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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 54061 | 12 May 2019 04:05:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Discovery of a Condition That Could Have Prevented Fulfillment of a Safety Function | On 5/11/19, Callaway Energy Center entered Mode 4 at 1217 (CDT). At 2305, the door from the Auxiliary Building to the RAM Storage building was found blocked open. This door is an Auxiliary Building pressure boundary for the Emergency Exhaust system. The Emergency Exhaust system is required in Modes 1,2,3,4, and during movement of irradiated fuel assemblies in the Fuel Building. The door was being blocked open with a large ramp. This rendered the Emergency Exhaust system not capable of performing its design safety function. LCO (Limiting Conditions for Operation) 3.7.13.B was entered, and preparations to move the ramp commenced. LCO 3.7.13.B is for two Emergency Exhaust trains being inoperable due to an inoperable auxiliary building boundary. The allowed outage time is 24 hrs. to restore the boundary to Operable. The door was closed and LCO 3.7.13.B was exited at 0111 on 5/12/19. This event is reportable per 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (C) control the release of radioactive material, or (D) mitigate the consequences of an accident. The NRC Senior Resident has been notified. | ||
| ENS 53759 | 28 November 2018 06:00:00 | 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 | Discovery of a Condition That Could Have Prevented Fulfillment of a Safety Function | On November 28, 2018, while performing an engineering review of the bases for environmental qualification (EQ) requirements for the Atmospheric Steam Dumps (ASDs), it was determined that applicable EQ requirements had not been applied to a key component of each of the ASDs. The result of this issue is that it the availability of the ASDs for a controlled plant cooldown following a postulated steam line break outside containment cannot be assured. Callaway is developing a compensatory action temporary plant modification to install insulation that will protect the affected ASD components from the post Main Steam Line Break temperature. This condition is reportable 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (B) remove residual heat, or (D) mitigate the consequences of an accident. The issue places the plant in a 24-hour Technical Specification (TS) Limiting Condition for Operations (LCO), 3.7.4. The licensee has notified the NRC Resident Inspector. | Main Steam Line | |
| ENS 53388 | 7 May 2018 18:35:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 | Discovery of a Condition That Could Have Prevented Fulfillment of a Safety Function | On May 7, 2018, during an engineering review of mission time requirements for Technical Specification related equipment, a deficiency was discovered regarding the Emergency Operating Procedure (EOP) guidance for natural circulation cooldown with a stagnant loop. This condition could be the result of a postulated Main Steam Line Break with a loss of offsite power. During a natural circulation cooldown with a faulted steam generator, flow in the stagnant reactor coolant system (RCS) loop associated with the isolated faulted steam generator (SG) could stagnate and result in elevated temperatures in that loop. This becomes an issue when RCS depressurization to residual heat removal system (RHR) entry conditions is attempted. The liquid in the stagnant loop will flash to steam and prevent RCS depressurization. In this condition, the time required to complete the cooldown would be sufficiently long that the nitrogen accumulators associated with Callaway's atmospheric steam dumps and turbine driven auxiliary feedwater pump flow control valves would be exhausted. The atmospheric steam dumps and turbine driven auxiliary feedwater pump would not be capable of performing their specified safety functions of cooling the plant to entry conditions for RHR operation. This issue has been analyzed by Westinghouse in WCAP-16632-P. This WCAP determined that to prevent loop stagnation, the RCS cooldown rate in these conditions should be limited to a rate dependent on the temperature differential present in the active loops. The WCAP analysis was used to support a revision to the generic Emergency Response Guideline (ERG) for ES-0.2 "Natural Circulation Cooldown." Figure 1 in ES-0.2 provides a curve of the maximum allowable cooldown rate as a function of active loop temperature differential which is directly proportional to the level of core decay heat. At the time of discovery of this condition, Callaway's EOP structure did not ensure that the ES-0.2 guidance would be implemented for a natural circulation cooldown with a stagnant loop. Callaway has issued interim guidance to the on-shift personnel regarding this concern and is in the process of revising the applicable EOPs. This condition is reportable per 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (A) Shutdown the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) mitigate the consequences of an accident." The licensee notified the NRC Resident Inspector of this condition. | Steam Generator Reactor Coolant System Auxiliary Feedwater Residual Heat Removal Main Steam Line | |
