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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 49990 | 2 April 2014 20:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Fire Related Unanalyzed Condition That Could Impact Credited Safe Shutdown Analysis | This is a non-emergency notification. As part of converting to NFPA (National Fire Protection Association) 805, Catawba determined that there are conditions that may not ensure required equipment remains available under certain postulated fire scenarios. The determination concluded that the effects of a postulated fire in specific fire areas could prevent critical systems or components from performing their intended functions, potentially resulting in the inability to achieve and maintain safe shutdown. Affected fire areas are as follows (elevation numbers are in parentheses): 1. Residual Heat Removal/Containment Spray Pump Room (522) 2. Unit 2 Auxiliary Feedwater Pump Room (543) 3. Unit 1 Auxiliary Feedwater Pump Room (543) 4. Auxiliary Building Common Area (543) 5. Unit 2 Electrical Penetration Room (560) 6. Unit 1 Electrical Penetration Room (560) 7. 2ETB Switchgear Room (560) 8. 1ETB Switchgear Room (560) 9. Unit 2 Battery Room (554) 10. Unit 1 Battery Room (554) 11. Auxiliary Building Common Area (560) 14. Unit 2 ETA Switchgear Room (577) 15. Unit 1 ETA Switchgear Room (577) 17. Unit 1 Cable Room (574) 18. Auxiliary Building Common Area (577) 20. Unit 1 Electrical Penetration Room (594) 45. Unit 1 Cable Room Corridor (574) 46. Unit 2 Cable Room Corridor (574) 48. Unit 2 Inside Doghouse 49. Unit 1 Inside Doghouse RB1. Unit 1 Reactor Building RB2. Unit 2 Reactor Building This is reportable as an unanalyzed condition that significantly degrades plant safety in accordance with 10 CFR 50.72(b)(3)(ii)(B). Fire watches have been implemented as appropriate in response to this determination. This condition has been entered into the Catawba Corrective Action Program. The NRC Resident Inspector will be notified. The licensee will notify the State of North Carolina and South Carolina and well as the local counties. | Auxiliary Feedwater | |
| ENS 48348 | 26 September 2012 19:15:00 | 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 | Technical Specification 3.0.3 Entered Due to Surveillance Testing Not Being Completed as Required | Pressurizer Pressure LO Instrument surveillance testing was discovered to not have been completed as required. Existence of an internal jumper on solid state protection system input to logic cards prevented complete circuit testing. Unit 1 and unit 2 entered Technical Specification (TS) 3.0.3 at 1515 hours. Alternate method of surveillance testing on one of the trains was completed and TS 3.0.3 was exited on Unit 1 at 2203 EDT and Unit 2 at 2132 EDT. The licensee plans to complete surveillance testing on the second train before 1515 hours on 9/27/12. The licensee is currently in TS 3.3.2. The licensee has notified the NRC Resident Inspector. The states of North Carolina and South Carolina will be notified. Local county governments of York, Gaston, and Mecklenberg counties will also be notified.
Catawba nuclear station is retracting EN #48348 for both units 1 and 2 based on completion of testing and investigation of system operability. The testing determined that both units respective train 'A' and 'B' pressurizer pressure low were operable and would have functioned as needed. This condition is not reportable under 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(D). The licensee has notified the NRC Resident Inspector. R2DO (O'Donohue) notified. | ||
| ENS 43945 | 30 January 2008 17:20:00 | 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 | Auxiliary Feedwater Pumps Declared Inoperable Based on Postulated Main Feedwater Line Break Scenario | On 01/30/08, it was discovered that restrictor plates were not installed as required in the Unit 1 and Unit 2 interior doghouse floor drains. The restrictor plates are indicated as required on applicable flow diagrams. The purpose of the plates is to limit the amount of flow into the interior doghouse floor drains following a Main Feedwater (MFW) line break event in the doghouse. Without the restrictor plates installed, the amount of flow into the floor drains following a MFW line break could overwhelm the capability of the floor drain pump. Water could then flood into the AFW pump rooms and could render all three AFW pumps on the affected unit inoperable. Operations declared the 3 AFW pumps on both Units inoperable at 1220 today. TS 3.7.5.D states that with 3 AFW trains inoperable, LCO 3.0.3 and all other LCO required actions requiring Mode changes are suspended until on AFW train is restored to operable status. Both Units will remain at 100% power until repairs are complete. This condition does not prevent the AFW pumps from starting on an automatic or manual signal for any required plant events. An investigation into how this condition occurred is underway. As a result of this discovery, all three AFW pumps on each unit have been declared inoperable. Plant personnel are currently in the process of fabricating the required restrictor plates or blind flanges as necessary and installing them to conform to the drawing requirements. The restrictor plates/blind flanges are being installed and upon completion the affected AFW pumps will be considered operable. The licensee notified the NRC Resident Inspector. State and local government agencies will also be notified by the licensee. | Feedwater Auxiliary Feedwater | |
