ML24263A142

MD 8.3 Evaluation for Prairie Island CVCS Letdown Isolation Valve

ML24263A142
Person / Time
Site:	Prairie Island
Issue date:	09/05/2024
From:	Billy Dickson NRC/RGN-III/DORS/RPB3
To:
Shared Package
ML24276A229	List: IR 05000282/2024004 ML24263A142
References
Download: ML24263A142 (11)

Text

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

PLANT:

Prairie Island EVENT DATE:

08/06/2024 DETERMINISTIC CRITERIA EVALUATION DATE:

09/05/2024 Brief Description of the Significant Operational Event or Degraded Condition:

On August 6, 2024, while making preps to restore the 23 inverter to service following corrective maintenance, Operations isolated the Chemical Volume Control System CVCS Letdown. Upon isolation, letdown isolation valve CV-31279 failed to indicate CLOSED (remained intermediate). Letdown Isolation valve CV-31230 and the CVCS orifice isolation valves, which are interlocked with CV-31230 and CV-31279, all closed as expected to secure letdown. Operations declared CV-31279 to be IST Inoperable.

Approximately 11 minutes after isolating letdown, Operations opened CV-31279, CV-31230, and the orifice isolation valves, returning the valve and the system to service. When returned to its fully OPEN position CV-31279 did indicate full OPEN as expected. The licensee has indicated that they still consider CV-31279 to be inoperable from an Inservice Testing IST perspective, and they are not relying upon it to perform its specified safety function.

Both CV-31279 and CV-31230 are safety-related valves in a Code Class 1 portion of the CVCS system immediately adjacent to the Reactor Coolant System RCS loop, and have the safety-related function to CLOSE to maintain the RCS pressure boundary in the event of a leak in the downstream Code Class 2 piping. The Prairie Island Update Safety Analysis Report USAR specifies the following: Redundant valves in series and redundant level channels powered by separate and independent buses are provided to ensure letdown isolation even in the event of a single active failure. In addition, the licensees Inservice Testing IST program indicates that, The RCS boundary extends to the second isolation valve. and This valve(s) has an Active safety function to close on loss of power, loss or air or low pressurizer level to maintain the RCS pressure boundary. It should be noted that this valve has no technical specifications associated with it.

Currently, the licensee has indicated that, With CV-31279 declared inoperable and left in its open position, the site currently does not meet the requirements of 50.55a(c)(2)(ii) nor can the letdown isolation functions be performed exactly as described in the USAR. This nonconformance with 10 CFR 50.55a and the USAR is recognized as a condition adverse to quality and is captured within the sites corrective action program.

Considering the single active failure criteria in the USAR, with a leak downstream of CV-31279, and assuming a single active failure of CV-31230, neither valve would perform their safety function and CLOSE to maintain the RCS pressure boundary, resulting in a loss of one of the fission product barriers.

The licensee believes that since this is entered into their Corrective Action Program (CAP),

they are ok to continue operating until their next cold shutdown, which is currently scheduled for September 26, 2025.

2 Y/N DETERMINISTIC CRITERIA Y

a. Involved operations that exceeded, or were not included in, the design bases of the facility Remarks: Valves CV-31279 and CV-31230 are installed in an in-series configuration, as required by the design basis, to ensure the reactor coolant system pressure boundary will isolate assuming an active single failure. The licensees decision to restore CV-31279 to an open position prior to repairing, replacing, or retesting it to ensure its functionality creates a condition where the licensee can no longer ensure the design basis of the letdown isolation valves (i.e., isolate the reactor coolant system pressure boundary assuming a concurrent active single failure) is being met N

b. Involved a major deficiency in design, construction, or operation having potential generic safety implications Remarks:

N

c. Led to a significant loss of integrity of the fuel, primary coolant pressure boundary, or primary containment boundary of a nuclear reactor Remarks:

N

d. Led to the loss of a safety function or multiple failures in systems used to mitigate an actual event Remarks:

N

e. Involved possible adverse generic implications Remarks:

N

f. Involved significant unexpected system interactions Remarks:

N

g. Involved repetitive failures or events involving safety-related equipment or deficiencies in operations Remarks:

Y

h. Involved questions or concerns pertaining to licensee operational performance Remarks: In 2021, Region III issued a green, non-cited violation to Prairie Island due to the licensees failure to meet the requirements of 10 CFR 50.55. Specifically, the licensee experienced a failure of a safety-related check valve while performing inservice testing activities required by the American Society of Mechanical Engineers (ASME) Code and 10 CFR 50.55a. The ASME Code states that if a check valve becomes inoperable, a licensee must repair, replace or retest the valve prior to

3 returning it to service. 10 CFR 50.55 also contains provisions for licensees to request relief from the ASME Code and 10 CFR 50.55a requirements via the NRCs relief request process. However, the licensee failed to repair, replace or retest the check valve and did not request the NRC to grant relief from the ASME Code and 10 CFR 50.55a requirements.

