Semantic search
| Entered date | Event description | |
|---|---|---|
| ENS 53610 | 18 September 2018 08:28:00 | At 1610 (EDT) on September 17, 2018, Southern Nuclear Operating Company (SNC) determined a contractor supervisor confirmed positive for drugs during a Fitness-for-Duty (FFD) test. The employee's unescorted access to the plant has been suspended. The NRC Resident Inspector has been notified. |
| ENS 53599 | 12 September 2018 14:14:00 | At 1115 EDT on September 12, 2018, Southern Nuclear Operating Company (SNC) determined a contractor supervisor confirmed positive for drugs during a Fitness-for-Duty (FFD) test. The employee's unescorted access to the plant has been suspended. The Resident Inspector has been notified. |
| ENS 53528 | 27 July 2018 13:41:00 | A non-licensed contractor supervisor had a confirmed positive for alcohol during a random fitness-for-duty (FFD) test. The employee's unescorted access to the plant has been suspended. The licensee notified the NRC Resident Inspector. |
| ENS 52269 | 28 September 2016 14:57:00 | The following was excerpted from an email received from WECTEC LLC: Nature of the defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply. The two flanges identified with deviations on Passive Core Cooling System pipe spools for the Vogtle Unit 3 AP1000r project had incorrect raised-face dimensions. This appears to have been caused by the two flanges being transposed due to an inadvertent fabrication error that occurred at the pipe spool supplier's facilities (CB&I Laurens). The error was subsequently discovered after delivery to the fabrication facility (Aecon Industrial). This error resulted in conditions where the two flanged connections would not have met the design configuration. If the flanged connections had been assembled in the delivered configuration, it is not known if system integrity and operability would have been maintained during operation. The incorrect configuration could have also led to subsequent failure after installation and operation. Hydrostatic testing of these connections is required, but had not yet been performed because the condition was discovered prior to the assembly and testing of these portions of the system. The condition is being corrected prior to the performance of that hydrostatic testing, therefore it is not known if the flanges in the incorrect configuration would have been able to pass hydrostatic testing. Due to the possibility that system integrity and operability could have been impacted by the use of the incorrect flanges, it has been conservatively concluded that this condition should be reported under 10 CFR Part 21. This conservative conclusion is based on the possibility that the Passive Core Cooling System could have been adversely impacted by the identified deviations, if the deviations had been left uncorrected. The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action. . . . The flange configuration was corrected and the Q223 Mechanical Module was delivered to the Vogtle Unit 3 site on September 23, 2016. A corrective action report has been entered into the Westinghouse/WECTEC system to further evaluate the circumstances that led to the identified deviations.
WECTEC LLC determined that additional pipe spools with incorrect flange configurations were fabricated for V.C. Summer Unit 3 and Vogtle Unit 4. None of the pipe spools were installed in either of the facilities. Corrective actions have been taken to prevent re-occurrence. Notified R2DO (Ehrhardt) and Part 21 group via email. |
| ENS 51704 | 1 February 2016 12:40:00 | The following information is summarized from notification received by WECTEC, LLC via email: The basic components being supplied are piping spools to be used for various safety-related systems for the V. C. Summer and Vogtle AP1000r projects. The pipe spools are classified as ASME Boiler and Pressure Vessel Code Section III. The specific conditions identified on various piping spools fall into four general categories: (1) Dimensional Length Deviations: These conditions include center-to-end and center-to-center lengths of straight pipe segments that exceed the minimum or maximum tolerances. (2) Bend Angle Deviations: These conditions include spool bend angles that exceed the minimum or maximum tolerances. (3) Minimum Wall Thickness Deviations: These conditions include pipe wall thickness below the applicable tolerance. These conditions typically have been found at some bends in the spools and at some ends of the pipe. (4) Other Conditions: These conditions include weld related attributes, potential carbon contamination, and tools marks. Uninstalled pipe spools are in storage and will be inspected, repaired or replaced prior to installation. Installed products will be identified and issues resolved as appropriate. Name and address of the individual or individuals informing the Commission: David Durham President WECTEC LLC 128 S. Tryon St., Suite 1000 Charlotte, NC 28202 |
