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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 57115 | 9 May 2024 12:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Containment Bypass Condition Due to Degraded River Water Piping | The following information was provided by the licensee via phone and email: At 0800 EDT on May 9, 2024, it was identified during leak rate testing that through-wall flaws existed on reactor plant river water piping inside the containment building. This determination resulted in a containment bypass condition such that a gaseous release could have occurred at a location not analyzed for a release in the loss of coolant accident dose consequence analysis. This condition is not bounded by existing design and licensing documents. Evaluation of the condition of the piping is ongoing to support repair prior to startup. With the plant currently in cold shutdown, the containment, as specified in Technical Specification 3.6.1, is not required to be operable. There was no impact on the health and safety of the public or plant personnel. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A), 10 CFR 50.72(b)(3)(ii)(B), and 10 CFR 50.72(b)(3)(v)(C). The NRC Resident Inspector has been notified. | ||
| ENS 56569 | 11 June 2023 05:30:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition | The following information was provided by the licensee via email: At 0130 EDT on June 11, 2023, it was discovered that the Beaver Valley Power Station, Unit No. 2 auxiliary building door A-35-5A, credited for tornado missile protection of the primary component cooling water system, was open and unlatched. Upon discovery, the door was shut and latched. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | ||
| ENS 56350 | 12 February 2023 13:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Control Room Emergency Ventilation System Inoperable | The following information was provided by the licensee via phone call and email: At 0800 on February 12, 2023, it was discovered that both trains of control room emergency ventilation system were simultaneously inoperable due to a safety injection relief valve discharging to a Unit 1 sump. This leakage in conjunction with design basis loss of coolant accident may result in radiological dose exceeding limits to the exclusion area boundary and to the control room, which is common to both Unit 1 and Unit 2. Therefore, this condition is being reported as an eight-hour, nonemergency notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(D) as an 'Unanalyzed Condition and a Condition that Could Have Prevented Fulfillment of a Safety Function.' There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
Retraction of EN56350, Control Room Emergency Ventilation System Inoperable: Based on subsequent evaluation, it was determined that the control room emergency ventilation system remained operable due to the maximum measured leak rate being within the bounds of the analysis. The maximum measured leak rate of 32,594 cc/hr from the safety injection system did not challenge the calculated maximum engineered safety features leak rate of 45,600 cc/hr and remained within the current dose analysis limits. As such, this was not an unanalyzed condition and did not prevent the fulfillment of a safety function to mitigate the consequences of an accident. The NRC Resident Inspector has been notified. Notified R1DO (Bickett). | Control Room Emergency Ventilation | |
| ENS 54366 | 2 November 2019 19:15:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Refueling Water Storage Tank Declared Inoperable Due to Low Level | At 1515 on November 2, 2019, the Refueling Water Storage Tank (RWST) was declared inoperable due to a Low Head Safety Injection relief valve discharging to the Safeguards Sump during routine surveillance testing. The leakage from the Low Head Safety Injection system in conjunction with a postulated Design Basis Accident (DBA) Loss of Coolant Accident (LOCA) with transfer to Safety Injection Recirculation may result in dose exceeding the Dose Analysis of the Exclusion Area Boundary (EAB) and the Control Room, which is common to both Unit 1 and Unit 2. This condition may not be bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. The Low Head Safety Injection relief valve has been isolated to prevent further leakage, and makeup to the RWST completed. At 1602 on November 2, 2019 the RWST was declared Operable. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(B), (C), (D) as an Unanalyzed Condition and a condition that could have prevented the Fulfillment of a Safety Function." The licensee notified the NRC resident inspector. | ||
| ENS 53548 | 12 August 2018 04:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | En Revision Imported Date 8/17/2018 | EN Revision Text: TECHNICAL SPECIFICATION REQUIRED SHUTDOWN - LOSS OF 480 VOLTAGE EMERGENCY BUS On 8-12-18 at 0158 EDT, Beaver Valley Unit 2 experienced a loss of 480 Volt 2P Emergency Bus. This resulted in a Loss of Safety Function due to the 2-2 Emergency Diesel Generator (EDG) being Inoperable coincident with the Residual Heat Release Valve (2SVS-HCV104). A Technical Specification shutdown is required per LCO 3.0.3. The Licensee also stated they were in an unanalyzed condition due to the EDG and Residual Heat Release Valve being inoperable at the same time. The Licensee is shutting down to Mode 5 (Cold Shutdown). The Licensee is notifying the Resident Inspector. The Licensee will be making a Press Release about the unplanned shutdown.