The NRC has concerns pertaining to the licensees operational performance because the position taken by the licensee is similar to the position taken in 2021. Specifically, the licensee did not repair, replace, or retest CV-31279 prior to placing letdown in service and placing CV-32179 in an open position.

CONDITIONAL RISK ASSESSMENT RISK ANALYSIS BY: Dariusz Szwarc DATE: 09/05/2024 Brief Description of the Basis for the Assessment (may include assumptions, calculations, references, peer review, or comparison with licensees results):

The Senior Reactor Analyst reviewed the guidance in IMC 0609 Appendix H, Containment Integrity Significance Determination Process to assist in determining the risk significance.

The SRA reviewed the Prairie Island SPAR model version 8.82 using SAPHIRE version 8.2.10 and determined that letdown isolation valve CV-31279 was not in the SPAR model. The SRA obtained risk information for the valve from the licensee whose model does contain the valve. The licensee stated that the change in core damage frequency (CDF) value for the valve was 7.08E-8 per year. Using the guidance in the Risk Assessment of Operation Events (RASP) Manual Volume 1 the SRA divided the delta CDF value by one inverse year to determine the incremental conditional core damage probability (ICCDP) associated with this failure.

ICCDP = delta CDF/(yr-1) 7.08E-8 = (7.08E-8/yr)/ (yr-1)

An exposure time of one year was assumed.

The letdown line is located inside containment and therefore containment performance and bypass are not a concern. Therefore, the conditional large early release probability (CLERP) is not considered.

The estimated conditional core damage probability (CCDP) is 7.08E-8 and places the risk in the No Additional Inspection range.

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 risk assessment results, the Branch will not recommend any reactive inspections and will pursue the issue through the baseline inspection process.

BRANCH CHIEF: Richard A. Skokowski ((signature: RAS DATE: 09/19/2024 SRA: Dariusz Szwarc ((signature: DXS4}} DATE: 09/20/2024 DIVISION DIRECTOR: Billy C. Dickson ((signature: BCD}} DATE: 10/01/2024 DIVISION DIRECTOR: DATE ADAMS ACCESSION NUMBER: ML24263A142 ADAMS PACKAGE ACCESSION NUMBER: ML24276A229 EVENT NOTIFICATION REPORT NUMBER (as applicable): Internal Distribution List is at the end of this document.

5 Decision Documentation for Reactive Inspection (Deterministic-only Criteria Analyzed) PLANT: Prairie Island EVENT DATE: 08/06/2024 EVALUATION DATE: 09/05/2024 Brief Description of the Significant Operational Event or Degraded Condition: On August 6, 2024, while making preps to restore the 23 inverter to service following corrective maintenance, Operations isolated the Chemical Volume Control System CVCS Letdown. Upon isolation, letdown isolation valve CV-31279 failed to indicate CLOSED (remained intermediate). Letdown Isolation valve CV-31230 and the CVCS orifice isolation valves, which are interlocked with CV-31230 and CV-31279, all closed as expected to secure letdown. Operations declared CV-31279 to be IST Inoperable. Approximately 11 minutes after isolating letdown, Operations opened CV-31279, CV-31230, and the orifice isolation valves, returning the valve and the system to service. When returned to its fully OPEN position CV-31279 did indicate full OPEN as expected. The licensee has indicated that they still consider CV-31279 to be inoperable from an Inservice Testing IST perspective, and they are not relying upon it to perform its specified safety function. Both CV-31279 and CV-31230 are safety-related valves in a Code Class 1 portion of the CVCS system immediately adjacent to the Reactor Coolant System RCS loop, and have the safety-related function to CLOSE to maintain the RCS pressure boundary in the event of a leak in the downstream Code Class 2 piping. The Prairie Island Update Safety Analysis Report USAR specifies the following: Redundant valves in series and redundant level channels powered by separate and independent buses are provided to ensure letdown isolation even in the event of a single active failure. In addition, the licensees Inservice Testing (IST) program indicates that, The RCS boundary extends to the second isolation valve. and This valve(s) has an Active safety function to close on loss of power, loss or air or low pressurizer level to maintain the RCS pressure boundary. It should be noted that this valve has no technical specifications associated with it. Currently, the licensee has indicated that, With CV-31279 declared inoperable and left in its open position, the site currently does not meet the requirements of 50.55a(c)(2)(ii) nor can the letdown isolation functions be performed exactly as described in the USAR. This nonconformance with 10 CFR 50.55a and the USAR is recognized as a condition adverse to quality and is captured within the sites corrective action program. Considering the single active failure criteria in the USAR, with a leak downstream of CV-31279, and assuming a single active failure of CV-31230, neither valve would perform their safety function and CLOSE to maintain the RCS pressure boundary, resulting in a loss of one of the fission product barriers. The licensee believes that since this is entered into their Corrective Action Program (CAP), they are ok to continue operating until their next cold shutdown, which is currently scheduled for September 26, 2025.

