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The Perry Nuclear Power Plant will be taking the Plant Computer out of service for scheduled maintenance which will take ERDS out of service. From 0500 hours EST on February 27, 2009 for approximately 15 hours, personnel will be performing inspection and cleaning activities on the Plant Computer power center panel. During the planned maintenance, the Safety Parameter Display System (SPDS) and the automatic mode calculation of the Computer Aided Dose Assessment Program (CADAP) will be unavailable. 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 Private Branch Exchange, the Off Premise Exchange, and various redundant intra-facility circuits throughout the emergency facilities. The dose assessment function will be maintained during the out of service time period by manual input of data into CADAP and, if required, by manual calculation. The ability to open and maintain an '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 period 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 maintenance activities are completed and the equipment is restored. The NRC Resident Inspector has been notified.

• * * UPDATE FROM TOM MORSE TO JOE O'HARA AT 1910 EST ON 2/27/09 * * *

The Perry Plant computer and the ERDS have been restored to service as of 1845 EST this evening. The licensee will notify the NRC Resident Inspector.

On March 30, 2009, at approximately 1843 hours EDT, the control room reported a loss of power to non-safety 480 volt busses F1C and F1D. Initial investigation revealed an instantaneous ground over-current (50G) relay had actuated. This relay actuation resulted in the supply breaker tripping and all associated loads being de-energized. This resulted in a loss of 120 VAC non-essential electrical power to a portion of the Emergency Response Data System (ERDS). 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 Private Branch Exchange, Off Premise Exchange, and various redundant intra-facility circuits throughout the emergency facilities. The ability to open and maintain an 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 period 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 emergency offsite communications capability. A follow-up notification will be made when the equipment is restored. The NRC Resident Inspector has been notified." The unit is currently defueled.

• * * UPDATE FROM TOM STEC TO JOE O'HARA AT 1600 EDT ON 3/31/09 * * *

Last evening, power was restored to ERDS following maintenance. ERDS is capable of performing its intended function. Notified R3DO(Orth)

On April 27, 2009, at approximately 1730 hours, with the plant in Mode 5 during refueling outage (RFO) 12, RHR 'A' pump tripped while operating in shutdown cooling. RHR 'A' was the primary decay heat removal shutdown cooling system. RHR 'B' was the backup decay heat removal shutdown cooling system. Preliminary investigation shows that jumper installation activities associated with plant testing resulted in a blown fuse and closure of the RHR shutdown cooling outboard common suction isolation valve (1E12F008). Closure of the 1E12F008 valve tripped the RHR 'A' pump and prevented the RHR 'B' pump from being used to initiate shutdown cooling from the control room. Operators were preparing to manually open 1E12F008 in parallel with activities to restore control from the control room. The plant entered Technical Specification 3.9.9, 'RHR - Low Water Level', Conditions A and C due to 1E12F008 isolating causing the loss of shutdown cooling. Action A.1 calls for verification of an alternate method of decay heat removal available for each inoperable RHR shutdown cooling subsystem in 1 hour and once per 24 hours thereafter. This was completed with the fuel pool cooling and cleanup system's two pumps and two heat exchangers cooled by NCC (Nuclear Closed Cooling System) (available due to the ability to reflood the upper pools with a hotwell pump through normal cavity reflood path) being one alternate system. A second alternate system was the utilization of the low pressure core spray to flood the vessel, returning to the suppression pool through safety relief valves, and a loop of RHR in suppression pool cooling. Actions for Condition C, to verify reactor coolant circulation by an alternate method and to monitor reactor coolant temperature were not met due to no reactor coolant flow past a valid temperature monitoring point. Approximate reactor coolant temperature was being trended using the reactor water cleanup system. The blown fuse was identified and replaced at approximately 1816 hours. The RHR 'B' pump was started at approximately 1834 hours. From the time that the RHR 'A' pump tripped (approximately 1730), until the RHR 'B' pump was started, the reactor temperature increased from 94 degrees F to 97 degrees F. Pre-determined time to boil had been calculated to be 9 hours. At approximately 1835 hours, TS 3.9.9 Condition C was exited due to the RHR 'B' shutdown cooling loop being placed in operation. This event is being reported as an event or condition that at the time of discovery could have prevented fulfillment of a safely function of structures or systems that are needed to remove residual heat under 10 CFR 50.72 (b)(3)(v)(B). The NRC Resident Inspector has been notified.