| ENS 51706 | 3 February 2016 13:29:00 | The following information is summarized from notification received by WECTEC LLC via email: (This report) provides a report in accordance with 10 CFR 21.21 pertaining to deviations of piping penetration sleeves for Vogtle Unit 4 and V. C. Summer Unit 3 AP1000r projects. During inspection of the listed piping penetration sleeves, corrosion was found. It has been conservatively determined that the penetrations could have failed if they were used in the corroded condition. Therefore, it is conservatively concluded that if left uncorrected this condition could have caused a substantial safety hazard for the V. C. Summer and Vogtle AP1000r projects. Name and address of the individual or individuals informing the Commission. Deborah A. Gustafson Vice President of Engineering WECTEC LLC 128 S. Tryon St., Suite 1000 Charlotte, NC 28202 |
| ENS 51507 | 29 October 2015 15:33:00 | The following information was excerpted from an email received from CB&I: This letter provides a report in accordance with 10 CFR 21.21 pertaining to the identification of a defect associated with a structural module piping penetration for the Vogtle Unit 3 AP1000 project. The defect is associated with a Pressurizer instrumentation piping penetration in structural module CA01 for Vogtle Unit 3. This condition was previously identified by interim report letters dated July 1, 2015, and September 2, 2015. The discovery date of these deviations is based on the date of the associated CB&I Power Inspection Report (IR). That IR was initiated on May 5, 2015. Interim Part 21 reports dated July 1, 2015, and September 2, 2015, were submitted to the NRC that state the evaluation of this condition was expected to be completed by October 30, 2015. The interim reports were identified as Accession Nos. ML15254A043 and ML15201A130, and Log Nos. 2015-50-01 and 2015-50-00, on the NRC 'Part 21 Reports' website. Name and address of the individual or individuals informing the Commission: Don DePierro CB&I Power 128 S. Tryon St., Suite 1000 Charlotte, NC 28202 |
| ENS 50734 | 14 January 2015 14:10:00 | The following report was received via email: This report is being provided in accordance with 10 CFR 21.21. (i) Name and address of the individual or individuals informing the Commission. Michael Hickey CB&I Nuclear 128 S. Tryon St., Suite 1000 Charlotte, NC 28202 (ii) Identification of the facility, the activity, or the basic component supplied for such facility or such activity within the United States which fails to comply or contains a defect. The basic components being supplied are pipe supports to be used inside the containment for the V. C. Summer and Vogtle AP1000 nuclear projects. The pipe supports are classified as ASME B31.1 Code and Quality Assurance (QA) Categories II and III. These supports are associated with various non-safety-related portions of several systems inside containment, including Component Cooling, Passive Core Cooling, Spent Fuel Pool Cooling, Waste Liquid, and others. The material being procured was not basic component materials and 10 CFR Part 21 was not applicable. The basic component aspect became inadvertently introduced based on the decision to use a coating that was based on inorganic zinc (IOZ) in lieu of the appropriate coating, which is a Self-Priming High Solids Epoxy (SPHSE). Use of the IOZ coating required the application to be performed as a safety-related application. Due to misinterpretation of the design specification requirements, the wrong safety-class was invoked and the wrong coating material was selected. The piping supports are not impacted by this use of the IOZ coating and would have not been impacted for meeting the pipe support design function. The use of the incorrect coatings, with the incorrect safety classification, could have impacted the ability of the required systems to perform the long-term cooling function, which is considered a safety-related functional impact. The approximate number of supports that are impacted for each unit is provided as follows: 952 Vogtle Unit 3, 275 Vogtle Unit 4, 967 V. C. Summer Unit 2, and 625 V. C. Summer Unit 3. (iii) Identification of the firm constructing the facility or supplying the basic component which fails to comply or contains a defect. The affected piping supports are being supplied by LISEGA Inc. USA, 370 East Dumplin Valley Rd., Kodak, TN 37764. Note that LISEGA supplied the pipe supports with coating material as specified in the procurement documents. Subsequent review has determined that the procurement documents specified incorrect information for many of the supports, which should have been coated with a different material. The procurement documentation was provided to the supplier by CB&I Power, 128 South Tryon Street Charlotte, NC 28202. (iv) Nature of the defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply. The use of the incorrect coating inside containment impacts debris generation and long-term cooling analyses performed for the AP1000 design. It is estimated that the amount of unqualified IOZ coating that could have been added to the containment would have eventually caused impairment of the long-term cooling function during events that require that capability. Therefore, if left uncorrected this condition could have caused a substantial safety hazard for the V. C. Summer and Vogtle AP1000 nuclear projects. (v) The date on which the information of such defect or failure to comply was obtained. The discovery date of these deviations is based on the date of the associated CB&I Power Corrective Action Report (CAR). That CAR was initiated on March 10, 2014. Interim Part 21 reports dated October 15, 2014, and December 11, 