Semantic search
| Discovered date | Reporting criterion | Title | Description | LER | |
|---|---|---|---|---|---|
| ENS 40118 | 29 August 2003 17:50:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection System Declared Inoperable | During performance of routine operability testing, the High Pressure Coolant Injection System (HPCI) tripped and restarted due to an as yet undetermined cause. The trip and restart sequence occurred twice in close succession approximately 20 minutes into a normal run before the operator took action to manually trip the turbine. Investigation into the cause of the malfunction is on-going. The HPCI system has been declared inoperable in accordance with Technical Specifications. The operability of all other Emergency Core Cooling System components has been verified. There was never any actual coolant injection by the HPCI system during this event. The NRC Resident Inspector has been notified by the licensee. The State of Massachusetts will also be notified. | |
| ENS 40138 | 6 September 2003 15:51:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Hpci Declared Inoperable Due to Loss of the 480V Bus. | While operating at 100% power, the plant sustained a loss of the 480 bus "B1". As a result of the loss of power, HPCI has been isolated due to the inability to auto isolate on a primary containment isolation signal. The SBGT was initiated to restore building ventilation. The Reactor Water Clean Up System was manually isolated due to the loss of power. The "A" recirculation pump tripped as a result of the loss of power. The loss of the 480V bus is being investigated at this time. The NRC Resident Inspector was notified.
Due to the loss of power to the RCIC quadrant coolers, the RCIC system is inoperable but available. The NRC Resident Inspector was notified. | |
| ENS 40211 | 1 October 2003 08:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Pressure Vessel Boundry Leakage Found During Inspection | At approximately 0430 on October 1, 2003 with the reactor Mode Switch in Refuel, reactor water level 210" above Top of Active Fuel, and reactor water temperature less than 110F, investigations and walkdowns were being performed to identify potential sources of drywell leakage. Drywell leakage was well within Technical Specification limits. During these walkdowns a small leak in a cap of a nozzle of the reactor pressure vessel boundary, nozzle N10, located at reactor vessel height 440" was identified. This location is 84" above Top of Active Fuel. This is a cut and capped 4" Control Rod Drive return line. The leak appears to be on the cap weld. This notification is being made per 10 CFR 50.72(b)3(ii)(A). All required ECCS systems are presently operable. Investigation is continuing. The (NRC) resident inspector has been notified. | |
| ENS 40448 | 14 January 2004 15:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Emergency Sirens Inoperable/Degraded Due to Potential Amplifier Problems | At 1049 EST on January 14, 2004 the Control Room was notified that 78% (87 out of 112) Emergency Sirens failed a 'Quiet Test.' The Quiet test is a status check between the Emergency Offsite Facility and each individual siren. Further investigation revealed that 53 sirens would not operate as required and 34 were experiencing some degree of degradation. Indications point toward amplifier problems within each affected siren. Efforts are ongoing to troubleshoot and repair the effected sirens. Contingency plans are in place for alternate notification. The resident inspector has been notified. This condition was discovered during routine monthly testing. The sirens were upgraded about a year ago using equipment supplied by Federal Signal. * * * UPDATE ON 1/20/04 AT 1422 EST FROM RANDY HAISLET TO GERRY WAIG * * * The following was received via facsimile: This is a follow up to event report 40448 concerning a loss of 53 sirens. As of this morning, 105 of the 112 sirens are considered operable. Six of the 7 inoperable sirens are due to rotational problems. The remaining siren is inoperable due to the impact of a vehicle on it's pole. Work is currently on going to restore these sirens. The siren failures reported in event report are believed to be the result of a manufacturing defect which resulted in temperature sensitivity. A temporary means of heating the subject equipment has been put in place. Investigation is continuing. Notified R1DO (Clifford Anderson) | |
| ENS 40547 | 26 February 2004 06:05:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor | Control Power Fuse Failure in Hpci System | The High Pressure Coolant Injection (HPCI) system had been removed from service to perform planned maintenance and testing. During post-maintenance testing, the HPCI gland seal condensate pump tripped due to a blown control power fuse. The maintenance scope did not include the pump or its power supply. The fuse was replaced and the surveillance completed satisfactorily. Investigation into the cause of the failure is ongoing. HPCI remains inoperable in support of additional pre-planned testing. The licensee notified the State and the NRC Resident Inspector. | |
| ENS 40909 | 30 July 2004 13:45:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Reactor Core Isolation Cooling (Rcic) Turbine Failed to Achieve Design Pressure and Flow During Surveillance Testing | The following information was obtained from the licensee via facsimile: During a surveillance of the Reactor Core Isolation Cooling system (RCIC), the RCIC turbine failed to achieve design pressure and flow. Other safety systems, including the High Pressure Coolant Injection (HPCI) system, are available. This failure is believed to be due to a RCIC flow controller problem. Investigation is continuing. Design pressure/flow is 1250 psig/400 gpm. Actual pressure/flow was 1220 psig/350 gpm. The licensee has notified the NRC Resident Inspector. | |
| ENS 41408 | 14 February 2005 00:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection System (Hpci) Declared Inoperable | The following text report was received from the licensee via facsimile: The High Pressure Coolant Injection (HPCI) system was declared inoperable on 2-13-05 at 19:00 EST due to loss of position indication to MO-2301-8 (HPCI injection valve #2) in the control room and at the system alternate shutdown panel. The other ECCS (Emergency Core Cooling) Systems remain operable. Position indication was restored after control power fuses were replaced. HPCI was returned to operable status at 21:50 EST, 2-13-05. The licensee has notified the NRC Resident Inspector and will be notifying Massachusetts Emergency Management Agency. | |
| ENS 41600 | 14 April 2005 15:30:00 | 10 CFR 26.73, Applicability | Fitness for Duty | A contract employee had a positive test result for alcohol during a for-cause test. The employee's access to the plant was terminated. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. | |
| ENS 41624 | 22 April 2005 11:00:00 | 10 CFR 26.73, Applicability | Contract Employee Tested Positive During a For-Cause Fitness for Duty Test | A contract foreman tested positive for alcohol during a for-cause alcohol test. The employee had not performed any work and was not allowed access to the site. The employee's access to the site was terminated. The NRC Resident Inspector was informed. | |
| ENS 41799 | 26 June 2005 14:55:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Emergency Diesel Generators Inoperable Due to High Ambient Temperature | The report is being made in accordance with 10 CFR 50.72(b)(3) due to both Emergency Diesel Generators (EDGs) being declared inoperable. (This is a 24-hour LCO.) The EDGs were declared inoperable due to indicated outside air temperature exceeding the established operability limit of 95F for two short periods of time. Indicated outside air temperature went above 95F during the following time frames: 10:55 to 11:00 and 12:15 to 12:20. Indicated outside air temperature subsequently decreased to less than 95F following these spikes. Both EDGs are currently operable. Outside ambient temperatures are continuing to be monitored. Offsite power is available. Further evaluation of this event is ongoing. The licensee notified the NRC Resident Inspector.
