.1 percent. This disparity was a result of a manufacturers design error. This resulted in a demand signal to open the valve and CV-2 went to 97 percent open from this demand signal. Due to the large difference between SPC-A and SPC-Bs position indication, the operators could not switch the servos back. This is by design.

Operators initially observed withdrawing control rods and promptly placed them in manual operation. Operators identified the increased steam flow and lowered turbine load to reduce thermal power below 100 percent. Following the immediate response, operators borated to bring control rods back to a favorable bank height for axial flux.

At 2030 on November 21, 2024, operators successfully implemented a special procedure to manually increase demand on SPC-Bs LVDT to bring the control valves back to equal positions.

This allowed the station to swap between SPCs. CV-2 was swapped back to the functional SPC-A, which allowed the operators to place the valve back into automatic control. Power ascension back to full power was then initiated and Unit 1 returned to full power at 0425 on November 22, 2024.

The site performed walk downs of their secondary systems and entered into a Failure Investigation Process evaluation.

2 Y/N DETERMINISTIC CRITERIA

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

Remarks: Although the licensee exceeded their Rated Thermal Power (max indicated power was 102.8 percent), they did not exceed any of the design bases of the facility.

According to the UFSAR, the design limit is 118 percent power with a protective feature set at 110 percent power.

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

Remarks: This event did not involve a major deficiency in design, construction, or operation having potential generic safety implications. While the LVDT provided incorrect valve indication resulting in unexpected valve opening, corresponding system interactions operated as expected. Operators took appropriate manual reactor control to control reactor power prior to any automatic reactor protection system actions.

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: There was no loss of any fission product barriers. No reactor coolant system fission product activity changes were identified.

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

Remarks: There were no loss of safety function or multiple failures in systems used to mitigate an actual event.

5. Involved possible adverse generic implications Y

Remarks: Current information indicates that Cook possesses only two LVDTs of this type onsite, which contains the inappropriate internal design. Both LVDTs have been quarantined. These LVDTs are not utilized in safety-related application at Cook. This LVDT was procured from another nuclear station that could possibly have these installed where they could have similar issues.

6. Involved significant unexpected system interactions N

Remarks: No unexpected system interaction occurred. Following control valve indication failure, all corresponding systems performed as expected.

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

Remarks: This event did not involve repetitive failures or events involving safety-related equipment or deficiencies in operations.

N

8. Involved questions or concerns pertaining to licensee operational performance

3 Remarks: There are no questions pertaining to licensee operation performance. Prompt action was taken within 2 minutes to reduce thermal power.

CONDITIONAL RISK ASSESSMENT RISK ANALYSIS BY: Steven Alferink DATE: 11/27/2024 Brief Description of the Basis for the Assessment (may include assumptions, calculations, references, peer review, or comparison with licensees results):

The senior reactor analyst considered the plants SPAR model and determined that the adequacy of the risk assessment models, assumptions, and uncertainties made it difficult to numerically quantify risk. Therefore, the senior reactor analyst performed a qualitative risk assessment. The senior reactor analyst determined that this event was not risk significant because operators took immediate action to reduce power. This resulted in the plant exceeding the thermal power limit for a short duration and limited the thermal power increase to 102.8 percent reactor thermal power, which was below the high-power trip setpoint.

The estimated conditional core damage probability (CCDP) is _______N/A____________ and places the risk in the range of a _____N/A________ and _______N/A___________ inspection.

4 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:

Based on the current information available regarding equipment and operation performance, a single deterministic criterion was met, however no detailed risk-significance assessment was necessary. Therefore, a reactive inspection is not warranted. This event will be followed up by the resident inspectors under baseline inspection. We will re-evaluate this decision If additional information is available that changes the answer to any deterministic criterion or risk assessment inputs.