2014, were submitted to the NRC. (vi) In the case of a basic component which contains a defect or fails to comply, the number and location of these components in use at, supplied for, being supplied for, or may be supplied for, manufactured, or being manufactured for one or more facilities or activities subject to the regulations in this part. The impacted materials are pipe supports with incorrect coatings intended to be used inside containment for the V. C. Summer and Vogtle AP1000 nuclear projects. The approximate number of supports that are impacted for each unit is provided as follows: 952 Vogtle Unit 3, 275 Vogtle Unit 4, 967 V. C. Summer Unit 2, and 625 V. C. Summer Unit 3. (vii) The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action. The nonconforming pipe supports were initially placed into a hold status and are being corrected. A corrective action report (CAR 2014-2574) has been entered in the CB&I Power Corrective Action Program that describes the circumstances that led to the identification of this potential substantial safety hazard. That CAR is identified as a Level 1, significant condition adverse to quality, and a root cause analysis of the condition is required by CB&I Power Corrective Action Program. The actions necessary to correct the identified conditions and the causes for these conditions will be established and tracked to completion under the CB&I Power Corrective Action Program. (viii) Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or will be given to purchasers or licensees. The condition was discovered by CB&I Power prior to installation of the affected components and the components are being corrected. Therefore, there is no additional action or advice needed for the licensees at this time. The condition has also been evaluated by CB&I Power for potential 10 CFR 50.55(e) reporting by the affected combined operating license holders. CB&I Power has recommended to the licensees that this condition is reportable under 10 CFR 50.55(e). (ix) In the case of an early site permit, the entities to whom an early site permit was transferred. Not applicable. This condition was previously identified by interim report letters dated October 15, 2014 (ML14296A427 - Log No. 2014-76-00), and December 11, 2014. If you have any questions pertaining to this information, please contact Curtis Castell, Licensing Manager, at 980-321-8314. |
| ENS 46045 | 24 June 2010 21:34:00 | At 14:34 hours on June 24, 2010, with the unit in MODE 5, Cold Shutdown, with approximately 50% pressurizer level, during Refueling Outage 26, Instrument Buses 3 and 8 unexpectedly de-energized during performance of testing in accordance with procedure OST-163, 'Safety Injection Test and Emergency Diesel Generator Auto Start on Loss of Power and Safety Injection.' The loss of Instrument Buses 3 and 8 occurred during the loss of power and Safety Injection testing of the 'A' Train. Instrument Buses 3 and 8 are normally powered from Inverter 'B' which is normally supplied by the Train 'B' DC Bus. During the test, it was noted that the power supply to Instrument Buses 3 and 8 had tripped. The cause of the failure of Inverter 'B' is not currently known. The failure of inverter 'B' caused the closure of the Residual Heat Removal (RHR) Heat Exchanger discharge valve (HCV-758) and the RHR Heat Exchanger bypass valve (FCV-605). Both trains of RHR continued to operate and reactor coolant system temperature remained in the range of approximately 93 to 96 degrees Fahrenheit. Abnormal Operating Procedure AOP-020, 'Loss of Residual Heat Removal (Shutdown Cooling)' was entered. Power was restored to Instrument Buses 3 and 8 by use of the alternate power supply at 14:49 hours. Normal configuration of the RHR system was restored and AOP-020 was exited at 14:51 hours. Currently Instrument Buses 3 and 8 are being powered from the alternate power supply which causes the associated 'B' EDG to be inoperable due to the inoperability of the automatic load sequencer that starts the associated Service Water and Component Cooling Water pumps. The 'A' EDG is inoperable due to the need to complete required post-maintenance testing. Therefore, both EDGs are currently inoperable. Both EDGs are currently considered available and are aligned for automatic starting. Both EDGs would be expected to automatically supply their respective buses if a loss of offsite power were to occur. Manual action would be required to start the required loads on the 'B' Train due to the current alignment of the Instrument Buses 3 and 8 on the alternate power supply. It is expected that the 'B' EDG will be restored to operable status when Inverter 'B' is restored to operable status and realigned to supply Instrument Buses 3 and 8. The Technical Specifications (TS) Action Statement currently in effect for loss of Inverter 'B' (TS 3.8.8 Condition A) requires initiation of action to restore AC instrument bus sources to OPERABLE status immediately. The actions to restore Inverter 'B' were initiated immediately and are continuing. This report is being made in accordance with 10 CFR 50.72(b)(3)(v)(D), for any event or condition that at the time of discovery could have prevented the fulfillment of structures or systems that are needed to mitigate the consequences of an accident. The licensee has notified the NRC Resident Inspector.
