Semantic search

This page or action requires JavaScript to work. Please enable JavaScript in your browser or use a browser where it is supported, so that functionality can be provided as requested. For further assistance, please have a look at the noscript help page.

Search Clear all entries

	Discovered date	Reporting criterion	Title	Event description
ENS 57019	10 March 2024 08:53:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Essential Chilled Water Trains Declared Inoperable	The following information was provided by the licensee via email: On 3/9/2024 at 2126 CST, train C essential cooling water was declared inoperable due to a through-wall leak on the discharge vent line. This would also cascade and cause train C essential chilled water to be inoperable. On 3/10/2024 at 0353 CDT, train B essential chilled water was declared inoperable due to chilled water outlet temperature greater than 52 degrees F following startup of essential chiller 12B. Chilled water outlet temperature was adjusted to less than 52 degrees F at 0440 CDT, and train B essential chilled water was declared operable. This condition resulted in the inoperability of two of the three safety trains required for the accident mitigating functions including: high head safety injection, low head safety injection, containment spray, electrical auxiliary building HVAC, control room envelope HVAC, and essential chilled water. This is an 8 hour reportable condition per 10CFR50.72(b)(3)(v)(D) because it could affect the ability to mitigate the consequences of an accident. The licensee notified the NRC Resident Inspector.
ENS 56861	16 November 2023 21:41:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Essential Chiller Trains Inoperable	The following information was provided by the licensee via phone and email: 11/05/23, 2200 CST: Essential Chiller 'B' train and associated cascading equipment were declared INOPERABLE for planned maintenance. Unit 2 entered the Configuration Risk Management Program as required by Technical Specifications on 11/12/23 at 2200. 11/16/23, 1541: Essential Chiller 'C' train and associated cascading equipment were declared INOPERABLE due to an unexpected material condition causing the Essential Chiller to trip. The most limiting (Limiting Condition of Operability) LCO is 3.7.7, Action c. This condition resulted in the INOPERABILITY of two of the three safety trains required for the accident mitigating function including: High Head Safety Injection, Low Head Safety Injection, Containment Spray, Electrical Auxiliary Building HVAC, Control Room Envelope HVAC, Essential Chilled Water. This is an 8 hour reportable condition per 10CFR50.72(b)(3)(v)(D) because it could affect the ability to mitigate the consequences of an accident. A risk analysis was performed for the equipment INOPERABILITY and mitigating actions have been taken per site procedures. All 'A' train equipment remains operable. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The 'B' train Emergency Diesel Generator was also inoperable due to planned maintenance and continues to be inoperable. It was considered in the Configuration Risk Management Program and it was determined this condition could be maintained. LCO 3.7.7, Action c requires reactor shutdown within 72 hours.
ENS 56848	10 November 2023 20:13:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Essential Chiller Trains Inoperable	The following information was provided by the licensee via email: On 11/10/23 at 0642 CST, essential chiller 'B' train and cascading equipment was declared inoperable due to chill water temperature exceeding limits. At 1413 CST, essential chiller 'C' train and cascading equipment was declared inoperable due to discharge pressure exceeding limits. This condition resulted in an inoperable condition on two out of the three safety trains for the accident mitigating function including the 'B' and 'C' train high head safety injection, low head safety injection, containment spray, electrical auxiliary building HVAC, control room envelope HVAC, and essential chill water. All 'A' train equipment remained operable. This was determined to be reportable within 8 hours as required by 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Plant is in a 72 hour limiting condition for operation per technical specification 3.7.7. Restoration of 'B' train anticipated on 11/11/23 mid day.
ENS 56480	18 April 2023 21:30:00	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	Safe Shutdown Capability and Accident Mitigation	The following information was provided by the licensee via email: Notification per 10 CFR 50.72 (b)(3)(v)(A) and (v)(D) At time 1630 CDT on 4/18/23, Comanche Peak Unit 1 entered TS (Technical Specification) 3.0.3 for 11 minutes due to declaring Train A component cooling water (CCW) inoperable in conjunction with a Train B centrifugal charging pump (CCP) inoperable for scheduled maintenance. This resulted in an event or condition that could have prevented fulfillment of a safety function, high head injection of the emergency core cooling system. CCP 1-02 and fan cooler were tagged out of service at 0400 CDT on 4/18/23 due to scheduled maintenance activities. Containment spray (CT) pump 1-03 seal oil cooler CCW leak was found by a watchstander at 0930 CDT on 4/18/23. Engineering determined that leakage was CCW from a pipe flange weld after insulation removal and could not (determine) operability and notified control room at 1630 CDT on 4/18/23. This placed unit 1 in a TS 3.0.3 condition from 1630 to 1641 CDT for approximately 11 minutes until CCP 1-02 was restored back to operable status. CCW was declared operable at 1912 after CT pump 1-03 seal oil cooler was isolated. CT pump 1-03 remained inoperable until weld repair completed. Train A CT pump 1-03 declared operable at 1211 CDT 4/19/23. ENS notification should have been made by 0030 CDT on 4/19/23. This report restores compliance. The NRC Resident Inspector has been notified. * * RETRACTION ON 6/7/2023 AT 1000 EDT FROM CASEY DAVIES TO BILL GOTT * * * The following information was provided by the licensee via phone and email: Although Comanche Peak Unit 1 conservatively entered a limiting condition for operation action statement and performed repairs immediately, further engineering inspection and evaluation concluded that the CCW system was fully able to provide the needed flow to the 1-03 CT pump seal coolers from the time of discovery (0930 CDT) until which time the piping was isolated for repairs. During this period, structural integrity of the joint was maintained, CCW inventory loss remained within acceptable limits, and CCW could perform its intended design and safety functions. Based on this revised operability determination, train A CCW was always operable, and TS 3.0.3 did not apply. Therefore, reportability requirements per 10 CFR 50.72 (b)(3)(v)(A) and (v)(D) did not apply, and a 60 day LER will not be submitted. The licensee notified the NRC Resident Inspector. Notified R4DO (Young)
ENS 55265	19 May 2021 10:15:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Automatic Reactor Trip Due to High Pressurizer Pressure	At 0315 MST on May 19, 2021, Unit 2 reactor automatically tripped during testing of the Plant Protection System. The Reactor Protection System actuated to trip the reactor on High Pressurizer Pressure, although no plant protection setpoints were exceeded. Main Steam Isolation Signal (MSIS), Safety Injection Actuation Signal (SIAS), and Containment Isolation Actuation Signal (CIAS) were received. No injection of water into the Reactor Coolant System occurred. Auxiliary Feedwater Actuation Signals (AFAS) 1 and 2 actuated on low Steam Generator water level post trip as designed. This event is being reported as a reactor protection system and a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). Following the reactor trip, all (Control Element Assemblies) CEAs inserted fully into the core. All systems operated as expected. No emergency plan classification was required per the Emergency Plan. Safety related busses remained powered during the event from offsite power and the offsite power grid is stable. Unit 2 is stable and in Mode 3. Steam Generator heat removal is via the class 1 E powered motor driven auxiliary feedwater pump and Atmospheric Dump Valves. The NRC Senior Resident Inspector has been informed. * * UPDATE ON 5/19/21 AT 1351 EDT FROM JASON HILL TO BRIAN P. SMITH * * * The Unit 2 reactor tripped because of actual High Pressurizer Pressure that occurred as a result of a Main Steam Isolation Signal actuation. At 0337 MST, both trains of Low Pressure and High Pressure Safety Injection (LPSI and HPSI) were made inoperable when the injection valves were overridden and closed in accordance with station procedures. At 0346 MST, in accordance with station procedures, both trains of Containment Spray, LPSI, and HPSI pumps were overridden and stopped, rendering Containment Spray inoperable as well. This represents a condition that would have prevented the fulfillment of a safety function required to mitigate the consequences of an accident per 10 CFR 50.72(b)(3)(v)(D). Additionally, at the time of the Safety Injection Actuation Signal (0315 MST), both trains of Emergency Diesel Generators actuated as required and both 4160 VAC busses remained energized from off-site power. The NRC Senior Resident Inspector has been informed. Notified R4DO (Young) * * UPDATE ON 7/02/21 AT 1943 EDT FROM YOLANDA GOOD TO JEFFREY WHITED * * * The inoperability of both trains of Low Pressure and High Pressure Safety Injection (LPSI and HPSI) and both trains of Containment Spray (CS) following the Unit 2 reactor trip has been determined to be an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). Additionally, inoperability of both trains of HPSI resulted in a reportable condition that could prevent fulfillment of its credited safety function to maintain the reactor in a safe shutdown condition per 10 CFR 50. 72(b)(3)(v)(A). The additional reporting criteria were discovered during review of the event and corresponding safety analyses. The NRC Senior Resident Inspector has been informed. Notified R4DO (Werner)
ENS 54417	30 November 2019 19:00:00	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	Both Trains of Containment Spray Removed from Service	On November 30, 2019, at 1100 PST, with Unit 2 in Mode 4, Operations identified that both trains of containment spray had been removed from service earlier at approximately 0217 hours as part of preparations for a planned Mode 5 entry. The containment spray pumps are required to be operable (along with the containment fan cooler units) in Modes 1 through 4 in accordance with Technical Specification 3.6.6. With both containment spray pumps inoperable, TS 3.6.6 Action F requires the Unit to be shut down in accordance with TS 3.0.3. At 1125 hours, both trains of containment spray were returned to operable and the required actions of TS 3.6.6 and TS 3.0.3 were exited. The five containment fan cooler units remained operable for the duration of the occurrence. This notification is being made in accordance with the requirement of 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented fulfillment of a safety function, and 10 CFR 50.72(b)(3)(ii) as an event or condition that may have resulted in the plant being in an unanalyzed condition. The NRC Senior Resident Inspector has been notified.
ENS 54307	2 October 2019 08:15:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Actuation of Containment Spray System	On October 2, 2019, at 0415 EDT, with Unit 2 in Mode 5 at 0 percent power, an actuation of the Unit 2 containment spray system occurred during valve strokes of the 2A train containment spray header isolations while the 2B train containment spray pump was in recirculation. The reason for the containment spray actuation was due to a conflicting procedural alignment with the 2B containment spray recirculation procedure. The containment spray system does not have an automatic function, and only receives manual actuation. The Unit 2 containment spray actuation was secured at 0416. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the containment spray system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 53777	9 October 2018 05:00:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification of Invalid Specified System Actuation	This 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to provide information pertaining to an invalid Engineered Safety Feature actuation signal. On October 9, 2018, Arkansas Nuclear One, Unit 2 was in refueling Mode 6, when a vital inverter failed while aligned from its alternate power source causing a loss of one of four vital instrument buses. The loss of the instrument bus resulted in one of the four engineered safety feature protection channels to enter a tripped state. Because one of the other four channels was already in a tripped state in support of a channel power supply replacement activity, two out of four protection channels were now in the tripped state resulting in a Safety Injection Actuation Signal, Containment Spray Actuation Signal, Containment Cooling Actuation Signal, Recirculation Actuation Signal, Emergency Feed Actuation Signal, and Containment Isolation Actuation Signal. In general, only one train of equipment is protected and assumed to be available during Mode 6 operations. Due to the defense-in-depth plant configuration in Mode 6, which is intended to avoid inadvertent start of emergency systems, the resulting actuations caused no adverse impact to Shutdown Cooling or Spent Fuel Pool cooling operations. At least one train of the following systems was aligned for automatic actuation: Service Water Emergency Diesel Generator Containment Penetration Room Exhaust Fan Other non-essential components which are shed or realigned upon safeguards actuation The few systems and components that were aligned for automatic operation responded as designed, including containment isolation valves and valves associated with the above systems (if aligned for automatic operation). The Service Water system was already in operation and, therefore, no Service Water pumps actuated. All systems and components which were capable of automatic operation performed as designed. The Emergency Diesel Generator started but did not synchronize to the bus. No safety injection occurred to the core. This actuation was caused by equipment failure and was not an actual signal resulting from parameter inputs. The affected actuation signals do not perform a safety function in Mode 6 and are not required to be available or operable. Therefore, this actuation is considered invalid. This event was entered into ANO's corrective action program for resolution. This event did not result in any adverse impact to the health and safety of the public. In accordance with 10 CFR 50.73(a)(i) a telephone notification is being made in lieu of submitting a written Licensee Event Report. The licensee has notified the NRC Resident Inspector.