The initial report was made in accordance with 10 CFR 50.72(b)(3) due to both Emergency Diesel Generators (EDGs) being declared inoperable. The EDGs were declared inoperable due to indicated outside air temperature exceeding the established operability limit of 95F for short periods of time on 6/26/05. Outside indicated air temperature subsequently decreased to less than 95F following these short timeframes. Both EDGs were operable when the notification was made, and outside ambient temperatures continued to be monitored. Offsite power was available. Further evaluation of this event has been performed. The initial report used temperature indication that was conservative with respect to actual conditions. The evaluation determined that the air temperature outside the EDG Building was a maximum of 91 F on 6/26105. This evaluation was based on other temperature reading taken during the periods of interest. Additionally, further evaluation has determined that both EDGs were not inoperable even at the higher temperatures initially indicated. Therefore, the EDGs were not inoperable during the periods reported and event notification #41799 is retracted. Notified the R1DO (Caruso). The licensee notified the NRC Resident Inspector. | |
| ENS 41948 | 25 August 2005 20:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Injection Inoperable Due to Flow Oscillations | This report is being made in accordance with 10CFR50.72 (b) (3) (v) due to the High Pressure Coolant Injection (HPCI) system being declared inoperable. HPCI was declared inoperable on 8/25/05 at 1630 EST due to oscillations at below rated flow during the scheduled operability testing. HPCI was restored to standby line-up when testing was completed and remains available for use. This event is an eight hour notification. Efforts are on going to determine the cause of the oscillations on the Flow Controller. This event had no adverse effect to the health and safety of the public. The resident NRC inspector has been notified of this event.
This follow-up notification is being made to retract the notification made to the NRC Operations Center on 8/25/05 at 2039 hours (notification #41948). The initial report was made in accordance with 10 CFR 50.72(b)(3) due to the HPCI system being declared inoperable. The system was declared inoperable due to oscillations in turbine speed, pump discharge pressure, and pump flow during a quarterly surveillance test of the HPCI pump. Further investigation and evaluation of this has been performed. The cause of the noted oscillations was the position of a hand operated valve that is located in the HPCI system full flow test line. The hand operated valve is located downstream of an in-series motor operated valve that automatically closes if an automatic HPCI system initiation signal occurs. The full flow test line is not part of the HPCI injection pathway to the reactor vessel. As a result, the position of this valve would not have impacted the ability of the system to perform its design function. After adjusting the position of the hand operated valve, the surveillance test of the HPCI pump was completed with satisfactory results. The evaluation has determined that the HPCI system was capable of performing the designed safety function. Therefore, the HPCI system was not inoperable and event notification #41948 is retracted. The licensee notified the NRC Resident Inspector. Notified R1DO (C. Cahill) | |
| ENS 42414 | 13 March 2006 23:08:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram After Offgas System Failure | During a power ascension, the non-safety Augmented Offgas System experienced a failure which caused a Recombiner high temperature condition. To protect components of the non-safety Augmented Offgas System, station procedures require a manual scram of the reactor when this condition occurs. Station procedures were followed and the reactor was manually scrammed at 18:08. All rods fully inserted and all safety systems performed per design. Primary containment isolation systems responded properly resulting in an automatic isolation of Primary Containment Isolation System Groups 2 and 6 valves and a Reactor Building Isolation due to the transient low reactor water level condition caused by the scram. The plant is in a stable condition. Investigation is continuing. No safety valves lifted on the scram, decay heat is being removed with the bypass valve, normal feed and condensate are maintaining reactor water level, and the electrical grid is stable on the startup transformer. The licensee notified the NRC Resident Inspector. | |
| ENS 42433 | 20 March 2006 15:25:00 | 10 CFR 26.73, Applicability | Fitness for Duty - Confirmed Positive for Non-Licensed Supervisor During Random Test | This report is being made in accordance with 10 CFR 26.73 due to a supervisor testing positive during a random fitness for duty test. The employee was fitness for duty tested on 3/16/06. The Medical Review Officer confirmed a positive test at 10:25 EST on 3/20/06. A review of previous work is being performed. The employee's access to the plant has been denied. The NRC Resident Inspector has been notified. | |
| ENS 42597 | 23 May 2006 02:00:00 | 10 CFR 74.11(a) | Special Nuclear Material (.003 Grams) Lost | During ongoing activities to remove non-fuel material from the Pilgrim Spent Fuel Pool it has been identified that an irradiated neutron detector containing a very small quantity (less than 0.003 grams) of special nuclear material is not in its expected location. Per the inventory sheets the neutron detector should have been enclosed in a "dry tube" in the Spent Fuel Pool. Processing of the "dry tube" for shipment identified that the neutron detector is not in its expected location. This condition is being conservatively reported under 10CFR74.11. There is no evidence of theft or diversion. Investigation is continuing. The licensee notified the NRC Resident Inspector.
In Event Notifications 42597 and 42599 on May 22 and May 23, 2006, respectively, Pilgrim Nuclear Power Station notified the NRC that during ongoing activities to remove non-fuel material from the Pilgrim Spent Fuel Pool it had been identified that 2 irradiated neutron detectors containing very small quantities (less than 0.003 grams each) of special nuclear material were not in their expected locations. Per the inventory sheets the neutron detectors should have been enclosed in 'dry tubes' in the Spent Fuel Pool. Processing of the 'dry tubes' identified that the neutron detectors are not in their expected locations. There were a total of twelve (12) 'dry tubes' that our records show enclosed neutron detectors. As part of the investigation associated with Event Notification 42597, all twelve (12) of these 'dry tubes' have been inspected. Three (3) of the 'dry tubes' were found to contain neutron detectors consistent with plant records. Nine (9) of the 'dry tubes' were found to contain no neutron detector which is not consistent with the plant records. The irradiated neutron detectors would have each contained a very small quantity (less than 0.003 grams each) of special nuclear material. This condition is being conservatively reported under 10CFR74.11. There is no evidence of theft or diversion. Investigation is continuing. The licensee notified the NRC Resident Inspector. Notified IRD (Blount and Leach), R1DO (Gray), NRR (Haney), ILTAB (English), DHS SWO (Cassandra), FEMA (S. Kimbrell), DOE (Ronnie), EPA (Crews), USDA (Margaret), and HHS (Marcy). | |
| ENS 42599 | 23 May 2006 14:08:00 | 10 CFR 74.11(a) | Lost Special Nuclear Material | In Event Notification 42597 on May 22, 2006 Pilgrim Nuclear Power Station notified the NRC that during ongoing activities to remove non-fuel material from the Pilgrim Spent Fuel Pool it had been identified that an irradiated neutron detector containing a very small quantity (less than 0.003 grams) of special nuclear material is not in its expected location. Per the inventory sheets the neutron detector should have been enclosed in a 'dry tube' in the Spent Fuel Pool. Processing of the 'dry tube' for shipment identified that the neutron detector is not in its expected location. There were a total of twelve (12) 'dry tubes' that our records show enclosed neutron detectors. Four (4) of these 'dry tubes' have been processed and a second 'dry tube' in which no neutron detector was stored has been found. This irradiated neutron detector would have contained a very small quantity (less than 0.003 grams) of special nuclear material. This condition is being conservatively reported under 10CFR74.11. There is no evidence of theft or diversion. Investigation is continuing. The licensee notified the NRC Resident Inspector. | |
| ENS 42661 | 26 May 2006 19:25:00 | 10 CFR 20.2201(a)(1)(ii) | Potential Loss of Licensed Nuclear Material | In Event Notifications 42597and 42599 on May 22, and 23, 2006, respectively, as updated on May 26, 2006 Pilgrim Nuclear Power Station notified the NRC that during ongoing activities to remove non-fuel material from the Pilgrim Spent Fuel Pool it had been identified that nine (9) irradiated neutron detectors containing very small quantities (less than 0.003 grams each) of special nuclear material were not in their expected locations. Per the inventory sheets, the neutron detectors were expected to be enclosed in "dry tubes" in the Spent Fuel Pool. Processing of the "dry tubes" identified that the neutron detectors are not in their expected locations. Subsequent investigation has determined that an error in the records was introduced in 1996 when the current base inventory sheets were developed. During the 1996 inventory, the assumption was made that these nine (9) "dry tubes" contained neutron detectors. The investigation has determined that the detectors, although moved to the spent fuel pool, were removed from the "dry tubes" prior to the "dry tubes" removal from the reactor vessel in 1987. This affected the validity of the 1996 base inventory. A written report will be submitted consistent with 10 CFR 20.2201(b). The licensee notified the NRC Resident Inspector.