* * * RETRACTION FROM C. ELBERFELD TO P. SNYDER AT 1727 ON 6/25/09 * * * 

The purpose of this call is to retract Event Number 45025. On April 28, 2009, at 0055 hours, notification was made to the NRC Operations Center by the Perry Nuclear Power Plant (PNPP) reporting a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system that is needed to remove residual heat (10 CFR 50.72(b)(3)(v)(B)). Jumper installation activities associated with plant testing resulted in a blown fuse and closure of the Residual Heat Removal (RHR) shutdown cooling outboard common suction isolation valve. Closure of the valve resulted in the RHR A pump tripping, as designed. The blown fuse was replaced and the valve reopened. The RHR B subsystem was then started as the primary decay heat removal shutdown cooling system. Additionally, it was initially questioned whether Technical Specification Limiting Condition for Operation 3.9.9 Required Actions were met to verify reactor coolant circulation by an alternate method and to monitor reactor coolant temperature. Based on further evaluation, it was determined that there was not a reasonable expectation of the loss of safety function of a system needed to remove residual heat (i.e., the RHR System). The redundant RHR B subsystem was manually aligned and operated in a timely manner to continue to meet the system requirements to fulfill the safety function. Since the condition reported in Event Number 45025 would not have prevented the fulfillment of the safety function of a system that is needed to remove residual heat, the condition is not reportable, and this notification is retracted. Additionally, it was determined that verification of reactor coolant circulation by an alternate method (i.e., Reactor Water Cleanup System) was performed and that monitoring of reactor coolant temperature was appropriate. Therefore, based on not meeting any 10 CFR 50.73 reporting criteria, no Licensee Event Report is required. The evaluations (i.e., Reportability Reviews) for this condition are documented in Condition Report 09-58110 and Condition Report 09-58123. The licensee notified the NRC Resident Inspector.

• * * UPDATE FROM JOHN PELCIC TO CHUCK TEAL AT 1515 EST ON 3/10/10 * * *

The Residual Heat Removal (RHR) system common suction isolation valve 1E12-F008 received an invalid isolation signal resulting in the operating RHR A pump tripping while the system was in the shutdown cooling mode of operation. The event was reported to the NRC Operations Center on April 28, 2009 in accordance with 10 CFR 50.72(b)(3)(v)(B). The event notification was subsequently retracted on June 25, 2009. This event was reevaluated for reportability via Condition Report 10-71293 using additional guidance and enforcement history related to safety system functional failure reporting. Based on results of this evaluation, the April 27, 2009, loss of shutdown cooling event is reportable as a Licensee Event Report (LER) under 10 CFR 50.73(a)(2)(v)(B), 'Any event or condition that could have prevented the fulfillment of the safety function of structures, or systems that are needed to remove residual heat.' Perry LER 2010-001 will be submitted to report this event. The NRC Resident Inspectors have been notified.

At 2145 (EDT) hours on April 5, 2010, a loss of electrical power to the Division 1 Loss of Coolant Accident (LOCA) initiation logic occurred. Control room annunciators for 'Residual Heat Removal (RHR) Out of Service', 'Reactor Core Isolation Cooling (RCIC) Out of Service', and 'RCIC/RHR D2 to D1 00 File Power Loss' were received. At 2250 hours, the Control Room staff determined a loss of isolation function existed. The operators entered Technical Specification (TS) action statements for Emergency Core Cooling System Instrumentation (TS 3.3.5.1), RCIC Instrumentation (TS 3.3.5.2), and Primary Containment and Drywell Isolation Instrumentation (TS 3.3.6.1). The power loss was caused by a blown fuse which occurred during surveillance testing. The surveillance test was suspended and plant personnel commenced troubleshooting and investigation efforts. A recovery plan is being developed to back out of the surveillance test and replace the fuse to re-energize the logic, while ensuring the plant does not experience an undesirable actuation of the logic. The power loss caused five containment isolation valves to lose automatic isolation function. These valves have no associated inboard automatic isolation valve powered by Division 2. As a result of the power loss, the valves cannot automatically close on demand to isolate and therefore cannot perform their automatic function to isolate the containment. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(C) as an 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 control the release of radioactive material. The NRC Resident Inspectors have been notified. The licensee entered a 12-hour shutdown action statement for the loss of containment isolation function as a result of the power failure.