This is a follow-up notification to Event Notification EN #46045 regarding Instrument Buses 3 and 8 that unexpectedly de-energized during performance of testing in accordance with procedure OST-163, 'Safety Injection Test and Emergency Diesel Generator Auto Start on Loss of Power and Safety Injection.' Power was restored to Instrument Buses 3 and 8 by use of the normal power supply at 08:03 hours on June 27, 2010. Inverter 'B' has been realigned to supply Instrument Buses 3 and 8. The restoration of inverter 'B' has returned the associated 'B' EDG to operable status at 11:09 hours with the return of the automatic load sequencer that starts the associated Service Water and Component Cooling Water pumps. The 'A' EDG continues to be inoperable due to the need to complete required post-maintenance testing. Currently 'A' EDG is considered available and aligned for automatic starting. The Technical Specifications (TS) Action Statement for loss of Inverter 'B' (TS 3.8.8 Condition A) that requires initiation of action to restore AC instrument bus sources to OPERABLE status immediately was exited at 11:09 hours on June 27, 2010. TS Action Statement for loss of 'B' EDG (TS 3.8.2 Condition B) that requires initiation of action to restore required DG to OPERABLE status immediately was also exited at 11:09 hours on June 27, 2010. The licensee is still investigating the cause of the failure. The licensee notified the NRC Resident Inspector. Notified the R2DO (Shaeffer). |
| ENS 45799 | 28 March 2010 22:47:00 | At approximately 1852 hours Eastern Daylight Time (EDT), on March 28, 2010, with the unit operating at approximately 99.5% power, the H. B. Robinson Steam Electric Plant, Unit No. 2, reactor protection system actuated resulting in an automatic trip of the reactor. At about the time of the reactor trip, there was indication of a loss of the power to the Train 'B' emergency bus. The Train 'B' Emergency Diesel Generator started and provided power to the Train 'B' emergency bus. The Reactor Coolant System pressure response after the reactor trip resulted in the actuation of the Safety Injection System. The reduction in Reactor Coolant System pressure allowed the safety injection system to provide flow to the reactor coolant system, although, there was no indication of conditions that would require the safety injection system to provide flow to the reactor coolant system. Specifically, diagnosis of the event determined that a loss of coolant event, steam generator tube rupture, or secondary system break were not occurring. The safety systems that actuated for this event included the Reactor Protection System, the Safety Injection System, both Emergency Diesel Generators started and the 'B' Emergency Diesel Generator provided power to the 'B' Train Emergency Bus, the Auxiliary Feedwater System actuated, and the three main steam isolation valves closed. The Steam Generator Power Operated Relief Valves are being used for decay heat removal because of the closure of the main steam isolation valves. The Startup Transformer is energized and providing power to the 'A' Train Emergency Bus. Investigation of the cause of the reactor trip and associated emergency system actuations is in progress. At the time of the event, it was noted that there was evidence of a fire condition at 4KV bus number 5 located in the Turbine Building. It is currently believed that this is the cause of the transient condition that resulted in the reactor trip and other emergency system actuations. The reactor is currently being maintained in MODE 3, Hot Standby, conditions. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(A) for ECCS discharge into the RCS and (B) for RPS actuation, and 10 CFR 50.72(b)(3)(iv)(A) for specified system actuation as described. Additionally, based on an inquiry from the local news media reporter, information regarding this event was discussed with the local media. This contact with the local news media is being reported in accordance with 10 CFR 50.72(b)(2)(xi). All control rods fully inserted on the trip. Volume Control Tank (VCT) level was not dropping consistent with no reactor coolant system leakage. Pressurizer level and pressure drop was indicative of shrink associated with the Reactor Coolant Pump stoppage and cooldown in one loop. The licensee observed no indication of reactor coolant system leakage from the containment sump water level monitors. The licensee informed the NRC Resident Inspector.