Worst case isotope totals for the 9 detectors combined would be the following (all other isotopes would be an order of magnitude or more lower): U 235 - 5.382E-08 Curies, Pu 239 - 2.376E-07 Curies, Pu-238 - 3.510E-08 Curies, Pu 241 - 4.572E-08 Curies The licensee notified the NRC Resident Inspector. Notified R1DO (Perry) and NRR EO (MJ Ross-Lee). THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks. This source is not amongst those sources or devices identified by the IAEA Code of Conduct for the Safety & Security of Radioactive Sources to be of concern from a radiological standpoint. Therefore is it being categorized as a less than Category 3 source | |
| ENS 42750 | 2 August 2006 09:45:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Motor Vehicle Fatality of Plant Employee | On Wednesday, August 2, 2006, at approximately 0545 EDT, a motor vehicle accident occurred at the entrance to the Owner Controlled Area that resulted in the fatality of a plant employee. The employee was reporting to work. Local law enforcement and emergency medical personnel were called and responded to the scene. An investigation by local law enforcement and plant security is ongoing. No formal press release is anticipated. This event presented no risk to public health and safety. The NRC Resident Inspector was notified of this report by the licensee. | |
| ENS 43093 | 11 January 2007 15:00:00 | 10 CFR 26.73, Applicability | Fitness for Duty | A non-licensed contract employee supervisor had a confirmed positive result during the initial fitness-for-duty test. The employee's access to the plant has been denied. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. | |
| ENS 43245 | 17 March 2007 20:58:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Group Isolations Following Manual Reactor Scram | On March 17 2007, at approximately 1658 EDT, Group 2 (Reactor Building Ventilation Isolation) and Group 6 (RWCU Isolation) automatic containment isolation signals were received due to low reactor water level following the insertion of a manual scram signal. The manual scram signal was inserted following the reaching of internal administrative limits on changes in unidentified drywell leakage. The receipt of these isolation signals is not unusual following the insertion of a scram signal. The reactor was manually shutdown due to reaching internal administrative limits on changes in unidentified drywell leakage. Prior to the manual shutdown, the leakage had not reached the applicable Technical Specification limits. Drywell leakage as of 1800 EDT was 1.3 GPM identified and 2.59 GPM unidentified with the unit in Hot Shutdown Mode. All safety systems responded as expected. The plant is in a stable condition. Investigation is continuing. The licensee notified the NRC Resident Inspector. | |
| ENS 43322 | 26 April 2007 23:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition During Weld Deposit Overlay Repair | On April 26, 2007, at approximately 1930 it was reported that the N2K recirculation system inlet nozzle had slight water seepage. The seepage developed during welding operations being performed to install a full structural weld overlay over the nozzle and stopped almost immediately during the welding. The weld overlay was being installed as a conservative measure following UT inspections that had been performed earlier during the outage. This event is being reported in accordance with 10 CFR 50.72(b)(3)(ii)A. The plant is in stable condition. The licensee notified the NRC Resident Inspector. | |
| ENS 43479 | 10 July 2007 23:54:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Main Turbine Trip Causing a Reactor Scram | During a planned power reduction to support thermal backwash of the main condenser, an automatic reactor scram resulted from a main turbine trip. The cause for the turbine trip is currently under investigation. Following the reactor scram, the expected reactor level shrink resulted in valid group 2, Reactor Building Ventilation, and group 6, Reactor Water Cleanup System isolations. All safety systems and equipment functioned as designed. A schedule for plant restart has not yet been determined. All control rods fully inserted and no safety relief valves lifted from the scram. Minimum level after the scram was -10 inches. Decay heat is being removed using the main turbine bypass valves to the main condenser and maintaining level using normal reactor feed water. The plant is using the normal shut down electrical lineup and slowly cooling down (currently at 820 psi). The licensee notified the NRC Resident Inspector. | |
| ENS 43642 | 14 September 2007 14:50:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification to Epa Required by Npdes Permit | At 0630 hours on September 14, 2007, while operating at 100% power, Pilgrim Station experienced a fish impingement on the intake structure traveling screens. Plant power was reduced to 50% and one of two seawater circulating pumps was secured. The impingement did not impact the operability of safety related cooling water systems. At 1050 hours, plant assessment determined that the impingement, involving a large school of Atlantic juvenile menhaden, is reportable to the Environmental Protection Agency (EPA) under conditions of the Pilgrim National Pollutant Discharge Elimination System (NPDES) permit. Following a substantial reduction in the numbers of fish discharged via the traveling screen outfall trough, the seawater pump was restored to service and restoration of the plant to 100% power is in progress. The licensee informed the NRC Resident Inspector and will inform both the State and EPA. | |
| ENS 43667 | 26 September 2007 18:55:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center and Operations Support Center Ventilation Inoperable | At 1455 on September 26, 2007 investigatory surveillance testing identified that the emergency ventilation system designed to be manually initiated upon facility activation to maintain the Technical Support Center (TSC) and Operations Support Center (OSC) at positive pressure was not functioning properly. The emergency fan motor operated discharge dampers were not fully opening preventing the ability to achieve rated system flow and therefore design positive pressure. Maintenance is currently in progress to restore the degraded dampers/damper controls. Pilgrim is also investigating a temporary modification that would allow manual damper alignment to support achieving design positive pressure. Responsible Pilgrim Emergency Response Organization (ERO) personnel have been notified of this condition. The alternate technical and operations support facilities remain fully functional. The licensee notified the NRC Resident Inspector. The licensee will notify the State of Massachusetts. | |
| ENS 43794 | 20 November 2007 11:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | High Pressure Coolant Inject Inoperable | On November 20, 2007 at 0630 hours, with the reactor at 100% core thermal power, a power supply failure was discovered in the high pressure coolant injection (HPCI) flow controller circuitry that may have precluded the system from performing its design basis function. Therefore, in accordance with 10 CFR Part 50.72(b)(3)(v) an eight-hour notification is being made. As background, on November 18, 2007, at 2145 hours, the high pressure coolant injection (HPCI) system was removed from service for planned maintenance. The required risk analysis was performed and the appropriate 14 day limiting condition for operation (LCO) was entered in accordance with Technical Specification (TS) 3.5.C. Later on November 19, 2007 at approximately 2100 hours the planned maintenance had been completed and HPCI was restored to the normal standby line-up in preparation for post maintenance testing (PMT). The HPCI valve quarterly operability and HPCI pump and valve quarterly operability tests were performed as the prescribed PMT. Upon initiation, the HPCI turbine was observed to come up to expected rated speed (~4,200 rpm) and expected HPCI pump discharge pressure (~1,300 psig). However HPCI pump indicated discharge flow was observed to be ~2,300 gpm, which is less than the Technical Specification requirement of 4,250 gpm. The HPCI system was secured and remained in the original TS 3.5.C LCO and a troubleshooting plan was initiated. On November 20, 2007, at 0630 hours, troubleshooting identified a power supply failure in the HPCI flow control circuitry. A replacement flow controller was identified and installed and it is anticipated that appropriate PMT will be initiated by 1600 hours. The impact of the power supply failure for the design basis operability for HPCl could not be definitively established before the eight-hour notification requirement of 10 CFR Part 50.72(b)(3)(v) was exceeded. The licensee notified the NRC Resident Inspector and the Commonwealth of Massachusetts.