* * * RETRACTION FROM LLOYD ZERR TO PETE SNYDER ON 6/4/2010 AT 1126 * * * 

On April 6, 2010, at 0611 hours, notification was made to NRC Operations Center of an event that at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material (10 CFR 50.72(b)(3)(v)(c)). A blown fuse during surveillance testing caused a loss of electrical power to the Division 1 Loss of Coolant Accident initiation logic. The power loss caused five containment isolation valves to lose their automatic isolation function. These valves have no associated inboard automatic isolation valve powered by Division 2. As a result of the power loss, the valves could not automatically close on demand and therefore, could not perform their automatic function to isolate containment. Following further evaluation, it was determined that a loss of safety function of structures or systems that are needed to control the release of radioactive material did not occur. Based on review of valve leak rate test history and compliance with 10 CFR 50, Appendix A, General Design Criteria, the penetrations and Containment were capable of performing their intended design function. The associated inboard valves and/or water seals were capable of automatically isolating containment. Additionally, all required Technical Specifications Limiting Conditions for Operation were complied with for the identified issue, and therefore, no operation or condition prohibited by the plant's Technical Specifications existed. Since the event or condition reported in Event Notification 45815 would not have prevented the fulfillment of the safety function needed to control the release of radioactive material, this event or condition is not reportable and the notification is retracted. Additionally, based on not meeting any 10 CFR 50.73 reporting criteria, no Licensee Event Report is required. The evaluation (i.e., Reportability Determination) for this event or condition is documented in Condition Report 10-74904. The NRC Resident Inspector has been notified. Notified the R3DO (Pelke).

On November 16, 2011, at 2000 hours (EST), control room operators accepted the results of an immediate investigation related to the adequacy of a calculation related to the flooding analysis for service water piping in the control complex. The conclusions of the original analysis assumed operator actions to mitigate the flooding. The existing procedural guidance at that time lacked specificity for the required operator actions. This was identified as a non-conforming condition with respect to a USAR flooding analysis. The immediate investigation determined that significant margin exists for mitigation of the service water leakage crack compared to the 30 minute actions required in the original analysis. A preliminary strategy for flood mitigation was identified in the immediate investigation. Guidance was provided to the operators for a leak mitigation strategy in a night order, and a prompt functionality assessment was requested for the control complex building with respect to the flooding analysis. On November 22, 2011, at 1957 hours (EST), the prompt functionality assessment was accepted by the control room operators. The assessment included compensatory measures that simplified guidance for mitigating the flooding from the service water system. The compensatory measures were implemented. On December 2, 2011, the NRC Component Design Basis Inspection team debriefed that the condition identified should have been called in to the NRC Operations Center within eight hours and that missing the call was a violation of 10 CFR 50.72(b)(3)(ii)(B). On December 7, 2011, at 1320 hours (EST), a call was received from the NRC Region III informing the compliance supervisor that the eight-hour notification should still be made. The NRC Resident Inspector has been notified.

• * * UPDATE FROM CHARLES ELBERFELD TO JOHN KNOKE AT 1412 EST ON 12/10/11 * * *

The licensee added further clarification to the event reported above as follows: Given that an unanalyzed condition existed until the appropriate measures were implemented, an eight-hour call in accordance with the aforementioned section of 10 CFR 50.72 should have been made. After further consideration, station management decided the eight-hour call was missed and it is being reported as required. The NRC Resident Inspector has been notified." R3DO (Skokowski) notified.

On December 21, 2011, at 1359 hours EST, it was determined that the test instruction for the weekly groundwater level readings contained non-conservative acceptance criteria. The test instruction acceptance criteria for groundwater level exceeds the initial assumptions used in the Updated Safety Analysis Report (USAR) Chapter 15 accident analysis for 'Postulated Radioactive Releases due to Liquid Containing Tank Failures.' The issue was identified during the conduct of a prompt functionality assessment evaluating the plant underdrain system performance and to ensure all USAR described functions were being met. The test instruction acceptance criteria is less than 575 feet. The calculation supporting the Chapter 15 accident analysis assumes an initial groundwater elevation of 568 feet in order to accumulate a sufficient volume of groundwater to dilute the tank inventory prior to exiting the underdrain system. The accumulation volume results in hold up time allowing for mixing and radioactive decay. With the degraded performance of the underdrain system pumps and high precipitation, the groundwater level had risen above 568 feet but was still less than 575 feet. A preliminary engineering evaluation has determined that significant margin exists in the calculation. A calculation revision is being pursued in parallel with this notification. Compensatory actions involving the repair of 1 permanent non-safety pump and installation of temporary pumps were previously initiated and the groundwater levels are decreasing. Restoration of the remaining permanent plant pumps continues. The NRC Resident Inspector has been notified.

