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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56165 | 15 October 2022 15:59:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Vessel Head Penetration Repair Degraded | The following information was provided by the licensee via email: On 10/15/2022 at 1159 (EDT), during the Catawba Nuclear Station Unit 2 refueling outage, it was determined that the results of a planned surface examination Liquid Penetrant test (PT) performed on a previous overlay repair on nozzle number 74 of the reactor vessel closure head (RVCH) did not meet applicable acceptance standards. The examination was being performed to meet the requirements of Relief Request RA-21-0144, 'Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant'. The penetration required repairs for the discovered indications. The repairs have been completed in accordance with the ASME Code of Record prior to returning the vessel head to service. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The original indication that led to the overlay repair was discovered in April 2021, during ultrasonic testing and reported to the NRC and assigned EN55201. | ||
| ENS 55275 | 25 May 2021 21:51:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Unit 2 Containment Local Leak Rate and As-Found Integrated Overall Leakage Rate Exceeded Acceptance Criteria | At 1751 EDT on May 25, 2021, it was determined the local leak rate test (LLRT) for the 2EMF-IN containment penetration did not meet 10 CFR 50 Appendix J requirements for both the inboard and outboard containment isolation valves (2MISV5230 and 2MISV5231). The LLRT was performed during the previous refueling outage at which time primary containment was not required to be operable. The leakage assigned to the penetration also resulted in total leakage exceeding the allowed overall leakage. The valves were repaired and retested satisfactory prior to entering the mode of applicability, This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A), There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Primary containment | |
| ENS 55201 | 21 April 2021 02:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System (Rcs) Pressure Boundary Degraded | During the performance of reactor vessel closure head (RVCH) examinations, at 2230 EDT on April 20, 2021, it was determined that the Unit 2 RVCH penetration nozzle number 74 did not meet the requirements of 10CFR50.55a(g)(6)(ii)(D) and ASME code case N-729-6 . All other RVCH penetration examinations have been completed per 10CFR50.55a(g)(6)(ii)(D) and ASME code case N-729-6 with no other relevant indications identified. The condition of the Unit 2 reactor vessel head penetration nozzle number 74 will be resolved prior to re-installation of the Unit 2 RVCH. This event is being reported as an eight-hour, non-emergency notification per 10CFR50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Reactor Coolant System | |
| ENS 54707 | 13 May 2020 02:20:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System Pressure Boundary Degraded | During the performance of reactor vessel closure head (RVCH) inspections, at 2220 EDT on May 12, 2020, it was determined that the Unit 1 RVCH penetration nozzle number 18 did not meet ASME code case N-729-4 requirements. A surface examination (penetrant test) identified a linear indication on nozzle number 18. The indication was not through-wall as determined by ultrasonic testing. The condition of the Unit 1 reactor vessel head penetration nozzle number 18 will be resolved prior to re-installation of the Unit 1 reactor vessel head. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Reactor Coolant System | |
| ENS 45710 | 19 February 2010 00:15:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Rcs Pressure Boundary Leakage Identified on Rtd Weld | Catawba Nuclear Station is making an 8 hour notification due to a pressure boundary leak on the A loop of the Unit 1 Reactor Coolant System. The leak was found to be originating from 1A Reactor Coolant System hot leg (Resistance Temperature Detector) RTD penetration. Since the allowance for pressure boundary leakage is zero, this is being considered a degraded principal safety barrier. No safety signals were received and no actuations occurred as a result of this leak. The amount of the leak is small and is contained inside containment. No release to the environment occurred and there is no danger to the public. Repair will require the plant to enter Mode 5. The pressure boundary leakage is 0.08 GPM. The licensee is in a 36 hour LCO 3.4.13(B)(2), and plans to be in Mode 5 this evening. The licensee has notified the NRC Resident Inspector. | Reactor Coolant System | |