NRC Notification 43794 was conservatively made to ensure that the Eight-Hour Non-Emergency reporting requirements of 10 CFR 50.72 were met pending the evaluation of an atypical condition (low reading) observed with the High Pressure Coolant Injection (HPCI) Flow Controller while performing scheduled surveillance testing for the HPCI System. During surveillance testing on 11/18/07, the HPCI System was started and met or exceeded the Technical Specification minimum requirements designed to demonstrate HPCI System Operability. While testing the specific components of the system, the HPCI Flow Controller was observed to be behaving erratically. Although the HPCI System was still capable of performing its required design safety function, the Shift Manager declared the system inoperable since he did not have definitive indication that the turbine was providing the required flow. Troubleshooting of the flow controller determined that the low flow indication was due to a degraded transmitter power supply located internal to flow controller FIC-2340-1. FIC-2340-1 is located in the main control room and is used to control HPCI system flow rate, and provide power to flow transmitter FT-2358. Although indicated flow rate was only 2300 gpm due to the degraded power supply, actual flow rate was approximately 5400 gpm based on pump hydraulic curves. The power supply in question only supplies power to FT-2358. Normal required supply voltage from this power supply is 28VDC to 36VDC. The degraded power supply could only supply 22.4VDC at the transmitter FT-2358 terminals. The degraded power supply voltage caused transmitter to output a lower than normal current for the actual measured flow rate giving a false low flow rate to FIC-2340-1. An Apparent Cause Evaluation and Past Operability Evaluation were performed in response to this event. These evaluations concluded that HPCI System was capable of performing its intended safety functions with the transmitter power supply degraded. HPCI system was capable of performing its intended safety functions during the time when FIC-2340-1 transmitter power supply exhibited low output voltage. HPCI would have started and supplied design basis flow to reactor vessel under design basis conditions. Thus there would have no impact on nuclear safety. Therefore, this event was not reportable pursuant to 10 CFR 50.72(b)(3)(v)(D). ENS Event Number 43794, made on 11/20/2007, is being retracted. The licensee will notify the NRC Resident Inspector and the Massachusetts Civil Defense Authority. Notified R1DO(Cobey) | |
| ENS 43885 | 8 January 2008 15:40:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Rcic Inoperable Due to Min Flow Valve Inability to Reposition | This report is being made in accordance with 10 CFR 50.72 (b) (3) (v) due to the Reactor Core Injection Cooling (RCIC) system being determined to be inoperable on 01/08/08 at 1040 EST. During a planned RCIC system outage, an instrument calibration surveillance identified a flow switch failure that would have prevented automatic closure of the pump minimum flow valve. Insufficient data is immediately available to assess the ability to achieve design basis flow rates with the minimum flow valve open. This event is an eight-hour notification. The RCIC instrument is currently under repair and will be completed prior to return to service. Plant is in a stable condition. Investigation is continuing. The resident NRC inspector has been notified of this event. This event places them in a 14-day LCO per ACT-1-08-002. HPCI verified operable.
BASIS FOR RETRACTION: Event Notification 43885 was conservatively made to ensure that the Eight-Hour Non Emergency reporting requirements of 10 CFR 50.72 were satisfied pending the evaluation of RCIC System operability. On 1/8/08, during performance of Attachment 5 to 8.E.13, 'RCIC System Instruments Calibration', RCIC flow switch FS-1360-7, contact number 2 failed to close as expected on increasing test pressure. This switch is expected to close while increasing test pressure between 13.7 to 14.3 inWC (inches of Water Column). Contact number 2 closes when RCIC flow exceeds 100 gpm signaling (minimum) flow valve MO-1301-60 to close. Failure of the switch to close prevents automatic closure of the (minimum) flow valve on a system flow of 100 gpm increasing. Failure of the (minimum) flow valve to close during RCIC system operation would allow about 70 gpm to 170 gpm of RCIC pump discharge flow to go directly to the torus bypassing the reactor vessel. The switch was replaced and the flow switch was returned to service. The defective switch was evaluated and the cause of the failure was determined to be carbon buildup on the switch contacts. A functional failure review was performed to assess the impact of the flow switch failure on the RCIC System design basis functions. The RCIC System is required to automatically provide makeup water to the reactor vessel following vessel isolation. This review identifies that 400 gpm is adequate to meet reactor vessel makeup requirements. With the flow controller in 'AUTO' and the minimum flow valve open, the flow controller would increase turbine speed until the flow rate setpoint of 400 gpm is achieved. Based on evaluation of the RCIC System flow controller configuration, turbine speed limits, and hydraulic modeling, it was determined that the required 400 gpm flow rate could have been delivered under worst case conditions with a failed open minimum flow valve. These evaluations concluded that the RCIC System was capable of performing its intended safety functions during the time when FS-1360-7 failure prevented automatic closure of the pump minimum flow valve. The RCIC System would have started and supplied design basis flow to reactor vessel under design basis conditions. Thus there would have (been) no impact on nuclear safety. Therefore, this event was not reportable pursuant to 10 CFR 50.72(b)(3)(v). Event Number 43885, made on 01/08/2008, is being retracted. The licensee will be notifying the NRC Resident Inspector. Notified R1DO (Caruso). | |
| ENS 44217 | 16 May 2008 00:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification to State of Massachusetts Regarding Press Release | Entergy Pilgrim Station has reached a tentative 4-year agreement with UWUA Local 369. This notification is in anticipation of media interest and Entergy Press Releases. Licensee notified the NRC Resident Inspector and the Massachusetts Emergency Management Agency. | |
| ENS 44509 | 22 September 2008 10:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center and Operational Support Center Unavailable | Unavailability of TSC/OSC Charcoal Filter for Scheduled Maintenance. At 0600 on Monday, September 22, 2008, the Pilgrim Nuclear Power Station (PNPS) Technical Support Center (TSC) / Operational Support Center (OSC) ventilation system charcoal was removed from service for planned charcoal bed maintenance. The balance of the TSC/OSC ventilation is not affected by the charcoal bed maintenance and remains available. Under certain accident conditions the TSC/OSC may become unavailable due to inability of the filtration system to maintain a habitable atmosphere. Compensatory measures exist to relocate TSC/OSC personnel to the Emergency Operations Facility (EOF) and to the Chiltonville Training Center. Charcoal filter maintenance is scheduled to be completed by 1700 hours on Monday September 22, 2008. The licensee will be notifying the NRC Senior Resident Inspector. This notification is being made in accordance with 10CFR50.72 (b)(3)(xiii) due to the loss of an emergency response facility.