| ENS 53223 | 20 February 2018 18:25:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 | All Three Auxiliary Feedwater Pumps Inoperable Due to Helb Door Being Open | At 1225 CST, all three Auxiliary Feedwater (AFW) pumps were declared inoperable at the Callaway Plant upon discovery that a door (DSK13311) credited for protection of equipment from the effects of a high-energy line break (HELB) hazard had come partially open due to vibration harmonics from the running turbine-driven Auxiliary Feedwater pump (TDAFP). Immediate investigation identified that play in the mechanism that holds the door closed had rendered it susceptible to movement from the vibration harmonics. The affected HELB door specifically protects safety-related instruments that provide a swap-over signal upon detection of a low suction pressure condition for the AFW pumps and thereby automatically effect a suction transfer for the AFW pumps from the condensate storage tank (normal/standby source) to the Essential Service Water (ESW) system (credited safety-related source). All of the AFW pump suction transfer instrument channels were declared inoperable. Per Technical Specifications (TS) 3.3.2, 'Engineered Safety Feature Actuation System (ESFAS) Instrumentation,' the applicable Condition(s) and Required Action(s) for inoperable AFW pump suction transfer instrumentation only addresses a single channel being inoperable. Thus, the condition of having all three instrument channels inoperable required entry into TS Limiting Condition for Operation (LCO) 3.0.3. At the same time, however, with the automatic suction transfer capability rendered inoperable, all three AFW pumps, i.e., the TDAFP and the 'A' and 'B' motor-driven AFW pumps, were declared inoperable. Although LCO 3.0.3 was applicable, entry into the Required Actions of LCO 3.0.3 was suspended per the Note attached to Required Action E.1 of TS 3.7.5, 'Auxiliary Feedwater (AFW) System,' which states, 'LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.' At 1336 CST, Operations took actions to prevent the TDAFP from running, so the remaining (AFW Pumps) could be returned to Operable status. Operations then declared the affected instrumentation and the 'A' and 'B' motor-driven AFW pumps Operable. This allowed LCO 3.0.3 and Conditions A, B, D, and E under TS 3.7.5 to be exited. With only the TDAFP inoperable, TS 3.7.5 Condition C and its Required Actions remain in effect. Due to the degraded HELB door rendering all three AFW pumps inoperable, the unidentified condition is being reported as an unanalyzed condition that significantly degraded plant safety (per 10 CFR 50.72(b)(3)(ii)(B)) as well as a condition that could have prevented the fulfillment of the safety functions of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, and mitigate the consequences of an accident (per 10 CFR 50.72(b)(3)(v)(A), (B), and (D), respectively). The NRC Senior Resident Inspector has been notified.
Event Notification (EN) # 53223, made on 2/20/2018, is being retracted because new information has been obtained that negates the original basis for reporting the unanalyzed condition. Specifically, an evaluation of the HELB that is postulated to occur in the TDAFP room has determined that without crediting door DSK13311 for protection, the affected safety-related instruments would not be exposed to environmental conditions beyond their analyzed capability. This resulted in a conclusion that the unanalyzed condition of door DSK13311 being open did not prevent the affected safety-related instruments or their supported AFW pumps from performing their required safety functions to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, and/or mitigate the consequences of an accident, nor did it significantly degrade plant safety. Consequently, the condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(A), (B), or (D). The licensee notified the NRC Resident Inspector. Notified R4DO (Drake). | Service water Auxiliary Feedwater | |
| ENS 52607 | 13 March 2017 18:58:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 | Discovery of Non-Conforming Conditions During Tornado Hazards Analysis | On March 13, 2017, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Callaway Plant identified non-conforming conditions in the plant design such that specific Technical Specification equipment is considered to not be adequately protected from tornado missiles. The Emergency Fuel Oil Truck Connection Lines for both redundant Emergency Fuel Oil trains extend through the Plant South wall of the Diesel Generator Building structure where they may be exposed to design bases tornado missile impact. The direct impact by the design basis missile could result in damage to the fuel oil transfer lines, thereby preventing delivery of the fuel supply from the Emergency Fuel Oil buried storage tank to the Fuel Oil Day Tank. This condition affects the fuel supply to both supported Emergency Diesel Generator trains. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (A) Shut down the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) Mitigate the consequences of an accident. These conditions are being addressed in accordance with EGM 15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents). The NRC Resident Inspector has been notified.