ENS 53067	12 November 2017 09:03:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Containment Spray Pump Control Switches Out of Service	At 2119 (CST) on 11/12/2017 a Control Room board walk down discovered that both of the Unit 2 Containment Spray Pump control switches were in pull-out. With the control switches in pull-out, the pumps would not automatically start as required. Unplanned TS (Technical Specifications) 3.0.3 was entered at 2119 as a result of not complying with TS 3.6.5, Containment Spray and Cooling Systems, which requires both trains of Containment Spray to be Operable while in Mode 4. Unit 2 had entered Mode 4 at 0303 on 11/12/2017. TS 3.0.3 was exited at 2127 on 11/12/2017 when both Containment Spray Pump control switches were placed in Automatic restoring Operability. Preliminary investigation determined that while Unit 2 was in Mode 5, Surveillance SP 2099, Main Steam Isolation Valve Logic Test, had taken the Containment Spray Pump control switches to pull-out but did not re-align the control switches to automatic after the test was complete. This 8-hour Non-Emergency report is being made per 10 CFR 50.72(b)(3)(v)(D), Accident Mitigation. The NRC Senior Resident Inspector has been informed.
ENS 52936	29 August 2017 17:00:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Technical Specification Required Shutdown	On August 22, 2017 at 2321 hours, Grand Gulf Nuclear Station entered Technical Specification conditions for three Limiting Condition for Operations (LCOs) not met due to Residual Heat Removal 'A' (RHR 'A') being declared inoperable. LCOs not met: 1) 3.5.1 for one low pressure ECCS (Emergency Core Cooling System) injection/spray subsystem. 2) 3.6.1.7 for one RHR containment spray subsystem, and 3) 3.6.2.3 for one RHR suppression pool cooling subsystem. The station has made the decision to shutdown the plant based on the results of troubleshooting performed on the RHR 'A' pump. The restoration of RHR 'A' pump will not be completed prior to the end of the 7 day LCO completion time. Grand Gulf Nuclear Station initiated plant shutdown required by Technical Specifications 3.5.1, 3.6.1.7, and 3.6.2.3 at 1200 hours CDT on 08/29/2017 due to expected restoration of RHR 'A' exceeding the completion time of 7 days prior to restoring operability. The licensee notified the NRC Resident Inspector.
ENS 52796	10 April 2017 17:47:00	10 CFR 50.73(a)(1), Submit an LER	Invalid System Actuation During Testing	This 60-day telephone notification is being made in accordance with the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of several safety systems. On April 10, 2017 at 1347 hours (EDT) with Unit 3 in Mode 6 during performance of the Train B Engineered Safeguards Integrated Test, safety system actuations occurred prior to the expected point in the test procedure when a loss of continuity resulted while the seismic clips were being removed from a fuse. The actuations were supposed to occur at a subsequent step when the fuse was to be pulled to actuate the Hi Containment Pressure signal. As a result, the following equipment actuated: 3B, 4A and 4B High Head SI pumps; 3B Containment Spray pump; Containment Isolation and Containment Ventilation Isolations; 3A and 3B Emergency Diesel Generators; Emergency Containment Coolers. Because an actual high containment pressure signal did not exist at the time of the actuation, the actuation is considered invalid. All equipment responded as expected. The NRC Resident Inspector has been notified.
ENS 52484	11 January 2017 21:00:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Teflon Found in Containment Spray Pump Components	On September 16, 2016, Comanche Peak reported an unanalyzed condition and potential loss of safety function per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v) related to Teflon (PTFE) installed in the pressure gauge diaphragm seal assemblies for all four of the Centrifugal Charging Pumps and both of the Positive Displacement Charging Pumps on Units 1 and 2 (EN#52244). On November 14, 2016, this event was subsequently retracted. On December 12, 2016, during the ongoing extent of condition review, Teflon was also found to be installed in the suction and discharge pressure gauge diaphragm seal assemblies for the Unit 1 and 2 Containment Spray Pumps. On January 11, 2017 at approximately 1500 CDT, the reportability evaluation determined that reasonable assurance did not exist that the Containment Spray system would have been able to fulfill its design function of removing heat from the containment environment without impacting the applicable dose limits. Teflon (PTFE) is a restricted material normally prohibited from use in contact with reactor coolant or in radiation environments. Teflon (PTFE) is not radiation tolerant and degrades in a radiation environment. The Teflon (PTFE) used in these diaphragm seal assemblies could fail during a postulated Loss of Coolant Accident (LOCA) which could cause the Containment Spray Pumps on Units 1 and 2 to be inoperable, and exceed system leakage limits. This could challenge dose limits and in plant post-accident accessibility. This represents an unanalyzed condition. The pressure gauges and diaphragm seals for all of the Unit 1 and 2 Containment Spray Pumps have been isolated and the Unit 1 and 2 Containment Spray Pumps are operable. The Teflon (PTFE) has likely existed in these diaphragm seals since initial plant licensing. Luminant Power is continuing to investigate the extent of this condition and potential repair techniques. The NRC Resident Inspector has been notified.
ENS 52335	26 October 2016 15:42:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Containment Sump Suction Valve Opened During Containment Spray Pump Testing	On October 26, 2016, the Harris Nuclear Plant was in Mode 6 with core reload complete, the reactor head removed, and reactor cavity water level greater than 23 feet. The refueling water storage tank (RWST) was less than 23.4% level as expected for the refueling conditions. During surveillance testing to adjust the eductor flow throttle position, the containment spray pump was started in recirculation mode with the discharge valve shut. With RWST level less than 23.4%, logic was satisfied to actuate Engineered Safety Features Actuation System (ESFAS) Functional Unit 8, containment spray switchover to containment sump. The containment sump suction valve opened in accordance with the design, however the action was unexpected by the operators. Therefore, operators secured the containment spray pump and shut the containment sump suction valve. ESFAS Functional Unit 2, Containment Spray, was not actuated and water did not flow through the containment spray nozzles. This event is reported as a specified system actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A) due to the opening of the containment sump suction valve. This event did not impact the health and safety of the public. The licensee has notified the NRC Resident Inspector.
ENS 52239	13 September 2016 22:30:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition Involving Station Service Water Trains	Based on a walk down in the Service Water Intake Structure (SWIS) with the NRC Resident (Inspector), it was observed that a vertical section of 4 inch Fire Protection pipe that provides a normally pressurized source of fire water supply to the overhead sprinkler system in the SWIS is not Moderate Energy Line Break (MELB) shielded similar to the horizontal segment of the same line near the ceiling. In the event of a MELB crack along any portion of the unshielded pipe, the MELB has a potential impact to the function of any one of the 4 Service Water pumps. Only one train at a time would be affected during the event. This is due to the physical characteristics of the postulated MELB and the configuration/separation relative to the source line and target pumps and/or associated Motor Control Centers (MCCs) that support pump operation. Since the Service Water trains have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, if the MELB were to have occurred during these times and affected the opposite train, then two Service Water trains could have been inoperable and this represents an unanalyzed condition. At the time of discovery, all four Service Water trains were operable, therefore, this condition is not reportable as a loss of safety function per 10 CFR 50.72(b)(3)(vi). Currently, Service Water Train B on each Unit has been declared inoperable per Technical Specification (TS) 3.7.8. This condition will be corrected within the 72-hour Completion Time of TS 3.7.8. Currently, Emergency Diesel Generator B on each Unit has been declared inoperable per Technical Specification (TS) 3.8.1. This condition will be corrected within the 72-hour Completion Time of TS 3.8.1. The NRC Resident Inspector was informed. * * UPDATE ON 10/6/2016 AT 2009 EDT FROM DAMON SCHROEDER TO DONG PARK * * * This is an update to Event Number 52239. On September 13, 2016 at 2228 EDT, Comanche Peak reported an unanalyzed condition involving station service water trains per 10CFR50.72(b)(3)(ii)(B). Specifically, the reported condition involved a vertical section of 4 inch Fire Protection pipe in the SWIS that was not adequately shielded for a Moderate Energy Line Break (MELB). In the event of a MELB crack along any portion of the unshielded pipe, the MELB had a potential impact to the function of any one of the 4 Service Water pumps. On October 6, 2016 at 1410 hours CDT, a section of eyewash station pipe in the Unit 2 Safeguards Building was identified as a result of extent of condition walkdowns that was not adequately shielded for a Moderate Energy Line Break (MELB). In the event of a MELB crack along any portion of this unshielded pipe, the MELB had the potential to impact Unit 2 Train B 480V Motor Control Center (MCC) 2EB2-1. This MCC provides power to Unit 2 Train B Emergency Core Cooling, Battery Charger, Containment Spray, and Containment Isolation Valve equipment. The affected eyewash station pipe was isolated shortly after it was discovered to not be adequately shielded for a MELB. Since 480V MCC 2EB1-1 and the Unit 2 Train A Emergency Core Cooling, Battery Charger, Containment Spray, and Containment Isolation Valve equipment trains have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, if the MELB were to have occurred during these times and affected the opposite train, then 2EB1-1, 2EB2-1 and both trains of the Unit 2 Emergency Core Cooling, Battery Charger, Containment Spray, and Containment Isolation Valve equipment could have been inoperable and this represents an unanalyzed condition. At the time of discovery, 2EB1-1 and the Unit 2 Train A Emergency Core Cooling, Battery Charger, Containment Spray, and Containment Isolation Valve equipment was operable. Therefore, this condition is not reportable as a loss of safety function per 10 CFR 50.72(b)(3)(vi). The NRC Resident Inspector was informed. Notified R4DO (Werner). * * UPDATE FROM ROBERT DANIELS TO DONALD NORWOOD AT 2233 EDT ON 10/10/2016 * * * This is an additional update to Event Number 52239. On September 13, 2016 at 2228 EDT and again on October 6, 2016 at 2009 EDT, Comanche Peak reported unanalyzed conditions involving Station Service Water System trains and a 480V Motor Control Center (MCC) per 10 CFR 50.72(b)(3)(ii)(B). The reported conditions involved sections of piping that were not adequately shielded for a Moderate Energy Line Break (MELB). In the event of a MELB crack along any portion of the unshielded piping, the MELB had a potential impact to the function of safety-related equipment in the Service Water Intake Structure and the Unit 2 Safeguards Building. On October 10, 2016 at 1708 CDT, as a result of ongoing extent of condition walkdowns, a section of fire protection pipe in the Unit 1 Safeguards Building was identified that was not adequately shielded for a MELB. In the event of a MELB crack along any portion of this unshielded pipe, the MELB had the potential to impact Unit 1 Train B Switchgear 1EA2, Unit 1 Train B 480V MCC 1EB4-2, and Unit 1 Train B Distribution Panel 1ED2-2. Only one of these power supplies at a time would be affected. 1EA2 provides 6.9KV electrical power to various Unit 1 Train B safety-related pumps, panels, sequencer, and transformers. 1EB4-2 provides 480V electrical power to various Unit 1 Train B safety-related pumps, valves, fans, panels, and transformers. 