On 8-12-18 at 0158 (EDT) Beaver Valley Unit 2 experienced a loss of 480 Volt 2P Emergency Bus. Per operational guidance, this was determined to be a Loss of Safety Function due to the Unit 2 Emergency Diesel Generator (EDG) being INOPERABLE coincident with the Residual Heat Release Valve (2SVS-HCV104) 10 CFR 50.72(b)(3)(v)(B) and (D). This was also reported as an Unanalyzed Condition 10 CFR 50.72(b)(3)(ii)(b). No Press Release was performed for this event. The NRC Resident Inspector was notified. At 0410 (EDT) a Technical Specification Shutdown was commenced 10 CFR 50.72(b)(2)(i). At 2011 (EDT) the 480 Volt 2P Emergency Bus was restored and energized. Further evaluation of the event has determined that this event was not an Unanalyzed Condition and did not result in a Loss of Safety Function. The classifications of Unanalyzed Condition and Loss of Safety Function are being retracted. The accuracy of the existing guidance relative to Safety Function has been entered in the Corrective Action Program and interim actions have been taken to provide accurate guidance. Notified R1DO (Young) via email. | Emergency Diesel Generator | |
| ENS 53419 | 22 May 2018 04:00:00 | 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 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | En Revision Imported Date 6/22/2018 | EN Revision Text: GAS VOIDS DISCOVERED IN BOTH TRAINS OF LOW HEAD SAFETY INJECTION On 5/22/2018, while operating at approximately 100 percent power, Ultrasonic Testing of the Beaver Valley Power Station (BVPS) Unit 1 Low Head Safety Injection (LHSI) pump suction piping identified gas voids in excess of the acceptable limit for void volume. Both trains of LHSI were declared inoperable. Technical Specification (TS) 3.5.2 for both trains of the LHSI system was entered along with TS 3.0.3 which requires the initiation of a plant shutdown. Time of TS entry was 12:56 (EDT). Plant shutdown was commenced at 15:56 (EDT) in accordance with plant procedures. At 15:59 (EDT) Train 'A' LHSI was restored to operable status, TS 3.0.3 Action was exited and the power reduction was stopped at approximately 99 percent. At 17:43 (EDT) Train 'B' LHSI was restored to operable status, TS 3.5.2 Actions were exited. This is reportable per 10 CFR 50.72(b)(3)(ii) Unanalyzed Condition, 10 CFR 50.72(b)(3)(v) Event or Condition that Could Have Prevented the Fulfillment of a Safety Function and 10 CFR 50. 72(b )(2)(i) TS Required Shutdown. The NRC Resident Inspector has been notified.
