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	Start date	Report date	Site	Reporting criterion	System	Event description
05000289/LER-2017-004	5 February 2018		Three Mile Island Three Mile Island Unit 1	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Emergency Diesel Generator	On December 6, 2017 Three Mile Island Unit 1, determined that both Emergency Diesel Generators do not conform with the licensing bases for protection against tornado generated missiles. The vent on the common Fuel Oil Supply Tank that serves both Emergency Diesel Generators could be damaged by debris generated from a tornado that could affect emergency diesel generator operation. An extent of condition review identified three additional items that are nonconforming to the design for tornado missile protection: vent stacks for each Emergency Diesel Generator Day Fuel Tanks, the Borated Water Storage Tank and steam piping near the main feed pumps that could affect Secondary Pressure Control. Upon determination of the initial nonconformance for the Fuel Oil Supply Tank Vent on December 6, 2017, both Emergency Diesel Generators were declared inoperable. Compensatory measures were put and verified in place in accordance with the NRC Enforcement Guidance Memorandum EGM 15-002, both emergency diesel generators were returned to an operable but nonconforming status and an 8 hour ENS Notification was made to the NRC. This condition has been in existence since original licensing of the plant. It is not known if it was overlooked or considered acceptable at the time of the original licensing process. There are no actual consequences as a result of the nonconforming conditions.
05000266/LER-2017-001	18 September 2017 16 November 2017	16 November 2017	Point Beach	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability		At 1724 (CDT) on 9/18/17 Door-061, South Control Room Door (DR) was inadvertently disabled. The door became wedged open against its backstop during control room ventilation testing. Door-061 is a barrier that functions to maintain the control room envelope (NA). The barrier was subsequently disengaged from the backstop allowing it to close. The door was inspected and returned to operable status at 1750 (CDT). While the door was stuck open, the control room was in an unanalyzed condition, a condition that could have prevented fulfillment of a safety function, and a common cause inoperability of independent trains or channels. This event is being reported pursuant to 10 CFR 50.73(a)(2)(ii)(B), 10 CFR 50.73(a)(2)(v)(A), 10 CFR 50.73(a)(2)(v)(D), and 10 CFR 50.73(a)(2)(vii) for a degraded barrier that affected the control room envelope.
05000483/LER-2017-002	15 August 2017 13 October 2017	13 October 2017	Callaway	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System	Steam Generator Service water Emergency Diesel Generator Auxiliary Feedwater Main Steam Safety Valve Decay Heat Removal Main Steam Line Main Steam	On August 15, 2017, Callaway Plant was in Mode 1 at 100 percent power. During evaluation of protection for safety-related equipment from the damaging effects of tornados, Callaway Plant personnel determined that the minimum-flow recirculation lines for the turbine-driven auxiliary feedwater pump (TDAFP) and both motor-driven auxiliary feedwater pumps (MDAFPs) could be damaged if a postulated tornado-generated missile were to penetrate the condensate storage tank (CST) valve house and strike the lines. In response, Operations declared all three auxiliary feedwater pumps inoperable. Compensatory measures were implemented consistent with Enforcement Guidance Memorandum (EGM) 15-002, "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance." Upon completion of the initial compensatory measures, the TDAFP and MDAFPs were declared Operable but nonconforming. Subsequent to the condition identified on August 15, 2017, continued investigation of tornado missile vulnerabilities led to discovery that the exposed steam exhaust stacks for the main steam safety valves and atmospheric steam dump valves, as well as the exposed vents for the diesel generator fuel oil storage and day tanks, were also susceptible to tornado missile damage to the extent that compliance with General Design Criterion 2 is not ensured. Compensatory measures were then promptly implemented for these conditions, as well, in accordance with EGM 15-002 such that the affected systems have been evaluated to be nonconforming but Operable. It has been determined that the identified noncomformances are an original plant design legacy issue. Long-term resolution for establishing compliance is under development and will be completed within the time frame described in the EGM.
05000382/LER-2017-002	17 July 2017 18 September 2017	18 September 2017	Waterford Waterford Steam Electric Station, Unit 3	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Steam Generator Reactor Coolant System Feedwater Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Main Transformer Main Turbine	On July 17, 2017, at 1606 CDT, Waterford 3 experienced an automatic reactor scram due to a loss of forced circulation, which was the result of a loss of off-site power to the safety and non-safety electrical busses. Prior to the scram, plant operators manually tripped the main turbine and generator due to overheating of the isophase bus duct due to the failure of a shunt assembly connection in the duct to Main Transformer 'B'. The automatic electrical bus transfer did not occur due to relay failures in the fast dead bus transfer system. Both 'A' and 'B' Emergency Diesel Generators started and loaded as designed to re-energize the 'A' and 'B' safety busses. The loss of off-site power caused a loss of both Main Feedwater pumps, resulting in an automatic actuation of the Emergency Feedwater system. The Root Cause of this event was the design change procedure used for modifications to the fast dead bus transfer circuitry did not include guidance to detect the susceptibility of the relays to DC coil inductive kick. The faulty relays in the fast bus transfer circuit were replaced prior to plant startup. An Unusual Event was declared at 1617 CDT due to loss of off-site power to safety buses for >15 minutes. All required safety-related equipment responded as expected during this event.
