ML24197A093
| ML24197A093 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 07/17/2024 |
| From: | Bradley Davis Region 2 Administrator |
| To: | |
| References | |
| EN 57215 MD 8.3 24-004 | |
| Download: ML24197A093 (1) | |
Text
Enclosure Table 1: Decision Documentation for Reactive Inspection (Deterministic and Risk Criteria Analyzed)
MD 8.3 24-004 PLANT: Vogtle Unit 3 EVENT DATE: 7/8/24 EVALUATION DATE: 7/9/24 Brief Description of the Significant Operational Event or Degraded Condition:
On June 8, 2024, Vogtle Unit 3 experienced a manual reactor trip with an automatic safeguards actuation, which activated portions of the safety-related passive core cooling system.
At 2122, on July 8, with unit 3 in Mode 1 at 100% power, the main feed pump (MFP) A minflow valve failed open diverting main feed flow from the steam generators (SGs) and causing SG water levels to lower. Operators took immediate actions to establish a critical parameter for SG narrow range at ~ 26 percent and entered the appropriate abnormal operating procedure (AOP), but before working through the AOP, the critical parameter was reached, and the operators manually tripped the reactor and placed the plant in a safe and stable condition using EOPs.
Although start-up feedwater (SFW) auto started, the SG wide range level dropped below 35 percent due to shrink from the reactor trip, which resulted in a passive residual heat removal (PRHR) actuation. The PRHR heat exchanger cooled the reactor coolant system (RCS) cold legs (Tcold) to 505 degrees Fahrenheit (°F) which automatically actuated a safeguards signal (containment isolation and core make up tank injection). The unit was cooled to Mode 4 using PRHR with the in-containment refueling water stork tank (IRWST) being cooled via the normal residual heat removal system (RNS). The licensee subsequently shutdown the plant and entered Mode 5 with normal RNS cooling configuration. The licensee submitted a 4-hour immediate notification to the NRC in accordance with 10 CFR 50.72 (EN 57215).
Timeline of Major Event Evolutions 7/8/2024 @ 2122 - Unit 3 MFP minflow valve failed open 7/8/2024 @ 2125 - Manual reactor trip 7/8/2024 @ 2126 - Passive residual heat removal actuation 7/8/2024 @ 2128 - Containment isolation and core make up tank injection actuation Y/N DETERMINISTIC CRITERIA N
- a. Involved operations that exceeded, or were not included in, the design bases of the facility
Enclosure Remarks: The event was within the design bases of the facility. The failure of the A MFP minflow valve, and resulting loss of feedwater flow, is bounded and less severe than a loss of feedwater event (LOFW), which is an anticipated operational occurrence (AOO) (Condition II event) analyzed in Chapter 15 of Vogtle Unit 3 UFSAR. As described in UFSAR 15.2.7, during a LOFW event the PRHR is expected to be actuated and provide the safety-related function of removing core decay heat. Furthermore, per UFSAR 15.2.7.1, the RCS cooldown from a PRHR actuation is expected to result in an additional safeguards signal, due to a Low-2 RCS TCOLD signal (505°F), to actuate the core makeup tanks which is consistent with what transpired in this event. Also, while not a safety-related function, the startup feedwater (SWF) pump auto-started as designed. It is noted that UFSAR 15.2.7.1 implies that PRHR would actuate only if SFW is not available, which was not the case in this event. When operators inserted a manual reactor trip at ~26 percent NR SG level, the WR SG level dropped below 35 percent (due to SG level shrink) which actuated the PRHR system as designed. While this represents a discrepancy between the actual plant response for this event and the expected response, as described in the UFASR, it did not involve operations that exceeded the safety design bases for the facility. The heat removal function of the SFW system is not safety-related. The licensees investigation will evaluate the response time of the SFW pump relative to the timing of the reactor trip and SG levels to determine acceptability of the SFW system performance and any corrective actions, if appropriate. This can be followed up via routine baseline inspection efforts.
Excessive Cooldown The actuation of PRHR and core makeup tanks resulted in exceeding the cooldown rate of 100°F/hour in the Pressure/Temperature (P/L) limit curve of the pressure and temperature limits report (PTLR). This large cooldown rate is expected whenever PRHR is actuated. The licensee recognized surpassing of the cooldown limit during the event and entered the applicable Technical Specification (TS) Limiting Condition of Operation (LCO) 3.4.3, RCS Pressure and Temperature (P/T) Limits. Although the TS cooldown rate was exceeded, the design bases of the facility was not exceeded. P/T limit curves define an acceptable region for normal operation and incorporate adequate margin to brittle failure for AOOs. In accordance with TS 3.4.3 Condition A, the licensee will perform an evaluation to determine if RCS operation can continue. The evaluation must verify the reactor coolant pressure boundary integrity remains acceptable and must be completed before continuing operation. This can be followed up via routine baseline inspection efforts.
- b. Involved a major deficiency in design, construction, or operation having potential generic safety implications N
Remarks: Did not involve a major deficiency having potential generic safety implications.
- c. Led to a significant loss of integrity of the fuel, primary coolant pressure boundary, or primary containment boundary of a nuclear reactor N
Enclosure Remarks: Did not lead to any loss of integrity of the fuel, primary coolant pressure boundary, or primary containment boundary.
- d. Led to the loss of a safety function or multiple failures in systems used to mitigate an actual event N
Remarks: There was no loss of safety function nor failures in systems used to mitigate an actual event. The PRHR actuated as designed on a valid Low-2 wide range (WR) SG level (35 percent). As noted in UFSAR 15.2.7.1, Loss of Normal Feedwater Flow, actuation of the PRHR system can lead to a safeguards signal on a Low-2 TCOLD signal (505°F), consistent with what transpired in this event.