Event Notification EN # 52607, made on 03/13/2017, is being retracted because new information has been obtained that negates the originally reported condition. Specifically, subsequent to the Event Notification, an engineering analysis was performed which confirmed that a design basis missile strike on either of the unprotected truck connection lines would not result in damage to the extent that the affected fuel oil transfer lines would be prevented from delivering the fuel supply from the Emergency Fuel Oil buried storage tank to the Fuel Oil Day Tank. (The analysis showed that although bending / deformation of the lines would occur in response to the postulated missile strike, integrity of the lines would remain.) Based on the above, the unanalyzed condition did not prevent the fulfillment of the safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, or mitigate the consequences of an accident, nor did it significantly degrade plant safety. Consequently, the condition does not meet the criteria for 8-hour notification that are provided in 10 CFR 50.72(b)(3)(ii)(B) or 10 CFR 50.72(b)(3)(v)(A), (B), or (D). As the condition does not require enforcement discretion, the provisions of EGM 15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents) need not be invoked. However, the immediate and long-term compensatory actions that have been taken following discovery of the condition will remain in place until the condition is fully resolved. The NRC Resident Inspector has been informed of this Event Notification retraction. Notified R4DO (Azua). | Emergency Diesel Generator | |
| ENS 50625 | 19 November 2014 01:00:00 | 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 | Unanalyzed Condition Due to All Four Safety Injection Accumulators Inoperable | Following shift turnover from days to nights on 11/18/2014, it was discovered that all (4) of the Safety Injection (SI) Accumulator Outlet Isolation Valve breakers were unlocked and closed. At the time of discovery, 3 of the safety injection accumulator valves were open and 1 was closed for testing. At that time the plant was in MODE 3 at normal operating pressure and temperature. The plant had been performing RCS pressure isolation valve testing prior to shift turnover. The condition was discovered during testing of valves associated with the 'C' safety injection accumulator. After discovery of the condition, Operations directed that the 'A', 'B', and 'D' SI Accumulator Outlet Isolation Valve breakers be opened and locked. This action was completed by approximately 1930 (CST) on 11/18/2014. The NRC Resident Inspector was notified. The plant entered T.S. 3.0.3 for approximately 30 minutes while restoring the 'A', 'B' and 'D' accumulators to operable (breakers opened and locked with their associated outlet valves open). | ||
| ENS 47783 | 28 March 2012 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 | Degraded Condition Identified in the B Train Containment Cooler Units | At 1500 on March 28, 2012, Callaway Plant personnel discovered that the installation of a modification on the two 'B' train containment cooler units had inadvertently introduced a potential failure mechanism to the 'B' train containment coolers. Specifically, with the containment cooling fans initially in fast-speed operation, combined with certain initial plant conditions, thermal overload tripping of the coolers could occur. In such an event, during some postulated accidents, slow-speed restart of the containment coolers by the Load Shedding and Emergency Load Sequencing system could be prevented. As a result, the 'B' train containment coolers could be rendered unavailable for a portion of a postulated accident. Thus, the safety function of the 'B' train containment cooling fans cannot be assured when this degraded equipment condition is present and the containment cooling fans are run in fast-speed operation. This condition existed for the 'B' train containment cooling units since they were restored to service from maintenance at 0400 on March 15, 2012. Upon identification of this condition, the 'B' train containment cooling fans were switched from fast-speed to slow-speed operation and restored to operable status at 1515. This action precludes this degraded equipment condition from adversely affecting containment cooling fan function during an accident. Concurrent with this condition, the opposite train of containment coolers was removed from service for scheduled maintenance at 0505 on March 27, 2012. As a result, from 0505 on March 27, 2012 until 1515 on March 28, 2012, the safety function of the containment cooling system could not be assured for certain postulated accident conditions. The NRC Resident Inspector has been notified. | ||
| ENS 46715 | 31 March 2011 19:32:00 | 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 | Unanalyzed Condition Identified for Inoperable Rwst Level | In response to a condition identified in late 2010 concerning the control and removal of hazard barriers in the plant, a review of the basis and analysis for high energy line breaks (HELBs) and the barriers for protecting against such events has been underway at Callaway in accordance with the plant's corrective action program. While following up on a question from the NRC Resident Inspector, and as a result of an additional question from the Nuclear Oversight organization at Callaway, it was identified that non-safety piping located in the valve room associated with the Refueling Water Storage Tank (RWST) could potentially (make) all four RWST low water level pressure transmitters