Semantic search
| Entered date | Event description | |
|---|---|---|
| ENS 49804 | 7 February 2014 08:19:00 | This event is being reported in accordance with 10CFR 50.72(b)(2)(i), 'Initiation of a Shutdown Required by Technical Specifications.' At 2043 hours (EST) on February 06, 2014, the Perry Nuclear Power Plant entered Technical Specification 3.6.1.3 Primary Containment Isolation Valves (PCIVs), action C.1, due to leakage identified during local leak rate testing of the containment penetration for the Containment and Drywell Purge system. Leakage was identified on the outboard containment isolation valve resulting in the plant exceeding the limit for secondary containment bypass leakage. The Containment and Drywell Purge system penetration is normally isolated and remains isolated in accordance with Technical Specifications. Action C.1 requires restoration of the leakage rate within four hours. At 0043 hours on February 7, 2014, the plant entered Technical Specification 3.6.1.3, 'Primary Containment Isolation Valves (PCIVs)', action E as the leakage rate was not restored. Action E requires the plant be in Mode 3 in 12 hours and Mode 4 in 36 hours. At 0600 hours on February 07, 2014, the Perry Nuclear Power Plant initiated a shutdown in accordance with Technical Specification 3.6.1.3, action E. Repairs to restore the penetration leakage to within allowable limits are in progress. The NRC Resident Inspector has been notified.
At 0943 hours (EST) the reactor shutdown to comply with Technical Specification 3.6.1.3 action E was terminated (with the reactor at 42% power). A blind flange was installed downstream of the outboard containment isolation valve. Local leak rate testing of the containment penetration for the Containment and Drywell Purge system verified that leakage was within the limits for secondary containment bypass leakage. The NRC Resident Inspector has been notified. The licensee has commenced increasing reactor power. Notified R3DO (Orlikowski) |
| ENS 49121 | 16 June 2013 02:42:00 | This event is being reported in accordance with 10 CFR 50.72(b)(2)(i) and 50.72(b)(3)(ii)(A). On June 16, 2013 at 0200 EDT, the Perry Nuclear Power Plant commenced a controlled plant shutdown. The shutdown was due to a small leak through the base of a vent line on the 'B' Reactor Recirculation Flow Control Valve. On June 15, 2013 at 2250 EDT, the leak was identified and was subsequently determined to require a plant shutdown in accordance with Technical Specification 3.4.5, Action (C) which requires the plant to be in Mode 3 within 12 hours. The NRC Resident Inspector has been notified." The licensee will also be notifying state and local authorities. The licensee had come down in power to make a drywell entry and investigate drywell leakage indications. Steam was observed to be coming from a vent line that comes off the top of the recirc flow control valve. The licensee was unable to characterize the leak rate other than a small leak. The licensee stated that the steam appeared be coming from a weld location where the vent line comes out of the flow control valve which would classify it as pressure boundary leakage. |
| ENS 47710 | 1 March 2012 05:51:00 | On March 1,2012, at approximately 0224 (EST), a manual Reactor Protection System (RPS) actuation was initiated due to 3 turbine bypass valves going open as a result of an automatic turbine runback signal. At the time of the event, the plant was in Mode 1 at 100% power. All control rods are inserted into the core and the plant is currently stable in Mode 3 (Hot Shutdown) with reactor pressure at approximately 930 psig. No Emergency Core Cooling Systems were required or utilized to respond to the event and there were no other reportable actuations. Reactor coolant level is being maintained in its normal band by the feedwater system and decay heat is being removed by the condenser. The plant is in a normal electrical line-up with all three Emergency Diesel Generators operable and available if needed. The cause of the automatic turbine runback has not been determined and is being investigated. During the transient, Reactor Water Cleanup System (RWCU) tripped. No automatic isolation signal was received. At the time of the event, restoration of a Stator Water Cooling pressure gauge was being performed (following maintenance). The NRC Resident Inspector has been notified. |
| ENS 44279 | 10 June 2008 01:29:00 | The Perry Nuclear Power Plant will be taking the ERDS out of service for scheduled maintenance. From approximately 0200 hours EDT, on June 10, 2008, until 2000 hours EDT, on June 10, 2008, personnel will be performing cleaning and inspection activities on the 120 VAC Emergency Response Information System (ERIS) Computer Power Center. During this planned preventive maintenance, the Integrated Computer System, the Safety Parameter Display System (SPDS), and the automatic mode calculation of the Computer Aided Dose assessment program (CADAP) will incur approximately two periods of unavailability of approximately two hours each. The unavailability periods are necessary to align a temporary power supply and reconnect the 120 VAC ERIS Computer Power Center upon completion of the activities. The dose assessment function will be maintained during the brief out of service time periods by manual input of data into CADAP and, if required, by manual calculation. In the event of an emergency, plant