| ENS 43405 | 5 June 2007 00:08:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Fire Safe Shutdown Capability Impacted by a Breaker Coordination Issue | An engineering analysis has identified existing breaker coordination issues and their effect on plant fire protection and resulting safe shutdown capability. Cables associated with certain breakers do not have qualified fire barriers in place to prevent an upstream breaker from tripping. The affected cables are in fire areas that credit the affected train as the assured post fire safe shutdown train. The failure mode of concern is that a fire in the area where the cable is located could cause a short circuit between the cable connectors." The fault current calculated for these cases is high enough to trip the upstream breaker(s) and it cannot be assured that the downstream load breaker(s) will act to clear the fault first. (This) event was initially discovered on 5/17/07. Mitigative actions were taken at that time, in the form of establishing fire watches for the affected fire areas. Reportability implications were discovered as a result of additional discussions that took place on 6/4/07. The following cables and power sources are impacted by this notification: CABLE NO. FIRE AREA AFFECTED PANEL 1CA581 Unit 1 Auxiliary Feedwater Pump Turbine Control Panel Room 1EDF 2CA581 Unit 2 Auxiliary Feedwater Pump Turbine Control Panel Room 2EDF 1IRE761 Unit 1 577 elevation Electrical Penetration Room 1EPD 2IRE761 Unit 2 577 elevation Electrical Penetration Room 2EPD Panels 1EDF and 2EDF are 125 VDC Vital Instrumentation and Control Auctioneered Distribution Centers. Loads fed from these distribution centers include Train 'B' circuits associated with the following: 4 KV Essential Switchgear Control Power, Diesel Generator Load Sequencer Control Power, and Auxiliary Feedwater System Control Power. Panels 1EPD and 2EPD are 125VDC Vital Instrumentation and Control Panel boards. These panel boards supply a variety of Train 'B' safety related control power loads. The licensee will notify the NRC Resident Inspector. | Auxiliary Feedwater | |
| ENS 40837 | 23 June 2004 21:00:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Related to Fire Protection Cable Separation | This is being reported under 50.72(b)(3)(ii) - Unanalyzed Condition, 50.72(b)(3)(v) - event or condition that could have prevented fulfillment of a safety function, and Catawba License Condition 2F - violation of License Condition 2.C.5 - Fire Protection Program. This report applies to Catawba Unit 2. On 6/23/04 an original design deficiency related to fire protection cable separation criteria was discovered. The condition relates to a postulated fire in the 'A' Train 4160 Volt Essential Bus switchgear room (ETA). The 'A' and 'B' Train Volume Control tank (VCT) outlet valves are in series providing the normal suction source for the centrifugal charging pumps (standard Westinghouse design). Cables related to the 'A' Train Volume Control Tank outlet valve are in the same fire area as the switchgear for the 'A' Train centrifugal charging pump. Therefore, if a hot short closes the VCT outlet valve, with the 'B' pump running, both centrifugal charging pumps could be rendered inoperable. Additional switchgear related to component cooling, and power cables for our Standby Shutdown Facility (backup reactor coolant pump seal cooling), are also located in this fire area. An all consuming fire scenario in ETA Switchgear Room could cause a temporary loss of seal cooling to at least 2 of 4 reactor coolant pumps. The actual distance between the ETA switchgear and SSF cables is approximately 20 ft; therefore the probability of an actual fire resulting in a loss of seal cooling is very low. The unanalyzed condition only exists when the 'B' centrifugal charging pump is in operation (both units currently have 'A' centrifugal charging pump in operation). This situation is analogous to a degraded fire rated assembly such that an all consuming fire in one area would affect multiple trains of safe shutdown equipment. Per our Fire Protection Program we are establishing appropriate fire watches as a remedial action until the issue is resolved. Catawba is currently evaluating if a similar concern exists on Unit 1. A follow-up report will be issued if necessary. The licensee will notify the applicable counties and states and has notified the NRC Resident Inspector.
On 6/23/04 an original design deficiency related to fire protection cable separation criteria for Unit 2 was discovered. The report stated that Catawba was evaluating if a similar concern existed on Unit 1 and that a follow-up report would be issued if necessary. A similar but slightly different condition exists on Unit 1. For Unit 1 the Standby Shutdown Facility (backup seal cooling) cables go through the Train "B" 4160 Essential Switchgear Room (1ETB) rather than 2ETA on Unit 2. This means that for Unit 1, an all consuming fire scenario in 1ETB Switchgear Room could cause a temporary loss of seal cooling. This situation is analogous to a degraded fire rated assembly such that an all consuming fire in one area would affect multiple trains of safe shutdown equipment. Per our Fire Protection Program we are establishing appropriate fire watches as a remedial action until the issue is resolved. Notified R2DO (Haag) | ||
| 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) | ||
| ENS 40359 | 28 November 2003 03:55:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 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 | Component Cooling System Surge Tank Supports May Fail During a Seismic Event | During an ISI inspection of the component cooling system, it was discovered that several of the surge tank support stiffener plates were missing due to a design change that was never fully implemented. This condition creates a question as to the adequacy of the tanks to withstand a seismic event. This deficiency affects both trains of component cooling on Units 1 and 2. The problem will be corrected on Unit 1 while it is in its refueling outage. Work on the Unit 2 "A" train is currently ongoing, making that train of the component cooling system inoperable, therefore, putting the plant in a 72 hour LCO. When work on the "A" train is completed, they will exit the LCO, but re-enter it when work begins on the "B" train. Licensee will notify the NRC Resident Inspector, the State and local agencies.
The subject EN #40359 was made on 11/28/03. 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 condition that could have prevented fulfillment of a safety function, as the tank supports were analyzed to withstand a seismic event. In addition, the event did not result in the plant being in an unanalyzed condition. The condition did not prevent the fulfillment of a safety function or system that is needed to remove residual heat or mitigate the consequences of an accident. The licensee plans to notify the NRC Resident Inspector. Notified R2DO Caudle Julian. |