1ED2-2 provides 125VDC electrical power to EDG 1-02 channel 1 starting circuit. The affected fire protection pipe was isolated shortly after it was discovered to not be adequately shielded for a MELB. Since Unit 1 Train A Switchgear 1EA1, Unit 1 Train A 480V MCC 1EB3-2, and Unit 1 Train A Distribution Panel 1ED1-2 have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, if the MELB were to have occurred during these times and affected the opposite train, then both trains of Unit 1 6.9KV power (1EA2 and 1EA1), both trains of Unit 1 480V power (1EB4-2 and 1EB3-2), and both trains of Unit 1 125VDC power (1ED2-2 and 1ED1-2) along with the safety-related equipment they supply could potentially have been inoperable and this represents an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). At the time of discovery, none of the affected Train A equipment was inoperable. Therefore, this condition is not reportable as a loss of safety function per 10 CFR 50.72(b)(3)(v). The NRC Resident Inspector was informed. Notified R4DO (Werner). * * UPDATE FROM HUNTER SCHILL TO DONALD NORWOOD AT 1457 EST ON 11/7/2016 * * * This is an update to Event Number 52239. On November 17, 2016 at 0730 CST, during ongoing extent of condition walkdowns in the Boric Acid Transfer Pump Area of the Auxiliary Building, two pressurized fire protection pipe segments were identified that did not contain Moderate Energy Line Break (MELB) shielding. In the event of a MELB crack along the unshielded portion of these pipes, the MELB had the potential to impact Unit 1 Train B 480V Motor Control Center (MCC) 1 EB4-1. This MCC provides 480V electrical power to various Unit 1 Train B safety-related pumps, valves, fans, battery chargers, and transformers. At 0743 CST, Technical Specification 3.8.9 Condition A was entered for one AC electrical power distribution subsystem inoperable. At 1021 CST, MCC 1 EB4-1 was declared Operable after MELB shielding was installed on the affected fire protection lines. Since Unit 1 Train A 480V MCC 1 EB3-1 and the associated Unit 1 Train A safety-related pumps, valves, fans, battery chargers, and transformers have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, given the MELB condition, 1 EB4-1, 1 EB3-1 and both trains of the Unit 1 safety-related pumps, valves, fans, battery chargers, and transformers they supply could have been inoperable and this represents an unanalyzed condition. At the time of discovery, 1 EB3-1 and the associated Unit 1 Train A safety-related pumps, valves, fans, battery chargers, and transformers were operable. Therefore, this condition is not reportable as a loss of safety function per 10 CFR 50.72(b)(3)(v). The NRC Resident Inspector was informed. Notified R4DO (Azua). * * UPDATE ON 12/05/2016 AT 1730 EST FROM HUNTER SCHILL TO STEVEN VITTO * * * This is an update to Event Number 52239. On December 5, 2016 during ongoing extent of condition walk downs in the Auxiliary Building, pressurized fire protection pipe segments (a flange and a pipe elbow) were identified which did not contain Moderate Energy Line Break (MELB) shielding. In the event of a MELB crack along the un-shielded portion of the pipes, a MELB had the potential to impact Unit 2 Train B 480V Motor Control Center (MCC) 2EB4-1. This MCC provides 480V electrical power to various Unit 2 Train B safety-related pumps, valves, fans, battery chargers, and transformers. At approximately 1355 CST Technical Specification 3.8.9 Condition A was entered for one AC electrical power distribution subsystem inoperable. At 1459 CST, MCC 2EB4-1 was declared Operable after MELB shielding was installed on the affected fire protection line locations. Since Unit 2 Train A 480V MCC 2EB3-1 and the associated Unit 2 Train A safety-related pumps, valves, fans, battery chargers, and transformers have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, given the MELB condition, 2EB4-1 , 2EB3-1 and both trains of the Unit 2 safety-related pumps, valves, fans, battery chargers, and transformers they supply could have been inoperable and this represents an un-analyzed condition. At the time of discovery, 2EB3-1 and the associated Unit 2 Train A safety-related pumps, valves, fans, battery chargers, and transformers were operable. Therefore, this condition is not reportable as a loss of safety function per 10CFR 50.72(b)(3)(v). The NRC Resident Inspector was informed. Notified R4DO (Gaddy). * * UPDATE ON 12/22/2016 AT 1649 EST FROM HUNTER SCHILL TO DONG PARK * * * This is an update to Event Number 52239. On December 22, 2016 at approximately 1046 (CST) during ongoing extent of condition walk downs in the common Auxiliary Building (AB) corridor room (X-179), several normally pressurized Waste Processing (WP) pipe segments and one Vent & Drain (VD) segment which are greater than 1" nominal pipe diameter, did not contain MELB shielding. In the event of a MELB crack along the unshielded portion of these pipes, a MELB could have had the potential to impact Unit 1, Train B 480V Motor Control Center (MCC) 1EB4-1. This MCC provides 480V electrical power to various Unit 1 Train B safety-related pumps, valves, fans, battery chargers, and transformers. Prior to the field walkdown, the subject WP and VD line segments were either isolated and depressurized (WP lines) and/or the AB sump discharges realigned (VD) such that the subject lines would pose no threat to the MCC 1EB4-1 if confirmed that shielding is required. As such, the identified condition does not adversely affect operability of 1EB4-1 and entry into a Technical Specification action statement was not required. Field activities continue to install MELB shielding in the affected locations. Since Unit 1 Train A 480V MCC 1EB3-1 and the associated Unit 1 Train A safety-related pumps, valves, fans, battery chargers, and transformers have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, given the MELB condition, 1EB4-1, 1EB3-1 and both trains of the Unit 1 safety-related pumps, valves, fans, battery chargers, and transformers they supply could have been inoperable and this represents an unanalyzed condition. At the time of discovery, 1EB3-1 and the associated Unit 1 Train A safety-related pumps, valves, fans, battery chargers, and transformers were operable. Therefore, this condition is not reportable as a loss of safety function per 10CFR 50.72(b)(3)(v). The NRC Resident Inspector was informed. Notified R4DO (Hay).
ENS 52225	8 September 2016 08:00:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Technical Specification Shutdown Due to Loss of Residual Heat Removal Pump	On September 4, 2016 at 0258 (CDT), Grand Gulf Nuclear Station entered three (Technical Specification) Limiting Conditions for Operations (LCOs) due to residual heat removal pump 'A' (RHR 'A') being declared inoperable. LCOs entered: 1) 3.5.1 for one low pressure ECCS injection/spray subsystem, 2) 3.6.1.7 for one RHR containment spray subsystem, and 3) 3.6.2.3 for one RHR suppression pool cooling subsystem. Station management has made the decision to shutdown the plant to repair the RHR 'A' pump prior to the end of the 7 day LCO completion time based on troubleshooting and testing performed on the RHR 'A' pump. Grand Gulf Nuclear Station initiated plant shutdown required by Tech Spec Actions 3.5.1, 3.6.1.7, and 3.6.2.3, at 0300 CDT on 09/08/2016 due to expected inability to restore RHR 'A' to operable status prior to exceeding the LCO time of 7 days. The unit is currently at 82 percent power. There are no other systems out of service that would complicate the orderly shutdown to Mode 4. The licensee will notify the NRC Resident Inspector.
ENS 51807	18 March 2016 16:28:00	10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat	Shutdown Cooling Pipe Void	During a scheduled surveillance test on 3/18/2016 at 1128 (CDT), Fort Calhoun ultrasonic testing technicians discovered a void on the common shutdown cooling heat exchanger discharge piping. This piping is normally isolated during power operation, and the void does not adversely affect the Containment Spray function, Low Pressure Safety Injection function, or High Pressure Safety Injection function. This isolated piping with the void is placed in service only during shutdown cooling operation. The fluid height measured was 10.8 inches, compared to the required height of 11.7 inches for the surveillance test. The void could potentially complicate the initiation of shutdown cooling in the required mode of operation. This piping was last tested satisfactory on 12/31/2015. The source of the void is still under investigation. Fort Calhoun maintenance was successful in venting the void on 3/18/2016 at 1704 CDT. The NRC Resident Inspector has been notified. * * RETRACTION AT 1701 EDT ON 05/16/16 FROM JAKE WALKER TO KARL DIEDERICH * * * Following the 8-hour 10 CFR 50.72 notification made on 3/18/16 (EN 51807), further engineering analysis has determined that the ensuing water hammer transient would not have prevented the shutdown cooling system from performing its required safety functions. Specifically, it was found that the resulting system pressure transient would not cause any relief valves to lift and that piping and supports would not be significantly challenged. Therefore, the common shutdown cooling heat exchanger discharge piping remained operable by the detailed analysis. As such, the safety function was not lost and the event notification is being retracted as it is not reportable pursuant to 10 CFR 50.72(b)(3)(v)(B). Notified the R4DO (G Miller).
ENS 51527	10 November 2015 20:02:00	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	Unanalyzed Condition Due to Debris Dropped Into Reactor Cavity Equipment Pit	On 11/10/2015 at 1502 (EST), Unit 2 MCR (main control room) was notified by workers in containment that 2 ice suits had been dropped into the Unit 2 Containment Reactor Cavity Equipment Pit. Based upon size and location of the dropped suits, Unit 2 entered LCO 3.6.15 (Containment Recirculation Drains) Condition B and LCO 3.0.3 for refueling canal drains being inoperable. The two refueling canal drains and the ice condenser drains function with the ice bed, Containment Spray System and ECCS to limit the pressure and temperature that could be expected following a DBA (Design Basis Accident). Following performance of a Safety Function Determination it was determined that, during the short duration when both coats were in the process of being retrieved, they could have potentially clogged the drains and prevented the fulfillment of safety functions if there was a DBA. Both suits were retrieved from the equipment pit by 1556 (EST) and all LCO conditions were exited. The licensee notified the NRC Resident Inspector.
ENS 51294	16 June 2015 05:00:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 Report - Abb 50H Instantaneous Over Current Protection Relay Failure	The following information was excerpted from a facsimile received from the Xcel Energy: Pursuant to 10 CFR 21.21(d)(3)(i) Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy, submits the attached initial notification of failure to comply or existence of a defect. If there is any question or if additional information is needed, please contact Dr. Glenn A. Carlson, P.E., at (651) 267-1755. Name and address of the individual or individuals informing the Commission: Scott M. Sharp, Site Operations Director, Prairie Island Nuclear Generating Plant (PINGP), Northern States Power Company - Minnesota, 1717 Wakonade Drive East, Welch, MN 55089 Identification of the facility, the activity, or the basic component supplied for such facility or such activity within the United States which fails to comply or contains a defect: ABB Power T&D Company Inc., Relay, Overload, Overcurrent, 58/125VDC, Type: 50H, Cat.: 468S0475, 1.B.: 7.2.1.7-3 Nature of the defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply: During bench testing to calibrate the relay, 125 VDC was applied to the relay with no noticeable effect. This was shop work on a spare relay and was not a plant installed piece of equipment. The relay protects against an instantaneous over current condition, which if undetected would cause substantial damage to the motor. This type of relay is installed in many locations; the failed relay was reserved to replace an existing relay protecting a Residual Heat Removal System pump motor. If the relay had been installed and required actuation during an accident, it would have resulted in major degradation in safety related equipment. The ABB 50H relays are installed in ten locations at PINGP. All of the 50H relays are installed on the Unit 2 safety-related buses. The equipment protected is 21 Aux Feedwater Pump, 21 Component Cooling Pump, 21 Residual Heat Removal Pump, 21 Safety Injection Pump, 21 Containment Spray Pump, 22 Safety Injection Pump, 22 Residual Heat Removal Pump, 22 Component Cooling Pump, 22 Containment Spray Pump, and 121 Cooling Water Pump. The failed relay was reserved to replace an existing relay protecting a Residual Heat Removal pump motor. PINGP Material Management has custody of the failed relay pending further investigation and notified the supplier of this notification by email on 8/6/2015. Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or will be given to purchasers or licensees: Continue bench testing relays prior to installation.