parameter data will be orally transmitted to the facilities through the Status Board Ring Down circuit with back-up by the Plant Branch Exchange, the Off Premise Exchange, and various redundant intra-facility circuits throughout the emergency facilities. The ability to open and maintain on 'open line' using the Emergency Notification System will not be affected and will be the primary means of transferring plant data to the NRC as a contingency until the ERDS can be returned to service during the two periods of unavailability. This event is being reported in accordance with 10 CFR 50.72(b)(3)(xiii), as a condition that results in a major loss of offsite communications capability. A follow-up notification will be made when the activities are completed and the equipment is restored. The Resident Inspector has been notified.
ERDS is still in service and will not be restored to normal until tomorrow 6/11/08. The scheduled maintenance on the normal power supply today was delayed. Notified R3DO(Lipa)
ERDS was on a temporary power supply. ERDS restored to normal power supply at 1650. Notified R3DO (Louden). |
| ENS 44111 | 31 March 2008 19:18:00 | At 1550 hours, a self identification call was made to the State of Ohio Environmental Protection Agency (EPA), offices in Columbus, Ohio to inform them of the recent discovery of non-compliance issues relating to the accumulation, storage, and shipment (i.e., 90 days to ship) of hazardous waste (reference 40 CFR 261, 'Identification and Listing of Hazardous Waste'). Guidance was requested from the state EPA for remediation and reporting of the condition. The Perry Nuclear Power Plant personnel were advised by the Ohio EPA to properly identify, package and ship the waste as soon as possible. The waste consists of three drums of floor grinding waste generated during the resurfacing of the Auxiliary Building floor in 2005, and approximately 100 bags of miscellaneous trash generated during disassembly of plant equipment in 2007. This issue was discovered during activities to prepare the waste for shipment off site. The waste will be shipped per the direction of the Ohio EPA and an out-of-cycle hazardous waste annual report will be made to the Ohio EPA. This report is being made in accordance with 10 CFR 50.72(b)(2)(xi) as an event or situation related to the protection of the environment for which a notification to another government agency has been made. The resident inspector has been notified. The hazardous material contains low level rad waste.
On March 31, 2008, in accordance with 10 CFR 50.72(b)(2)(xi), notification was made for an event or situation related to the protection of the environment for which a notification to another government agency was made. This report was made when a self-identification call was made to the State of Ohio Environmental Protection Agency (EPA) offices in Columbus, Ohio, for the discovery of potential non-compliance issues with 40 CFR 261 requirements for the accumulation, storage, and shipment of hazardous waste. Based on further investigation using a Toxicity Characterization Leachate Procedure analysis of the waste, it was determined that the waste was, in fact, non-hazardous and could be shipped for off-site disposal. The requirements of 40 CFR 261 had been and continue to be met. The licensee has notified the NRC Resident Inspector and the State of Ohio. Notified R3DO ( Pelke). |
| ENS 43860 | 21 December 2007 12:31:00 | Annulus Exhaust Gas Treatment System (AEGTS) 'A' was removed from service (INOPERABLE) to obtain a charcoal sample. AECTS 'B' Train was the OPERABLE Train. At 0825 hours, the charcoal plenum for the 'A' train was opened to obtain a charcoal sample resulting in alarms for AECTS 'B' low flow and Annulus low differential pressure. Based on this indication AEGTS 'B' and secondary containment were INOPERABLE. AEGTS 'B' flow was restored to normal following closure of the charcoal plenum and annulus differential pressure was restored to normal at 0833 hours, restoring secondary containment. AEGTS 'B' was declared operable at 0845 hours. Early indication is that a discharge damper in the AEGTS 'A' train had not operated properly. This condition was determined to be reportable in accordance with 10 CFR 50.72(b)(3)(V)(C) and (D), as condition that could have prevented 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 licensee notified the NRC Resident Inspector. |
| ENS 43363 | 15 May 2007 01:53:00 | An automatic reactor scram occurred due to lowering reactor water level. Digital feedwater tuning activities were in progress at the time of the scram. All systems functioned as designed. The reactor water level has been restored to normal level band. The plant is stable in Hot Shutdown. There were no ECCS injections. At the time of the scram, the feedwater pump was in manual control for feedwater tuning. When water level started going down quickly, the operator was not able to restore sufficient feedwater flow before the level 3 (water level low) actuation. All control rods fully inserted on the scram. No valves repositioned and no safety or relief valves lifted after the scram. Reactor water level is being maintained with the motor feed pump and decay heat is being removed to the main condenser. The plant is in the normal shutdown electrical lineup. Reactor pressure is 509 psi and stable. The licensee notified the NRC Resident Inspector.