The original Non-Emergency classification was reclassified as an ALERT based on the following: A fire on 4kV Bus 5 affected 4kV Bus 4 which caused a loss of 'B' Reactor Coolant Pump which caused a Reactor trip and Turbine Trip. That fire was extinguished without a required event declaration, but a subsequent fire on 4kV Bus 4 required declaration of an ALERT at 2300 (EDT) based on the fire affecting the safety-related 'A' and 'B' DC Buses. The fire was out at 2301 (EDT). The primary systems are available and control of the Reactor Coolant System has been maintained. The Safety Injection termination Emergency Operating Procedure has been performed and the general procedure for post-trip stabilization is being performed. The licensee informed state/local agencies and the NRC Resident Inspector. Notified R2DO(Guthrie), IRD(Gott), EO(Blount), ET(Leeds), and R2RA(Reyes).
Event Termination Notice for Event 45799. ALERT declaration is no longer required. Event Termination Notice: Current plant Conditions are stable and the conditions that required declaration of the ALERT are no longer present. The fire causing the ALERT classification was extinguished at 2301 (EDT). There was no explosion or steam line break. The last safety-related DC bus ground was identified and cleared at 0001 (EDT). The licensee informed state/local agencies and the NRC Resident Inspector. Notified R2DO(Guthrie), IRD(Gott), EO(Blount), DHS(Moore), FEMA(Casto), DOE(Moorone), USDA(Hovey), and HHS(Nunn). |
| ENS 43450 | 27 June 2007 12:56:00 | This telephone notification to report an invalid actuation is provided in accordance with 10 CFR 50.73(a)(1), which states, 'In the case of an invalid actuation reported under Sec. 50.73(a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER.' The specific reporting requirement in 10 CFR 50.73(a)(2)(iv)(A), states, 'Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B).' For this report, the affected system was the emergency AC electrical power systems as listed in 10 CFR 50.73(a)(2)(iv)(B)(8). Specifically, the Train A emergency diesel generator (EDG) automatically started due to an invalid system actuation. On May 1, 2007, at approximately 2339 hours (EDT), with H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, in MODE 5 with reactor coolant system temperature at approximately 84 degrees F, during activities associated with surveillance testing as part of Refueling Outage 24, the Train A EDG automatically started from the standby condition. The Train A EDG did not automatically connect to the associated emergency bus (E-1), because no E-1 bus undervoltage (UV) signal was present. The automatic start was caused by the inadvertent grounding of a test connection lead that was being used to monitor the Train A EDG during testing. The grounding of the test lead resulted in a drop in voltage on the starting circuitry that caused the actuation of the 4A diesel start relay. The Train A EDG successfully started and achieved the required speed and voltage. As previously stated, the EDG did not load because no UV signal was present on the associated E-1 bus. No failures or abnormalities were noted. The Train B EDG remained operable during this event and the Train A EDG was able to provide power to E-1 emergency bus, if necessary. The invalid EDG start event was entered into the corrective action program for HBRSEP, Unit No. 2. The investigation of this event has been completed. The results of the investigation determined that the EDG start was caused by improper isolation of a safety significant power source from the field installed equipment. The Plant Nuclear Safety Committee has reviewed the investigation associated with this reportable event and concurred with the investigation. Corrective actions for this event are in progress and are being tracked via the corrective action program." The licensee informed the NRC Resident Inspector. |
| ENS 42084 | 27 October 2005 09:39:00 | This telephone notification to report an invalid actuation is provided in accordance with 10 CFR 50.73(a)(1), which states, 'In the case of an invalid actuation reported under Sec. 50.73(a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER.' The specific reporting requirement in 10 CFR 50.73(a)(2)(iv)(A), states, 'Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B).' For this report, the affected system was the emergency AC electrical power systems as listed in 10 CFR 50.73(a)(2)(iv)(B)(8). Specifically, the Train A emergency diesel generator (EDG) automatically started due to an invalid system actuation. On August 31, 2005, at approximately 1805 hours (EDT), with H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, operating in MODE 1 at approximately 100% power, the Train A EDG automatically started from the standby condition. The EDG did not automatically connect to the associated emergency bus (E-1), because no E-1 bus undervoltage (UV) signal was present. The automatic start was caused by a failed solenoid-operated valve (SOV) in the air start system for the EDG (valve number DA-23A). The SOV failed-open, which is the designed failure-mode condition for this valve. This admitted air to the EDG air-start distributor, which started the Train A EDG. No additional failures or abnormalities were noted. The