• * * RETRACTION ON 2/15/12 AT 1315 EST FROM LLOYD ZERR TO ERIC SIMPSON * * *

The calculation supporting the USAR Chapter 15 accident analysis was reviewed and subsequently revised. The revised calculation verifies and initial groundwater elevation of 575 feet is consistent with the preliminary assessment that substantial margin existed in the calculation for the underdrain system. Because the condition reported in Event Number 47545 was not an event or condition that results in the nuclear power plant being in an unanalyzed condition that degrades safety, the condition is not reportable, and this notification is retracted. The evaluation for this condition is documented in condition report 2011-07169. The NRC Resident Inspector and R3DO (Passehl) were notified.

Beginning at approximately 1100 hours EDT on May 10, 2012, plant personnel will be taking the plant integrated computer system (ICS) out-of-service for planned maintenance. During the time ICS is out-of-service, the Safety Parameter Display System (SPDS), the Emergency Response Data System (ERDS), and the automatic mode calculation of the Computer Aided Dose Assessment Program (CADAP) will be unavailable. The computer outage is scheduled for two hours. 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 Private Branch Exchange, Off Premise Exchange, and various redundant intrafacility circuits throughout the emergency facilities. The dose assessment function will be maintained during the out-of-service time period by manual input of data into CADAP and, if required, by manual calculation. The ability to open and maintain an '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. This event is being reported in accordance with 10 CFR 50.72(b)(3)(xiii). A follow-up notification will be made when the maintenance activities are completed and the equipment is restored. The NRC Resident Inspector has been notified.

• * * UPDATE AT 1245 EDT ON 5/10/12 FROM MORSE TO HUFFMAN * * *

The maintenance activities were completed as scheduled and the integrated computer system and associated systems SPDS, ERDS and CADAP has been returned to service as of 1238 EDT. The licensee will notify the NRC Resident Inspector. R3DO (Giessner) notified.

At approximately 0800 hours EDT on October 17, 2012, computer engineering personnel will be taking the plant integrated computer system (ICS) out-of-service for planned maintenance. During the time ICS is out-of-service, the Safety Parameter Display System (SPDS) and the Emergency Response Data System (ERDS) will be unavailable. The computer outage is scheduled for twelve hours. In the event of an emergency, plant parameter data will be communicated to the facilities through the Status Board Ring Down circuit with back-up by the Private Branch Exchange, Off Premise Exchange, and various redundant intra-facility circuits throughout the emergency facilities. The dose assessment function will be maintained during the out-of-service time period by manual input of data into the Meteorological Information and Dose Assessment System (MIDAS). The ability to open and maintain an '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. This event is being reported in accordance with 10 CFR 50.72(b)(3)(xiii). A follow-up notification will be made when the maintenance activities are complete and the equipment is restored. The (NRC) Resident Inspector has been notified.

• * * UPDATE FROM THOMAS MORSE TO VINCE KLCO ON 10/17/12 AT 2146 EDT * * *

As of 2140 EDT on 10/17/12, the ERDS system was tested and restored to service. The licensee notified the NRC Resident Inspector. Notified the R3DO (Orth).

Entered an Unusual Event (under Emergency Action Level) MU1, toxic gas, carbon monoxide (CO), detected in the Radwaste Control Room. Levels rose to 34 ppm and the Radwaste Control Room was evacuated prior to reaching the First Energy exposure limit of 35 ppm. The source of the CO has not been determined. There is no radiation release from this event. There were no personnel injuries and offsite assistance was not requested. There was no effect on plant operations. The licensee has notified state and local authorities and the NRC Resident Inspector. Notified DHS SWO, FEMA, DHS NICC and NuclearSSA via email.

• * * UPDATE AT 1504 EST ON 11/28/12 FROM MARIKIO BLOUNT TO HUFFMAN * * *

(At 1452 EST, the licensee) terminated the Unusual Event for MU1, due to toxic gas - carbon monoxide, detected in the Radwaste Control Room.  The source of the carbon monoxide readings was determined to be from a leaking acetylene bottle.  The acetylene bottle has been removed from the building.  Carbon monoxide readings have returned to normal.

The licensee noted that the acetylene is detected as carbon monoxide by the toxic gas monitoring devices. The licensee has notified State and local authorities and the NRC Resident Inspector. Notified R3DO (Stone), NRR (EO) Lubinski) , IRD (Marshall), DHS SWO, FEMA, DHS NICC and NuclearSSA via email.