This call is to clarify information provided in notification Event Number 43363 made by the Perry Nuclear Power Plant on May 15, 2007. The second paragraph, fourth sentence of the notification states: 'No valves repositioned and no safety or relief valves lifted after the scram.' The intent of the portion of the sentence 'No valves repositioned' was supposed to be in reference to the reporting requirement in 10 CFR 50.72(b)(3)(iv) to report any event or condition that results in valid actuation of general containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs). The event on May 15, 2007, did not meet the 'containment isolation valves in more than one system / MSIVs criterion', however, subsequent review of the event identified that Residual Heat Removal 'B' Heat Exchanger Second Vent to Suppression Pool Containment Isolation Valve 1E12-F073B, closed, as designed, in response to the reactor coolant level 3 (water level low) condition that was present during the event. No additional reporting criteria have been identified and this update has been provided to clarify the earlier statement. The NRC Resident Inspector has been notified." Notified R3 RDO, (Ring), and NRR EO, (Ross Lee). |
| ENS 43049 | 13 December 2006 05:49:00 | A loss of instrument air caused a transient in feed system causing a lowering Hot Surge Tank Level. The Reactor was manually scrammed when Reactor Feed Booster Pumps were cavitating. Control Rod 42-55 did not insert on the initial scram. Control Rod 42-55 did insert when ARI was manually initiated. The loss of instrument air was the result of an air line rupture, and the licensee is steaming to the condenser while maintaining vessel level with feedwater. There was no ECCS injection, and the licensee is investigating the cause of the line rupture. The licensee notified the NRC Resident Inspector.
This update is being made in accordance with 10CFR50.72(c)(2) to immediately report the results of ensuing evaluations or assessments of plant conditions, the effectiveness of response or protective measures taken, and information related to plant behavior that is not understood. On December 13, 2006, at 0549 hours, the Perry Nuclear Power Plant notified the NRC Operations Center (Event Number 43049) of a manual Reactor Protection System actuation associated with a loss of instrument air. In the initial report, it was stated that control rod 42-55 did not insert on the initial scram, but did insert when Alternate Rod Insertion was manually initiated. Subsequently, reactor engineering review of control rod performance during the event determined that the control rod scram time for control rod 4255 was satisfactory and that the problem was with the control rod indication. The plant remains stable at this time. Control rod indication has two channels. Control rod 42-55 displayed one fully inserted indication and one blinking green indication (i.e. not fully inserted). When the system was reset, all rods indicated fully inserted with a "00" position indication. The licensee notified the NRC Resident Inspector. Notified the R3DO (Lara). |
| ENS 42693 | 8 July 2006 09:22:00 | On 6/29/06 at 1630 Neutralization Basin A was pumped to Lake Erie. On 7/7/06 the analysis of suspended solids for this discharge was completed with results outside of the daily limit. Results for the suspended solids for the basin discharge was 142 ppm with a limit of 100 ppm. State of Ohio EPA was notified for this non-compliance. This 4 hour notification is being made due to the notification to the State of Ohio EPA. Notification to the Ohio EPA was made at 0918 on 7/08/06. Specific. Reporting details required by Part 12.A are as follows: Discharge occurred- 6/29/06 at 16:30; Discovered 7/7/06 at 14:38 (analysis reported by First Energy Beta Drive Lab) Approximate amount 18,300 gallons; 7.94 Ph; 142 ppm Total Suspended Solids (backup sample 148 ppm) The licensee notified the NRC Resident Inspector.
