|Entered date||Event description|
|ENS 51874||20 April 2016 22:04:00||Recent Operating Experience at Callaway has shown that the pressure transient experienced in the Essential Service Water (ESW) system during Engineered Safety Feature Actuation System (ESFAS) testing can result in gasket failure on the Control Room Air Conditioners rendering the units nonfunctional. Previously, this pressure transient was considered to be the result of the system alignment used to perform the surveillance test, which is not the same as the system lineup which would occur on an actual Loss of Offsite Power (LOOP) or Safety Injection Signal Design Basis Accident (DBA) event. However, on April 20th, 2016, Callaway received preliminary analysis results that predict the Control Room Air Conditioners would actually experience a greater pressure transient during a DBA than what is currently experienced during ESFAS testing. This condition could result in the Control Room Air Conditioners not being capable of performing their safety function following a DBA event, and challenge Control Room Habitability. Therefore, this condition meets the reporting criteria of 10 CFR 50.72(b)(3)(ii)(B). Based on current conditions (i.e., the plant is not in Power Operation), this condition does not present an immediate safety concern. The analysis of the pressure transient experienced by the ESW system during a postulated DBA event is preliminary and further evaluation of the analysis is ongoing. The NRC Resident Inspector has been notified.|
|ENS 51846||3 April 2016 07:13:00|
At 2302 (CDT) on April 2, 2016, with the plant shutdown, (with) all control rods inserted in the reactor and while attempting to reset the reactor trip breakers to support outage activities (reset of the main turbine), the reactor trip breakers reopened. This was identified to be due to having both trains of Solid State Protection System (SSPS) out of service while in Mode 5. With both trains of SSPS out of service, a condition was met that would cause a reactor trip signal due to having a general warning condition on both trains. Per procedure, the control rods were incapable of withdrawal and fully inserted. Reactor Coolant System boron was 2280 ppm. There were no actuations as a result of the reactor trip breakers opening due to SSPS being removed from service. The licensee will be notifying the NRC Resident Inspector.
At 0713 EDT on April 3, 2016, EN #51846 provided notification of a Reactor Protection System actuation as revealed by the reactor trip breakers opening. Upon further investigation, it has been determined that the system actuated during maintenance activities due to a reactor trip signal caused by both trains of the Solid State Protection System (SSPS) being in test. This signal was not in response to actual plant conditions or parameters satisfying the requirements for initiation of the system and was therefore invalid. As such, the notification made by EN #51846 for a valid actuation of a specified system is hereby retracted. In addition, an editorial change to the first sentence of the original notification description is hereby made. The first sentence is revised to read as follows: At 2303 EDT on April 2, 2016, with the plant shut down and all control rods inserted into the reactor, while attempting to reset the reactor trip breakers to support outage activities (reset of the main turbine), the reactor trip breakers reopened. The NRC Resident Inspector will be notified. Notified the R4DO (Kellar).
|ENS 50650||3 December 2014 05:05:00||An unexpected main turbine trip causing a reactor trip occurred on 12/03/2014 (at 0022 CST) with the plant operating in Mode 1 at 100 percent power. As part of the plant design, an expected, valid actuation of the Auxiliary Feedwater System occurred in response to the reactor trip. As part of the Auxiliary Feedwater actuation, the 'B' Motor Driven Auxiliary Feedwater Pump to 'D' Steam Generator throttle valve did not throttle as expected and was manually isolated. All other systems functioned normally in response to the plant conditions. The plant is currently stable in Mode 3 at 0 percent power. Safety related buses are receiving normal off-site power and the grid is currently stable. The NRC Resident Inspector was notified. All control rods fully inserted on the reactor trip. Steam generator levels are being maintained by the AFW system and decay heat is being removed by the main condenser. No primary or secondary safety relief valves lifted during the transient.|