The TSC/OSC ventilation system was returned to functional status on Monday, September 22, 2008. This follow-up notification is being made to provide closure from the initial notification under 10CFR50.72 (b)(3)(xiii) due to the loss of an emergency response facility. The licensee has notified the NRC Resident Inspector. The licensee also notified the State of Massachusetts Emergency Management Agency. Notified R1DO (Conte) | |
| ENS 44545 | 7 October 2008 02:24:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Unexpected Reactor Core Isolation Cooling Isolation During Test | During performance of PNPS Procedure 8.M.2-2.6.3 Attachment 1 step (65) the RCIC System isolated on a Group 5 signal when relay contact blocking devices (boots) were removed. All isolations went to completion. This isolation was not part of the planned evolution. The Group 5 isolation was reset and RCIC was placed in stand by line-up at 2327 on 10/6/2008. Investigation is continuing. The licensee notified the NRC Resident Inspector. | |
| ENS 44587 | 21 October 2008 23:44:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Failure of 250V Dc Hpci Injection Valve Undervoltage Relay | A 250 Volt DC undervoltage relay for HPCI injection valve MO-2301-8 failed. The reason for the failure is still under investigation. The injection valve is a normally closed valve that opens on an initiation signal. The failure of the under voltage relay would prevent the HPCI injection valve (MO-2301-8) from opening and would prevent HPCI from performing its safety function. Pilgrim Station has entered a 14 day LCO due to Technical Specification 3.5.C.2. The NRC Resident Inspector has been notified. | 05000293/LER-2008-004 |
| ENS 44593 | 22 October 2008 16:17:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Rcic Declared Inoperable Due to Aging Concern of Several Flow Controller Components | On October 22, 2008, at 1217 hours, with the reactor at 100% core thermal power and steady state conditions, Pilgrim Nuclear Power Station (PNPS) conservatively declared the Reactor Core Isolation Cooling System (RCIC) inoperable in response to a concern regarding the reliability of aged capacitors that are installed in the RCIC flow controller. As background, the RCIC flow controller was calibrated and successfully tested on October 7th, 2008 as part of normal surveillance activities, however several of the capacitors installed in the controller were noted to be between 21 to 30 years of age. Industry recommended replacement interval for the capacitors is typically between 7 to 10 years of age. PNPS engineering review in conjunction with Entergy fleet consultation concluded today (10/22) that there was no definitive technical bases to provide a reasonable expectation that the RCIC flow controller function can be assured throughout it's mission time due to the capacitor aging concern. Therefore, RCIC was declared inoperable and a 14 day limiting condition for operability action statement was entered in accordance with TS 3.5.D.1. A replacement controller is being prepared for installation, with post maintenance testing projected to be completed by 2100 hours this evening. Ultimately the suspect controller will be the subject of further evaluation and this notification will be updated as appropriate. This notification has no impact on the health and safety of the public. The NRC Senior Resident Inspector is onsite and has been notified. This is an 8 hour notification made in accordance with 50.72(b)(3)(v)(D).
Basis for Retraction: Event Notification 44593 was conservatively made to ensure that the eight-hour non-emergency reporting requirements of 10 CFR 50.72 were satisfied pending the evaluation of RCIC System operability. On 10/22/08, RCIC flow controller FIC-1340-1 was declared inoperable due to engineering uncertainty for controller operability. The controller's electrolytic capacitors appeared to be aged beyond the expected useful life, and the resultant degrading power supply voltage indicated that the controller may not operate for the required FSAR mission time of eight hours. The controller was replaced on 10/23/08 with a refurbished controller and subsequent post-maintenance RCIC system flow testing demonstrated RCIC system operability. The controller that was removed from service was evaluated. Controller bench testing was performed on 11/6 and 11/7, 2008. This testing demonstrated that the controller could provide a full demand output signal for a minimum of 15 continuous hours. During this testing, it was also determined that the power supply output voltage was not degrading. Based on this post-service controller testing, and the successful in-service RCIC flow controller calibration and system performance test conducted on 10/07/08, the controller was operable when installed. The RCIC system was capable of performing its intended safety functions and would have started and supplied design basis flow to the reactor vessel under design basis conditions. Thus there would have (been) no impact on nuclear safety. Therefore, this event was not reportable pursuant to 10CFR50.72(b)(3)(v)(D). Event Number 44593, made on 10/22/2008, is being retracted. The licensee notified the NRC Resident Inspector. Notified R1DO (Bellamy). | |
| ENS 44611 | 30 October 2008 00:50:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event - Fire in Hp Calibration Lab | At 2050 EDT a fire was discovered on the 2nd floor of an outbuilding inside a Health Physics calibration lab inside the protected area. The fire was reported as starting in a "drawer or cabinet" in the room. Plymouth Fire Department was notified and responded along with the on-site fire brigade. An Unusual Event was declared at 2058 EDT. The fire was extinguished at 2109 EDT. A re-flash watch was established. Electrical breakers to the room were opened. Health Physics personnel determined that no sources had been impacted by the fire and that no radiological concerns existed. The licensee notified the Commonwealth of Massachusetts and local authorities. The licensee notified the NRC Resident Inspector.
The licensee has terminated the Unusual Event as of 2202 EDT on 10/29/2008. The licensee notified the NRC Resident Inspector. | |
| ENS 44672 | 20 November 2008 21:57:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Unexpected High Pressure Coolant Injection Isolation During Test | The High Pressure Coolant Injection (HPCI) system was declared inoperable on 11/20/08 at 1657 EST due to a Group 4 isolation signal generated during scheduled surveillance testing in accordance with PNPS (Pilgrim Nuclear Power Station) Procedure 8.M.2-2.5.3, Attachment 1. The HPCI testing was stopped to determine the cause of the isolation which was not part of the planned evolution. HPCI isolation was reset and HPCI was restored to standby lineup at 1804 EST. This event is an eight-hour notification. Efforts are ongoing to determine the cause of the error during testing. This event had no adverse effect to the health and/or safety of the public. The licensee has notified the NRC Resident Inspector of this event. This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(D) due to the loss of a single train system required to mitigate the consequences of an accident. | 05000293/LER-2008-005 |
| ENS 44735 | 19 December 2008 23:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Load Reject | On December 19, 2008, at 1831 hours with the reactor at 100% core thermal power (CTP) an automatic reactor scram occurred. It appears the scram occurred as a result of a load reject experienced during a severe winter storm. Three of the four safety relief valves opened in response to the event. Primary Containment Isolation System (PCIS) Group 2 (sample valves) and Group 6 (reactor water cleanup system) and the reactor building isolation system (RBIS-secondary containment) isolated as designed (all have since been restored) on the reactor water level +12 inches setpoint due to normal vessel shrink. Initial review indicates all safety-related systems responded as designed. Off-site power has been maintained; however the two 4kv safety related buses (A5&A6) were conservatively placed on the station emergency diesel generators due to potential grid stability concern. Currently, the reactor mode select switch (RMSS) is in the Shutdown position; the reactor scram has been reset; the reactor is being maintained at approximately 940 psig, with pressure being maintained by the main turbine by-pass valves; and reactor water level is being maintained at normal levels of 29 inches with the main condensate and feedwater system. Station maintenance personnel are assessing the status of the station switchyard. Once this activity is complete the duration of the forced outage and recovery plans will be determined. All control rods fully inserted on the scram. The licensee has notified the NRC Resident Inspector. | |
| ENS 44737 | 20 December 2008 15:45:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Momentary Loss of 345Kv Offsite Power | At approximately 1045 hours on Saturday, December 20, 2008, while in a Hot Shutdown condition, Pilgrim Station experienced a momentary loss of all 345kV off-site power to the Startup Transformer (SUT). As a result, the following safety system actuations occurred: automatic Reactor Protection System (RPS) actuation (all control rods were previously inserted), automatic start of both Emergency Diesel Generators (EDG) and loading of their respective emergency buses, automatic actuation of Primary Containment Isolation Systems (PCIS) Groups I, II, VI and Reactor Building Ventilation, manual initiation of the High Pressure Coolant Injection (HPCI) system for reactor pressure control, and manual initiation of the Reactor Core Isolation Cooling (RCIC) system for reactor level control. All systems functioned as designed and expected. Because power from an off-site source remained available from the 23kV Shutdown Transformer, the criteria for declaration of an Unusual Event were not met (Pilgrim EAL 6.3.2.1). 345kV power was automatically restored to the Startup Transformer via breaker reclose logic. Restoration of normal on-site power to all 4kV buses is complete and efforts to return systems to normal shutdown alignments are in progress. Current plans are in place to maneuver the plant to a Cold Shutdown condition as necessary to support recovery actions. The NRC Resident was on-site at the time of the event and has been notified. There was no threat to the health or safety of the public as a result of this event. The loss of power was caused by a breaker icing in the onsite switchyard. The switchyard breakers were inspected for icing. The licensee notified the State of Massachusetts. | 05000293/LER-2008-007 |
| ENS 44998 | 18 April 2009 05:20:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Courtesy Notification Made to State of Massachusetts Involving an Employee Fatality | On April 18, 2009 an informational notification was made to the US NRC. At approximately 0120 EDT on April I8, 2009, the Pilgrim main control room received an emergency medical call stating that a Pilgrim employee had been found unconscious inside the radiologically controlled area of the plant. The individual was treated by the on-site first responders who administered CPR and AED treatment. Off site medical assistance was requested and the individual was transported to the local hospital. At approximately 0230 EDT, Pilgrim was notified that the individual had died of an apparent heart attack. A courtesy notification was made to the Massachusetts Emergency Management Agency. The NRC Resident Inspector has been notified. The event had no adverse effect to the health and safety of the public. The licensee confirmed that the employee transported offsite was not radiologically contaminated. No press release is planned.