This 4 hour notification is being made due to the additional notification to the State of Ohio EPA. This was follow-up information regarding event number 42693 ('Offsite Notification - exceeded limit for suspended solids in discharge'). On 7/10/06 final data analysis for NPDES Monthly Report was obtained and the 6/29/06 exceedance resulted in an exceedance in the monthly average of total suspended solids (TSS). Results for monthly average suspended solids were 44.2 ppm with a limit of 30 ppm. State of Ohio EPA was notified of this non-compliance at 0925 on 7/11/06. The licensee informed the NRC Resident Inspector. Notified R3DO (Steve Orth). |
| ENS 42329 | 10 February 2006 16:30:00 | At 1345, a notification was made to the Ohio EPA. This notification was based on 1) Any unanticipated bypass which exceeds any effluent limitation in the NPDES Permit and 2) Any discharge of water to the storm drains that is not covered by the following excerpt from the permit. The event itself was the pumping of rain water that had sulfuric acid (pH approximately 2.6) in it to a storm drain. Approximately 10 gallons of this mixture was pumped to the drain. The 10 gallons pumped contained approximately 1 milliliter of sulfuric acid. Sample taken at the storm drain out fall indicated normal pH value of 8.03. The licensee notified the NRC Resident Inspector. |
| ENS 42116 | 4 November 2005 03:32:00 | At 0200 on 11/03/05 a clearance was authorized that defeated the DW (Drywell) pressure high and Rx (Reactor) vessel low isolation features to valves in the Nuclear Closed Cooling System and Instrument Air Systems. The required T.S. (Technical Specification) actions after this discovery are that the plant should have been in Mode 3 at 1600 on 11/3/05. The clearance was removed and the circuit restored to operability at 0142 on 11/4/05. The time of discovery for the loss of safety function was 2345 on 11/3/05. The clearance was to perform pre-planned maintenance activities. The licensee plans on entering this incident into their corrective action program and will issue a Condition Report. The licensee informed the NRC Resident Inspector.
An 8-hour notification was made on November 3, 2005, in accordance with 10CFR50.72(b)(3)(v)(D), Accident Mitigation. This report was made when it was discovered that a clearance had unintentionally deenergized a portion of the containment isolation logic. This logic would have prevented a containment isolation valve for the nuclear closed cooling system and a containment isolation valve for the instrument air system from closing on a signal due to high drywell pressure or reactor vessel low level as designed. The condition was determined not to be a loss of containment (leakage) function since each containment penetration also has an inboard containment check valve which is leak tested and is credited for preventing leakage. The check valves in both penetrations were successfully tested to be in conformance with 10 CFR 50 Appendix J Option B criteria in March 2005. The containment isolation (instrumentation) function was also not lost. Only the logic for Group 2A outboard valves was deenergized. This was a small portion of the outboard logic and had no impact on the inboard logic. Since neither the containment function nor the containment isolation function was lost, there was no loss of safety function for an accident mitigation function and ENF 42116 Is retracted. As discussed in ENF 42116, a Technical Specification violation occurred and is reportable per 10CFR50.73(a)(2)(i)(B), as a condition prohibited by Technical Specifications." The licensee will be submitting a written LER for the 10CFR50.73(a)(2)(i)(B) event. The licensee notified the NRC Resident Inspector. Notified the R3DO (Hills). |
| ENS 42072 | 24 October 2005 12:41:00 | This event is being reported as 10CFR50.72(b)(3)(ii)(B), an unanalyzed condition that significantly degrades plant safety. A postulated fire water system line break did not take into consideration the movement of a fire door boundary. Consequently, the break may affect safe shutdown of the plant. The door boundary was moved in May 1999. While assessing a calculation to update fire pump curves (CR 04-00422-04) it was found that the calculation does not appear to have been updated for the tornado depressurization event modifications (DCP 99-05014). Specifically, door DG-112 was moved to the control complex wall from the diesel generator wall. This door movement isolates the rattle space that was previously credited as a relief path for this internal flood. A full break is postulated for this line as the result of a safe shutdown earthquake. As the result of the postulated pipe break, both fire protection pumps are expected to auto start resulting in a break flow rate into the hallway of approximately 6,000 gpm. Standard Perry design practice is to assume a 30 minute duration for the pipe break flow isolation unless justified otherwise. The line in question was isolated at 2027 on 10/21/05. The valves are being maintained closed under administrative controls. The licensee notified the NRC Resident Inspector.