Further engineering evaluation has determined that the gas voids that existed at the time of discovery would not have rendered the LHSI (Low Head Safety Injection) system inoperable if it were required to actuate. The engineering evaluation concluded that filling of the containment sump during a Design Basis Accident would result in a void volume reduction such that the void in the LHSI suction piping would not be large enough to significantly impact the operability of the system. Therefore, the system remained operable but degraded. No TSs (Technical Specifications) were required to be entered and no shutdown was required. As such, all three reporting criteria do not apply and are being retracted. The NRC Resident Inspector has been notified. Notified R1DO (Burritt). | ||
| ENS 52571 | 23 February 2017 15:40:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Potential Tornado Missile Vulnerabilities | In order to address the concerns outlined in RIS (Regulatory Information Summary) 2015-06 'TORNADO MISSILE PROTECTION', an evaluation of tornado missile vulnerabilities and their potential impact on Technical Specification (TS) plant equipment was conducted. This evaluation concluded that the following Structures, Systems, and Components (SSCs) are potentially vulnerable to tornado generated missiles: The BVPS Unit 1 (BV-1) and BVPS Unit 2 (BV-2) Main Steam Safety Valve (MSSVs) discharge flow paths to atmosphere (reference TS 3.7.1) are potentially vulnerable to tornado generated missiles. A tornado could generate missiles capable of striking the exhaust piping of the MSSVs potentially crimping the piping and resulting in reduced flow capacity. In the worst case, all MSSV's could be rendered inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B), and 10 CFR 50.72(b)(3)(v)(A). The BV-1 and BV-2 Atmospheric Steam Dumps (ADVs) discharge flow paths to atmosphere (reference TS 3.7.4) are potentially vulnerable to tornado generated missiles. A tornado could generate missiles capable of striking the exhaust piping of the ADVs potentially crimping the piping and resulting in reduced flow capacity. In the worst case, all ADVs could be rendered inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B), and 10 CFR 50.72(b)(3)(v)(A). The BV-2 Auxiliary Building tornado missile shield door (A-35-5A), credited for tornado missile protection of the Primary Component Cooling Water (PCCW) system, was found to not be fully closed and latched. A tornado could generate missiles capable of striking the PCCW system with the missile door open rendering both trains of the PCCW system inoperable. This door is now maintained closed and latched except when opened under administrative controls. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B). The BV-2 Spent Fuel Building tornado missile shield door (F-66-3), credited for tornado missile protection of the irradiated fuel assemblies in the Spent Fuel Pool including the Cask Pit, was found to not be fully closed and latched. This door is now maintained closed and latched except when opened under administrative controls. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B). The potential tornado missile vulnerabilities for the MSSVs and ADVs (discussed above) are being addressed in accordance with EGM-15-002 Rev 1 and DSS-ISG-2016-01 (NRC enforcement discretion and interim guidance documents). Immediate compensatory measures were taken to reduce the likelihood and mitigate the potential consequences of an onsite tornado generated missiles. The NRC Resident inspector has been notified. | Main Steam Safety Valve | 05000334/LER-2017-002 05000334/LER-2017-001 |
| ENS 52132 | 29 July 2016 01:20:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Fire Event That Could Adversely Impact Safe Shutdown Equipment | A review of the Beaver Valley Unit 2 Fire Protection Safe Shutdown Report (FPSSR) found that a postulated fire had the potential to spuriously open all three individual steam generator atmospheric dump valves in addition to a common residual heat release valve. Previous analysis did not consider all of the valves spuriously opening from a fire. The potential impact of these valves spuriously opening is a cooldown that could adversely affect shutdown margin. Hourly fire tours have been put in place for those fire areas that have the potential to initiate this condition. This condition is reportable as an 8 hour report in accordance with 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector was notified. This condition is not applicable to Unit 1.