BRANCH CHIEF: Matthew Learn DATE: 12/09/2024 SRA: Steven Alferink DATE: 12/09/2024 DIVISION DIRECTOR: Jason Kozal DATE: 12/09/2024 DIVISION DIRECTOR:

DATE:

RA (if reactive inspection is initiated)

DATE:

ADAMS ACCESSION NUMBER: ML24344A106 ADAMS PACKAGE ACCESSION NUMBER: ML24346A099 EVENT NOTIFICATION REPORT NUMBER (as applicable): N/A Internal Distribution List is at the end of this document Signed by Learn, Matthew on 12/09/24 Signed by Alferink, Steven on 12/09/24 Dickson, Billy signing on behalf of Kozal, Jason on 12/09/24

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

PLANT:

D.C. Cook EVENT DATE:

11/21/2024 DETERMINISTIC CRITERIA EVALUATION DATE:

11/27/2024 Brief Description of the Significant Operational Event or Degraded Condition:

At 1022 on November 21,2024, licensee staff performed testing on Unit 1 control valve CV-2.

During testing CV-2 opened greater than the expected demand of the valve resulting in a transient to 102.8 percent reactor thermal power. Operators took prompt (less than 2 minutes) action to reduce reactor thermal power and power was restored below 100 percent within approximately 8 minutes.

Specifically, the licensee was investigating servo conditions on each of the control valves after previously identifying that CV-4 had automatically swapped servo coils from SPC-A to SPC-B after a load change. It was later discovered that one of the two servos on CV-4 had faulted, presumably due to age. From the concern of failing servos, the licensee wanted to confirm that the other control valves did not have the same issue and began testing the servos by manually swapping between SPC-A and SPC-B. After switching the servo, if the servo control automatically switched back, that would be a sign of degradation in the servo similar to what was observed identified in CV-4.

While testing on CV-2, immediately after swapping from SPC-A to SPC-B, the Linear Variable Differential Transformer (LVDT) for SPC-B indicated 16 percent valve position when its actual position was 41.1 percent. This disparity was a result of a manufacturers design error. This resulted in a demand signal to open the valve and CV-2 went to 97 percent open from this demand signal. Due to the large difference between SPC-A and SPC-Bs position indication, the operators could not switch the servos back. This is by design.

Operators initially observed withdrawing control rods and promptly placed them in manual operation. Operators identified the increased steam flow and lowered turbine load to reduce thermal power below 100 percent. Following the immediate response, operators borated to bring control rods back to a favorable bank height for axial flux.

At 2030 on November 21, 2024, operators successfully implemented a special procedure to manually increase demand on SPC-Bs LVDT to bring the control valves back to equal positions. This allowed the station to swap between SPCs. CV-2 was swapped back to the functional SPC-A, which allowed the operators to place the valve back into automatic control.

Power ascension back to full power was then initiated and Unit 1 returned to full power at 0425 on November 22, 2024.

The site performed walk downs of their secondary systems and entered into a Failure Investigation Process evaluation.

6 REACTOR SAFETY Y/N IIT Deterministic Criteria 1.

Led to a Site Area Emergency N

Remarks: This event did not lead to a Site Area Emergency.

2.

Exceeded a safety limit of the licensees technical specifications N

Remarks: This event did not result in the licensees Technical Specification Safety Limit being exceeded.

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: The event was from a single valve response to a control input, and corresponding system interactors are well understood.

Y/N SI Deterministic Criteria 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: This event did not involve any failure to implement the emergency preparedness program.

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.

N Remarks: This event did not involve any deficiencies in operational performance.

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: This event did not involve in radiological release to unrestricted areas.

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: This event did not involve any occupational exposure or exposure to a member of the public.

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: This event did not involve any deliberate misuse of byproduct, source, or special nuclear material that resulted in the exposure of individuals.

4.

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

Remarks: This event did not involve any byproduct, source, or special nuclear material, that resulted in a fatality.

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: The event was from a single valve response to a control input. It did not involve any complex, unique or not well understood phenomena.