05000346/LER-2017-001	20 July 2017 18 September 2017	18 September 2017	Davis Besse	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Emergency Diesel Generator	On July 20, 2017, with the Davis-Besse Nuclear Power Station (DBNPS) operating at approximately 100 percent power, it was identified that the Emergency Diesel Generator (EDG) fuel oil storage tank vents were not adequately protected from potential tornado-generated missiles. If a missile crimped the vent it could disable the transfer pump or tank, potentially impacting the seven-day fuel supply for the affected train(s) of EDG. While the storage tanks were protected from tornado missiles when installed, the vents were not provided with any such protection. Compensatory measures were established to ensure a vent path remained following a tornado event, and actions will be taken to ensure the vents for each EDG fuel oil storage tank are adequately protected from tornado missiles. This issue is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degraded plant safety, in accordance with 10 CFR 50.73(a)(2)(v) as a condition that could have prevented the fulfillment of the safety function, in accordance with 10 CFR 50.73(a)(2)(vii) as an event where a single cause or condition caused two independent trains to become inoperable in a single system, and in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.
05000334/LER-2017-002	19 July 2017 13 September 2017	13 September 2017	Beaver Valley	10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Emergency Diesel Generator	In order to address the concerns outlined in NRC Regulatory Issue Summary (RIS) 2015-06 "TORNADO MISSILE PROTECTION", evaluations of tornado missile vulnerabilities and their potential impact on Technical Specification (TS) plant equipment were conducted. This particular evaluation concluded that the following Structures, Systems, and Components (SSCs) are potentially vulnerable to tornado generated missiles: The Beaver Valley Power Station Unit 1 (BV-1) Emergency Diesel Generators (EDGs) engine exhaust piping is potentially vulnerable as a result of tornado generated missiles striking and subsequently crimping or crushing this piping rendering the EDGs inoperable. On July 19, 2017, both of the BV-1 TS required EDGs were declared inoperable and Enforcement Guidance Memorandum (EGM) 15-002 Rev 1 "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance," was applied. Compensatory measures were implemented within the time allowed by the applicable Limiting Condition(s) for Operation and both EDGs were then declared operable but nonconforming. The apparent cause of this issue was a lack of clarity during the original design and licensing of the plant that led to inadequate understanding of the tornado missile protection regulatory requirements. Actions will be taken to establish compliance for BV-1 EDGs either by a plant modification or employing a methodology for addressing tornado missile non- conformances for the EDG exhaust piping. This issue is reportable under 10 CFR 50.72 for a loss of safety function. However, enforcement discretion is being applied. As stated in EGM 15-002, Rev. 1, the NRC will exercise enforcement discretion for subsequent tornado missile 10 CFR 50.72 notifications. On February 23, 2017, FENOC provided the NRC the initial 10 CFR 50.72 notification in Event Notification (EN) number 52571 concerning tornado missile protection issues known at that time.
05000298/LER-2017-005	17 August 2017		Cooper	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Primary Containment Isolation System Primary containment	On June 21, 2017, Traversing In-core Probe (TIP) C failed to stop at its in-shield position when being withdrawn from the core. Cooper Nuclear Station (CNS) Operations personnel declared the associated TIP C ball valve inoperable as a Primary Containment Isolation Valve (PCIV) at 0524 and entered Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.1.3, Condition A. On June 22, 2017, TIP D failed to stop at its in-shield position when being withdrawn from the core. CNS Operations personnel declared the associated TIP D ball valve inoperable as a PCIV at 0445 and entered TS LCO 3.6.1.3, Condition A. Subsequent investigation determined the cause of the failures was inadequate mounting and securing of the in- shield limit switch to the chamber shield. Corrective actions included repair of the mounting of the in-shield limit switches for all TIP channels, and improved procedure guidance for properly mounting the in-shield limit switches. Both valves were declared operable on July 13, 2017, at 1133. There were no safety consequences associated with this condition.