Train A EDG successfully started and achieved the required speed and voltage, but as stated previously, the EDG did not load because no UV signal was present on the associated bus. The Train B EDG remained operable during this event. Also, the EDGs each have two starting SOVs that are in parallel in the starting air system. A start signal or a loss of power to either valve is sufficient to start the associated EDG. The failed SOV was replaced. The Train A EDG was tested after replacement of the SOV and returned to service at 1016 hours on September 1, 2005. The invalid EDG start was entered into the corrective action program for HBRSEP, Unit No. 2. The investigation of this event has been completed. The results of the investigation determined that the EDG start was caused by a short-circuit failure of the coil for the SOV. Additional corrective actions for this condition, beyond replacement of the failed SOV, are in progress and are being tracked via the corrective action program. The licensee notified the NRC Resident Inspector. |
| ENS 40617 | 28 March 2004 15:00:00 | At approximately 1343 hours on March 27, 2004, during a routine hand - rotation check of the "C" high pressure safety injection pump (HPSI), it was discovered the pump shaft was exhibiting some binding. At the time, the "C" HPSI pump was in service as the "B" Train HPSI pump. The "C" HPSI pump was declared inoperable and Condition A of LCO 3.5.2 was entered. This LCO condition requires restoration of the inoperable safety injection train within 72 hours. At that time, the "B" HPSI pump was in service as the Train "A" HPSI pump. The "A" HPSI pump was out of service due to a previously discovered condition of minor leakage observed near two of the casing bolts. The H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, HPSI system has three safety injection pumps. The "A" HPSI pump is the normal Train "A" pump, the "C" pump is the normal Train "B" pump, and the "B" pump is capable of serving as either the Train "A" or the Train "B" pump in the place of the "A" or "C" HPSI pump. At the time of discovery of the binding in the "C" HPSI pump, the "B" HPSI pump was in service as the Train "A" HPSI pump. It was determined that the "A" pump could be restored to operable status and placed back in service to restore two Trains of HPSI. In order to do so, the "B" HPSI pump was removed from service as the Train "A" pump and placed in service as the Train "B" pump. Therefore, at 1026 hours on March 28, 2004, for approximately 25 minutes, during the process of placing the "B" HPSI pump in service on Train "B," which was necessary to allow the "A" HPSI pump to be returned to service, there was no HPSI pump automatically available to provide HPSI to the Reactor Coolant System, if an accident were to occur. Therefore, this condition is reportable in accordance with 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function of systems needed to mitigate the consequences of an accident. An additional 8-hour reporting criterion associated with the plant being in an unanalyzed condition that significantly degrades plant safety 10 CFR 50.72(b)(3)(ii) has also been identified due to the inoperability of the HPSI system. It was known prior to the switching the "B" HPSI pump to Train "B" that this would cause inoperability of both trains of HPSI. It was also known that LCO 3.0.3 would be entered due to this circumstance. This situation was not avoidable, based on the sequence of events. The time in this condition was minimized, the operators were fully aware of plant conditions, and no other system inoperabilities were known of that would have complicated the situation. Both trains of HPSI are operable, although the "C" HPSI pump remains inoperable and out-of-service, pending investigation and repair. The licensee notified the NRC Resident Inspector |
| ENS 40477 | 26 January 2004 03:50:00 | At approximately 22:36 hours EST, on January 25, 2004, the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, Emergency Response Facility Information System (ERFIS) computer system became inoperable due to a loss of the uninterruptible power supply (UPS). The ERFIS computer system provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), Meteorological Data link system, FAX modems for Emergency Preparedness functions, and the Inadequate Core Cooling Monitor (ICCM). The loss of ERFIS requires alternate methods as described in plant procedures to be used for these functions, as necessary. Therefore, it is expected that appropriate assessment of plant conditions, notifications, and communications could still have been made as necessary, if required, during the time that the ERFIS was inoperable. The ERFIS computer system was restored at approximately 01:47 hours EST on January 26, 2004, using a back-up power supply. The cause of the loss of the ERFIS UPS has not yet been determined. Plant personnel are continuing to investigate the loss of the power supply and are in the process of repairing the UPS. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. As previously stated, alternate means remained available to assess plant conditions, make notifications, and accomplish required communications, as necessary. The Licensee notified the NRC Resident Inspector. |