Retraction of EN 52132 'Postulated Fire Event That Could Adversely Impact Safe Shutdown Equipment'. On 07/29/2016, an 8-hour notification (EN 52132) was made describing the discovery of a postulated fire event that could adversely impact safe shutdown equipment, specifically the spurious opening of the three Atmospheric Steam Dump Valves (ASDVs) and the Residual Heat Release (RHR) Valve, simultaneously. This notification was made pursuant to 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition because of the potential effect on shutdown margin. Further engineering evaluation has determined that the required shutdown margin is not challenged by the event, as bounded by previous analysis of a similar scenario as well as validated operator actions. Therefore, this does not result in a reportable condition. The Resident NRC Inspector has been notified. Notified the R1DO (Cook). | Steam Generator | |
| ENS 50075 | 30 April 2014 15:40:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Hot Short Fire Event Could Adversely Impact Safe Shutdown Equipment | Based on a review of industry operating experience, it was identified that each unit has two un-fused DC control circuits for non safety-related DC motors which are routed from the turbine building through other separate fire areas including the Control Room. The DC breakers used to protect the motor power conductors are insufficient to protect the control conductors for these circuits. It is postulated that a fire induced short in one fire area could adversely impact safe shutdown equipment by overheating the cable and causing a secondary fire in other fire areas where the cable is routed. At Unit 1, cables for the affected circuits are routed in the Turbine Building, Cable Spreading Area and Control Room. At Unit 2, cables for the affected circuits are routed in the Turbine Building, Normal Switchgear, Service Building Cable Tray Area, Cable Vault, Instrument Relay Room, Control Building West Communication Room, Control Building Cable Spreading Area and Control Room. The postulated secondary fires or cable failures are outside the assumptions of each unit's fire protection analysis. A preliminary investigation of the issue indicates that existing fire protection safe shutdown procedures could be used to safely shut down the plant if needed. This condition is reportable as an 8-hour report in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures will be implemented for affected areas of the plant. The NRC Resident Inspector has been notified. | ||
| ENS 47403 | 2 November 2011 18:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Caused by Non-Conforming Fire Barriers | Ongoing field inspections as a result of Industry Operating Experience have identified certain fire barriers that are not in conformance with required tested configurations. Specifically, some of the flexible conduit that penetrates these barriers has a coating that does not exhibit flame retardant characteristics and therefore does not meet the requirements for fire barriers at Beaver Valley Power Station, Unit 1. In the event of a postulated fire, this non-conformance has the potential to affect fire barriers separating the two independent trains required for post fire Safe Shutdown equipment. This issue is being reported per 10CFR 50.72(b)(3)(ii)(B). Compensatory actions have been established in accordance with the approved Fire Protection Program. The NRC Resident Inspector has been notified. | 05000334/LER-2011-001 | |
| ENS 46744 | 10 April 2011 07:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Manual Reactor Trip While Performing a Technical Specification Required Shutdown | On April 09, 2011, at 2349 (EDT), Beaver Valley Power Station (BVPS) Unit No. 2 was operating at 15% power while preparing to synchronize the main unit generator to the grid. At that time, the 'A' Auxiliary Feedwater Injection Header was declared inoperable due to a water leak identified from a vent valve fillet weld between the inside and outside Containment Isolation Valves (outside of containment) for containment penetration X-79. In accordance with Technical Specification 3.7.5, Auxiliary Feedwater (AFW) System, Condition D, at 0345 (EDT), April 10, 2011, BVPS Unit 2 commenced a Reactor Shutdown to Mode 3. Required action is to be in Mode 3 within 6 hours. This event is being reported as a Technical Specification required shutdown pursuant to 10CFR50.72(b)(2)(i), 4 hour notification. Repairs are in progress. The following additional shutdown actions may be required from the time the Injection Header/Containment Penetration was declared inoperable: Technical Specification 3.7.5, Condition D, Mode 4 in 18 hours and Technical Specification 3.6.1, Condition A, Mode 5 within 37 hours. This event is also being reported as a degraded condition for Containment pursuant to 10CFR50.72(b)(3)(ii)(A), 8 hour notification. Additionally, at 0357 (EDT), during the Reactor Shutdown, at 4.6% Reactor Power, the BVPS Unit 2 Reactor was manually tripped due to reaching a pre-established manual trip criteria of 25% Steam Generator Level for the 21A Steam Generator. This was conservative criteria set above automatic actuation setpoint of 20.5% level. This event is being reported as a RPS Actuation pursuant to 10CFR50.72(b)(2)(iv)(B), 4 hour notification. Control room personnel entered Emergency Operating Procedure E-0, 'Response to Reactor Trip and Safety Injection.' Safety systems and equipment functioned as designed following the manual reactor trip. Due to the cooldown and subsequent shrink of level in the 21A Steam Generator, an automatic start of the Steam Driven Auxiliary Feedwater Pump (2FWE-P22) occurred at 20.5%. This event is being reported as an Auxiliary Feedwater System Actuation pursuant to 10CFR50.72(b)(3)(iv)(A), 8 hour notification. All control rods fully inserted into the core. The plant electrical system is aligned to normal offsite power sources. Decay heat from the reactor coolant pumps is being directed to atmospheric dump valve. There is no primary to secondary leakage. There was no impact on Unit 1. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater Control Rod | |