While Pilgrim does not plan to issue a news release regarding this event, it is anticipated that media coverage of the event will ensue based upon the off-site response to the medical emergency. This 10 CFR 50.72(b)(3)(xi) notification is conservatively being made in advance of any media attention. Notified R1DO (Powell). | |
| ENS 45219 | 21 July 2009 18:05:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Response Capability - Tsc/Osc Is Unavailable Due to Hvac System Trouble | Unavailability of TSC/OSC Heating, Ventilation and Air Conditioning (HVAC) System. At 1405 hours on Wednesday, July 21, 2009, the Pilgrim Nuclear Power Station (PNPS) Technical Support Center (TSC) / Operations Support Center (OSC) HVAC system was discovered to be nonfunctional. During initial troubleshooting, the breaker providing power to the supply fan mechanically tripped and will not be reset until troubleshooting certifies the breaker is acceptable for use. This event occurred during scheduled preventative maintenance (PM) of the system. Under certain accident conditions, the TSC/OSC may become unavailable due to inability of the filtration system to maintain a habitable atmosphere. Compensatory measures exist to relocate TSC/OSC personnel to alternate locations. The licensee has notified the NRC Senior Resident Inspector/Resident Inspector. This notification is being made in accordance with 10CFR50.72(b)(3)(xiii) due to the loss of an emergency response facility. The licensee notified the Commonwealth of Massachusetts Emergency Management Agency.
This is a follow-up courtesy notification to EN #45219. All corrective maintenance activities on the TSC/OSC HVAC system are complete and the TSC/OSC is now functional and available for use. The NRC Resident has been notified. The licensee will inform the Commonwealth of Massachusetts Emergency Management Agency. Notified R1DO (Rogge). | |
| ENS 45586 | 22 December 2009 13:45:00 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Loss of Water Seal to Secondary Containment | On December 22, 2009, at 0845 hours, with the reactor at 100% core thermal power and steady state conditions, Pilgrim Nuclear Power Station (PNPS) declared Secondary Containment inoperable due to the loss of a water seal on two 14 inch drain lines in the Torus Compartment designed to mitigate the consequences of a flood in the Reactor Auxiliary Bay. An initial assessment of the condition indicates that the cross-sectional area of pipes, as found, have the potential to exceed the analytical value of allowable Secondary Containment leakage pathway size documented in the design calculations. The Limiting Condition for Operation (LCO) for Technical Specification (TS) 3.7.C.2.a, was immediately entered at 0845 until the condition was corrected. The TS LCO was subsequently exited at 0945 hours at which time Secondary Containment was declared operable. As background, each Reactor Auxiliary Bay (2) is equipped with a sump containing two 14 inch drain lines that discharge into separate trough in the Torus Compartment in the event of a flooding condition. The trough is approximately 4 feet by 4 feet by 4 feet deep and is equipped with a high and low level alarm which annunciates in the Control Room. One of the functions of the alarm is to indicate a low level condition prior to the pipes losing their water seal which maintains Secondary Containment for those penetrations. When the water level in the trough dropped below the low level setpoint, the alarm failed to annunciate. This condition was discovered during an engineering walk down of the Torus Compartment. Immediate actions taken were to refill the trough to the correct level, initiate repairs to the level switch, which was found to be defective, and to enhance the weekly tour requirement of the Torus Compartment performed by plant operations. This notification has no impact on the health and safety of the public. The NRC Resident Inspector is onsite and has been notified. This is an 8 hour notification made in accordance with 50.72(b)(3)(v)(C). | |
| ENS 45793 | 26 March 2010 02:55:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Standby Gas Treatment Declared Inoperable After Discovery of Open Demister Door | System Affected: Standby Gas Treatment (SBGT) Condition: Demister Door Discovered open rendering both trains of SBGT INOP At 2255 (EDT) on 3/25/2010, a Demister Door on 'B' Train of Standby Gas Treatment (SBGT) was found open. The door was closed upon discovery. The door was opened during performance of a scheduled surveillance during the dayshift and it appears that the door was not closed at the completion of work. With this door being open the 'B' Train of SBGT was INOP. Due to the physical configuration of the SBGT System it cannot be immediately verified that the 'A' Train of SBGT would have been able to perform its Safety Related Function since there is a probability that it could have drawn suction through the demister door of the 'B' Train and a normally open crosstie between the Trains. A 10 CFR 50.72 notification is conservatively being made based on the information currently available. This condition has no impact on the health and safety of the public. Both trains of SBGT are currently operable. The licensee informed the NRC Resident Inspector. | |
| ENS 45840 | 14 April 2010 21:03:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Response Organization In-Plant Notification System | At 1703 EDT on Wednesday, April 14, 2010, the Pilgrim Nuclear Power Station (PNPS) determined that the in-plant paging/notification systems used to notify the Emergency Response Organization (ERO) were not functioning as designed or required by the Emergency Plan. Specifically, the in-plant notification system which is comprised of the Computerized Automated Notification System (CANS) and back-up system (BEEPS) could not be determined to function reliably or consistently during an Emergency Plan drill or event. Compensatory measures exist to contact members of the ERO by procedure via call-trees through the operations and security organizations at the site. Immediate actions are being taken to restore the system to functional status working with the service provider for the in-plant paging/notification systems. It cannot be determined at the time of this notification as to when the system will be restored to full, functional status. The licensee has notified the NRC Senior Resident Inspector. | |
| ENS 45879 | 28 April 2010 18:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Rcic Declared Inoperable Due to Oil Leak on Governor System | On 04/28/10, at 1400 EDT, with the reactor at 100% power, the Reactor Core Isolation Cooling (RCIC) system was declared inoperable by the Shift Manager (SM) due to an oil leak on the RCIC governor control oil system that could have impacted the system performance during the accredited 24 hour mission time. The fitting where the oil leakage was observed was tightened and the machine was placed in service with no leakage identified. Currently the system is operable and in its normal standby lineup. The system was available for use during this time. At no time was there an impact to the health and safety of the public. The licensee has notified the NRC Resident Inspector.
