ML24276A045
| ML24276A045 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 09/23/2024 |
| From: | Erin Carfang, Blake Welling Division of Operating Reactors |
| To: | |
| References | |
| Download: ML24276A045 (1) | |
Text
Issue Date: 12/14/23 Encl1-1 0309
- Decision Documentation for Reactive Inspection (Deterministic and Risk Criteria Analyzed)
Decision Documentation for Reactive Inspection (Deterministic and Risk Criteria Analyzed)
PLANT: Nine Mile Point Unit 2 EVENT DATE: 09/23/2024 EVALUATION DATE: 09/23/2024 Brief Description of the Significant Event or Degraded Condition:
On 09/23/2024 at approximately 0720, from mode 1 at 100 percent, NMP Unit 2 experienced a SCRAM because of a main turbine trip and resultant turbine stop valve closure. The main turbine trip cause was the result of a localized grid disturbance caused by high resistance in the exciter breaker. The sight and smell of smoke was reported to the main control room by plant personnel at 0738. A responding operator reported a visible flame (confirmed a fire) of the exciter breaker at 0739. Licensee fire brigade extinguished the fire at approximately 0803. The fire reflashed at approximately 0806, and was extinguished again at approximately 0820. Offsite Scriba Fire Department was called for support, but did not assist in extinguishing the fire. Damage was localized to the NMP2 exciter enclosure and did not impact the ability to place the unit in a safe and stable condition. Additionally, at the time of the event, the neighboring nuclear plant, James A.
FitzPatrick (JAF), also experienced a SCRAM as a result of the localized grid disturbance. At this time, each event will be treated separately. There was no impact to Nine Mile Point Unit 1 (NMP1),
which continued to operate in mode 1 at 100 percent.
NMP2 operator and plant response to the SCRAM was not complicated. All control rods inserted.
Reactor water level lowered below level 2 (108.8 inches) during the transient resulting in the initiation and injection of High-Pressure Core Spray (HPCS) and Reactor Core Isolation Cooling (RCIC) to restore water in band and provide decay heat removal. Reactor water level control was maintained using RCIC. In addition, with reactor water level below level 2 during the transient, primary containment isolation signals actuated resulting in group 2 recirculation sample system isolation, group 3 traveling in-core probe (TIP) isolation valve isolation, group 6 and 7 reactor water cleanup isolation, group 8 containment isolations, and group 9 containment purge isolations.
All ECCS and electrical systems responded as expected. Main steam isolation valves remained open including availability of feedwater, condensate and main condenser for pressure control and decay heat removal (via the turbine bypass valves). Offsite power remained stable and available.
Review of MVAR data showed small swings following by a large MVAR drop from NMP2 that caused a large demand in NMP1 and FitzPatrick. This demand in MVAR resulted in an overvoltage condition for NMP1 and FitzPatrick. The stations have a breakover diode, which functions as an overvoltage trip associated with the AVR. The overvoltage trip is set at 800V for FitzPatrick and 1200V for NMP1. Additionally, NMP1 has had a generator rewind recently, unlike FitzPatrick, which would show differences in generator capabilities.
Issue Date: 12/14/23 Encl1-2 0309 Y/N DETERMINISTIC CRITERIA N
Involved operations that exceeded, or were not included in, the design bases of the facility Remarks: No design bases were exceeded during this event N
Involved a major deficiency in design, construction, or operation having potential generic safety implications Remarks: For this event, there was no identified design, construction, or operation deficiency having potential generic safety implications.
N Led to a significant loss of integrity of the fuel, primary coolant pressure boundary, or primary containment boundary of a nuclear reactor Remarks: There was no impact or challenge to the integrity of the fuel or primary containment boundary N
Led to the loss of a safety function or multiple failures in systems used to mitigate an actual event Remarks: This event did not lead to loss of a safety function or cause any failures in systems used to mitigate an actual event.
N Involved possible adverse generic implications Remarks: Based on current knowledge of the cause of the event, there are no adverse generic implications.