| ENS 46421 | 16 November 2010 05:01:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Control Room Envelope Inoperable Due to Excessive Emergency Core Cooling System Relief Valve Leakage During Eccs Pump Surveillance Testing | Emergency Core Cooling System Control Room Envelope | 05000334/LER-2010-003 | |
| ENS 46208 | 26 August 2010 20:49:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Both Trains of High Head Safety Injection Pumps Declared Inoperable | Beaver Valley Power Station Unit No.1, while operating at 100% full power, performed Ultrasonic Testing on the High-Head Emergency Core Cooling System Pump suction headers (6" and 8") following a fill and vent of the 'A' High Head Emergency Core Cooling System Pump. This is required because the fill point for the out of service Emergency Core Cooling System Pump 1CH-P-1A, is it's suction valve from the 8" Charging header, 1CH-19, which was opened. Preliminary indications were such that the 6" suction header was full of water and an air void did exist in the 8" suction header, but the size was indeterminate and had to be calculated by System Engineering. Operations commenced additional monitoring for cavitation of the running charging pump with none identified. Operations then vented the 8" suction header multiple times. At 1649 hours today, the results of the Ultrasonic Test (UT) was provided by System Engineering indicating that an air void existed in the Emergency Core Cooling System Pump 8" Suction header that was in excess of the Acceptance Criteria. At 1649 hours, both trains of High Head Safety Injection pumps were declared Inoperable but remain Available. Technical Specification (TS) 3.5.2 - ECCS operating is not met. Required Action C.1 requires entry of TS LCO 3.0.3 immediately. Per 10CFR50.72(b)(3)(v)(A)&(D) - Event or Condition That Could Have Prevented Fulfillment of a Safety Function, and 10CFR50.72(b)(3)(ii)(B) Unanalyzed Condition, this event is reportable to the NRC within 8 hours. At 1715 additional venting of the 6" Charging Pump suction header revealed no air present. Additional venting of the 8" Charging Pump suction header revealed additional air pockets. The 8" header was then vented multiple times, with a short delay in between each venting, until no air was identified. At 1718, follow up UT on both Charging Pump Suction headers revealed it remained full of water with no voids present. Both trains of High Head Safety Injection are declared OPERABLE restoring compliance to TS LCO 3.0.3 and 3.5.2. Reactor power remained at 100% during these evolutions. The NRC Resident Inspector was informed. | Emergency Core Cooling System | 05000334/LER-2010-001 |
| ENS 44597 | 23 October 2008 22:00:00 | Information Only 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Information Only Notification Regarding Plant Shutdown | This is an information only notification that Beaver Valley Power Station Unit No. 2 has initiated a plant shutdown as of 1800 on 10/23/08, a proactive measure, to complete repairs on the Train 'A' Low Head Safety Injection (LHSI) Pump 2SIS-P21A. On 10/19/08 at 2326, Beaver Valley Power Station Unit No. 2 removed 2SIS-P21A from service for routine preventive maintenance. Technical Specification 3.5.2, Condition A was entered during this maintenance activity with the Required Action A.1 to restore 2SIS-P21A to operable status within 72 hours (2326 on 10/22/08). On 10/20/08, during this maintenance; the pump shaft was unable to be rotated by hand after approximately 3/4 revolution. The pump was then disassembled and the apparent cause was determined to be that the pump rotating assembly was not centralized in the casing bore resulting in minimal clearance between the pump wear rings. On 10/22/08, First Energy Nuclear Operating Company (FENOC) requested enforcement discretion from the NRC to permit an additional 36 hours of time to complete repairs on the pump. The NRC verbally granted FENOC's request for enforcement discretion on 10/22/08 at 1105. The enforcement discretion period will expire on 10/24/08 at 1126. Since unexpected difficulties have delayed the reassembly of the pump, FENOC management decided to proactively shutdown Beaver Valley Power Station Unit No.2 beginning at 1800 on 10/23/08. While pump reassembly and testing to restore operability may still be completed by the end of the enforcement discretion period, FENOC management determined that it was more prudent to shutdown the unit at this time so that the Operations crews are not placed under undue time pressure to reach Mode 3 conditions. The licensee notified the NRC Resident Inspector.