On April 28, 2010, at 1940 hours, Pilgrim Nuclear Power Station (PNPS) made an 8-hour non-emergency 50.72 notification, Event Notification EN# 45879. The notification was made in accordance with 50.72 (b)(3)(v)(D), Accident Mitigation. Earlier on April 28, 2010, at 1400 hours, a minor oil leak had been identified on the Reactor Core Isolation Cooling (RCIC) system at a lubricating oil vent fitting. The leak was immediately repaired by properly tightening the fitting, then running RCIC to verify no active leak existed. However in the interim, the Shift Manager conservatively declared RCIC inoperable when the high standard for operability could not be assured by initial system engineering judgment for the impact of the oil leak on RCIC system performance in consideration of mission time. Subsequent engineering evaluation concluded that the observed leak, conservatively assumed to be one drop per 3 minutes, would not have impacted RCIC operability for the duration of its required 24 hour mission time. All relevant technical information is documented in the PNPS corrective action system. Therefore PNPS is retracting the event notification EN# 45879. The USNRC Resident Inspector Office has been notified of this retraction. Notified R1DO (Dwyer). | |
| ENS 46083 | 9 July 2010 04:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Elevated Levels of Tritium Found in a Well Sample | Entergy Pilgrim Station has twelve groundwater monitoring wells used to sample for tritium and other radioactive nuclides in accordance with the Nuclear Energy Institute's (NEI) voluntary Groundwater Protection Initiative (GPI). One of the wells, MW-205, located in the vicinity of the Condensate Storage Tanks (CST) indicated an elevated level of tritium, however well below the limits established by the NEI Groundwater Protection Initiative, the Nuclear Regulatory Commission's (NRC) limits for liquid effluent release and the Environmental Protection Agency's (EPA) limits for tritium in drinking or non-drinking water wells. The latest sample taken on June 21, 2010, returned a test result of 11,072 picocuries per liter of tritium. To date, tritium is the only isotope detected in the samples collected at the site. This information has been communicated to federal, state and local stakeholders and a press advisory is expected to be issued by the Massachusetts Department of Public Health (MDPH). On that basis and the anticipated interest to the general public this notification is being made. This event has no impact on the health and/or safety of the public. The NRC Resident Inspector is on-site and has been notified.
Entergy Pilgrim Station has received the results of its most recent weekly tritium sample taken on July 7, 2010 for groundwater monitoring well, MW-205. The sample results have shown an increase in the tritium concentration to 25,552 picocuries per liter (pCi/L) from the previous sample taken on June 30, 2010 which had a test result of 8,477 pCi/L. The latest results remain below any regulatory reporting requirements and the Environmental Protection Agency's (EPA) limits for tritium in non-drinking water wells. This information has been communicated to federal, state and local stakeholders. There remains no threat to drinking water sources and no impact on the health and/or safety of the public. The NRC Resident Inspector is on-site and has been notified of this update. This is an update to the 4-hour non-emergency notification made in accordance with 50.72(b)(2)(xi) on July 9, 2010 at 1555 hours. Notified R1DO (Doerflein). | |
| ENS 46206 | 25 August 2010 12:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification - Release of Hydrogen Gas in Excess of the Reportable Quantity of Ten Pounds | On August 25, 2010 (the licensee) made a notification to) the Massachusetts Department of Environmental Protection and the Plymouth Massachusetts Fire Department in accordance with 310CMR40.300, Massachusetts Contingency Plan Notification of Oil and Hazardous Material; Identification and Listing of Oil and Hazardous Material. (The licensee made the notification) due to a release of hydrogen gas to the environment exceeding the reportable quantity of ten pounds via the designed system release path. The release occurred when restoration of a system vacuum pump following planned maintenance resulted in higher than normal system makeup rates. Combustible gas sampling of equipment areas confirmed that the excess hydrogen gas had been released to the environment via a remote roof vent designed to continuously exhaust (hydrogen) removed from the seal oil system by the vacuum pump. The release of hydrogen gas was secured when the vacuum pump was removed from service. The estimated quantity of hydrogen gas released was approximately twenty pounds. This event posed no danger to the health and safety of plant personnel or members of the general public. The NRC Senior Resident Inspector is on-site and has been notified. The licensee is investigating the cause of the excessive makeup rate. | |
| ENS 46271 | 23 September 2010 10:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Response Capability - Tsc/Osc Is Unavailable | Unavailability of TSC/OSC Ventilation System Due to Scheduled Maintenance. At 0600 (EDT) on Thursday, September 23, 2010, the Pilgrim Nuclear Power Station (PNPS) Technical Support Center (TSC) / Operations Support Center (OSC) emergency ventilation system was removed from service for planned maintenance. The balance of the TSC/OSC ventilation is not affected by the maintenance and remains available. Under certain accident conditions the TSC/OSC may become unavailable due to inability of the filtration system to maintain a habitable atmosphere. Compensatory measures exist to relocate TSC/OSC personnel to the Emergency Offsite Facility (EOF) and to the Control Room. In the event of a declared emergency, all OSC personnel will report to the Control Room Annex. TSC minimum staffing will report to the Control Room and all other TSC staff will report to the EOF. The maintenance is scheduled to be completed by 1900 hours on Thursday September 23, 2010. The licensee has notified the NRC Senior Resident Inspector. This notification is being made in accordance with 10 CFR 50.72 (b) (3) (xiii) due to the loss of an emergency response facility. The licensee also informed the Commonwealth of Massachusetts.
TSC/OSC ventilation was restored to operable status at 1600 EDT. The licensee notified the NRC Resident Inspector. Notified the R1DO (Dentel). | |
| ENS 46521 | 5 January 2011 06:20:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Reactor Core Isolation Cooling Declared Inoperable | On January 5, 2011, at 0120 hours, with the reactor at 100% thermal power and steady state conditions, Pilgrim Nuclear Power Station (PNSP) declared the Reactor Core Isolation Cooling (RCIC) system inoperable due to the RCIC suction isolation valve from the Torus/Suppression Pool (RCIC-26) failing to go fully closed during planned surveillance testing. The RCIC-26 is a motor-operated valve (MOV) and its normal position is closed. The RClC-26 valve is redundant to the RCIC-25 valve, and is not the credited containment isolation valve. The RCIC-26 valve has a safety function to be (manually) opened during certain event mitigation scenarios requiring a transfer of suction sources from the Condensate Storage Tank (CST) to the Torus. Based on the valve failing to fully close during MOV stroke time testing per PNPS Procedure 8.5.5.4, the RCIC system was declared inoperable at 0120 hours and the appropriate LCO was entered. The RCIC-26 was subsequently returned to a full open position, caution tagged and the RCIC system was declared operable. The LCO was exited at 0200 hours. An investigation of the event is underway and continuing. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector is on-site and has been notified. This is an 8-hour notification made in accordance with 50.72(b)(3)(v)(D). The licensee will notify the State of Massachusetts.