inoperable in the event of a malfunction of the non-safety piping concurrent with a design-basis loss-of-coolant accident (LOCA) and/or following a seismic event. The RWST water level transmitters (which are located in the RWST valve room) perform a safety-related function for the emergency core cooling system (ECCS) by automatically swapping suction sources for the ECCS during a LOCA from the RWST to the containment sumps when a low water level condition is reached in the RWST. These instrument channels are required to be OPERABLE in Modes 1, 2, 3 and 4 per Callaway Technical Specification 3.3.2, 'Engineered Safety Feature Actuation System (ESFAS) Instrumentation.' The subject non-safety piping delivers steam supplied by the Auxiliary Steam system to (and from) heaters surrounding the RWST for maintaining RWST contents above the minimum required temperature during winter conditions. The piping passes through the RWST valve room containing the noted RWST water level transmitters which were designed only for a mild environment. It has been identified, however, that the non-safety Auxiliary Steam piping constitutes a high energy line and that its failure could create harsh (hot and wet) conditions in the valve room to which the RWST water level instrumentation was not designed. Per the Callaway FSAR, where non-safety piping interfaces with safety-related piping or systems, the design must be such that failure of the non-safety piping does not adversely affect the safety function(s) of the interfacing safety-related piping or system (since non-safety piping may be assumed to malfunction in conjunction with a design-basis accident). In this case, and based on a conservative interpretation of the FSAR, if the non-safety piping in the RWST valve room is assumed to malfunction (i.e., break), a failure of the RWST instrumentation could occur, thereby preventing the ECCS suction swap over from occurring as required or assumed for LOCA mitigation. This condition required declaring all four RWST water level channels inoperable. In light of recognizing that the RWST water level instruments could be subject to a harsh environment when they were only designed for a mild environment, and could thus fail as a result, this condition represents an unanalyzed condition that significantly degrades plant safety. With regard to the impact on the required ECCS suction swap over function that requires the RWST water level channels to be operable, the inoperability of all four instrument channels is a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to shutdown the reactor and maintain it in a safe condition, remove residual heat, control the release of radioactive material, and mitigate the consequences of an accident. Upon declaring the RWST water level instrument channels inoperable, TS Limiting Condition for Operation (LCO) 3.0.3 was entered at time 1432 CDT on 3/31/2011. At 1634 CDT, the Auxiliary Steam system was isolated and depressurized. This removed the energy that could be released from a break in the non-safety piping, thereby restoring Operability for the RWST water level instruments. The NRC Senior Resident Inspector was notified.
On March 31, 2011, event notification EN 46715 documented that a harsh environment from a postulated High Energy Line Break (HELB) in the Refueling Water Storage Tank (RWST) valve room could affect RWST level transmitters. These level transmitters provide RWST water level indication in the main control room, which is identified as a safe shutdown function in the Callaway FSAR. They also provide low RWST water level signals for effecting automatic swap over of suction sources for the Emergency Core Cooling System in the event of a loss-of-coolant accident (LOCA). This break may be postulated to occur on non-safety related auxiliary steam lines that run through the RWST valve room and on to the RWST heaters. This condition was initially reported both as an unanalyzed condition that significantly degraded plant safety and as a condition that could have prevented fulfillment of a safety function. When EN 46715 was reported, it was assumed that breaks were required to be postulated at any intermediate fitting, welded attachment, or valve on the subject auxiliary steam lines. Subsequent analysis shows that the sections of auxiliary steam piping in the RWST valve room are able to withstand safe shutdown earthquake (SSE) loadings and rupture loadings. For piping of this qualification, breaks at all intermediate fittings, welded attachments, and valves do not need to be postulated. Instead, line breaks are only required to be assumed at the terminal ends of the lines and at the locations specified for ASME Class 2 and 3 piping. None of these postulated break locations are located inside the RWST valve room, and a postulated auxiliary steam line break outside of the room would not adversely affect the RWST level transmitters. Since none of the postulated break locations are located inside the RWST valve room, there exists reasonable assurance that the RWST level transmitters would have remained capable of performing their safe shutdown function following a postulated break of the subject auxiliary steam lines. Further, there is no adverse effect on the assumed response to a postulated design basis LOCA since a hazard (such as a break in an auxiliary steam line) is not assumed to occur concurrently with the LOCA. Therefore, this condition does not meet the reporting requirements for an unanalyzed condition that significantly degraded plant safety or a condition that could have prevented fulfillment of a safety function. Event notification 46715 is hereby retracted. The NRC Senior Resident Inspector has been notified. Notified R4DO (Haire). | Emergency Core Cooling System | |