| ENS 41048 | 17 September 2004 14:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Apparent Pressure Boundary Leakage Identified on Steam Generator Bowl Drains | On September 16, 2004 examinations of the steam generator (SG) bowl drains (SGBDs) for the 2A, 2C and 2D SGs were performed as part of the Alloy 600 program during 2EOC13. These examinations determined that the 2C and 2D SGBDs had leakage and the leakage appeared to be pressure boundary leakage. The 2A SGBD examination determined that the ASME code acceptance limits were satisfied. An evaluation of the leakage indications determined that the leakage had existed for some time prior to the Unit 2 shutdown. The evaluation of this condition on September 17, 2004, determined that this met the reportability criteria of 10 CFR 50.72 (b)(3)(ii)(A). Unit 2 is currently in MODE 6 and this event has no impact on current plant operation. Engineering and plant management are evaluating repair methods to be completed during this refueling outage. This event is not applicable to Unit 1 because the Unit 1 steam generators are of a different design and do not have a drain line in the bottom channel head. The licensee will notify the NRC Resident Inspector. | Steam Generator | |
| ENS 40372 | 3 December 2003 20:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Backward Installation of Containment Loop Seal Penetration Vacuum Breakers | Vacuum breakers 1WL980 and 2WL980 are installed backwards. In their current orientation, the valves will not lift from their seats to break a siphon into the corresponding unit's Ventilation Unit Condensate Drain Tank (VUCDT). The VUCDT input line is a 6-Inch pipe. There is a loop seal between the outboard containment isolation valve and the VUCDT. Since the VUCDT is vented to the auxiliary building environment, the purpose of the loop seal is to provide a barrier between the containment atmosphere and the auxiliary building atmosphere during normal unit operations. The purpose of the vacuum breaker is to prevent siphoning water out of the loop seal. In its current configuration, the vacuum breaker will not open. The loop seal is not needed to provide a barrier between the containment atmosphere and the auxiliary building atmosphere during a large break Loss of Coolant Accident (LOCA) because valves 1(2)WL867A and 1(2)WL869B will close on a Phase B containment isolation signal on high-high containment pressure (3.2 psig in containment, accounting for instrument error). During certain small break LOCAs, however, a high-high containment isolation signal may not occur, since pressure might not reach the setpoint. In this scenario, the loop seal is needed to isolate the containment atmosphere from the auxiliary building atmosphere. Given the size of the VUCDT inlet piping, the only mechanism that could form a siphon out of the loop seal is a large flow of water that would push the air out of the top of the loop seal. In this instance, a siphon could form and pull water out of the low point of the loop seal. If this were to occur, a vent path from the containment atmosphere to the auxiliary building atmosphere would be open. However, during normal operation, there is not sufficient flow into the tank to make this a plausible scenario. For a large break LOCA, containment pressure would rise quickly to the high-high setpoint; then the inoperable VUCDT loop seal would be isolated by its containment isolation valves. For smaller LOCAs, particularly, for a rod ejection accident resulting in a LOCA, containment pressure would rise slowly- from 2.81 psig (the pressure at which the loop seal isolation function would fail), until 3.2 psig (the maximum high-high containment pressure setpoint, accounting for instrument error), the inoperable loop seal would represent a containment leak path. The rod ejection accident does result in a high level of fuel clad failure; therefore, the unisolated containment leak path represents a source of release to the environment until such time as the high-high containment pressure setpoint is reached (if it is reached). The dose consequences associated with this potential leak path have not been evaluated. Upon discovery of the incorrectly installed vacuum breakers, the containment isolation valves associated with this penetration flow path were closed to isolate the path. The Unit 2 loop seal configuration has since been modified to correct this situation. The Unit 1 loop seal configuration will be modified prior to the completion of the current end of cycle 14 refueling outage. The incorrect installation of the vacuum breakers was identified on 11/03/03, and it is being investigated on how long this condition has existed. It is possible that it has existed since construction. The licensee will notify the NRC Resident Inspector, state and local regulatory agencies.
The subject EN was made on 12/3/03. Following additional review by the licensee, this event was determined to not meet the reportability requirements of 10 CFR 50.72. The event was determined to not result in a degraded or unanalyzed condition, as the consequences of the event were determined to be bounded by transients currently analyzed and described in the Updated Final Safety Analysis Report (UFSAR). In addition, the event did not represent a failure of structures, systems, or components utilized to control the release of radiological material or to mitigate the consequences of an accident. The licensee is therefore retracting the subject EN. The licensee notified the NRC Resident Inspector. Notified R2DO (P. Fredrickson) |