05000286/LER-2017-001	14 May 2017 13 July 2017	13 July 2017	Indian Point Indian Point Unit 3	10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability	Reactor Coolant System Residual Heat Removal Emergency Core Cooling System	On May 14, 2017, at 0233 hrs, Indian Point Unit 3 entered Mode 4 as part of coming out of outage 3R19 and preparing for power operations. Operations test group was preparing for performance of 3-PT- CS004, Residual Heat Removal (RHR) Check Valve Testing. The team gathered for a pre job brief in accordance with the requirements of EN-HU-102, Human Performance Traps &Tools Procedure. At the time the only allowable access point to the Inner Crane Wall was through the double gate combination of Gates D and E, which require one gate to be maintained closed and secured at all times. Workers needed to enter inside of the Crane Wall to perform a portion of the valve lineup required by 3-PT-CS004. After unbolting and opening the gate, the two operators and a contract Radiation Protection (RP) Technician went through gate C despite a posted sign stating that the gate was not to be utilized in modes 1 through 4. While the valve manipulations were in progress the NRC Resident Inspector was also conducting a tour of the Vapor Containment (VC) and identified that gate C was opened. This gate being open in this plant condition resulted in a safety system functional failure, since with the gate unsecured this made the containment sumps inoperable.
05000334/LER-2017-001	3 February 2017 18 April 2017	18 April 2017	Beaver Valley	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System	Auxiliary Feedwater Main Steam Safety Valve Decay Heat Removal Main Steam	In order to address the concerns outlined in NRC Regulatory Issue Summary (RIS) 2015-06 "TORNADO MISSILE PROTECTION", an evaluation of tornado missile vulnerabilities and their potential impact on Technical Specification (TS) plant equipment was conducted. This evaluation concluded that the following Structures, Systems, and Components (SSCs) are potentially vulnerable to tornado generated missiles: The steam discharge flow paths to atmosphere of the Beaver Valley Power Station Unit 1 (BV-1) and Unit 2 (BV-2) Main Steam Safety Valves (MSSVs) (reference TS 3.7.1) are potentially vulnerable to tornado generated missiles. The steam discharge flow paths to atmosphere of the BV-1 and BV-2 Atmospheric Dump Valves (ADVs) (reference TS 3.7.4) are potentially vulnerable to tornado generated missiles. On February 23, 2017, the BV-1 and BV-2 TS required MSSVs and ADVs were declared inoperable and Enforcement Guidance Memorandum (EGM) 15-002 Rev 1 "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance," was applied. Compensatory measures were implemented within the time allowed by the applicable Limiting Condition(s) for Operation and the associated systems were then declared Operable but nonconforming. The apparent cause of this issue was a lack of clarity during the original design and licensing of the plants that led to inadequate understanding of the tornado missile protection regulatory requirements. In addition, as part of the evaluation of tornado missile vulnerabilities, two BV-2 tornado missile barrier doors were found to be open. Specifically, Auxiliary Building door (A-35-5A) was found open and Fuel Building door (F-66-3), was found to be partially open. These doors were then closed and latched. Actions will be taken to establish compliance for the MSSVs and ADVs either by plant modification or by employing a methodology for addressing tornado missile noncompliance for the MSSVs and the ADVs. These conditions (as applicable) were reported to the NRC on February 23, 2017 in Event Notification (EN) number 52571 under 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(A).
05000277/LER-2017-001	8 March 2017		Peach Bottom	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Emergency Diesel Generator	On 1/9/17, it was determined that the site's Emergency Diesel Generators do not conform with the licensing basis for protection against tornado generated missiles. The exhaust stacks for the four on-site diesel generators extend approximately seven feet above the roof of the diesel generator building. In the event of a tornado, debris generated from the tornado could strike the exhaust stacks and, if at a sufficient mass and velocity, could crimp the exhaust stacks in a manner that would affect diesel generator operation. As a result of the non-conforming condition, on 1/9/17 at 1530, all four emergency diesel generators were declared inoperable. Compensatory measures were put in place and, in accordance with NRC guidance contained in EGM 15-002, the diesel generators were returned to an operable but non-conforming status. This condition has been in existence since original licensing of the plant. It is not known if it was overlooked or considered acceptable at the time of the original licensing process. There are no actual consequences as a result of the non-conforming condition.
05000445/LER-2016-002	22 February 2017		Comanche Peak	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System	Service water Emergency Diesel Generator Containment Spray	On September 13, 2016, and September 14, 2016, during plant walk downs by Engineering and the NRC Senior Resident inspector, pressurized fire protection piping in the Service Water Intake Structure was found to not be shielded against a Moderate Energy Line Break (MELB), resulting in inoperability of one train of Service Water for both units. During extent of condition walk downs conducted on October 6, 2016, October 10, 2016, November 17, 2016, December 5, 2016, and December 22, 2016, additional piping in the Unit 1 and Unit 2 Safeguards and Auxiliary Buildings was found to not be shielded against a MELB, resulting in inoperability of one train of various.safety related equipment for both units. The most likely cause of this event was the methodology used to conduct the initial MELB walk downs was flawed and allowed some MELB threats to be missed. Corrective actions include shielding the affected piping, performing a 100 percent walk down of rooms containing MELB piping identified for shielding, and revising the systems interaction program maintenance procedure. I All times in this report are approximate and Central Time unless noted otherwise.