Computer engineering personnel will be taking the plant integrated computer system (ICS) out-of-service for planned maintenance. During the time ICS is out-of-service, the Safety Parameter Display System (SPDS) and the Emergency Response Data System (ERDS) will be unavailable. The computer outage is scheduled for six hours. Contingency plans have been established to transmit plant parameter data and perform the dose assessment function in the event of an emergency while ERDS is unavailable. This event is being reported in accordance with 10 CFR 50.72(b)(3)(xiii). A follow-up notification will be made when the maintenance activities are complete and the equipment is restored. The licensee notified the NRC Resident Inspector.

• * * UPDATE FROM JIM CASE TO HOWIE CROUCH AT 1309 EST ON 12/20/12 * * *

At 1300 EST, the plant integrated computer system was restored and SPDS and ERDS was returned to service. Notified R3DO (Cameron) and ERDS Group email.

Release of toxic or flammable gas affecting the Protected Area boundary deemed detrimental to the safe operation of the plant. Emergency Action Level entered: MU-1. The leak is Trichloroethylene (TCE) gas used in the Off-Gas building. The Off-Gas building ground and basement levels were evacuated due to the leak. There is no safe-shutdown equipment located in the Off-Gas building. The licensee is working to isolate the leak. The licensee informed the NRC Resident Inspector. The licensee notified the State of Ohio and the local counties. Notified DHS SWO, FEMA Ops Center, NICC Watch Officer, DOE Ops Center, USDA Ops Center, HHS Ops Center, and Nuclear SSA via email.

• * * UPDATE AT 1630 EDT ON 4/2/14 FROM DON ROGERS TO S. SANDIN * * *

The licensee notified the following outside agencies: U.S. EPA National Response Center, Ohio EPA, Perry Township Fire Department, Lake County Emergency Planning Committee, and the U.S. Coast Guard. Notified R3DO (Passehl).

• * * UPDATE FROM MICHAEL ADLER TO DANIEL MILLS AT 0115 EDT ON 04/05/2014 * * *

Unusual Event has been terminated on 4/5/2014 at 0059 EDT. The trichloroethylene leak has been stopped. Access has been restored to all normally accessible areas. Unit 1 remains in Mode 1 at 100% power. The licensee notified the NRC Resident Inspector and the Local and State emergency agencies. Notified the IRD MOC (Gott), R3DO (Passehl), and NRR EO (McGinty). Notified DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, NICC Watch Officer, USDA OPS Center, EPA EOC, FDA EOC, and Nuclear SSA via email.

At 1313 (EDT) on 7/26/14, the plant experienced an electrical transient on bus EK-1-B1 (safety-related 120 volt AC distribution panel) that resulted in partial Balance of Plant Division 2 isolation signals and alarms received in the Control Room. The following component actuations occurred: valve 1P50F140 closed, resulting in a trip of Containment Vessel Chilled Water C; valve 1G41F140 closed, isolating the Fuel Pool Cooling and Clean-up return from the containment building upper pools; valve 1B33F019 closed, isolating Reactor Water sampling; valve 1D17F071B closed, isolating the Drywell Atmosphere Radiation Monitor; valve 1D17F081B closed, isolating the Containment Atmosphere Radiation Monitor; valves 1G61-F030, 1G61-F150, 1G61-F075, and 1G61-F165 closed, isolating the Containment and Drywell Floor and Equipment drain sumps; valve 1G50-F272 closed isolating the Reactor Water Cleanup Backwash Receiving Tank: 1M25F020B, Control Room HVAC Inboard supply damper, closed and Division 2 indicated an auto initiation (M25-S12, Auto Initiate Active amber light was on). This event is considered an invalid system actuation reportable under 10 CFR 50.73(a)(2)(iv)(A).

The isolation was not initiated in response to actual plant conditions or parameters, and was not a manual initiation. Therefore, this notification is provided via a 60 day optional phone call in accordance with 10 CFR 50.73(a)(1) instead of submitting a written Licensee Event Report. The event meets reporting criteria specified in 10 CFR 50.73(a)(2)(iv)(B)(2) as a general containment isolation valve signal affecting containment isolation valves in more than one system. All affected systems functioned as expected in response to the electrical transient on bus EK-1-B1 that resulted in the partial Balance of Plant Division 2 isolation signals. The valves were reopened in accordance with plant procedures. The failure mechanism that caused the electrical transient was a failed capacitor in regulating transformer EFB1B2. The capacitor was replaced and tested with satisfactory results. The NRC Resident Inspector has been notified.