An 8-hour notification was made on October 24, 2005 under 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades plant safety. The report was made due to a postulated break for a non-safety non-seismic fire water pipe that could possibly affect safe shutdown of the plant. An evaluation was completed on October 28, 2005. This evaluation confirmed that the current plant configuration is consistent with the design basis. The evaluation used for the original event notification assumed a full break of the involved piping. Perry design basis for this moderate energy system is a leakage crack. The postulated leakage from the crack in the piping remains within design basis and does not significantly degrade plant safety. Since Perry remains in compliance with design basis and there is no unanalyzed condition that significantly degrades plant safety, there is no reportable condition. Therefore, ENF 42072 is retracted. The licensee notified the NRC Resident Inspector. Notified R3DO (Lipa). |
| ENS 41085 | 2 October 2004 15:22:00 | At 1500 hours the plant commenced reactor shutdown from 100% power for entering T.S. 3.0.3 due to both trains of the Emergency Recirc Vent System being declared inoperable. The reactor will be in mode 2 by 2000 hours, mode 3 by 0200 hours on 10/03 and mode 4 by 0200 hours on 10/04. If the problem is corrected, they will terminate the shutdown. The NRC Resident Inspector was notified. HOO Note: see event 41084
Update to (Event) Notifications 41084 and 41085: At 1840 (EDT) on 10/02/04 it was determined that the apparent slow response times of the Control Room Emergency Recirculation (CRER) dampers was due to a malfunctioning relay in the initiation circuit, not due to failure of the dampers. The LCO actions associated with the CRER system were exited and the actions associated with the initiation instrument were entered. Therefore the plant is no longer required to shutdown per T.S. 3.0.3. The plant shutdown has been terminated. Plant power will be returned to 100%. The licensee has notified the NRC resident Inspector. Notified R3DO (Thomas Kozak).
At 1300 on 10/02/04, results of a surveillance indicated that dampers in both trains of the Control Room Emergency Recirculation System (CRERS) were slower than allowed by Technical Specifications (TS) requiring both trains of the CRERS to be declared inoperable. With both trains inoperable, Technical Specification 3.0.3 was entered which required a plant shutdown. The shutdown was commenced at 1500. This condition was reported as required in Event Notification 41085. Additionally, with both CRERS trains inoperable, this condition was determined to be reportable as a loss of safety function (accident mitigation) and was reported as required in Event Notification 41084. Subsequently, it was determined that the failure was the result of a defective time delay relay in the radiation monitor initiation circuit. Other inputs that would have caused the dampers to reposition in an accident were not impacted. The appropriate TS (3.3.7.1), for the radiation monitor, was entered and TS 3.0.3 was exited. The significant actions required by this TS were to restore the function within 7 days or place the system in emergency recirculation. It did not require entry into an action to shutdown. When this condition was identified, TS 3.0.3 was exited, the shutdown was terminated, and the plant was restored to full power. Since a TS required shutdown was not required, Event Notification 41085 is being retracted. The licensee has notified the NRC Resident Inspector. NRC R3DO(Gardner) has been notified. |
| ENS 41084 | 2 October 2004 15:22:00 | A surveillance test was being performed on the Emergency Recirc vent System and all six dampers on both trains failed to stroke in the required Tech. Spec. times. Therefore, both trains of the Emergency Recirc Vent System were declared inoperable and the plant entered T.S. 3.0.3. The LCO action statement requires the plant to be in mode 2 in 7 hours and mode 3 in the following six hours and mode 4 in the following 24 hours. They are currently troubleshooting the problem. The NRC resident Inspector was notified HOO Note: see event 41085
Update to (Event) Notifications 41084 and 41085: At 1840 (EDT) on 10/02/04 it was determined that the apparent slow response times of the Control Room Emergency Recirculation (CRER) dampers was due to a malfunctioning relay in the initiation circuit, not due to failure of the dampers. The LCO actions associated with the CRER system were exited and the actions associated with the initiation instrument were entered. Therefore the plant is no longer required to shutdown per T.S. 3.0.3. The plant shutdown has been terminated. Plant power will be returned to 100%. The licensee has notified the NRC resident Inspector. Notified R3DO (Thomas Kozak).