| ENS 46693 | 24 March 2011 04:54:00 | 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 | Pressure Transmitters Needed for Auxiliary Feedwater Suction Path Not Analyzed for Potential High Energy Line Break | While performing an extent of condition review of high energy line break (HELB) analyses, a detailed review of the auxiliary steam system was being performed. During this review, sections of pipe that run through rooms 1206/1207 in the Auxiliary Building were identified that have design ratings indicating that they could possibly be classified as high energy lines. The pipes were verified to have not been considered in the current HELB analyses. This condition affects pressure transmitters ALPT0037, 38, & 39 which are not qualified for operation in a harsh environment. These pressure transmitters provide the Auxiliary Feedwater Pump (AFW) Suction Transfer signal on low suction pressure from the non safety Condensate Storage Tank to the Safety Related supply (Essential Service Water). Technical Specification (TS) 3.3.2-6.h bases state: "since these detectors are in an area not affected by HELBs or high radiation, they will not experience any adverse environmental conditions and the Trip Setpoint reflects only steady state instrument uncertainties." Based upon the above bases, with the identified aux steam lines in service, the pressure transmitter's operability could not be assured. This represented an unanalyzed condition and had the potential to affect equipment used for accident mitigation. TS 3.0.3 was entered at time 2354 (CST) on 3/23/2011. At 0009 (CST) on 3/24/2011, Aux Steam valves FBV0158, FBV0I48, FAV0002, and FAV0003 were isolated, removing the HELB concern (TS 3.0.3 was exited at this time). These are the active feed (isolation valves) to the lines passing through the Aux Building Rooms 1206/1207. The licensee notified the NRC Resident Inspector.
On March 23, 2011, event notification EN 46693 documented that a harsh environment from a postulated High Energy Line Break (HELB) could affect pressure transmitter ALPT0037, 38 and 39. These pressure transmitters provide the Auxiliary Feedwater Pump suction transfer signal on low suction pressure from the Condensate Storage Tank to the safety-related water supply (Essential Service Water). This break was postulated to occur on auxiliary steam lines in Auxiliary Building rooms 1206 And 1207. This condition was initially reported both as an unanalyzed condition that significantly degraded plant safety and as a condition that could have prevented fulfillment of a safety function. When EN 46693 was reported, it was assumed that breaks were required to be postulated at any intermediate fitting, welded attachment, or valve on the subject auxiliary steam lines. Subsequent analysis shows that these sections of auxiliary steam piping are able to withstand safe shutdown earthquake (SSE) loadings and rupture loadings. For piping of this qualification, breaks at all intermediate fittings, welded attachments, and valves do not need to be postulated. Instead, line breaks are only required to be assumed at the terminal ends of the lines and at the locations specified for ASME Class 2 and 3 piping. None of these postulated break locations are located inside rooms 1206 and 1207. Analysis has been performed on these auxiliary steam lines for the remaining break locations that are required to be postulated. This analysis demonstrates reasonable assurance that safety related equipment, including pressure transmitters ALPT0037, 38 and 39, would have performed their safety functions following a postulated break of these auxiliary steam lines. Therefore, this condition does not meet the reporting requirements for an unanalyzed condition that significantly degraded plant safety or a condition that could have prevented fulfillment of a safety function. Event notification 46693 is hereby retracted. The NRC Resident Inspectors have been notified. Notified the R4DO (Shannon). | Service water Auxiliary Feedwater | |
| ENS 43240 | 14 March 2007 19:35:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Two Out of Three Auxiliary Feedwater Pumps Out of Service | A pinhole leak was discovered on B train Essential Service Water (ESW) system piping while preparing the pipe surface for non-destructive examination. Control room personnel were notified of the leak at 1435. B ESW was immediately declared inoperable. At the time of control room notification, surveillance testing on the Turbine Driven Auxiliary Feedwater Pump (TDAFP) was in progress. This surveillance testing made the TDAFP inoperable and non-functional. The surveillance activities were terminated and the TDAFP was returned to operable status at 1438. B ESW is the safety related water source for B train of auxiliary feedwater (AFW). For the three minute period between notification of the pinhole leak until the TDAFP was restored to operable status, there were two auxiliary feed pumps inoperable. This met the conditions for entry into T/S LCO Action 3.7.5.D which requires a plant shut down to Hot Standby within 6 hours. This action was exited when the TDAFP surveillance testing was terminated. Additionally, with 2 of 3 auxiliary feedwater pumps non-functional for 3 minutes, there was a condition which could have prevented fulfillment of a safety function for those 3 minutes. The licensee notified the NRC Resident Inspector.