This 4-hour notification is to report that Beaver Valley Power Station Unit No. 2 has initiated a manual reactor shutdown at 0600 on 10/24/08 in accordance with Technical Specification 3.5.2, 'ECCS - Operating'. On 10/19/08 at 2326 Beaver Valley Power Station Unit No. 2 removed 2SIS-P21A (Train 'A' Low Head Safety Injection (LHSI) Pump) from service for routine preventive maintenance. Technical Specification 3.5.2 was not met during this maintenance activity for two operable trains of Emergency Core Cooing System and Condition A was entered with the Required Action A.1 to restore 2SIS-P21A to operable status within 72 hours (2326 on 10/22/08). Problems were unexpectedly encountered during this maintenance; the pump shaft was unable to be rotated by hand after approximately 3/4 revolution. The pump was disassembled and the apparent cause was determined to be that the pump rotating assembly was not centralized in the casing bore resulting in minimal clearance between the pump wear rings. On 10/22/08 FirstEnergy Nuclear Operating Company requested enforcement discretion from the NRC to permit an additional 36 hours of time to complete reassembly of the pump. The NRC verbally granted FENOC's request for enforcement discretion on 10/22/08 at 1105. The enforcement discretion period would have expired on 10/24/08 at 1126. Since unexpected difficulties have delayed the reassembly of the pump, FENOC management decided to proactively shutdown Beaver Valley Power Station Unit No.2 beginning at 1800 on 10/23/08 (reference Information Only EN#44597). Beaver Valley Power Station Unit No. 2 entered Mode 3 at 0519 on 10/24/08 as part of the planned controlled shutdown. On 10/24/08 at 0600, Technical Specification 3.5.2, Condition B was entered based on a determination that the projected work completion time for restoring 2SIS-P21A to operable status would not meet the conditions of the enforcement discretion. Technical Specification 3.5.2, Required Action B.1 was met since the plant was in Mode 3, the plant is continuing the cooldown to be in Mode 4 within 12 hours per Required Action B.2. This event is being reported as a Technical Specification required shutdown pursuant to 10CFR50.72(b)(2)(i). No radiological release occurred due to this event. The licensee notified the NRC Resident Inspector. Notified R1DO (Cahill). | ||
| ENS 42606 | 26 May 2006 18:15:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Solid State Protection System Train Inoperable | Beaver Valley Unit 1 commenced a Technical Specification required shutdown at 1415 hours due to Train B Solid State Protection System inoperable. The condition was discovered during testing and the Unit is required to be in Mode 3 by 1846 hours. Train A of Solid State Protection System remains operable and in service. Investigation of the Solid State Protection problem is in progress. The licensee believes that the problem is caused by a faulty memory card. They are continuing to investigate the problem. The unit was at 18% power at 1727 EDT on the way to mode 3. The licensee notified the NRC Resident Inspector.