At 1300 on 10/02/04, results of a surveillance indicated that dampers in both trains of the Control Room Emergency Recirculation System (CRERS) were slower than allowed by Technical Specifications (TS) requiring both trains of the CRERS to be declared inoperable. With both trains inoperable, Technical Specification 3.0.3 was entered which required a plant shutdown. The shutdown was commenced at 1500. This condition was reported as required in Event Notification 41085. Additionally, with both CRERS trains inoperable, this condition was determined to be reportable as a loss of safety function (accident mitigation) and was reported as required in Event Notification 41084. Subsequently, it was determined that the failure was the result of a defective time delay relay in the radiation monitor initiation circuit. Other inputs that would have caused the dampers to reposition in an accident were not impacted. The appropriate TS (3.3.7.1), for the radiation monitor, was entered and TS 3.0.3 was exited. The significant actions required by this TS were to restore the function within 7 days or place the system in emergency recirculation. It did not require entry into an action to shutdown. When this condition was identified, TS 3.0.3 was exited, the shutdown was terminated, and the plant was restored to full power. Since a TS required shutdown was not required, Event Notification 41085 is being retracted. The condition was also reported as a loss of safety function for the accident mitigation function of CRERS. The CRERS is automatically activated by a Loss of Coolant Accident (LOCA) signal or a Control Room Ventilation (CRV) high radiation signal. The LOCA instrumentation circuitry was not affected by the defective time delay relay and thus the damper stroke times were not impacted. The CRV airborne radiation monitor signal is considered a "diverse" signal to the LOCA signal. Since the LOCA signal would have properly initiated the CRERS and the CRV high radiation signal is redundant, there was no loss of safety function. Since there was no loss of safety function, Event Notification 41084 is being retracted. The licensee has notified the NRC Resident Inspector. NRC R3DO(Gardner) has been notified. |
| ENS 40326 | 17 November 2003 15:26:00 | While aligning the Emergency Service Water (ESW) to the swale, the sluice gates were opened without the ESW being aligned to the swale. This condition made both Div 1&2 Diesel Generators and all ECCS systems inoperable. The probable cause of this event is that the wrong procedure was used in this evolution. Procedure SOI-P45/49 sec 7.3.1 "ESW Pumphouse Forebay Emergency Supply Initiation" was used instead of the correct procedure SOI-P45/49 section 7.3.3. "Alignment of ESW to Pumphouse Forebay Emergency Supply". The condition was discovered at 1215 and corrected at 1220 after section 7.3.1 was exited and section 7.3.3 was used to complete the lineup to the swale. The NRC Resident Inspector was notified.
Further engineering analysis documented (that) the safety function of the Emergency Service Water (ESW) system was not lost while the system was inappropriately aligned. Also, plant design accounts for operator action to restore proper alignment and ensure continued functionality. Since the safety function of ESW was, and would continue to be maintained, the ECCS systems did not and would not lose the ability to perform their safety functions. Therefore, this condition is not reportable under 10CFR50.72(b)(3)(v)(B) and (D) as an event or condition that could have prevented fulfillment of a safety function and ENF 40326 is retracted. The licensee notified the NRC Resident Inspector. Notified R3DO Bruce Burgess. |