Y Involved significant unexpected system interactions Remarks: NMP2 exciter breaker failure led to a SCRAM at NMP2. JAF experienced a SCRAM at nearly same time as the NMP2 event due to a localized grid disturbance. NMP1 did not experience a generator trip / SCRAM due to different setpoints on generator protection system. The licensee has reviewed the timeline of events and determined the failure of the breaker at NMP U2 caused the localized grid disturbance. Inspectors have not independently validated this information.
N Involved repetitive failures or events involving safety-related equipment or deficiencies in operations Remarks: The event did not involve repetitive failures N
Involved questions or concerns pertaining to licensee operational performance Remarks: Initiation of the fixed CO2 system to suppress the fire was delayed because of a human performance error. System was subsequently successfully initiated.
Issue Date: 12/14/23 Encl1-3 0309 CONDITIONAL RISK ASSESSMENT RISK ANALYSIS BY: David Werkheiser DATE:9/24/2024 Brief Description of the Basis for the Assessment (may include assumptions, calculations, references, peer review, or comparison with licensees results):
A senior reactor analyst (SRA) evaluated the reactor trip event using the Nine Mile Point 2 SPAR model 8.82 and SAPHIRE program version 8.2.11.
Though the initiator is an apparent failure of the generator exciter breaker and resulting fire, no secondary damage was experienced, and it resulted in a SCRAM (i.e. IE-TRANSIENT); The IE-TRANS basic event was set to a probability of 1.0 with all other initiating events set to 0.0. Test and maintenance basic events were set to zero. The SRA assessed that operators maintained reactor water level via RCIC past post-trip response and longer than expected versus a timely transition to normal feedwater. Based on discussions with onsite inspectors, this was due to known feedwater regulating valve leakage that may challenge level control. Operators transitioned to normal feedwater once group isolation was reset and a letdown path was established. Though feedwater was available, for this assessment, the SRA set feedwater function (using common failure of motor driven feedwater pumps, MFW-MDP-CF-FR; operator start/control MFW injection, MFW-XHE-XO-ERROR) conservatively to 0.1 failure probability as a surrogate to failure / recovery. A nominal human error probability for feedwater restoration action-only is 1E-3. A SPAR-H estimate for high-stress (diagnosis and action) is 2.2E-2. The value of 0.1 used is considered a bounding value for this event that also involved a fire. The resulting mean conditional core damage probability (CCDP) was estimated at 5E-7 (5% 2E-8, 95% 2E-6).
The dominant core damage sequences involved a reactor trip with failure of main feedwater, failure of HPCS and RCIC to provide sufficient flow to the reactor vessel, and failure of operators to manually depressurize to support low pressure injection.
As a sensitivity, the SRA modeled the event as a loss of main feedwater (a plausible turbine building fire consequence) with similar basic event adjustments as above, except IE-LOMFW was set to 1.0 and other initiating events were set to zero. This sensitivity is considered a bounding case since normal feedwater and condensate were available during the event, but prudent since initial discussions with inspectors determined there were initial challenges (but eventually successful) manual initiation of the installed cardon dioxide fire suppression system at the exciter breaker location. Challenges in fire suppression systems and response may increase fire risk and is not directly assessed. This sensitivity considered the restoration surrogate event by operator action for a LOMFW event. The resulting mean conditional core damage probability (CCDP) was estimated at 1.5E-6 (5% 6E-7, 95% 6E-6).
The dominant core damage sequences involve a loss of main feedwater (and reactor trip),
failure of HPCS and RCIC to provide sufficient flow to the reactor vessel, and failure of operators to manually depressurize to support low pressure injection.
Due to the delay in the initial licensee response to the fire, there may be additional qualitative fire risk, which is not documented in this assessment.
Issue Date: 12/14/23 Encl1-4 0309 The estimated conditional core damage probability (CCDP) is 5E-7 to 1.5E-6 and places the risk in the range of a No Additional Inspection and Special Inspection overlap area.