ENS 51170	22 June 2015 18:06:00	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	Sodium Hydroxide Below Technical Specification Required Concentration	At 1406 (EDT) on 6/22/15, the chemistry sample results for the Salem Unit 1 Containment Spray Additive tank were below the minimum Technical Specification requirement for NaOH (Sodium Hydroxide) concentration. The Containment Spray Additive tank was declared inoperable at that time in accordance with Technical Specification 3.6.2.2.a. Based on the fact that there is no redundant equipment for the Containment Spray Additive tank, this condition could have prevented this Safety System from Controlling the Release of Radioactive Material and Mitigating the consequences of an accident. This event is being reported under the requirements of 10 CFR 50.72(b)(3)(v)(C) and (D) as 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of systems that are needed to Control the Release of Radioactive Material and Mitigate the Consequences of an Accident.' Actions are being taken to restore the NaOH concentration to within the Technical Specification limits. The Licensee has notified the NRC Resident. No one was injured as a result of the event. The licensee will notify the Lower Alloways Township.
ENS 51149	11 June 2015 20:54:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Degraded Containment Spray Flow Check Valve	During containment spray pump inservice testing the minimum flow recirculation line recorded negative flow indicating reverse flow in the line. After the troubleshooting, it was determined that a degraded minimum flow check valve was allowing a path to the refueling water storage tank (RWST) for certain post loss of coolant accident (LOCA) conditions. The minimum flow isolation has been closed to eliminate the path. No actuation occurred during this time. The NRC Resident Inspector, Connecticut Department of Energy and Environmental Protection (DEEP) Hartford, and Watertown Dispatch have been notified. * * UPDATE FROM TOM CLEARY TO STEVEN VITTO ON 06/16/15 AT 1349 EDT * * * The purpose of this call is to correct the record for the 10 CFR 50.72(b)(3)(v)(C) event report provided on June 11, 2015. Event report number 51149 states: 'During containment spray pump inservice testing ...' The condition of the degraded containment spray flow check valve was identified: 'During high pressure safety injection pump inservice testing...'." Notified R1DO (Bickett) * * UPDATE FROM WILLIAM McCOLLUM TO DANIEL MILLS AT 1724 EDT ON 7/10/15 * * * The purpose of this call is to update the record for the 10 CFR 50.72(b)(3)(v)(C) event report provided on June 11, 2015. During subsequent investigation, it has been determined that a second release path may have existed through ECCS system relief valves under certain post LOCA conditions as a result of the degraded check valve described in the original report. The licensee has informed the NRC Resident Inspector. Notified R1DO (Cahill).
ENS 50282	15 July 2014 14:42:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition - Low Containment Spray Flow Rate	At 0942 (CDT) on 7/15/2014, the 2A Containment Spray chemical additive flow was found out of tolerance low during surveillance testing. This resulted in an unanalyzed condition in that insufficient chemical additive flow might have resulted in lower than assumed containment spray pH values during past periods. Based on the above, this is being reported in accordance with 10CFR50.72(b)(3)(ii)(B). Actions are in progress to restore the 2A Containment Spray chemical additive flow to within tolerance. The 2B Containment Spray system is operable per Technical Specification 3.6.6 and is capable of providing required chemical additive flow. The required flow is 18 to 67 gallons per minute (gpm), however, the measured flow was 17.96 gpm. The licensee has notified the NRC Resident Inspector. * * RETRACTION FROM JAMES PETTY TO JOHN SHOEMAKER AT 1729 EDT ON 8/1/14 * * * The purpose of this report is to retract ENS report #50282 (July 15, 2014). This ENS report was made for the 2A Containment Spray chemical additive flow which was found out of tolerance low during surveillance testing. At the time of reporting, it was concluded that this was an unanalyzed condition in that insufficient chemical additive flow may have resulted in lower than assumed containment spray pH values during past periods. This was reported in accordance with 10CFR50.72(b)(3)(11)(B) as an unanalyzed condition that significantly degrades plant safety. On Wednesday, July 29, 2014, Braidwood Generating Station concluded that the prior ENS notification could be retracted based on the completion of Engineering Change 398884, 'Evaluation Of 2A CS NaOH Spray Additive Test Results And Discussion of IRs 1682209 and 1683413.' The Engineering Change concluded that the approval of the alternate source term (AST) license amendment resulted in the elimination of a minimum containment spray (CS) spray pH value. The current containment release analysis does not credit the addition of sodium hydroxide (NaOH) to CS spray for fission product removal from the containment atmosphere. The long-term retention of captured fission products in the sump water assumes the sump water pH is greater than 7. This is established by the transfer of the containment spray additive tank (CSAT) contents to the sump during CS system operation. To transfer the maximum CSAT inventory to the sump within 8 hours, a minimum NaOH eductor flow of approximately 10 gpm is required. The minimum NaOH injection flow for the 2A CS eductor system exceeded 10 gpm so the eductor injection flows meet the criteria to transfer CSAT inventory to the containment recirculation sump within the expected minimum CS system operating time. The out of tolerance flow values recorded at the time of the initial ENS notification are acceptable. The licensee has notified the NRC Resident Inspector. Notified the R3DO (Lara).
ENS 50189	11 June 2014 19:34:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	One Containment Spray Train Chemical Additive Flow Out of Specifications	At 1434 (CDT) on 06/11/14, the 1A Containment Spray chemical additive flow was found out of tolerance low during surveillance testing. This resulted in an unanalyzed condition in that insufficient chemical additive flow might have resulted in lower than assumed containment spray Ph values during past periods. The impact of the unanalyzed condition has not been fully evaluated. Based on the above, this is being reported in accordance with 10CFR50.72(b)(3)(ii)(B). Engineering analyses are in progress to evaluate the past condition. Actions are in progress to restore the 1A Containment Spray chemical additive flow to within tolerance. The 1B Containment Spray system is operable per Tech Spec 3.6.6 and is capable of providing required chemical additive flow. The required flow is 30 to 63 gpm. Measured flow was 27 gpm. The last measurement was six years ago. The licensee notified the NRC Resident Inspector. * * RETRACTION PROVIDED BY JOHN LOGAN TO JEFF ROTTON AT 1803 EDT ON 06/13/2014 * * * The purpose of this report is to retract EN report #50189 (June 11, 2014). This EN report was made for the 1A Containment Spray chemical additive flow which was found out of tolerance low during surveillance testing. At the time of reporting, it was concluded that this was an unanalyzed condition in that insufficient chemical additive flow may have resulted in lower than assumed containment spray pH values during past periods. This was reported in accordance with 10CFR50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. At approximately 1700 CDT on Thursday, June 12, 2014, Braidwood Generating Station concluded that the prior EN notification could be retracted based on the completion of Engineering Change 398472, 'Evaluation of As-Found Results for 1A Containment Spray NaOH Additive Flow.' The Engineering Change concluded that the approval of the alternate source term (AST) license amendment resulted in the elimination of a minimum containment spray (CS) spray pH value. The current containment release analysis does not credit the addition of sodium hydroxide (NaOH) to CS spray for fission product removal from the containment atmosphere. The long-term retention of captured fission products in the sump water assumes the sump water pH is greater than 7. This is established by the transfer of the containment spray additive tank (CSAT) contents to the sump during CS system operation. To transfer the maximum CSAT inventory to the sump within 8 hours, a minimum NaOH eductor flow of approximately 10 gpm is required. The minimum NaOH injection flow for the 1A CS eductor system (27.0 ' as-found' and 27.7 'as-left' gpm) exceeded 10 gpm so the eductor injection flows meet the criteria to transfer CSAT inventory to the containment recirculation sump within the expected minimum CS system operating time. The out of tolerance flow values recorded at the time of the initial notification are acceptable. The licensee has notified the NRC Resident Inspector. Notified R3DO (Daley).
ENS 49425	10 October 2013 15:09:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Low Pressure Core Spray Declared Inoperable	At 0809 PDT on 10/10/2013, after starting Standby Service Water (SW) pumps, Columbia Generating Station (Columbia) received a flow low alarm for the Low Pressure Containment Spray (LPCS) pump motor cooling water. The flow indicator SW-FIS-19 was reported too low to support pump function. The LPCS system was declared inoperable, and the appropriate Technical Specification action statement was entered. The cause of the low flow alarm has not been determined. This event is reportable under criterion 10 CFR 50.72(b)(3)(v)(D) 'Any Event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (D) Mitigate the consequences of an accident.' Columbia is continuing to troubleshoot and repair as appropriate to restore the SW flow to the LPCS pump. The licensee has notified the NRC Resident Inspector. * * UPDATE ON 10/11/13 AT 1653 EDT FROM MATT HUMMER TO DONG PARK * * * Subsequent to receipt of the low flow alarm, flushing of the flow indicating switch sensing lines was conducted. It has been determined that the instrument sensing lines are partially blocked providing a flow indication that is slow to respond to actual flow conditions. The flow is currently reading normally. The LPCS pump was declared operable on 10/10/13 at 1447 PDT. The initial notification incorrectly stated 'Low Pressure Containment Spray (LPCS)', the correct description is 'Low Pressure Core Spray (LPCS)'. The licensee has notified the NRC Resident Inspector. Notified R4DO (Hay). * * RETRACTION FROM DAVID HOLICK TO JOHN SHOEMAKER AT 2030 EDT ON 10/18/13 * * * Subsequent evaluation of this event found there was no actual low flow condition to the LPCS bearing cooler. A flow scan on the SW outlet line from the LPCS pump bearing cooler was conducted on 10/12/13 which confirmed the actual flow conditions were reading normally. The problem resides in the installed flow indication switch. Since there was no actual low flow condition to the LPCS bearing cooler, the LPCS pump could perform its safety function to mitigate an accident. Therefore, this event notification is being retracted. Notified R4DO (Azua). The licensee has notified the NRC Resident Inspector.
ENS 49324	5 September 2013 13:31:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Potential Low Pressure Safety Injection Pump Run Out Condition	Current design basis calculations indicate the Low Pressure Safety Injection (LPSI) pumps could potentially operate in a run-out condition under certain worst case design basis conditions. The LPSI pumps could operate in a run-out condition beyond the analyzed time by 20 minutes. Current design basis calculation assumes LPSI Pump would be shutdown by (the) RAS (Recirculation Actuation Signal) in less than one hour, however due to past changes to Containment Spray Pump Start Logic, the time was lengthened to 80 minutes which is beyond the one hour analyzed. This represents a reportable unanalyzed condition. The licensee notified the NRC Resident Inspector. * * RETRACTION FROM LUKE JENSEN TO HOWIE CROUCH AT 1722 EDT ON 10/31/13 * * * Fort Calhoun completed additional analysis which verified that the LPSI pumps will not go into run-out as previously reported. Therefore Fort Calhoun is withdrawing the event notification. The licensee will notify the NRC Resident Inspector. Notified R4DO (Drake).