6 REACTOR SAFETY Y/N IIT Deterministic Criteria N Led to a Site Area Emergency Remarks: N Exceeded a safety limit of the licensees technical specifications Remarks: N 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 Remarks: Y/N SI Deterministic Criteria N Significant failure to implement the emergency preparedness program during an actual event, including the failure to classify, notify, or augment onsite personnel Remarks: N 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: RADIATION SAFETY Y/N IIT Deterministic Criteria N 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 Remarks: N 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)

7 Remarks: N 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 Remarks: N Involved byproduct, source, or special nuclear material, which may have resulted in a fatality Remarks: N 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 Remarks: Y/N AIT Deterministic Criteria N 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) Remarks: N 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 Remarks: N 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 Remarks: N Involved the failure of the dam for mill tailings with substantial release of tailings material and solution off site Remarks:

8 Y/N SI Deterministic Criteria N 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 Remarks: N 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) Remarks: N 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 Remarks: N 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 Remarks: N 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 as low as reasonably achievable ALARA guidelines contained in Appendix I to 10 CFR Part 50 Remarks: N 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 Remarks:

9 n 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 Remarks: 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 Remarks: SAFEGUARDS/SECURITY Y/N IIT Deterministic Criteria N 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 Remarks: N Failure of licensee significant safety equipment or adverse impact on licensee operations as a result of a safeguards-initiated event (e.g., tampering). Remarks: N Actual intrusion into the protected area. Remarks: Y/N AIT Deterministic Criteria N Involved a significant infraction or repeated instances of safeguards infractions that demonstrate the ineffectiveness of facility security provisions Remarks: N Involved repeated instances of inadequate nuclear material control and accounting provisions to protect against theft or diversions of nuclear material Remarks: N Confirmed tampering event involving significant safety or security equipment Remarks: N 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 Remarks:

10 Y/N SI Deterministic Criteria N 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) Remarks: N Involved a significant safeguards infraction that demonstrates the ineffectiveness of facility security provisions Remarks: N Confirmation of lost or stolen weapon Remarks: N Unauthorized, actual non-accidental discharge of a weapon within the protected area Remarks: N Substantial failure of the intrusion detection system (not weather related) Remarks: N Failure to the licensees package/personnel search procedures which results in contraband or an unauthorized individual being introduced into the protected area Remarks: N Potential tampering of 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 Remarks:

11 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 risk assessment results, the Branch will not recommend any reactive inspections and will pursue the issue through the baseline inspection process. BRANCH CHIEF: Richard A. Skokowski ((signature: RAS}} DATE: 09/19/2024 SRA: Dariusz Szwarc ((signature: DXS4}} DATE: 09/20/2024 DIVISION DIRECTOR: Billy C. Dickson ((signature: BCD}} DATE: 10/01/2024 DIVISION DIRECTOR: DATE: ADAMS ACCESSION NUMBER: ML24263A142 ADAMS PACKAGE ACCESSION NUMBER: ML24276A229 EVENT NOTIFICATION REPORT NUMBER (as applicable): 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; Geoffrey.Miller@nrc.gov; Michael.Hay@nrc.gov; Michelle.Garza@nrc.gov; Doris.Chyu@nrc.gov; Joshua.Havertape@nrc.gov; Dariusz.Szwarc@nrc.gov; NRR_Reactive_Inspection.Resource@nrc.gov; Richard.Skokowski@nrc.gov; Roy.Elliott@nrc.gov}}