Y/N AIT Deterministic Criteria N

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)

8 Remarks: This event did not involve a radiological release of byproduct, source, or special nuclear material to unrestricted areas.

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: This event did not involve a deliberate misuse of byproduct, source, or special nuclear material.

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: This event did not involve a failure of radioactive material packaging.

9.

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

Remarks: This event did not involve a failure of the dam for mill tailings.

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: This event did not involve any exposure via the radiological release of byproduct, source, or special nuclear material to the unrestricted area.

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: This event did not involve any unplanned occupational exposure.

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: This event did not involve any unplanned changes in dose rates.

N

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

9 accessible to personnel and where the airborne radioactivity level was not promptly recognized and/or appropriate actions were not taken in a timely manner Remarks: This event did not involve an unplanned change in restricted area airborne radioactivity levels in an area where personnel were present or was accessible to personnel.

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: This event did not involve an uncontrolled, unplanned, or abnormal release of radioactive material to the unrestricted area.

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: This event did not involve a unplanned release of radioactive liquid inside the restricted area that had the potential for ground-water, or offsite, contamination.

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: This event did not involve a failure of radioactive material packaging.

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: This event did not involve an event that was expected to cause significant, heightened public or government concern.

10 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: This event did not involve any complex, unique, or not well enough understood, or safeguards concerns.

2.

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

N Remarks: This event did not involve a safeguards-initiated event.

3.

Actual intrusion into the protected area N

Remarks: This event did not involve an intrusion into the protected area.

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: This event did not involve any instances of safeguards infractions.

5.

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

Remarks: This event did not involve any nuclear material control and accounting provisions.

6.

Confirmed tampering event involving significant safety or security equipment N

Remarks: This event did not involve a tampering event.

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: This event did not involve a failure in the licensees intrusion detection or package/personnel search procedures.

11 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: This event did not involve any inadequate nuclear material control and accounting provisions.

9.

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

Remarks: This event did not involve a safeguards infraction.

10. Confirmation of lost or stolen weapon N

Remarks: This event did not involve any lost or stolen weapon.

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

Remarks: This event did not involve a discharge of a weapon.

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

N Remarks: This event did not involve a failure of the intrusion detection system.

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: This event did not involve a failure to the licensees package/personnel search procedures.

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: This event did not involve tampering or vandalism.

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:

Based on the current information available regarding equipment and operation performance, a single deterministic criterion was met, however no detailed risk significance assessment was necessary. Therefore, a reactive inspection is not warranted. This event will be followed up by the resident inspectors under baseline inspection. We will re-evaluate this decision If additional information is available that changes the answer to any deterministic criterion or risk assessment inputs.

BRANCH CHIEF: Matthew Learn DATE: 12/09/2024 SRA: Steven Alferink DATE: 12/09/2024 DIVISION DIRECTOR: Jason Kozal DATE: 12/09/2024 DIVISION DIRECTOR:

DATE:

ADAMS ACCESSION NUMBER: ML24344A106 ADAMS PACKAGE ACCESSION NUMBER: ML24346A099 EVENT NOTIFICATION REPORT NUMBER (as applicable): N/A Distribution: Robert.Ruiz@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; Anthony.Masters@nrc.gov; Jason.Kozal@nrc.gov; Billy.Dickson@nrc.gov; David.Curtis@nrc.gov; Jonathan.Feibus@nrc.gov; Jared.Heck@nrc.gov; Geoffrey.Miller@nrc.gov; Michael.Hay@nrc.gov; Michelle.Garza@nrc.gov; Doris.Chyu@nrc.gov; Joshua.Havertape@nrc.gov; Lionel.Rodriguez@nrc.gov; Steven.Alferink@nrc.gov; Matthew.Learn@nrc.gov; NRR_Reactive_Inspection.Resource@nrc.gov Signed by Learn, Matthew on 12/09/24 Signed by Alferink, Steven on 12/09/24 Dickson, Billy signing on behalf of Kozal, Jason on 12/09/24