Beaver Valley Unit 1 entered mode 3 at 1831 hours on 5/26/06. Train B Solid State Protection was repaired and tested satisfactorily. It was returned to OPERABLE status at 1915 hours on 5/26/06. The failure was in the memory portion of the universal logic card in the low power trip block circuitry. A root cause investigation is ongoing. All technical specifications action statements were exited when Train B Solid State Protection was declared OPERABLE. Reactor start up is planned for 0500 on 5/27/06. The licensee will notify the NRC Resident Inspector. Notified R1DO (M. Gray). | 05000334/LER-2006-002 | |
| ENS 42491 | 11 April 2006 13:24:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown | At 0924 Unit 2 experienced multiple Fire Protection System Deluge Valve actuations for the Main Filter Banks, System Station Service Transformers (SSSTs) 2A and 2B, the Main Transformer, Condensate Polishing building ventilation charcoal filter and Decontamination building ventilation charcoal filter. The Transformers were unaffected by the spray actuation and remained operable. No actual fire occurred or was observed for these components or areas. The Motor driven and Engine Driven Fire Pumps automatically started upon the actuations. These above noted areas and components received water spray down until manually isolated. It was verified that both of the Charcoal Main Filter Banks (TS 3.7.8.1) had become wetted down by the spray actuation. With the charcoal main filter banks wet, their filtering capability becomes impacted and is being evaluated. These filter components were declared inoperable. At 0924, Unit 2 entered the actions of Tech. Spec. 3.0.3. The requirements of Tech Spec 3.0.3 are: Within 1 hour action shall be initiated to place the unit in Hot Standby within the following 6 hours, Hot Shutdown within the following 6 hours, and Cold Shutdown within the subsequent 24 hours. At 1055 Unit 2 commenced shutdown at 20%/hr. The cause of the multiple actuations is not understood at this time. An Event Response Team has been formed to investigate the issue. All other safety related systems are operable. The action required for the licensee to exit this Tech Spec required shutdown will be to restore the charcoal filter banks to operable status. The licensee estimates the investigation, troubleshooting, and repair efforts to last approximately 40 hours. The licensee notified the NRC Resident Inspector. | Main Transformer | |
| ENS 40434 | 10 January 2004 06:27:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Initiated and Subsequently Terminated | During a surveillance of the Beaver Valley Power Station Unit No. 1 Solid State Protection System (SSPS) train "B" on 01/09/2004, a master relay contact exhibited unacceptable high resistance and was declared inoperable at 2042 hours. Corrective action was initiated to replace the relay, which requires that the "B" train of the reactor trip system be transferred from the reactor trip breaker to the bypass reactor trip breaker and the "B" train of SSPS be de-energized. Action 40.b of Technical Specification 3.3.1.1, Table 3.3-1, Item 21 requires the Unit to be in at least Hot Standby within 6 hours with one reactor trip breaker inoperable as a result of something other than an inoperable diverse trip feature. Bypassing the reactor trip breaker in order to replace the SSPS master relay makes the reactor trip breaker inoperable as a result of something other than an inoperable diverse trip feature. The 'B' train Reactor Trip System reactor trip breaker was bypassed at 00:28 hours on 01/10/2004. Beaver Valley Power Station Unit No. 1 initiated a plant shutdown at 01:27 hours on 01/10/2004 in response to entering Action 40.b of Technical Specification 3.3.1.1, Table 3.3-1, Item 21. Action to replace the SSPS master relay was completed and the reactor trip breaker was returned to operability at 02:55 hours and the plant shutdown was terminated at 72% power. The Unit is scheduled to return to full power operation. Source Range N-31 re-energized when source range high voltage switch was restored to "Normal" position. Investigation is currently underway to determine the cause of this. The licensee notified the NRC Resident Inspector. | ||