05000390/LER-2017-001	10 November 2016 14 February 2017	9 January 2017	Watts Bar	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability		On November 10, 2016, Watts Bar Nuclear Plant personnel identified a failure of the non-reverse clutch key on Emergency Raw Cooling Water (ERCW) motor B-A. While performing a lubrication work order, it was discovered that the clutch key was sheared. Subsequent investigation identified that other clutch key failures had occurred in the recent past. The non-reverse clutch prevents the ERCW pump from rotating in the reverse direction after pump trip, which could cause the motor to develop a higher than normal in-rush current if the motor was subsequently started, such as following an accident. Based on the potential common mode failure of the non-reverse clutch, immediate corrective actions were put in place to ensure that the safety function of the ERCW pumps to start following an accident would not be impaired. The cause of the failure is under investigation.
05000341/LER-2016-001	6 January 2016 23 January 2017	23 January 2017	Fermi	10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability	Reactor Protection System	At 1514 EST on January 6, 2016, while operating at 100 percent Reactor Thermal Power (RTP), the East and West Turbine Bypass Valves (TBV) automatically opened as expected for 3 minutes and 32 seconds in response to the number one High Pressure Turbine Stop Valve (TSV) drifting from full open to 25 percent open due to an actuator malfunction. Per Technical Specification (TS) Bases 3.3.1.1, TBVs must remain shut while RTP is at or above 29.5 percent to consider all channels of the TSV closure and Turbine Control Valve (TCV) fast closure Reactor Protection System (RPS) functions operable. Reactor Operators lowered RTP to 91.0 percent and at 1518 EST the TBV automatically closed and the TSV closure and TCV fast closure RPS functions were no longer considered inoperable. TS 3.3.1.1 requires that the TSV closure and TCV fast closure RPS functions be operable at or above 29.5 percent RTP. In this event, during the period of time while TBVs were open, reactor power was maintained above 91 percent and the RPS functions were confirmed to be enabled. The actuator malfunction was caused by faulty connectors within the actuator. The faulty connectors were replaced.
05000400/LER-2016-001	7 July 2016 1 September 2016	1 September 2016	Harris	10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Emergency Diesel Generator Safety Relief Valve Main Steam	On July 7, 2016, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Shearon Harris Nuclear Power Plant identified nonconforming conditions in the plant design such that specific TS equipment did not meet current design basis for protection against potential tornado missile impact. Identified systems were declared inoperable and Enforcement Guidance Memorandum (EGM) 15-002, "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance," was implemented. Compensatory measures were implemented within the time allowed by the applicable Limiting Condition(s) for Operation and the associated systems were then declared operable but nonconforming. The two systems identified with credible impacts were the 'A' train Emergency Diesel Generator and the Main Steam Safety Relief Valves. Actions will be taken to establish compliance, either by plant modification or by employing a methodology for addressing tornado missile noncompliances. Due to the historical nature of the issue, a specific cause for the identified vulnerabilities was not determined.
05000317/LER-2016-004	13 November 2015 20 July 2016	20 July 2016	Calvert Cliffs	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability	Service water Emergency Diesel Generator Auxiliary Feedwater	The Unit 1 Service Water (SRW) Room high energy line break (HELB) door was opened on 10/19/2015 and 10/21/2015 to conduct a maintenance activity. The door was opened twice for approximately three and a half minutes each time. With the HELB door blocked open, the affected equipment located behind the barrier door should have been considered inoperable. Affected equipment included the 13 motor driven auxiliary feedwater pump, both trains of saltwater air compressors and both trains of SRW. Since both SRW trains were inoperable, a loss of safety function occurred, an unanalyzed condition existed and a common cause failure of independent trains existed for the time that the HELB door was opened. The HELB door was blocked opened because Maintenance Planners did not include the proper barrier controls in the work order that opened the HELB door, because they were unaware of the barrier control procedure requirements. The apparent cause of the event was that the assigned change management agent (an Engineer) failed to inform Maintenance Planners of the implementation of the barrier procedure. Affected groups were briefed about the barrier control procedure. Corrective action for the human performance issue was handled through the performance management system.