Issue Date: 12/14/23 Encl1-5 0309 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: Table 1, Deterministic Criterion 6, Involved Significant Unexpected Interactions, was met. A risk assessment was performed resulting in the no-inspection/SIT overlap region. Based on the known facts at this time and evaluated risk, no reactive inspection is planned at this time. Follow up will be through baseline inspection procedures IP 71153 and IP 71152 to inspect significant unexpected system interactions described in Criterion 6.
BRANCH CHIEF REVIEW:
DIVISION DIRECTOR REVIEW:
ADAMS ACCESSION NUMBER: ML24276A045 EVENT NOTIFICATION REPORT NUMBER (as applicable):
Note: The above tables are provided as examples only. The regions have discretion to modify these tables in their implementing procedures or office instructions.
ERIN CARFANG Digitally signed by ERIN CARFANG Date: 2024.10.02 11:42:10 -04'00' BLAKE WELLING Digitally signed by BLAKE WELLING Date: 2024.10.02 11:45:38 -04'00'
Issue Date: 12/14/23 Encl2-1 0309
- Decision Documentation for Reactive Inspection and Examples (Deterministic-only Criteria Analyzed)
Decision Documentation for Reactive Inspection (Deterministic-only Criteria Analyzed)
PLANT: Nine Mile Point 2 EVENT DATE: 9/23/24 EVALUATION DATE: 9/23/24 Brief Description of the Significant Event or Degraded Condition:
See description in Enclosure 1 REACTOR SAFETY Y/N IIT Deterministic Criteria N
Led to a Site Area Emergency Remarks:
N Exceeded a safety limit of the licensee's 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:
Issue Date: 12/14/23 Encl2-2 0309 RADIATION SAFETY Y/N IIT Deterministic Criteria N/A 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/A 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)
Remarks:
N/A 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/A Involved byproduct, source, or special nuclear material, which may have resulted in a fatality Remarks:
N/A 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/A 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:
Issue Date: 12/14/23 Encl2-3 0309 N/A 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/A 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/A Involved the failure of the dam for mill tailings with substantial release of tailings material and solution off site Remarks:
Y/N SI Deterministic Criteria N/A 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/A 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/A 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/A 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:
Issue Date: 12/14/23 Encl2-4 0309 N/A 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 Remarks:
N/A 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:
N/A 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:
N/A 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:
Issue Date: 12/14/23 Encl2-5 0309 SAFEGUARDS/SECURITY Y/N IIT Deterministic Criteria N/A 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/A 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/A Actual intrusion into the protected area Remarks:
Y/N AIT Deterministic Criteria N/A Involved a significant infraction or repeated instances of safeguards infractions that demonstrate the ineffectiveness of facility security provisions Remarks:
N/A Involved repeated instances of inadequate nuclear material control and accounting provisions to protect against theft or diversions of nuclear material Remarks:
N/A Confirmed tampering event involving significant safety or security equipment Remarks:
N/A 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:
Issue Date: 12/14/23 Encl2-6 0309 Y/N SI Deterministic Criteria N/A 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/A Involved a significant safeguards infraction that demonstrates the ineffectiveness of facility security provisions Remarks:
N/A Confirmation of lost or stolen weapon Remarks:
N/A Unauthorized, actual non-accidental discharge of a weapon within the protected area Remarks:
N/A Substantial failure of the intrusion detection system (not weather related)
Remarks:
N/A Failure to the licensees package/personnel search procedures which results in contraband or an unauthorized individual being introduced into the protected area Remarks:
N/A 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 Remarks:
Issue Date: 12/14/23 Encl2-7 0309 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: No deterministic criteria were met. A reactive inspection is not planned, based on the criteria evaluated.
BRANCH CHIEF REVIEW:
DIVISION DIRECTOR REVIEW:
ADAMS ACCESSION NUMBER: ML24276A045 EVENT NOTIFICATION REPORT NUMBER (as applicable):
Note: The above tables are provided as examples only. The regions have discretion to modify these tables in their implementing procedures or office instructions.
ERIN CARFANG Digitally signed by ERIN CARFANG Date: 2024.10.02 11:42:48 -04'00' BLAKE WELLING Digitally signed by BLAKE WELLING Date: 2024.10.02 11:46:17 -04'00'