ML23311A424
| ML23311A424 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 11/06/2023 |
| From: | Hironori Peterson NRC/RGN-III/DORS/RPB3 |
| To: | |
| References | |
| LER 2023-001-01, LER 2023-001-00 | |
| Download: ML23311A424 (1) | |
Text
MD 8.3 Evaluation Decision Documentation for Reactive Inspection (Deterministic and Risk Criteria Analyzed)
PLANT:
Prairie Island EVENT DATE:
10/19/2023 DETERMINISTIC CRITERIA EVALUATION DATE:
11/02/2023 Brief Description of the Significant Operational Event or Degraded Condition:
At 11:10 on 10/19/2023 Unit 1 tripped due to both generator output breakers opening coincident with a loss of the 1R transformer (offsite power) and all non-safeguards busses. All reactor coolant pumps (RCP) and main feed pumps (MFP) de-energized, as well as circulating water pumps, condensate pumps, and other non-vital loads. No loss of offsite power to the safety related buses.
Unit 1 decay heat removal was via natural circulation, due to loss of power to RCPs, via safety related auxiliary feed water (AFW) pumps, and Steam Generator Power Operated Relief Valves.
Plant personnel responded per their Emergency Operating Procedures to place the plant in a safe condition. The licensee commenced cooldown to Mode 5. Offsite power was maintained to the vital busses and therefore both Unit 1 Emergency Diesel Generators (EDG) remained available and did not start (no demand). Unit 2 was shutdown, defueled for refuel outage with D5 EDG available and D6 EDG in maintenance. The licensee considered alternate power supplies for support systems to maintain operability of the D5. One train of spent fuel pool (SFP) cooling deenergized during the fast transfer and was manually restarted approximately 70 minutes after the trip with no noticeable impact on SFP temperatures. Both trains of SFP were operable throughout the event. There was no loss of SFP cooling.
No Emergency Action Level entries were required based on the loss of offsite power to non-safeguards buses. The plant was stabilized with offsite power to all safety buses.
Preliminarily the licensee believes cause is due to underground directional boring that was being done outside but near the switchyard that may have impacted DC control power lines running between the plant and the switchyard. This caused grounds on the DC busses as well as spurious operation of several switchyard breakers, relays, and indicating lights.
As of 10/25/2023, 1R Transformer remains deenergized, some non-safeguards loads have been restored using the D3 EDG. Efforts are being made to restore the 1R transformer utilizing a temporary modification.
2 Y/N DETERMINISTIC CRITERIA N
- a. Involved operations that exceeded, or were not included in, the design bases of the facility Remarks: No operations that exceeded or were not included in the design basis.
N
- b. Involved a major deficiency in design, construction, or operation having potential generic safety implications Remarks: No deficiencies having potential generic safety implications.
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: No known loss of integrity to any fuel barrier.
N
- d. Led to the loss of a safety function or multiple failures in systems used to mitigate an actual event Remarks: Power was lost to support systems (CRDM Cooling Fans, EDG Air start, EDG Keep warm), but did not result in any T.S. inoperability. The safeguards busses remained energized and capable of performing their safety function.
N
- e. Involved possible adverse generic implications Remarks: No known adverse generic implications.
N
- f. Involved significant unexpected system interactions Remarks: There were unexpected system interactions with spurious operation of breakers and relays in the switchyard as the initiating event resulting in a loss of 1R offsite power source to non-vital bus, and Unit 1 turbine and reactor trip. There were no unexpected plant responses or system interactions following the trip and during the recovery.
N
- g. Involved repetitive failures or events involving safety-related equipment or deficiencies in operations Remarks: No repetitive failures involving safety-related equipment or deficiencies in operations during the event.
Y
- h. Involved questions or concerns pertaining to licensee operational performance Remarks: Operational performance during the mitigation of the event was satisfactory in dealing with loss of non-safety equipment, natural circ cooldown, and safely placing the reactor in cold shutdown. However, the control of T&D operations as the initiating factor and potential quality/enhancements needed on procedures were of concern.