05000293/LER-2016-002	19 April 2016 20 June 2016	18 August 2016	Pilgrim Pilgriin Nuclear Power Station	10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System	Emergency Diesel Generator	, On April 19, 2016, at approximately 1450 hours, it was discovered that a maintenance activity performed between 2010 hours on August 26, 2014 and 0143 hours on August 27, 2014, had rendered the Startup Transformer (X4) and the standby Emergency Diesel Generators (EDG) (X-107A&B) unable to automatically supply power to Buses A5 and A6, due to the breaker interlock that would prevent Startup Transformer breakers (152-504 and 152-604) and standby EDG breakers (152-509 and 152-609) from closing, when Bus A8 to Bus A5 breaker (152-501) and Bus A8 to Bus A6 breaker (52-601) are in the TEST position and CLOSED. During the maintenance activity, the plant was operating at 100 percent power and the_ Unit Auxiliary Transformer (X3) was providing power to Emergency Buses A5/A6. The functional testing of negative sequence relays (146-600/A and B) and 23kV feed undervoltage relays (127-600A/1 and 2, and 127-600B/1 and 2) created a test configuration, lasting less than 1-hour, whereby power to Buses A5 and A6 was not automatically available from either the startup transformer or from the EDGs. As a result, Limiting Conditions for Operation (LCO) Action Statement 3.9.6.2 was not met. The root cause is that the decision to perform the described surveillance testing online, instead of during cold shutdown, lacked sufficient rigor to ensure compliance with Technical Spepifications. Corrective actions will establish and institutionalize expectations and accountability for station leadership regarding consequence-biased decision-making and effective risk manaaement. There was no impact to_public health and safety.
05000391/LER-2016-001	14 April 2016 13 June 2016	13 June 2016	Watts Bar	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Steam Generator Auxiliary Feedwater	From March 18, 2016, when Watts Bar Nuclear Plant Unit 2 first entered Mode 4 to April 14, 2016 with the plant in Mode 3, it was determined that a condition prohibited by Technical Specifications (TS) existed. During this time both automatic and manual closure of the containment isolation valves and the sample isolation valves for the Steam Generator Blowdown (SGBD) sampling lines were disabled due to improperly installed electrical jumpers in the valve control circuits. The misplaced jumpers bypassed the Phase A containment isolation signals, the auto/manual start signals for the Auxiliary Feedwater (AFW) pumps, and the control valve seal-in circuits. Containment isolation on a Phase A signal is used to control potential release of radioactive material to the environ in the event of a Design Bases Accident. The AFW pump auto/manual start signals are used to isolate the SGBD sampling lines to preserve steam generator inventory. The seal-in circuits are used to allow the operator to manually position the valves in either the open or closed position from the main control room. This event occurred prior to initial reactor criticality. There was no loss of safety function. The isolation valves for the SGBD sample lines were returned to service on April 14, 2016. This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(vii)(B) and (C).
05000247/LER-2016-007	10 June 2016	9 August 2016	Indian Point	10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability	Reactor Coolant System Residual Heat Removal Emergency Core Cooling System	On June 20, 2016, Entergy management was advised by the NRC that during a tour in containment while the unit was in Mode 4, the inspector identified two open barrier gates for the Emergency Core Cooling System (ECCS) sump. Personnel were moving scaffolding from inside the crane wall to areas outside the crane wall through the two open barrier gates. Having both sump barrier gates open violated ECCS operability basis which requires the sump barrier system to be operable in Modes 1-4. The inspector notified the operator touring with him of the observation. The operator subsequently coached the Radiation Protection (RP) door guard to ensure that one of the gates be closed at all times. The apparent cause was a latent organizational weakness associated with the use of procedure OAP-007 (Containment Entry and Egress) which had not been communicated well within the organization. The failure mode was personnel not being aware of all available information. The scaffold supervisor was not aware of his requirement to serve as containment coordinator and provide the required briefing on gate closure. The RP brief was focused on the locked high radiation requirements not gate control. Corrective actions included closing and securing one gate, briefing RP personnel on the event, the lessons learned and management expectations. This event will be included in all 3R19 supplemental supervisors qualifications required reading list. Procedure OAP- 007 will be revised to include a checklist for entry briefings to include GS-191 requirements. The event had no significant effect on public health and safety. Indian Point 2 05000-247 APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by intemet e-mail to Note: The Energy Industry Identification System Codes are identified within the brackets (). Contents 1 DESCRIPTION OF EVENT 2 CAUSE OF EVENT 3 CORRECTIVE ACTIONS 4 EVENT ANALYSIS 5 PAST SIMILAR EVENTS 6 SAFETY SIGNIFICANCE DESCRIPTION OF EVENT On June 20, 2016, Indian Point management was advised by the NRC that during a tour in containment (NH) while the unit was in Mode 4, the