Event Notification 46521 was conservatively made to ensure that the Eight-Hour Non-Emergency reporting requirements of 10 CFR 50.72 were satisfied pending the evaluation of RCIC System operability. On 01/05/11, at 0120 hours the RCIC System was declared inoperable due to uncertainty of RCIC System Operability when the Torus/Suppression Pool Suction Valve (RCIC-26) failed to go fully closed during planned surveillance testing. The valve was restored to the full open position and the valve was declared operable based on capability to meet the required safety function to fully open when RCIC pump suction from the suppression pool is required. The apparent cause evaluation concluded that valve failure was the result of high relay contact resistance in the closing control circuit components of the valve breaker. This failure prevented the valve from fully closing but had no affect on capability to open the valve. Surveillance testing verified that capability to open the valve was not affected. Corrective action was completed to clean or replace the control circuit relay contacts. Post work testing confirmed capability to open and close the valve. An extent of condition for similar breaker control circuit components was also performed. All relevant technical information is documented in the corrective action system. The failure observed did not affect the valve's required safety function and did not impact RCIC System operability. Thus there was no impact on nuclear safety. This event is not reportable pursuant to 10 CFR 50.72(b)(3)(v)(D) . Event Number 46521, made on 01/05/2011, is being retracted. The licensee has notified the NRC Resident Inspector. Notified R1DO (Anthony Dimitriadis) | |
| ENS 46628 | 20 February 2011 05:00:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown-Reactor Building Closed Cooling Water Train Inoperable | At 0000 hours on Sunday, February 20, 2011, the Pilgrim Nuclear Power Station (PNPS) commenced a controlled shutdown of the reactor due to the 'B' train of Reactor Building Closed Cooling Water (RBCCW) being declared inoperable at 1250 hours on February 18, 2011 and expected to exceed its 72-hour Limiting Condition for Operability (LCO) as required by TS prior to return to operable status. With the plant operating at 100% power, leakage of Salt Service Water (SSW) was detected in the RBCCW system due to high chloride levels and increased inventory in the system. An investigation into the event determined that the source of the SSW was isolated to the 'B' RBCCW heat exchanger which is designed to cool RBCCW under normal and post-accident conditions. The quantity of the leakage was determined to exceed the design limits established to ensure post-accident operation of the system and the 'B' train of RBCCW was subsequently declared inoperable. The leak identification and repair activities are expected to exceed the TS LCO for the RBCCW system and therefore a controlled shutdown of the reactor was initiated. This event had no impact on the health and/or safety of the public. The licensee has notified the NRC Senior Resident Inspector. This 4-hour notification is being made in accordance with 10 CFR 50.72 (b)(2)(i). | |
| ENS 46631 | 20 February 2011 15:34:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System Actuation While Shutdown | On Sunday, February 20, 2011, at 1034 EST with the reactor shutdown and all control rods fully inserted a valid Reactor Protection System (RPS) low reactor water initiation signal (+12 inch) was received. The RPS actuation signal resulted in reactor scram and actuation of the Primary Containment Isolation Systems for Group II - Primary Containment Isolation and Reactor Building Isolation System (RBIS), and Group VI - Reactor Water Cleanup System (RWCU). At the time of the event, a controlled reactor shutdown and cooldown was in progress. The Reactor Mode Selector Switch was in 'Startup' and a cooldown was being controlled using the Mechanical Hydraulic Control (MHC) System. Initial event review indicates that the turbine by-pass valve controlling the cooldown closed causing a 'shrink' on indicated reactor water level to briefly lower to +12 inches (lowest observed water level). Reactor water level was immediately restored, the isolations (Group II and VI) were reset, and the RPS signal was reset at 1135 EST. All systems operated as expected, in accordance with design. This event had no impact on the health and/or safety of the public. The US NRC Senior Resident Inspector was in the Main Control Room at the time of the event. The licensee will be notifying the State. | |
| ENS 46837 | 10 May 2011 17:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Protection System Actuation During Start-Up | At 1315 hours on Tuesday, May 10, 2011, with the Reactor Mode Select Switch in Startup, the Reactor Protection System (RPS) actuated due to a valid Hi-Hi trip signal from the Intermediate Range Monitors (IRMs) of the Neutron Monitoring System. At the time of the RPS actuation, reactor thermal power was four (4)% and control rods were being withdrawn as part of a planned reactor startup from a refueling outage. The RPS actuation signal resulted in a reactor scram. The IRMs were in range 7 when the scram occurred. All control rods automatically inserted in accordance with design. No other ESF actuation signals were generated. The reactor scram was not part of pre-planned test or reactor operation. The cause of the Hi-Hi IRM trip signal is under investigation. This event had no impact on the health and/or safety of the public. The licensee has notified the NRC Senior Resident Inspector. This 4-hour notification is being made in accordance with 10CFR50.72(b)(2)(iv)(B). The unit is in a normal post trip electrical alignment. Decay heat is being removed via the main steam line drains to the main condenser. The licensee will notify the Massachusetts Emergency Management Agency. | |
| ENS 46852 | 14 May 2011 06:45:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to Inability to Maintain Specified Differential Pressure Between Drywell and Suppression Chamber | Following startup from RFO (Refueling Outage) 18, at approximately 14% reactor power, PNPS (Pilgrim Nuclear Power Station) was unable to set the conditions required to demonstrate the drywell to suppression chamber decay rate is less than 25% of the differential pressure decay rate for the maximum allowable bypass area of 0.2 ft2 (as required by) TS (Technical Specification) 3.7.A.4(b). Prior to exceeding 15% power, PNPS Technical Specifications require differential pressure between the drywell and suppression chamber to be greater than 1.17 psid. The licensee was not able to achieve this delta pressure and initiated troubleshooting to determine the cause. A plant shutdown is being initiated to continue this effort. The plant is in a stable condition. Investigation is continuing into the failure to establish test conditions. There is no threat to the health and safety of the public as a result of this condition. The licensee suspects leakby in the Drywell to Torus vacuum breaker but will not be able to verify until the plant is shutdown. The licensee has notified the NRC Resident Inspector and will be notifying the State of Massachusetts. | |
| ENS 46933 | 7 June 2011 12:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Errors in Methodology Used for Emergency Core Cooling System Performance Requirements | On Tuesday, June 7, 2011, at 0800 hours, Pilgrim Station reviewed General Electric Hitachi (GEH) 10 CFR 50.46 Notification Letters 2011-02 and 2011-03. These letters indicate that certain errors were discovered in the methodology application and inputs used by GEH for nuclear fuel core configurations with GE14 and GNF2 fuel, and when corrected may increase the Peak Cladding Temperature (PCT) limits in excess of 2200 degrees F under Loss of Coolant Accident (LOCA) conditions. Pilgrim's core contains both GE14 and GNF2 fuel. 10 CFR 50.46 paragraph (b) defines the acceptance criteria for the LOCA analysis process. The Pilgrim licensing basis PCT is evaluated for compliance with the criterion 50.46(b)(1) and must not exceed a PCT of 2200?F. GEH had provided a compensatory measure in the form of multiplier to be applied to the MAPLHGR (Maximum Average Planar Linear Heat Generation Rate) limit so that Pilgrim operates within 50.46 limits. Entergy/Pilgrim implemented the compensatory measure and as a result the errors reported have no impact on current plant operation or public health and safety. This 8-hour notification is being reported for conservative purposes in accordance with 10 CFR 50.72 (b)(3)(ii)(B). Based on 50.46(a)(3)(ii) criteria, Entergy/Pilgrim will submit a report within 30 days. Entergy/Pilgrim has notified the NRC Senior Resident Inspector." The licensee also plans to notify the State of Massachusetts. | |
| ENS 47199 | 26 August 2011 17:45:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Press Release Related to Hurricane Preparations at the Pilgrim Site | At approximately 1345 EDT, on August 26, 2011, Entergy Corporation issued a press release regarding ongoing preparations at Pilgrim Station for hurricane Irene currently located 300 miles SSW of Cape Hatteras, North Carolina traveling in a northerly direction at approximately 14 miles per hour. The purpose of the press release is to re-assure the public regarding the safety and security of the facility and the extent of preparations that are in progress. Current plant operation is not impacted and the release of this information does not indicate a current threat to the facility or the health and safety of the public. The licensee has notified the NRC Resident Inspector. This four (4) hour report is being made in accordance with 10 CFR 50.72(b)(2)(xi). State and local notifications will be made coincident with the Issuance of this Event Report. |