|ENS 49857||26 February 2014 18:18:00||Contrary to the requirements in 10 CFR 26.137(b), a DHHS (Department of Health and Human Services) certified laboratory returned results for a blind specimen that were inconsistent with what was expected. On 02/25/2014, dilute blind specimens from the same lot # were sent to the three contracted DHHS laboratories. Upon review by the Callaway MRO (Medical Review Officer) at approximately 07:30 (CST) on 2/26/2014, it was discovered that one of the laboratories (Toxicology) reported results of negative. That result was inconsistent with the certification received from the blind provider (ProTox) certifying the specimen as negative and dilute. Later in the day on 2/26/14, the remaining two labs (Quest and CRL) also returned results of negative instead of negative and dilute. 10 CFR 26.719(c)(3), reporting requirements requires that 'If a false negative error occurs on a quality assurance check of validity screening tests, as required in � 26.137(b), the licensee or other entity shall notify the NRC within 24 hours after discovery of the error.' While it initially appears that the blind specimen certification provided by ProTox may be in error, since all three DHHS labs obtained the same testing result, additional investigation is necessary. This report is being made conservatively until the cause can be determined. The licensee informed the NRC Resident Inspector.|
|ENS 47291||25 September 2011 19:48:00|
At 1804 on Sunday, September 25, the Callaway Plant Technical Support Center (TSC) will undergo planned maintenance to replace the building's heating, ventilation, and air conditioning (HVAC) system. This maintenance is currently scheduled to last for approximately five days, at which time the TSC will be restored to service. During this period, the TSC's HVAC system will not be able to provide positive pressure to the TSC, thus rendering it non-functional. If an emergency is declared requiring TSC activation while the TSC is non-functional, TSC emergency response personnel will report to their backup locations in accordance with Callaway Plant emergency planning procedures. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to the planned unavailability of an emergency response facility. The NRC Resident Inspector has been notified.
At 1804 CDT on September 25, 2011, the Callaway Plant Technical Support Center (TSC) underwent planned maintenance to replace the building's heating, ventilation, and air conditioning (HVAC) system. While this maintenance was being performed, the TSC was unable to maintain positive pressure within the building, thus rendering it non-functional. This ENS update is to document that, at 1800 on September 30, 2011, the Callaway Plant TSC was returned to service following successful completion of planned maintenance on the building HVAC system. The NRC Senior Resident Inspector has been notified. Notified R4DO (Werner).
|ENS 46693||24 March 2011 06:02:00|
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).
|ENS 45747||5 March 2010 16:32:00|
Valve FBV0147, Boric Acid Batch Tank Auxiliary Steam Supply Isolation Valve, is credited with being closed in the Callaway FSAR. This eliminated the need to analyze lines FB-081-HBD and FB-082-HBD for high energy line breaks (HELB). However, FBV0147 was found to be kept normally open to allow steam service for the boric acid batching tank. This is contrary to the normal position assumed in the FSAR and HELB analyses. With valve FBV0147 open, the lines must considered high energy lines. The lines are in the auxiliary building and they traverse rooms containing several components including the flow transmitters for Residual Heat Removal (RHR) to train `A' accumulator injection supply header, RHR train `A' and 'B' SIS hot leg recirculation supply header, and several safety related auxiliary feedwater components. These instruments are used to provide indications during post-accident conditions. This configuration is therefore outside of current HELB analysis and potentially represents a condition that significantly degrades plant safety. The condition was identified to Operations at 0740. Valve FBV0147 was closed at 0810. At 1325 CST, the issue was determined to be reportable. The NRC Resident Inspector has been notified.
On 03/05/2010, EN #45747 provided notification that valve FBV0147 was found to be kept normally open to allow steam service for the boric acid hatching tank. This configuration was not consistent with the normal position assumed in the FSAR and HELB analyses. An engineering evaluation was subsequently performed for the auxiliary steam inlet and outlet piping for the boric acid batching tank. This valuation identified four postulated break locations which have all been analyzed. The analyses determined that all safety-related components in the affected rooms would be able to perform their safety functions in the event of a line break. Pipe whip, jet impingement, compartment temperature and over pressurization, flooding, and internal missiles resulting from a postulated line break would not prevent any safety-related equipment from performing its design function. As supported by this evaluation, this condition does not meet the criteria for an unanalyzed condition that significantly degrades plant safety as stated in 10 CFR 50.72(b)(3)(ii)(B). Therefore, the notification made on 03/05/2010 is hereby retracted. The NRC Resident Inspector will be notified. R4DO (Farnholtz) notified.