ENS 48921	14 April 2013 11:20:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Containment Spray Chemical Addition Flow Path Isolated	At 0620 CDT on 4/14/13, the Unit 1 Sodium Hydroxide Tank outlet valve was found to be shut. This valve isolated the flow path for both trains of containment spray chemical addition and resulted in LCO 3.6.7 (Spray Additive System) not being met, which resulted in a condition reportable pursuant to 10 CFR 50.72(b)(3)(v)(D). At 0651 CDT on 4/14/13, the Unit 1 Sodium Hydroxide Tank outlet valve was restored to its required locked open position and TSAC (Technical Specification Action Conditions) 3.6.7.8 was exited. The licensee has notified the NRC Resident Inspector.
ENS 48806	4 March 2013 20:00:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition Due to Mechanical Seal Material Design	It has been determined that the mechanical seals used in two Low Pressure Safety Injection Pumps and three Containment Spray Pumps are made of a material that may not maintain the designed integrity of the systems under certain accident conditions. These seals have been installed since original plant construction. This issue was discovered by plant personnel while researching requirements for the replacement parts during scheduled outage activities. The licensee notified the NRC Resident Inspector.
ENS 48661	29 November 2012 04:05:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification of Invalid System Actuation	This telephone notification is being made for an invalid actuation under 10 CFR 50.73(a)(2)(iv)(A) following the reporting guidance of 10 CFR 50.73(a)(1) and is not considered a Licensee Event Report. With Unit 2 at Cold Shutdown (about 105?F and depressurized), an invalid actuation of the Unit 2 'A' train of the High High Consequence Limiting Safeguards (Hi-Hi CLS) system occurred at 2305 (EST) during reinstallation of fuses in preparation for return to service testing. The fuses were pulled to implement a design modification to replace existing relays with a new design. The 'A' train of the High High CLS actuated as soon as the fuses were installed. Plant systems and components responding to the Hi-Hi CLS 'A' train signals started and functioned successfully as designed with the exception of those systems and components procedurally rendered inoperable due to the RCS being below 350?F and 450 psig. Shutdown cooling was not lost due to safety injection leads being tagged out. The signal could not be reset from the Main Control Room due to system design in this configuration requiring manual local manipulations to address affected components. The relays on both trains were replaced with the original design and the fuses reinstalled. The affected systems were restored to their pre-event configuration. Specific trains and systems that actuated as a result of the "A" train of Hi-Hi CLS signal are described below: -- Component Cooling from the A Reactor Coolant Pump isolated. -- Containment Spray realigned and gravity flowed the Refueling Water Storage Tank and Caustic Addition Tank to the Containment Sump. The level did not reach the point where any components in the containment basement were affected. -- Service Water flowed to the A and C Recirculation Spray Heat Exchangers. -- Containment Instrument Air isolated. -- Emergency Diesel Generator (EDG) No.2 started but did not load since its associated Emergency Bus remained energized by offsite power. The EDG was stopped and returned to automatic. A root cause evaluation is in progress. The licensee notified the NRC Resident Inspector
ENS 48615	26 October 2012 04:08:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification for an Invalid Actuation of the Essential Spray Pond System	The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73. This telephone notification is being made pursuant to the reporting requirements of 10 CFR 50.73(a)(2)(iv)(A) and 50.73(a)(1) to describe an invalid actuation of the Palo Verde Nuclear Generating Station (PVNGS) Unit 2 essential spray pond system (SP) which serves as the ultimate heat sink as described in 10 CFR 50.73(a)(2)(iv)(B)(9). On October 25, 2012, at approximately 2108, Mountain Standard Time, during refueling outage 2R17, Unit 2 experienced an invalid actuation of the train B essential spray pond system. Testing of the train B engineered safeguards features actuation system (ESFAS) was in progress. During the test, following activities to reset the train B containment spray actuation signal (CSAS), the procedure required a check of contact status on a relay contact which provides input to the train B Load Sequencer. The guidance required the use of a digital multi-meter to perform the contact status check. When the digital multi-meter test leads were landed and removed from the circuit, the train B Load Sequencer changed output modes which resulted in automatic starting of the train B essential chilled water system, essential cooling water system, essential fuel building air filtration unit and essential spray pond system. This was a partial actuation of the train B ESF equipment and all the affected equipment responded as designed. No equipment failures resulted from the event. The event was entered into the PVNGS corrective action program. The invalid actuation was caused by an incorrect testing methodology in the procedure instructions which resulted in the unintended interaction of the digital multi-meter and the digital train B Load Sequencer. The inadequate procedures will be revised to modify the testing methodology for verification of relay contact status. The licensee has notified the NRC Resident Inspector.
ENS 48237	27 August 2012 17:00:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Both Emergency Diesel Generators Declared Inoperable	At 12:00 CDT maintenance personnel identified a pinhole leak from the Division 1 Service Water System piping in the Service Water Pump Room. Division 1 Service Water (SW) was declared inoperable and LCO 3.7.2 Condition A was entered due to a potential loss of structural integrity. This directs entry into LCO 3.8.1 for Diesel Generator #1 made inoperable by SW. DG 2 was previously made inoperable at 05:39 CDT on 8/25/2012 due to an unrelated issue regarding rain water inleakage into the DG 2 Room. Control Room Emergency Filtration system (CREFs) is aligned to Div 1 power. LCO 3.7.4 Condition A is applicable, requiring restoration of CREFs to operable status within 7 days. TRM LCO 3.6.1 condition A and B also apply, requiring (A) Restoration of containment spray subsystem A to OPERABLE status within 7 days and (B) Restore one RHR containment spray subsystem to operable status within 8 hours. DG 1 and DG 2 comprise the onsite emergency power systems. Both DGs inoperable is reportable per 10CFR50.72(b)(3)(v)(D) as a condition that could prevent fulfillment of the safety function of structures or systems needed to mitigate the consequences of an accident. Actions were taken to expedite repairs of the DG 2 roof leak and to further characterize the Division 1 SW piping leak. LCO 3.8.1 Condition E allows 2 hours to restore one DG to operable status or enter Condition F, to be in Mode 3 in 12 hours, which was entered at 14:00. Repairs to the roof leak on the DG 2 room were completed, after which DG 2 was declared Operable at 18:30. LCO 3.8.1 Conditions E and F required shutdown were exited at this time. LCO 3.8.1 Condition B for DG 1, and 3.7.2 for SW Loop A, continue to be active. Planning to repair the SW piping pinhole leak is continuing. There were no adverse grid conditions during the period both DGs were inoperable. The NRC Resident has been informed of the condition. No media or press release is planned at this time. * * RETRACTION FROM FRED SCHIZAS TO DONALD NORWOOD AT 1802 EDT ON 8/31/12 * * * This notification is being made to retract Event Notification EN #48237 which reported a loss of safety function due to both onsite Emergency Diesel Generators (DGs) being simultaneously INOPERABLE. On 8/25/12 at 0539 CDT, DG#2 was declared INOPERABLE due to rain water in-leakage into the DG#2 room. Condition B of LCO 3.8.1 was entered. and Required Actions were being taken to restore the INOPERABLE DG within 7 days. Subsequently, on 8/27/12 at 1200 CDT Maintenance personnel identified a pinhole leak from the Division I Service Water System piping in the Service Water Pump Room. Division I Service Water was declared INOPERABLE and LCO 3.7.2 Condition A was entered due to a potential loss of structural integrity. This prompted entry into LCO 3.8.1 Conditions E and F which require shutdown, because DG#1 was made INOPERABLE by SW. Both DG's INOPERABLE is reportable per 10CFR50.72.b.3.v.D as a condition that could prevent fulfillment of a safety function of structures or systems needed to mitigate the consequences of an accident. Repairs to the roof leak on the DG#2 room were completed, after which DG#2 was declared Operable at 1830 CDT on 8/27/2012. LCO 3.8.1 Conditions E and F were exited at the time. LCO 3.8.1 Condition B for DG#1, and 3.7.2 for SW Loop A, continued to be active, and planning to repair the pin-hole leak continued. Subsequent investigation and UT examinations provided data which enabled SW Division 1 and DG#1 to be declared OPERABLE on 8/30/2012 at 0528 CDT. An evaluation of this condition concluded that further characterization of the SW Piping Pin-Hole leak enabled ASME Code Case N-513-3 to be applied to determine piping structural integrity was maintained. SW ability to supply required flows to its loads is not adversely diminished, because the flaw is small. Water leaking from the flaw will not adversely affect any other equipment important to safety by spray or flooding. The piping is compliant with ASME code. Based on this evaluation, the Division 1 SW system can perform its safety function and is OPERABLE. With Division 1 SW subsystem capable of fulfilling its safety function, DG#1 was therefore also capable of fulfilling its safety function during the period of SW subsystem INOPERABILITY. Consequently, during the period of DG#2 INOPERABILITY, DG#1 was capable of fulfilling the safety function of the onsite emergency power system. NPPD therefore retracts Event Notification EN 48237. The NRC Resident has been informed of this retraction. Notified R4DO (Azua).
ENS 47473	22 November 2011 23:00:00	10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat	Temporary Loss of Shutdown Cooling	During walkdown of scheduled work it was discovered that HCV-335 (Shutdown Cooling Heat Exchanger Inlet Header Isolation Valve) would not be able to be manually positioned open due to a missing idler gear key. Upon a loss of instrument air, HCV-335 would have failed closed, interrupting shutdown cooling flow with no ability to open HCV-335 manually. Alternate shutdown cooling pump and paths were available at the time of discovery. No loss of instrument air or interruption in shutdown cooling flow occurred while preparing to align alternate shutdown cooling. An 8 hour LCO under Technical Specification 2.8.1(3)2 was entered at 1700 CST. Alternate shutdown cooling was established on a containment spray pump as allowed by procedure. The 8 hour LCO was exited at 2306 CST. A replacement idler key has been fabricated for HCV-335. The licensee notified the NRC Resident Inspector. * * * RETRACTION FROM ERICK MATZKE TO PETE SNYDER AT 1654 EST ON 12/16/11 * * * Additional analysis has determined that the shutdown cooling system was capable of performing its design safety functions during the time that the idler key was missing. Therefore this event is being retracted. The licensee notified the NRC Resident Inspector. Notified R4DO (Walker).
ENS 47357	20 October 2011 01:02:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition Due to Pressure Instruments Containing Aluminum Material	With Unit 1 in Mode 4 and Unit 2 in Mode 1 an issue was identified with the body material of existing installed pressure instruments for both the Personnel and Emergency Airlocks of both units. The pressure instruments were determined to have an aluminum body which is not suited for safety related use in containment. Aluminum is a restricted/limited material in containment because it is not compatible with accident conditions and has failures with multiple adverse effects. Due to this condition, the pressure instruments would potentially lose pressure integrity during a LOCA with containment spray actuation. These pressure instruments are located inside containment and are connected to tubing that penetrates the airlock barrel. In event of a failure of any pressure instrument the integrity of the airlock would be compromised. The containment air locks form part of the containment pressure boundary and, as such, a loss of pressure boundary integrity would no longer meet general design criteria. Compensatory measures have been taken to prevent a failure of the airlock integrity due to containment spray actuation and at this time the airlock is operable. Luminant power determined this to be reportable at 2002 on 10/19/11 per 50.72(b)(3)(ii)(B) Comanche Peak Units 1 and 2 being in an unanalyzed condition that significantly degrades plant safety. The licensee notified the NRC Resident Inspector.