The licensee provided the following information as the basis for retracting this report: On March 14, 2007, (Event Number 43240) Callaway Plant reported a condition that, at the time, was believed to be a condition which could have prevented fulfillment of a safety function. At that time the 'A' motor driven auxiliary feedwater pump (MDAPP) was operable, the turbine driven auxiliary feedwater pump was not functional because of a surveillance test in progress. The 'B' MDAFP was presumed to be non-functional because of a pinhole leak in the 'B' train essential service water (ESW) system piping. 'B' ESW is the safety related water source for the 'B' MDAFP. Subsequent inspection, non-destructive examination, analysis and evaluation of the 'B' train ESW piping determined that the structural integrity of the pipe was retained. 'B' ESW pump was able to provide the required flow to the train. 'B' train ESW was functional with the pinhole leak. With the 'B' ESW train functional, the 'B' train of auxiliary feedwater had its emergency water source. The auxiliary feedwater system would have been able to fulfill its safety function. This event is not reportable per 10CFR50.72(b)(3)(v). The licensee will inform the NRC Resident Inspector. Notified R4DO (O'Keefe). | Service water Auxiliary Feedwater | |
| ENS 42454 | 30 March 2006 22:50:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Inadequate Operator Response Time for Component Cooling Water System Realignment During a Large Break Loca | At 1650 on March 30, 2006, a concern was identified where the operators in the training simulator could not complete realignment of the component cooling water (CCW) flow to the residual heat removal (RHR) heat exchanger in a timely manner under certain accident scenarios. This could result in exceeding the maximum design temperature of the CCW system. In addition, assumptions made in the containment pressure and temperature analysis following a large break loss of coolant accident (LOCA) are non-conservative with respect to when CCW flow to the RHR heat exchangers is manually established in accordance with emergency operating procedures. Callaway plant FSAR indicates CCW system flow is manually aligned to the RHR heat exchangers prior to the recirculation phase of emergency core cooling system (ECCS). If the automatic transfer of the RHR pumps to cold leg recirculation, which happens at the Lo-Lo-1 level of the refueling water storage tank (RWST), occurs before CCW flow has been manually aligned to the RHR heat exchanger, containment sump water at temperatures up to 270F can be circulated through the RHR heat exchanger without CCW flow on the other side of the heat exchanger. The CCW side of the heat exchanger would contain stagnant water. This water can heat up quickly with no established flow and exceed the design rated temperature of the system. Recent simulator scenarios of large break LOCAs have shown that the CCW alignment is not reached before the Lo-Lo-1 RWST alarm level is reached. The CCW alignment is completed as part of procedure ES-1.3, Transfer to Cold Leg Recirculation. A review of two large break LOCA scenarios completed on 3-20-06 show that it takes between 1:00 and 1:30 minutes to initiate the step to align CCW to the RHR heat exchangers and takes between 3:00 and 4:30 minutes to complete the alignment. In addition to CCW system temperature concerns, an assumption that CCW flow is established to the RHR heat exchanger prior to reaching the Lo-Lo-1 level in the RWST is made in the containment temperature and pressure response analyses. As a result, a failure to establish CCW flow to the RHR switchover would result in an adverse impact on the inputs used in the Licensing Bases Containment Analysis. However, preliminary sensitivity runs using containment analyses codes indicate that post-peak temperature and pressure are not significantly affected by this issue. Actions taken: 1650 Declared both trains of CCW inoperable. Declared both trains of ECCS inoperable and entered Technical Specification 3.0.3 1710 Both trains of CCW aligned with flow through the RHR heat exchangers 1711 Exited Technical Specification 3.0.3 The licensee notified the NRC Resident Inspector.