3 CONDITIONAL RISK ASSESSMENT RISK ANALYSIS BY: Carey Bickett DATE: 11/3/2023 Brief Description of the Basis for the Assessment (may include assumptions, calculations, references, peer review, or comparison with licensee=s results):
Regional senior reactor analysts (SRAs) performed an analysis of this issue using SAPHIRE 8, version 8.2.8, and the Prairie Island SPAR model version 8.82, with the following assumptions and adjustments:
The SRAs set IE-TRANS to 1.0 to model the reactor trip. All other initiators were set to zero.
Because all 4KV non-vital buses were lost, the SRAs set the basic events for failure of the non-vital buses to TRUE. 4KV non-vital bus 12 is not included in the SPAR model.
Loss of the non-vital 4KV buses resulted in loss of power to the main feedwater pumps, condensate pumps, reactor coolant pumps, and circulating water pumps. The SRAs set the basic events for failure of the condensate pumps and main feedwater pumps to TRUE. The reactor coolant pumps and circulating water pumps are not included in the SPAR model.
Given the nature of the failure, power was not easily restored to the non-vital buses. As such, the SRAs set operator actions related to failure to recover main feedwater to TRUE.
The SRAs determined the conditional core damage probability (CCDP) for this event to be approximately 2E-6. The dominant core damage sequences included anticipated transients without scram (ATWS) and failure of the auxiliary feedwater system.
The risk for this event is in the overlap region of no additional inspection and SIT.
The estimated conditional core damage probability (CCDP) is 2 E-6 and places the risk in the range of overlap between no additional inspection and SIT 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:
Region III has determined that the licensees operational performance adequately mitigated the reactor trip event for loss of offsite power to only the non-vital buses and dealt with challenges, and the operators satisfactorily placed the reactor in safe cold shutdown condition, inspection should continue under the baseline inspection program versus a special inspection (SIT). The basis for this decision is as follows:
The equipment degradation was limited to non-safety related equipment. No safety related systems or components were lost during mitigation of the loss of offsite power to the non-vital buses, which limits the overall risk of the issue.
The inspectors, due to prompt response to the event and subsequent inspection activities related to the response, have obtained a significant amount of information regarding potential causes of the degraded condition which initiated the event.
Licensee performance before, during, and after the degradation was identified, and the licensees root cause and corrective actions are being followed by the inspectors.
Based on these items, continued inspection in the baseline inspection program is the appropriate approach to resolving this issue.
BRANCH CHIEF: Hironori Peterson /RA/
DATE: 11/6/2023 SRA: Carey Bickett /RA/
DATE: 11/6/2023 DIVISION DIRECTOR: Jason W. Kozal /RA/
DATE: 11/6/2023 RA (if reactive inspection is initiated)
DATE:
ADAMS ACCESSION NUMBER: ML23311A424 ADAMS PACKAGE ACCESSION NUMBER: ML24284A359 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:
10/19/2023 EVALUATION DATE:
11/2/2023 Brief Description of the Significant Operational Event or Degraded Condition: See Page 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:
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
6 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)
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:
7 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 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:
8 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:
N 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:
9 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:
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:
10 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:
Region III has determined that although licensees operational performance adequately mitigated the reactor trip event for loss of offsite power to only the non-vital buses and dealt with challenges, and the operators satisfactorily placed the reactor in safe cold shutdown condition, inspection should continue under the baseline inspection program versus a special inspection (SIT). The basis for this decision is as follows:
The equipment degradation was limited to non-safety related equipment and no safety related systems or components were lost during mitigation of the loss of offsite power to the non-vital buses, which limits the overall risk of the issue.
The inspectors, due to prompt response to the event and subsequent inspection activities related to the response, have obtained a significant amount of information regarding potential causes of the degraded condition which initiated the event.
Licensee performance before, during, and after the degradation was identified, and the licensees root cause and corrective actions are being followed by the inspectors.
Based on these items, continued inspection in the baseline inspection program is the appropriate approach to resolving this issue.
BRANCH CHIEF: Hironori Peterson /RA/
DATE: 11/6/2023 SRA: Carey Bickett /RA/
DATE: 11/6/2023 DIVISION DIRECTOR: Jason W. Kosal /RA/
DATE: 11/6/2023 ADAMS ACCESSION NUMBER: ML23311A424 ADAMS PACKAGE ACCESSION NUMBER: ML24284A359 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; Laura.Pearson@nrc.gov; LaDonna.Suggs@nrc.gov; Ravi.Penmetsa@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