ENS 47352	18 October 2011 20:45:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Aluminum Valve Discovered Installed in Containment	While Unit 1 was in Mode 5, for the 15th refueling outage, an issue was identified with the valve material of an existing installed airlock hydraulic system valve. The valve was determined to have an aluminum body which is not suitable for safety related use in containment. The Unit 1 airlock hydraulic system penetrates the containment pressure boundary. The airlock hydraulic system achieves containment integrity by being a closed system under GDC-57. A loss of pressure boundary integrity would no longer meet General Design Criteria 57 (GDC-57) for a closed system. Aluminum is a restricted/limited material in containment because it is not compatible with accident conditions and has failures with multiple adverse effects. Due to this condition, the valve would potentially lose pressure integrity during a LOCA with containment spray actuation. Compensatory measures have been taken to prevent containment spray from affecting this valve and at this time, the airlock is operable. Luminant Power determined this issue to be reportable at 1545 (CDT) on 10/18 per 50.72(b)(3)(ii)(B) Comanche peak Unit 1 being in an unanalyzed condition that significantly degrades plant safety. The Unit 2 airlock is a different design and this condition does not apply to Unit 2. This material was installed during original construction and discovered during a licensee self-assessment. The licensee will notify the NRC Resident Inspector.
ENS 47210	28 August 2011 01:30:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Automatic Actuation of an Emergency Diesel Generator	At 2130 (EDT) on August 27, 2011, an automatic actuation of the Unit 1 Train A emergency diesel generator occurred due to an actuation signal from the load sequencer. The Train A 4160 kV emergency bus transferred to the emergency diesel generator and all Train A emergency loads required for Mode 2 started and sequenced onto the Train A 4160 KV emergency bus except the 480 volt center breaker to the bus E1A2 that did not close. The load sequencer is designed upon the receipt of a safety injection actuation and/or loss of offsite power to provide a signal to strip loads from the 4160 kV emergency bus and then, in sequence, to re-energize the associated 480 volt buses and to load engineering safety feature components onto the 4160 kV and associated 480 volt emergency buses in a predetermined sequence. Per 10 CFR 50.72(b)(3)(iv)(B), additional emergency safety features loads that actuated were the Train A reactor containment fan coolers and auxiliary feedwater system. Unit 1 remains critical at 100 percent power. No emergency core cooling system injection occurred into the reactor coolant system. The event occurred during surveillance testing when the Train A sequencer was taken from the AUTO Test position to the local position. It is not understood why the actuation occurred. In addition, it is not understood why the 480 volt load center breaker to the bus E1A2 did not close. The 480 volt bus E1A2 was re-energized at 2308 (EDT) on August 27, 2011. The Train A 4160 kV bus was restored to the offsite power source at 0150 (EDT) on August 28, 2011 and the Train A emergency diesel generator and engineering safety features loads were restored to their normal condition at 0201 (EDT) on August 28, 2011. With the Train A sequencer non-functional, the following Train A components are inoperable: 1) High Head Safety Injection Pump 1A; 2) Low Head Safety Injection Pump 1A; 3) Containment Spray Pump 1A; 4) RCFC (Reactor Containment Fan Cooler) Fan 11A; 5) RCFC Fan 12A; 6) Component Cooling Water Pump 1A; 7) Essential Cooling Water Pump 1A; 8) Aux Feedwater Pump 11; 9) Control Room/Elect. Aux Bldg HVAC; 10) Ess (Essential) Chiller 12A; and 11) ESF Diesel Generator 11. Although these components will not automatically start on a safety injection signal or loss of offsite power, these loads can be manually actuated. Engineered Safety Features Trains B and C remain operable. The licensee notified the NRC Resident Inspector. * * UPDATE ON 10/13/11 AT 1221 EDT FROM MORRIS TO HUFFMAN * * * The licensee is updating this event report to retract the originally reported valid specified system actuation and report it instead as a 60-day invalid specified system actuation report made by telephone: This update is a 60-day telephone notification in lieu of a written licensee event report being made under 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1). This event was originally reported per 50.72(b)(3)(iv)(A) as a valid actuation of the Unit 1 Train A emergency diesel generator and sequencing of Mode II (Loss of Offsite Power) emergency loads. The actuation occurred during surveillance testing when the Train A load sequencer was taken from the AUTO test position to the local position. Subsequent investigation has determined that the actuation occurred due to a faulty integrated chip within the sequencer's load sequence auto test module, and was not due to sensed or simulated plant conditions that would require a Mode II actuation. Unit 1 was at 100% power and no loss of offsite power occurred. The Train A equipment response to this invalid actuation is described in the original notification information provided on 08/28/2011. Additionally, the 10CFR50.72 Notification originally reported under Event Number 47210 is being retracted, since the actuation has been determined to be not valid. The licensee will notify the NRC Resident Inspector. R4DO (Whitten) notified.
ENS 47157	6 August 2011 15:55:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of Motor Driven Auxiliary Feedwater Pump	This telephone notification is being made in lieu of submitting a written LER for an invalid actuation of the 1A Motor Driven Auxiliary Feedwater (MDAFW) pump under 10 CFR 50.73(a)(2)(iv)(A). On August 6, 2011 at 1055 CDT, during the performance of surveillance testing on the 1A Containment Spray (CS) pump, an unexpected automatic start of the 1A MDAFW pump occurred. While installing jumpers in the B1F sequencer to cause an automatic start for the 1A CS pump per the test procedure, the jumpers were inadvertently connected to the wrong relay which started the 1A MDAFW pump instead. This automatic start was considered invalid since the start signal was not initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the safety function of the system. The 1A MDAFW pump was restored to a normal standby alignment at 1131 CDT. The following required information is being submitted per NUREG-1022, Rev. 2. (a) The 1A MDAFW pump is an 'A' train component. (b) The 1A MDAFW pump automatic start is considered a complete train actuation. (c) Once the 1A MDAFW pump inadvertently started, the system functioned per design. The licensee has informed the NRC Resident Inspector.
ENS 47120	2 August 2011 18:46:00	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	Vulnerability from a Potential High Energy Line Break	The following condition is being reported by Arkansas Nuclear One, Unit 2 (ANO-2) in accordance with 10CFR 50.72(b)(3)(ii)(B), 'Unanalyzed Condition' and in accordance with 10CFR 50.72(b)(3)(v)(D), 'A Condition That Could Have Prevented Fulfillment of a Safety Function.' On 08/02/2011 at 1346 CDT, the ANO Unit 2 Control Room was notified by Engineering that a postulated High Energy Line Break (HELB) could potentially cause both the Red and Green Train Emergency Safeguard Features (ESF) Rooms to exceed their environmentally qualified temperature limits. This postulated condition would be possible due to normally open room purge dampers exposing ESF equipment in these rooms to a common area impacted by HELB conditions. The ESF Rooms contain the Red and Green Trains of High Pressure Safely Injection Pumps, Low Pressure Safety Injection Pumps, Containment Spray Pumps, and Shutdown Cooling Heat Exchangers. Until further Engineering evaluation can be performed to validate this postulated scenario, ANO-2 has closed ESF room purge dampers to provide Red and Green ESF train separation during a potential HELB event. Refer to (ANO-2) Condition Report CR-ANO-2-2011-02772 for further information. The NRC Resident has been notified. * * RETRACTION FROM STEVE COFFMAN TO HOWIE CROUCH AT 1514 EDT ON 8/18/11 * * * The purpose of this notification is to retract a previous report made by Arkansas Nuclear One, Unit 2 (ANO-2) on 08/02/2011 at 2127 (EDT) (EN# 47120). The initial report documented that a postulated High Energy Line Break (HELB) could potentially cause rooms containing both trains of Emergency Safeguard Features (ESF) equipment to exceed their environmentally qualified temperature limits. The ESF rooms contain the High Pressure Safely Injection Pumps, Low Pressure Safety Injection Pumps, Containment Spray Pumps, and Shutdown Cooling Heat Exchangers. Specifically, normally open ESF room purge dampers exposing both trains of ESF equipment to a common area impacted by postulated HELB conditions were not modeled in the ANO-2 HELB analysis. This condition was reported in accordance with 10CFR 50.72(b)(3)(ii)(B), 'Unanalyzed Condition' and 10CFR 50.72(b)(3)(v)(D), 'A Condition That Could Have Prevented Fulfillment of a Safety Function'. Since the initial report, Engineering has revised the ANO-2 HELB model to include the effects of the open ESF room purge dampers. The resulting analysis shows that a HELB event will not cause the required ESF equipment to exceed analyzed temperature limits with the room purge dampers in the open configuration. Therefore, the condition did not result in 'a condition that could have prevented the fulfillment of a safety function' and did not result in an 'unanalyzed condition that significantly degrades plant safety'. Based on the revised HELB analysis, the previous report (EN#47120) describes a condition that does not meet the reporting requirements of 10CFR 50.72(b)(3)(v)(D) or 10CFR 50.72(b)(3)(ii)(B) and is therefore retracted. The NRC Resident Inspector has been informed of the retraction. Notified R4DO (Hay).
ENS 46824	6 May 2011 07:09:00	Other Unspec Reqmnt	Discovery of an After-The-Fact Unusual Event	Operators were making preparations to fill the Containment Spray System riser to support outage activities. At 0204 CDT, when Containment Spray Riser Isolation Valve CS-125A was opened, pressurizer level began to lower. The licensee suspects leak by of a valve in the Shutdown Cooling System. At 0214 CDT, the leak was stopped. Pressurizer level was lowered by 2.6%. After reviewing the event, the licensee determined that the leak rate was greater than 25 gpm which would have resulted in a declaration of an Unusual Event under EAL CU1. Since the event had concluded, no declaration was made. The licensee notified the NRC Resident Inspector.
ENS 46786	26 April 2011 05:20:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Safety Related Piping Had Unacceptable Air Void	Engineering reported that an ultrasonic (test) had identified an unacceptable air void in the horizontal run of suction piping from the Refueling Water Storage Tank (RWST) to the Safety Injection pumps, Residual Heat Removal pumps, and Containment Spray pumps. Technical Specification 3.0.3 was entered at 0020 CDT (on) 04/26/2011, and actions were started to remove the (air) void by venting of the suction pipe. Unit 2 was in progress of low power physics testing following refueling outage 2RF12. Inspection for (air) voids is in response to NRC generic letter on voiding in ECCS piping. The void was removed from the system by venting. Engineering confirmed with ultrasonic (test) that the void was removed, and Technical Specification 3.0.3 was exited at 0143 CDT (on) 04/26/2011. The licensee has notified the NRC Resident Inspector. * * RETRACTION FROM MIKE NIEMEYER TO DONALD NORWOOD AT 1628 EDT ON 6/23/2011 * * * Further analysis demonstrated that the declaration of inoperability was a conservative action, and RHR and SI would have fulfilled their respective safety functions and been operable. This analysis is based on a detailed study of the void transport and piping design. The analysis confirmed that one (1) train of containment spray was inoperable, however the containment spray safety function would have been fulfilled with the remaining operable (containment spray) train. Based on the above, it is concluded that the health and safety of the public were unaffected by this condition and this event has been evaluated to not meet the definition of a safety system functional failure per 10CFR50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified of this retraction. Notified R4DO (Deese).