The purpose of this notification is to retract a previous notification made on 3/30/06 (EN# 42454). That report was made per 50.72(b)(3)(v)(D) - Accident Mitigation. An engineering evaluation has determined the RHR and CCW systems would have fulfilled their safety functions had they been necessary to respond to an event. Since the safety functions would have been performed there are no applicable reporting criteria under 50.72 or 50.73 and Event Notification 42454 is retracted. The NRC Resident Inspector will be notified. Notified R4DO (Shaffer). | Residual Heat Removal Emergency Core Cooling System | |
| ENS 42257 | 13 January 2006 14:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Pressurizer Porv Stroke Times Exceed Analysis Assumptions | The analyses for the Callaway Cold Overpressure Mitigating System (COMS) assumes an opening time and delay time for the pressurizer Power Operated Relief Valves (PORV)s. Evaluations performed by Callaway engineering personnel have determined that PORV stroke times measured during surveillance testing do not account for all of the delay times credited in the Design Bases COMS Analyses of Record. Further reviews determined the allowed delay times could not be met by the control loop. This results in potentially non-conservative PORV lifts settings relative to those specified in the PTLR for satisfying Technical Specification (TS) 3.4.1.2.a and 3.4.12.c. At the time the determination was made, the plant was in Mode 1. Technical Specification 3.4.12, COMS, is applicable in Mode 4 with RCS temperature less than or equal to 275 F, Mode 5 and Mode 6 with the head on the reactor vessel. The TS allows other methods of providing cold overpressure protection (ex. Residual Heat Removal system suction relief valves). Remedial actions have been initiated to ensure the COMS TS function is maintained when required, pending completion of corrective actions. The licensee notified the NRC Resident Inspector. | Residual Heat Removal | |
| ENS 41877 | 28 July 2005 02:25:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Two Trains of Control Room Emergency Ventilation Systems (Crevs) Inoperable and Not Restored within 24 Hours | At 2125 on 7/27/05, Door 33021 (B Engineering Safety Feature Switchgear to B Emergency Diesel Generator) (B ESF Switchgear to B EDG) was found not to be latched. Reviewing history; the door was first discovered not to latch at 1215 on 7/26/05, by Security. The Control Room was notified at 2125 on 7/27/05 by an Equipment Operator, who found the door unlatched. Door was subsequently latched closed at 2155 on 7/27/05. Due to this door not being able to be verified latched, T/S LCO 3.7.10.B should have been entered at 1215 on 7/26/05. This renders 2 trains of Control Room Emergency Ventilation Systems (CREVS) inoperable, and if not restored within 24 hours, a plant shutdown is required; being in Mode 3 (Hot Standby) in 6 hours and Mode 5 (Cold Shutdown) in 36 hours. The plant should have been in Mode 3 at 1815 on 7/27/05. This time was not met. As stated previously, the door was verified to be latched at 2155 on 7/27/05. A plant shutdown is not being made due to the LCO 3.7.10.B being satisfied at 2155 on 7/27/05. Door 33021 (B ESF Switchgear to B Emergency Diesel Generator) was repaired at 0022 on 7/28/05. This issue has been entered in the licensee corrective action program. The NRC Resident Inspector was notified of this event by the licensee.
Upon further review, it was concluded that this event is not reportable. The design functions of this door are pressure boundary and fire protection. Based upon the following criteria, this event was determined to be not reportable: 1) Pressure boundary: Actual duration of door inoperability did not result in a violation of Control Room Emergency Ventilation System Technical Specification Action completion time limits. 2) Fire protection: Only one fire suppression system was impacted and the inoperability of a fire protection suppression system for a single area is not reportable with regards to the Fire Protection Program. The loss of one fire suppression system was bounded by Callaway licensing basis. The licensee will notify the NRC Resident Inspector. Notified R4 DO (Powers) | Emergency Diesel Generator Control Room Emergency Ventilation |