inspector identified two open barrier gates for the Emergency Core Cooling System (ECCS) (BQ) sump. Personnel were removing disassembled scaffolding from inside the crane wall on the 46 foot elevation of containment and moving it through the ECCS sump barrier gates (GATE) to areas outside the crane wall through the two open barrier gates. Having both sump barrier gates open violated ECCS operability basis which requires the sump barrier system to be operable in Modes 1-4. The plant had entered Mode 4 on July 10, 2016, at 23:30 hours. The inspector notified the operator touring with him of the observation. The operator subsequently coached the Radiation Protection (RP) door guard to ensure that one of the gates were closed at all times. No condition report recorded the event at the time. Subsequently, on June 20, 2016, after an NRC inspector advised a site manager the condition was recorded in the Indian Point Energy Center (IPEC) Corrective Action Program (CAP) as Condition Report CR-IP2- 2016-04036. On June 21, 2016, CR-IP2-2016-04037 recorded the failure to initiate a CR at the time of the event. For postulated breaks in the Reactor Coolant System (RCS) (AB) there are two recirculation related sumps within the containment, 1) the Recirculation Sump and, 2) the Containment Sump. Both sumps collect liquids discharged into the containment during a design basis accident. As part of the resolution of Generic Safety Issue (GSI)-191 (Assessment of Debris Accumulation on PWR Sump Performance) and Generic Letter 2004-02 (Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors) various flow barrier debris interceptors were installed in the containment to channel the recirculation flow into the reactor cavity sump area, up and out of the Incore Instrumentation Tunnel, through the crane wall and containment sump labyrinth wall via specially designed openings, and into the annulus area outside the crane wall. The recirculation flow will migrate towards the Recirculation Sump or the Containment Sump depending on which pump(s) are operating. Flow channeling barriers are installed on the Reactor Cavity Sump around the Incore Instrumentation Tunnel, on the Recirculation Sump trenches, and at the Containment Sump. Flow channeling barrier gates are installed in the northeast and northwest quadrant openings of the Crane Wall. In addition, flow channeling barrier gates are installed in the north and south entrances to the Recirculation Sump area. There is one dual access gate (gates 17 and 23) to allow access without violating the flow barrier integrity during transient through the flow barrier system. comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by intemet e-mail to Investigation of the event determined that the scaffold work group was the largest group within containment during the time the scaffold job on the 46 foot elevation was being worked. In accordance with procedure OAP-007 (Containment Entry and Egress) the scaffold group was to assume the role of containment coordinator and be responsible for performing a containment entry briefing. However, instead of providing a containment entry briefing, a regular or standard HU pre-job brief was given to the scaffold workers by their contractor supervisor. Because the area being worked was a radiation controlled area the scaffold workers then met with Radiation Protection (RP) personnel at the Health Physics access (HP1) for a Locked High Radiation Area (LHRA) RP pre-job brief. The RP brief included discussion that one of the ECCS barrier gates had to be closed at all times per GSI-191 and OAP-007 requirements. The job also required an RP door guard whose only function was to ensure that anyone entering into the inside crane wall had to have an HP individual with them. There was no ECCS barrier gate monitor assignment as required by OAP- 007. As work progressed the scaffold workers left both ECCS barrier gates open to enhance removal of scaffold material to storage. Although the scaffold workers were told during the RP briefing that one ECCS barrier gate must remain closed at all times, it was discovered in interviews with workers that they thought that none of the other gates could be opened while they were using their gate location to remove scaffold. It was determined that the supplemental scaffold supervisor was not aware of a specific entry procedure that was required to be used prior to a containment entry. A specific OAP-007 procedure containment entry brief was not given to the workers nor were they and their supervisor aware of the procedure. The supplemental scaffold group supervisor stated that in previous containment entries he and his group were never the main group going into containment so they were always briefed by operations or RP and didn't recall using OAP-007. Procedure OAP-007 is specifically written to cover many aspects of containment entries. The procedure contains sections and steps discussing the ECCS barrier gates with diagrams of the crane wall and all of the gates with their locations and numbers. If the procedure had been used in addition to a pre-job and RP brief, the requirements would have been clearer to the workers and this event most likely would not have occurred. An extent of condition (EOC) review was performed and it determined that both units are similar and both would be vulnerable in an event if there was a direct flow path for accident debris to enter the containment/internal recirculation sump. The condition is bounded by the crane wall gates as these are the only types of gates in the containment installed in the crane wall that protect