ENS 46723	4 April 2011 20:00:00	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)(v)(C), Loss of Safety Function - Release of Radioactive Material	Scaffolding Affecting Safety Related Equipment	At 1500 (CDT), a concern was raised with regard to scaffolding that had been constructed around safety related equipment in the Auxiliary Building which contains both trains of safety injection and containment spray. As a result T.S. 2.0.1 was entered (which is the Fort Calhoun equivalent to standard T.S. 3.0.3). The scaffolding in question was removed and the equipment was returned to operable status and T.S. 2.0.1 was exited at 1726 (CDT). The NRC Resident Inspector has been notified. * * RETRACTION FROM ERICK MATZKE TO HOWIE CROUCH @ 2027 EDT ON 5/27/11 * * * Following the initial report, Fort Calhoun performed a seismic analysis of the impact of the scaffolding previously reported to determine if the equipment in the room would be capable of performing its required safety functions. The evaluation determined that the safety related function of the affected equipment would be able to be accomplished. Therefore, this event is being retracted. Notified R4DO (Haire). The licensee has notified the NRC Resident Inspector of this retraction.
ENS 46562	21 January 2011 21:39:00	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	Steam Exclusion Door Declared Inoperable	On 1/21/2011 at 1539 CST, the NRC Resident Inspector informed the Control Room that the lower Cane bolt was disengaged on Steam Exclusion Door 3, between Emergency Diesel Generator Room B and the Cardox Room. While the Cane bolt was not engaged, the barrier was Non-Functional and, in accordance with TRM 3.0.9, all equipment supported by that steam exclusion barrier was immediately declared inoperable. This included both Emergency Diesel Generators A & 8, safety-related 4160 V Busses 5 & 6, Service Water Trains A & B, and safety-related 480 V Busses 51, 52, 61. & 62. In addition, with Service Water inoperable, the following equipment was also inoperable in accordance with TRM 3.3.1: Component Cooling Trains A & B, Safety Injection Trains A & B, Residual Heat Removal Trains A & B, Containment Spray and Cooling Trains A & B, Auxiliary Feedwater Pumps A & B, and the Turbine Driven Auxiliary Feedwater Pump. With all three AFW pumps inoperable. TS 3 A.b.2 was entered to immediately initiate action to restore one AFW Train to operable status and suspend all LCOs requiring mode changes until one AFW Train is restored to operable status. Steam Exclusion Door 3 was properly secured at 1545 CST on 1/21/2011, and LCO 3.0.c and TS 3 A.b.2 were exited at that time. All equipment affected by the steam exclusion barrier is operable. This is reportable under 10 CFR 50.72 (b)(3)(v)(B), 'Any event or condition that at the time of discovery could have prevented the fulfillment of a safety function,' and under 10 CFR 50.72(b)(3)(ii)(B) 'any event or condition that results in the nuclear plant being in an unanalyzed condition that significantly degrades plant safety.' The licensee notified the NRC Resident Inspector. * * RETRACTION FROM CRAIG J. NEUSER TO DONALD NORWOOD AT 1427 EDT ON 3/22/2011 * * * Retraction of EN #46562 Non-Functional Steam Exclusion Door. On January 21, 2011, EN #46562 provided notification that both trains of ESF equipment (e.g., SI, RHR, ICS, etc ) were inoperable following discovery that the lower cane bolt was disengaged on steam exclusion Door 3, between emergency diesel generator Room B and an adjacent equipment room in the turbine building. With the lower cane bolt disengaged, the steam exclusion barrier was considered non-functional. A subsequent engineering evaluation determined that the Door 3 lower cane bolt was not required for Door 3 to fulfill its function as a steam exclusion barrier. The previously reported condition would not have resulted in an environment that would have adversely impacted the equipment protected by Door 3. Therefore, the door remained functional and the supported ESF equipment remained operable. Consequently, this condition did not meet the reportability criteria in 10CFR50.72. As a result, the notification made on January 21, 2011, in EN #46562 is hereby retracted. The NRC Senior Resident Inspector has been notified. Notified R3DO(Cameron).
ENS 46508	23 October 2010 00:49:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition - Battery Chargers May Lockup During a Design Basis Event	On October 22, 2010, the station confirmed the following: On a safety injection (SI) event with no loss of offsite power (LOOP), calculations show that at various points in the load sequence, the voltage at the charger Motor Control Center will dip low enough on the SI unit to lockup both of the chargers. Additionally, when considering out of sequence loads, such as the containment spray pump, starting on the emergency diesel generator during normal load sequencing, the potential exists that the voltage for any of the chargers on an SI event with a LOOP may drop to the point that would cause the chargers to lockup. On October 22, 2010, the station completed an operability evaluation of the 11, 12, 21, and 22 battery chargers, implemented compensatory measures to restore battery chargers in the event of an SI to maintain operability (a simple manual action to restart the charger), and determined that the battery chargers were operable but non-conforming. The NRC Resident (Inspector) has been informed. The licensee is planning on correcting this issue during the next refueling outage.
ENS 46368	27 October 2010 17:22:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Unit Commenced Shutdown After Declaring Containment Spray Inoperable	During pressure drop testing of the backup nitrogen supply for HCV-345 (Containment Spray Header isolation valve) the accumulator failed its test. This renders HCV-345 inoperable. HCV-344 the opposite header isolation valve also has an air leak that appears to be of similar magnitude to the leak on HCV-345. Fort Calhoun Station is conservatively considering both valves inoperable and has entered technical specification 2.0.1 which requires shutting the plant down to hot standby within 6 hours. The plant shutdown began at 1513 CDT. Repair efforts are underway on HCV-345 and HCV-344. The leakage on both valves was identified by systems engineering during testing. The licensee informed the NRC Resident Inspector. * * UPDATE AT 1756 EDT ON 10/27/10 FROM AARON CHLADIL TO S. SANDIN * * * At 1647 CDT, the licensee exited technical specification 2.0.1 after declaring HCV-344 operable. The Unit is currently at 74% power. The licensee will stabilize power at 70% and then commence power escalation. The licensee informed the NRC Resident Inspector. Notified R4DO (Proulx). * * RETRACTION FROM ERICK MATZKE TO HOWIE CROUCH AT 1155 EST ON 12/17/10 * * * Following the original notification, the Fort Calhoun Station reviewed and reanalyzed the acceptance criteria for the subject valve air accumulators. The analysis determined that the valves had been and were operable during the event. Therefore, the report under 10CFR50.72(b)(3)(v)(D) is being retracted. The licensee has notified the NRC Resident Inspector. Notified R4D0 (Howell).
ENS 46146	3 August 2010 20:06:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Potential Safety System Functional Failure of the Accident Mitigating Function	On 8/3/10 South Texas Project Unit 2 was in a scheduled A Train work week with the following equipment inoperable for planned maintenance; Essential Cooling Water Pump, Essential Chiller, Component Cooling Water Pump, Engineered Safety Function (ESF) Diesel Generator (DG), High Head Safety Injection (HHSI) pump, Low Head Safety Injection (LHSI) pump, and Containment Spray (CS) pump. At 0754 (CDT) on 8/3/10 the B train sequencer trouble alarm was received. The immediate operability determination was the sequencer remained operable. It was later identified during testing that the sequencer was inoperable. The B train sequencer was declared inoperable at 1506 (CDT) on 8/3/10. Due to loss of the automatic load sequencing support function, all associated train B safety equipment that is sequenced on the B train 14.16 kv bus during a Mode 1 Safety Injection (SI) was also declared inoperable. This condition resulted in an inoperable condition on two out of three safety trains for the accident mitigating function including the A and B train HHSI, LHSI, and CS pumps. All C train safety injection pumps remained operable. Pending a formal operability determination, this is conservatively considered to be a safety system functional failure of the accident mitigating function. This was determined to be reportable within 8 hours as required by 10 CFR 50.72(b)(3)(v)(D). The B train trouble alarm, an auto test feature, was discovered by operators during their rounds. The licensee entered their configuration risk management plan within the 1 hour as required. Currently, the licensee is working on completing the scheduled A train maintenance and restoring operability sometime in the morning. Also, a work package is under development to repair the faulty B train sequencer. The risk based time limit for restoring operability requires completion by 0449 (CDT) on 8/8/10. Unit 1 is unaffected and continues to operate at 100% power. The licensee informed the NRC Resident Inspector. * * RETRACTION AT 1638 EDT ON 08/26/2010 FROM JIM MORRIS TO S. SANDIN * * * The purpose of this update is to retract the notification made in ENS Report #46146 (August 3, 2010). Following the ENS notification, troubleshooting determined the cause of the Train B sequencer alarm to be the failure of an Output Mode I Actuation Timing Switch Module. An engineering evaluation of the event has been completed and determined that a failure of this module did not affect the ability of the ESF load sequencer to perform its design function. Therefore, the Train B sequencer and associated Train B ESF equipment remained technically operable during the time that Train A equipment was inoperable due to scheduled maintenance, and a condition reportable per 10 CFR 50.72(b)(3)(v) did not exist. The licensee will notify the NRC Resident Inspector." Notified R4DO (Walker).
ENS 46021	17 June 2010 10:19:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition Due to Containment Spray Recirc Sump Isolation Valve Failure to Stroke Closed	At 0519 CDT on June 17th, Unit 2 was closing the 2CS009B, Containment Spray Recirc Sump Isolation Valve, as part of post maintenance testing when the valve stopped stroking (i.e. mid position). The 2CS009B valve was being stroked closed for restoration from a successful timed stroke in the open direction. The 2CS009B valve was manually closed and verified closed via limit switch indication. With the 2CS009B valve unable to be closed from the Main Control Room, an unanalyzed condition may have existed where, during a large break LOCA requiring cold leg recirc, the Refueling Water Storage Tank (RWST) had an additional flow path to the containment recirc sump. This potentially challenges the operators to complete the switchover prior to the RWST reaching 9%, the point at which pumps taking a suction from the RWST only are shutdown. This condition is still being evaluated. The licensee notified the NRC Resident Inspector. * * * RETRACTION FROM P. MOODY TO P. SNYDER AT 0404 ON 2/1/11 * * * At 0509 on June 17, 2010, Unit 2 was closing the 2CS009B, Containment Spray (CS) Recirculation Sump Isolation Valve, as part of post maintenance testing when the valve stopped stroking (i.e., mid-position). The 2CS009B was being stroked closed for restoration from a successful timed stroke in the open direction. The 2CS009B was manually closed and verified closed via limit switch indication. With the 2CS009B unable to close from the Main Control Room, an unanalyzed condition may have existed where, during a large break LOCA requiring cold leg recirculation, the Refueling Water Storage Tank (RWST) had an additional flow path to the recirculation sump. This potentially challenged the operators to complete the switchover prior to the RWST reaching 9%, the point at which pumps taking a suction from the RWST only are shutdown. While this condition was being evaluated, an ENS notification was made per ENS 46021 under 10CFR50.72(b)(3)(ii)(B). As the evaluation approached the 60-day reporting period, LER 2010-002 was issued in accordance with 10 CFR 50.73(a)(2)(ii)(B), assuming the results would yield an unanalyzed condition. Since then, an evaluation was completed. The results concluded the operators would have performed the switchover steps within the allowed time, before reaching the RWST empty alarm set point. Therefore, the Emergency Core Cooling System (ECCS) and CS system would have performed their design functions. The evaluation also determined the RWST outflows with 2CS009B in the open position during the ECCS switchover sequence did not affect the RWST vortex analysis. Based on no loss of design function, the plant was not in an unanalyzed condition and this event is not reportable per 10CFR 50.72(b)(3)(ii)(B) or 10CFR 50.73(a)(2)(ii)(B). This event was screened for additional reportability criteria contained in the Exelon Reportability Manual. Again, since there was no loss of design function there is no reportability requirement. Therefore ENS notification 46021 is being retracted. The licensee notified the NRC Resident Inspector.