- e. Involved possible adverse generic implications N
Remarks: The event does not involve possible adverse generic implications.
- f. Involved significant unexpected system interactions N
Remarks: Did not involve unexpected system interactions.
- g. Involved repetitive failures or events involving safety-related equipment or deficiencies in operations N
Remarks: Did not involve repetitive failures or events involving safety-related equipment or deficiencies in operations.
- h. Involved questions or concerns pertaining to licensee operational performance N
Remarks: Did not involve questions or concerns pertaining to licensee operational performance.
2 CONDITIONAL RISK ASSESSMENT RISK ANALYSIS BY: Andy Rosebrook DATE: 7/10/2024 Brief Description of the Basis for the Assessment (may include assumptions, calculations, references, peer review, or comparison with licensee's results):
Per IMC 0309, Reactive Inspection Decision Basis for Power Reactors, dated 12/14/2023, a risk evaluation for this event is not required because none of the deterministic criteria in Table 1 were met. However, for informational purposes a risk evaluation was conducted by conservatively assuming a complete failure of the SFW pump, which represented a risk in the range of 1E-8.
The estimated conditional core damage probability (CCDP) is ___________________ and places the risk in the range of a _______________ and ____________________ inspection.
3 Table 2: Decision Documentation for Reactive Inspection (Deterministic-only Criteria Analyzed)
PLANT: Vogtle Unit 3 EVENT DATE: 7/8/2024 EVALUATION DATE: 7/9/2024 Brief Description of the Significant Operational Event or Degraded Condition:
See Table 1 REACTOR SAFETY Y/N IIT Deterministic Criteria Led to a Site Area Emergency N
Remarks: No EAL criteria were exceeded.
Exceeded a safety limit of the licensee's technical specifications N
Remarks: No safety limit was exceeded 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: Did not involve complex or unique circumstances.
Y/N SI Deterministic Criteria Significant failure to implement the emergency preparedness program during an actual event, including the failure to classify, notify, or augment onsite personnel N
Remarks: No EAL thresholds were exceeded that would have required execution of the emergency preparedness program.
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: Did not involve significant deficiencies in operational performance.
4 RADIATION SAFETY Y/N IIT Deterministic Criteria 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: There were no radiological releases.
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: There were no occupational or significant exposures.
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: Did not involve the deliberate misuse of nuclear material.
Involved byproduct, source, or special nuclear material, which may have resulted in a fatality N
Remarks: Did not result in a fatality.
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: Did not involve complex or unique circumstances.
Y/N AIT Deterministic Criteria 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: Did not lead to a radiological release of byproduct, source, or special nuclear material to unrestricted areas.
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: Did not involve the deliberate misuse of nuclear material.
5 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: Did not involve radioactive material packaging.
Involved the failure of the dam for mill tailings with substantial release of tailings material and solution off site N
Remarks: Did not involve the failure of a mill tailing dam.
Y/N SI Deterministic Criteria 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: Did not lead to an exposure in excess of the applicable regulatory limits.
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: Did not lead to an unplanned occupational exposure.
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: Did not lead to unplanned changes in restricted area dose rates.
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: Did not lead to unplanned changes in restricted area airborne radioactivity levels.
6 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: Did not lead to a release of radioactive material.
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: Did not result in the unplanned release of radioactive liquid.
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: Did not result in the failure of radioactive material packing.
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: Did not involve an emergency or non-emergency event or situation.
7 SAFEGUARDS/SECURITY Y/N IIT Deterministic Criteria 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 is not considered a complex or unique issue.
Failure of license significant related equipment or adverse impact on licensee operations as a result of a safeguards initiated event (e.g., tampering).
N Remarks: The event did not involve tampering.
Actual intrusion into the protected area N
Remarks: Did not involve an intrusion into the protected area.
Y/N AIT Deterministic Criteria Involved a significant infraction or repeated instances of safeguards infractions that demonstrate the ineffectiveness of facility security provisions N
Remarks: Did not involve safeguards.
Involved repeated instances of inadequate nuclear material control and accounting provisions to protect against theft or diversions of nuclear material N
Remarks: Did not involve inadequate nuclear material control.
Confirmed tampering event involving significant safety-or security-equipment N
Remarks: The event did not involve tampering.
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: Was not a failure of the intrusion detection or package/personnel search procedures Y/N SI Deterministic Criteria 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: Did not involve nuclear material control and accounting.
8 Involved a significant safeguards infraction that demonstrates the ineffectiveness of facility security provisions N
Remarks: Did not involve safeguards.
Confirmation of lost or stolen weapon N
Remarks: Did not involve the loss of a weapon.
Unauthorized, actual non-accidental discharge of a weapon within the protected area N
Remarks: Did not involve the intrusion detection system.
Substantial failure of the intrusion detection system (not weather related)
N Remarks: Did not involve the intrusion detection system.
Failure to the licensees package/personnel search procedures which results in contraband or an unauthorized individual being introduced into the protected area N
Remarks: Did not involve the package/personnel search procedures.
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 conditions that tampering or vandalism was not a factor in the condition(s) identified.
N Remarks: Did not involve tampering.
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 review of the event and the analysis none of the deterministic or risk criteria were marked YES. This was developed by the senior resident inspector and a senior project engineer from DCO/CIB2. Based on this information it is recommended that this issue be reviewed by the resident inspectors during routine baseline inspection.
BRANCH CHIEF REVIEW:
Bradley Davis, DCO/CIB2 DIVISION DIRECTOR REVIEW:
Nicole Coovert, DCO ADAMS ACCESSION NUMBER: ML24197A093 EVENT NOTIFICATION REPORT NUMBER (as applicable):
E-mail to NRR_Reactive_Inspection@NRC.GOV Signed by Davis, Bradley on 07/15/24 Signed by Coovert, Nicole on 07/17/24