| ENS 40049 | 6 August 2003 20:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition | On 8/06/2003 at 1630 hours, the Beaver Valley Power Station (BVPS) Unit 2 control room was notified by Engineering that the manual operator actions specified in Unit 2 post-fire procedures are not adequate for defeating a potential fire-induced spurious operation of the Power Operated Relief Valves (PORVs). The procedure step de-energizes the PORVs by opening breakers at the d-c distribution panel. This action alone would not be sufficient to prevent a cable-to-cable hot short from re-energizing the circuit since the de-energized circuit is routed in cable trays with other energized circuits in the affected fire areas. Preliminary reviews have identified the following potentially affected fire areas: CB-1 Control Building Instrumentation and Relay Area CB-2 Control Building Cable Spreading Room CB-3 Control Building Main Control Room CB-6 Control Building West Communication Room CT-1 Cable Tunnel CV-1 Cable Vault and Rod Control Area SB-1 Emergency Switchgear Room (Orange) RC-1 Reactor Containment An hourly roving fire watch patrol has been established for the affected fire areas as compensatory measures, with the exception of the Main Control Room and the Reactor Containment area, until the condition is fully evaluated and resolved. The Main Control Room is continuously manned and does not require an hourly fire watch patrol. The Reactor Containment area is not accessible during normal power operations and, as such, compensatory measures for this area will include a once-per-shift verification of remote instrumentation by operations personnel to confirm that there are no abnormal conditions or indications for this area. This condition was discovered during the review of manual operator actions for fire-induced spurious operations to confirm the safe shutdown circuit analysis is consistent with the manual actions identified in the procedures. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii) as an unanalyzed condition that significantly degraded plant safety since the failure to assure the PORVs remain in the closed position could result in the failure to meet the fire protection safe shutdown criteria. BVPS Unit 2 was licensed to the NUREG 0800 Standard Review Plan 9.5.1 "Fire Protection Program", and License Condition 2.F requires compliance to the provisions of the approved fire protection program as described in the Final Safety Analysis Report. UFSAR Section 9.5A.1.2.1.6 states that the safe shutdown capability should not be adversely affected by a fire in any plant area which results in spurious actuation of the redundant valves in any one high-low pressure interface line. Even though there is a very low likelihood of multiple shorts causing a spurious PORV actuation in an ungrounded DC circuit, the existing actions to prevent re-energizing the circuit and causing a spurious actuation of the PORVs is not consistent with the NRC guidance for high-low pressure interface valves. The NRC resident inspector has been notified.
On 8/07/03 at 1515, the following additional areas were identified as being potentially affected by this condition: CV-2 Cable vault and Rod Control Area (East) CV-3 Cable vault and Rod Control Area (Elev 755'6") SB-2 Emergency Switchgear Room (Purple) SB-3 Service Bldg. Cable Tray Area ASP Alternative Shutdown Panel Room Interim compensatory measures in the form of hourly roving watch patrols have been expanded to include the above noted areas until the condition is fully evaluated and resolved. The NRC Resident Inspector has been notified. Notified the Region 1 Duty Officer (C. Anderson)
The licensee is retracting this report based on the following: Beaver Valley Power Station (BVPS) Unit No. 2 retracts the notification made on 08/06/2003 at 19:12 hrs regarding the event reported under 10 CFR 50.72(b)(3)(ii)(B) (ENS #40049). The previous notification identified a potential unanalyzed condition that significantly degraded plant safety since the failure to assure the PORVs remain in the closed position could result in the failure to meet the Fire Protection Program safe shutdown criteria. The postulated event for spurious PORV opening due to an external cable-to-cable hot short from a fire was subsequently analyzed. It was determined that since the reactor core does not experience an unrecoverable plant condition and the radiological conditions remained bounded by analyzed DBAs, the plant's safe shutdown capability for a fire event is maintained. Although the now-credible spurious opening of a PORV is a deviation from the current Fire Protection Program criteria, the results of the spurious PORV opening does not adversely impact the plant's safe shutdown capability and hence is not a significant degradation of plant safety. Therefore this condition is not reportable pursuant to either 10 CFR 50.72(b)(3)(ii)(B) nor 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degraded plant safety. Compensatory measures in the form of shiftly fire watch patrols for the affected fire areas, with the exception of the Main Control Room and Reactor Containment area (as described in prior notifications), will remain in place until the Fire Protection Program non-conformance issue is dispositioned. The licensee informed the NRC Resident Inspector. Notified R1DO(Gray). |