ENS 45767	15 March 2010 13:46:00	10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded	Secondary Containment Bypass Leakage Exceeded Technical Specification Limit	Appendix J Local Leak Rate Testing has determined that Secondary Containment Bypass Leakage (SCBL) has been exceeded for Unit 1. During performance of leak rate test SE-159-045, the combined SCBL limit of 15 scfh (standard cubic feet per hour) for as-found minimum pathway was exceeded as specified in Tech Spec SR 3.6.1.3.11. Acceptance Criteria Test results were within Acceptance Criteria for the 10CFR50 Appendix J limits of 0.6 La (maximum allowed leakage rate). This event is being reported as a degraded or unanalyzed condition pursuant to 10CFR50.72(b)(3)(ii)(A). The RHR system containment spray penetration isolation valve was being tested when the failure occurred. The valve will be repaired and re-tested. The licensee has notified the NRC Resident Inspector.
ENS 45461	23 October 2009 15:00:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Post Loca Recirculation Valve Position Interlocks	During the current Unit 2 fifteenth refueling outage (2R15), with the core off-loaded to the spent fuel pool, the containment recirculation sump valve interlock position switches were found not to perform correctly. Initial investigation has determined that the valve position sensing interlock switches were left incorrectly set during the Unit 2 fourteenth refueling outage (2R14). In the event of a Loss of Coolant Accident, (LOCA) this condition would have prevented Residual Heat Removal (RHR) flow from reaching containment spray, high head and intermediate head safety injection pumps following alignment to long term recirculation from the containment sump without additional operator action. The RHR function, including the long term recirculation from the containment sump, was not adversely affected by this condition. Also, containment spray, high head and intermediate head safety injection from the refueling water storage tank (RWST) was not adversely affected by this condition. Unit 2 containment recirculation sump valve interlock position switches have been correctly set and are being retested during the current 2R15. Unit 1 containment recirculation sump valve interlock position switches have been verified as correctly set to be capable of performing their design function by testing performed during the last refueling outage. This condition will be reported in a Licensee Event Report (LER) within 60 days in accordance with 10 CFR 50.73(a)(2)(ii). The licensee notified the NRC Resident Inspector.
ENS 45434	15 October 2009 16:25:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Unit Commenced Shutdown Due to Inoperable Division 2 Emergency Service Water System	This event is being reported in accordance with 10CFR50.72(b)(2)(i). On October 15, 2009, at approximately 1225 hours, the Perry Nuclear Power Plant commenced a Technical Specification required plant shutdown. On October 14, 2009, At 1747 hours, the Division 2 Emergency Service Water (ESW) system was declared inoperable and unavailable for planned work. The plant entered Technical Specification (TS) 3.7.1 Action A for one inoperable ESW Division. Other supported TSs were also entered (TS 3.8.1 for the Division 2 diesel generator, TS 3.6.1.7 for Containment Spray 'B', TS 3.6.2.3 for Suppression Pool Cooling 'B', TS 3.7.10 for Emergency Closed Cooling 'B', TS 3.5.1 for LPCI 'B' & 'C', among others). On October 15, 2009, at 0601 hours, the ESW 'B' pump was started for a planned pump run. At 0718 hours, the ESW 'B' pump tripped for unknown reasons. The determination was made to commence a controlled plant shutdown and power reduction commenced at 1225 hours. This decision was based on the anticipated investigation and repair time of ESW 'B' pump exceeding the TS 3.7.1 Action A 72 hour LCO completion time, and therefore, Action B requires the plant to be in MODE 3 within 12 hours and in MODE 4 in 36 hours. The TS 3.7.1 Action A 72 hour LCO completion time coincides with October 17, 2009, at 1747 hours. Currently, the plant is expected to be in MODE 3 at approximately 2300 hours on October 15, 2009. The NRC Resident Inspector has been notified. The State and Counties will also be notified.
ENS 44814	28 January 2009 19:25:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition Related to a Postulated Appendix R Fire Scenario	A postulated Appendix R fire scenario has been identified which could result in the plant being in an unanalyzed condition that may degrade plant safety. The unanalyzed condition involves a potential loss of power to the credited train of fire safe shutdown equipment for postulated fires in two areas of the Auxiliary Building. Safety-related loads on the 6.9KV Shutdown Boards (Centrifugal Charging Pumps, Safety Injection Pumps, Containment Spray Pumps, Motor-Driven Auxiliary Feedwater Pumps, and Residual Heat Removal Pumps) are equipped with local control switches to allow starting or stopping the pumps from the Auxiliary Building. The DC control circuit wiring for these local switches is routed in the same area as the power cables for pump motors. If the fire damages the control circuit cables, the control circuit fuses may fail, resulting in loss of trip capability for the associated 6.9KV breaker. If the breaker is already closed (due to a spurious signal or a previous valid start signal), then the breaker could be disabled in the closed position. If the fire damages the power cables resulting in a phase-to-phase fault with the load breaker trip capability disabled, then the Shutdown Board feeder breaker is designed to trip open on overcurrent to provide backup protection. This could result in the shutdown board being de-energized. In two areas in the Auxiliary Building (elevation 669 and 690 common areas), the above scenario could result in a condition which is outside the fire safe shutdown analysis due to loss of power to the credited train of fire safe shutdown equipment (e.g. Centrifugal Charging Pump, Essential Raw Cooling Water Pump, and motor-operated valves). As a result of this condition, Sequoyah has entered the Fire Protection Report Limiting Condition for Operation 3.7.12 for inoperable fire barriers. In accordance with this LCO action, the operability of fire detectors in the affected areas has been verified and an hourly fire watch has been established in the affected areas. This issue has been entered into the corrective action program. A permanent resolution is being evaluated. The NRC Resident Inspector has been notified.
ENS 44598	24 October 2008 01:26:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Reactor Scram Due to Loss of the Electronic Pressure Regulator	Control Room Operators observed slight (reactor) pressure rise during panel walk down. Investigation of pressure indication led Control room staff to determine that (the) EPR (Electronic Pressure Regulator) was not functioning properly (noise in the output signal). Control Room Staff entered Special Operating Procedure for failed pressure regulator. EPR could not be moved and this was confirmed by operators in the field. Control Room Staff (then) inserted a manual scram. Immediately after the scram reactor water level reached a low of 36", Emergency Operating Procedures for Level (EOP-2) were entered. HPCI initiated on the turbine trip to control water level. After the turbine tripped, all turbine bypass valves failed open; MSIVs (main steam isolation valve) were manually shut to control pressure. (The) EPR eventually disengaged from control, allowing the operator control of the turbine bypass valves. MSIVs were then reopened. (The) Scram has been reset. (The) turbine driven shaft pump did not initially disengage, pump (was) manually tripped after turbine speed reduced to 1500 rpm. All other systems responded correctly. (The) plant is not currently in any SOPs or EOPs and is proceeding to cold shutdown using normal operating procedures. All control rods fully inserted as expected. The plant is in a normal shutdown electrical lineup. At the time of the event, containment spray loop 1-12 was out of service for routine surveillance. The plant is currently cooling down and is at 365 psi. The licensee notified the NRC Resident Inspector.
ENS 43807	27 November 2007 15:15:00	10 CFR 50.72(a)(1)(i), Emergency Class Declaration	Alert Declared Due to Oxygen Deficient Atmosphere	An ALERT was declared at 0915 CST. A nitrogen freeze seal was being used on a 3 inch pipe in the 1B RHR and 1B Containment Spray pump room. A leak developed on the freeze seal jacket. This created an oxygen deficient atmosphere (less than 19.5 percent oxygen). The nitrogen supply was isolated. And the room was evacuated. The alert was declared in accordance with Emergency Action Level (EAL) "hazard alert #7 (HA7)." Additional ventilation was being placed in the area to restore normal oxygen levels. EAL table entry HA7 is for "Release of toxic or flammable gases within or restricting access to a vital area which jeopardizes operation of systems required to maintain safe operations or establish or maintain safe shutdown. No personnel injuries occurred. Oxygen levels were returned to acceptable levels at 0933 CST. The licensee notified the NRC Resident Inspector. * * UPDATE AT 1205 ON 11/27/2007 FROM PAUL CANTWELL TO MARK ABRAMOVITZ * * * The ALERT was terminated at 1103 CST. Notified IRD (McDermott), R3DO (Lipa), DHS SWO (Mary Anne), FEMA (Dan Sullivan), DOE (Anthony Parsons), EPA (Ms Ross), USDA (Andrew Watts), and USDA (Rick Turner).
ENS 43782	14 November 2007 19:00:00	10 CFR 21.21, Notification of failure to comply or existence of a defect and its evaluation	Part 21 Notification - Areva 4Kv Cutler Hammer Breakers	The licensee provided the following information via facsimile: In accordance with 10CFR21.21(d)(3), Southern Nuclear Operating Company (SNC) is making notification of a defect in a basic component supplied to Joseph M. Farley Nuclear Plant (Farley). A 10CFR21 report regarding a defect associated with Model MA-VR-350 4160 V circuit breakers supplied by AREVA was made by AREVA to SNC on October 3, 2007. The breaker design incorporates the use of a C-clip which may not have been properly installed or that can become dislodged from its groove on the Main Link Assembly pin which holds the Banana Link in place. If the Banana Link becomes disengaged from the Main link Assembly pin, the breaker will charge, but not close or it will leave the breaker in a 'trip-free' condition. The Model MA-VR-350 4160 V circuit breakers are used in the plant safety related 4160 V switchgear and serve as pump motor supply breakers for multiple safety related applications, e.g., component cooling water, low-head and high-head safety injection, containment spray, auxiliary feedwater, as well as the emergency diesel generator output breakers. Currently, there are breakers in stock and installed. Consequentially, their postulated failure in these critical applications could create a substantial safety hazard. Existing plant procedures already included pre-installation inspection steps for the Model MA-VCR-350 4160 V circuit breakers to identify loose nuts, bolts, retaining rings, or other hardware. In response to this concern, SNC revised plant procedures to add the C-clips to the inspection list to verify they are properly seated on the main link. Given the multiple examinations that were being conducted on the breakers in accordance with existing procedures, and the subsequent procedure enhancements that have been made to examine the C-clips, SNC determined that the installed breakers would continue to operate as designed on demand. As recommended by AREVA, a visual inspection, of the Model MA-VR-350 4160 V circuit breakers should be performed at regular maintenance intervals to insure proper installation of the C-clip on the main link assembly. SNC has been in contact with NRC Region II (Scott Shaffer, Chuck Casto) and has notified the NRC Resident Inspector.

Semantic search

Query log

Navigation menu