the ECCS sumps from debris. comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by intemet e-mail to CAUSE OF EVENT The apparent cause was a latent organizational weakness associated with the use of procedure OAP-007 (Containment Entry and Egress) which had not been communicated well within the organization. The failure mode was personnel not being aware of all available information. The scaffold supervisor was not aware of his requirement to serve as containment coordinator and provide the required briefing on gate closure. The RP brief was focused on the locked high radiation requirements not gate control. CORRECTIVE ACTIONS The following corrective actions have been or will be performed under the Corrective Action Program (CAP) to address the causes of this event: Dual gates were closed and applicable gate secured. The RP door guard was coached on requirement to have at least one gate closed and secured at all times. A HU meeting was held and interviews conducted with the work crew, supervisor and RP personnel and the requirements of OAP-007 were reviewed and requirements of the ECCS barrier reinforced. A Department Clock Reset/Yellow memo was prepared and the lessons learned on the event and management expectations communicated with Projects organizations, all site departments, and the Fleet. The event will be included in all 3R19 supplemental supervisors qualifications required reading list. Procedure OAP-007 will be revised to include a checklist for entry briefings to include GS-191 requirements. EVENT ANALYSIS The event is reportable under 10CFR50.73(a)(2)(v)(D) as a safety system functional failure as the condition could have prevented adequate post accident core cooling due to DBA debris blockage of the recirculation and/or the containment sump. An ECCS train is inoperable if it is not capable of delivering design flow to the RCS. Individual components are inoperable if they are not capable of performing their design function or supporting systems are not available. Technical Specification (TS) 3.5.2 (ECCS-Operating) requires three ECCS trains to be operable in Modes 1, 2 and 3, and TS 3.5.3 (ECCS-Shutdown) requires one ECCS residual heat removal (RHR) subsystem and one ECCS recirculation subsystem to be operable in Mode 4. comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by intemet e-mail to Indian Point 2 05000-247 The licensing and design basis of the ECCS per UFSAR Section 6.2.2 (ECCS System Design and Operation) credits flow channeling barriers installed in containment in response to the resolution of GL-2004-02. The two flow barrier gates that were used for removing scaffolding were not closed and secured to prevent it from being forced open during a DBA. The unsecured gates were not in accordance with design and not a sufficient robust barrier to prevent debris from entering the recirculation and containment sumps had a DBA occurred while in Mode 4. The condition is also reportable under 10CFR50.73(a)(2)(vii) (common cause inoperability of independent trains or channels) as the condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to (D) mitigate the consequences of an accident. The NRC inspector tour occurred during the time the unit was in Mode 4. The unit entered Mode 4 on June 10, 2016, at 23:30 hours. However, no CR was initiated for this condition at that time. CR- IP2-2016-04037 recorded that condition that while performing a walkdown with an NRC inspector on June 11, 2016, the NRC raised a question about an activity in the field and no condition report was initiated. PAST SIMILAR EVENTS A review was performed of the past three years of Licensee Event Reports (LERs) for events that involved SSFFs and/or common cause inoperability of an Engineered Safety Feature System that had a similar cause. No LERs were identified at Unit 2. A review of all reported events during the past three years at both units identified one LER at Unit 3 that was similar. Unit 3 LER-2013-002 reported on April 29, 2013, a Safety System Functional Failure and Common Cause Inoperability of the Emergency Core Cooling System due to violation of containment sump debris barrier integrity. The LER reported that on March 4, 2013, during shutdown for a refueling outage, Radiation Protection (RP) personnel entered the reactor containment building to install plastic RP fencing for the Reactor Coolant Drain Tank (RCDT). After receiving clearance at Mode 4 to enter the Inner Crane Wall (ICW) to install fencing around the RCDT and post it as a Locked High Radiation Area (LHRA). The RP work crew assumed they could enter the ICW area through any sump barrier gate for the Emergency Core Cooling System (ECCS). The RP work crew chose to use a single gate access point due to its proximity to the RCDT. Subsequently, a RP Technician identified that personnel had not entered the area using the double access gate and had brought in plastic fencing which was inappropriate material for the sump area. The apparent causes were an inadequate pre-job brief and inadequate procedure for Containment Entry and Egress (OAP-007, 0-RP-RWP-405) due to poor change management. The pre-job brief failed to cover the requirement to use the dual sump barrier gate access point when in Modes 1-4, nor did it address the type of fencing allowed. Corrective actions included revision of Procedure OAP-007 to clearly state that within the procedure's attachments that only the sump barrier dual access gate for 46 foot Containment ICW entries shall be used in Modes 1-4. comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by intemet e-mail to The revision specified the use of the double entry gate and that one gate is to remain shut and secured at all times. Securing the gates at unit 3 which uses a slide latch does not state the use of a gate monitor. The step for Unit 2 includes statements that the gates be secured with a padlock or nut and bolt closure from the outside. This condition requires posting of a gate monitor to allow exit. SAFETY SIGNIFICANCE This event had no significant effect on the health and safety of the public. There were no actual safety consequences for the event because there were no accidents or transients during the time of the event. The analysis performed in response to GL- 2004-02 included debris transport analysis conservatisms for transport of debris to both the IR sump and the Containment sump in excess of quantities that would be generated. Establishing normal RHR cooling to the RCS has RCS temperature below 350 degrees F and pressure less than 400 psig. In Mode 4 the reactor is not critical and reactivity is stable. In Mode 4 there is significantly less energy in the RCS to generate debris. At the time the actual RCS pressure (pressurizer pressure) was approximately 355 psig. An evaluation of a LOCA during Mode 3 and 4 operation was performed by Westinghouse (WCAP-12476) that showed a direct reduction in break probability for Mode 4. The evaluation concluded that Mode 4 LOCAs are not a significant contributor to shutdown risk. During this event the entire flow barrier was not disabled because only two debris barrier gates were unsecured and only for the time scaffold workers were allowed to perform assigned work. The exact time the gates were open cannot be determined as the barrier gates have no electronic timing devises. However, the scaffold workers were assigned three entries with stay time limitations for heat stress of 45 minutes each for a total job time of 135 minutes. The scaffold work assignment took place in Mode 4. The unit entered Mode 4 on June 10, 2016, at 23:30 hours. For the first two of three entries, the scaffold workers had to go inside the crane wall and disassemble erected scaffolding. Per the scaffold supervisor at least one door was closed during scaffold disassembly. Therefore approximately 60 minutes were left for moving disassembled scaffolding from inside the crane wall through the open gates to the outside crane wall storage areas. Therefore most debris would have been intercepted by the flow barrier system. Also, the barrier gates swing into the crane wall so that DBA flow and forces would tend to close the gate when pressure is applied (e.g., DBA debris loads) therefore limit flow barrier bypass and sump debris loading.
05000346/LER-2016-004	5 April 2016 6 June 2016	30 August 2016	Davis Besse	10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Steam Generator Reactor Coolant System Reactor Protection System Remote shutdown Control Rod	On April 5, 2016, at 0243, with the Davis-Besse Nuclear Power Station (DBNPS) in Mode 5, it was discovered that the six dual-element Resistance Temperature Detectors .(RTD) on both Reactor Coolant SyStem (RCS) Hot Legs had varying degrees of wire insulation degradation. The cause of the RTD insulation degradation was accelerated aging due to high temperatures as a result of improper configuration of piping insulation on the RTD during the previous refueling outage. Corrective actions taken include replacing five of the six RCS Hot Leg RTDs. The sixth RTD was determined to be suitable for continued operation and is planned to be replaced during the next Refueling Outage. The Electrical Conductor Seal Assemblies (ECSA) taken from the removed RTDs were evaluated for reuse on replacement RTDs. It was discovered that the Midlock Ferrules were installed backwards on two RTDs during the past refueling outage. The cause was determined to be less than adequate installation instructions. Corrective actions are to revise the Maintenance procedure and implement further training. This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B), 10 CFR 50.73(a)(2)(v)(A), and 10 CFR 50.73(a)(2)(vii).
05000333/LER-2016-002	25 February 2016 25 April 2016	25 April 2016	FitzPatrick	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability	Reactor Coolant System Main Steam Isolation Valve Primary containment Main Turbine Main Steam Line	On January 23, 2016, James A. FitzPatrick Nuclear Power Plant (JAF) initiated a manual Scram in response to lowering screenwell water level due to frazil ice blockage, and subsequently closed the Main Steam Isolation Valves (MSIV). A post Scram review identified that MSIV 29A0V-8613 closed slowly. On January 26, 2016, testing per ST-1B identified that MSIV 29A0V-86C closed slowly. In both cases, the inboard MSIVs performed satisfactorily. Troubleshooting identified that the problem originated in the solenoid valve cluster assemblies (SVCA) and they were replaced and tested successfully. A failure analysis was performed by Exelon PowerLabs on the SVCAs. On February 25, 2016, the Exelon PowerLabs analysis concluded that the DC pilot valves, 2950V-86B3 and 2950V-86C3, exhibited slow vent times. Additional corrective actions include changing the preventative maintenance frequency from 8 years to 6 years and initiating further investigation through the component's vendor. Two MSIVs exceeded the closing time of Technical Specification Surveillance Requirement (SR) 3.6.1.3.6. This condition caused two independent channels of a system used to control the release of radioactive material to become inoperable; reportable per 10 CFR 50.73(a)(2)(vii).
05000341/LER-2016-002	22 March 2016		Fermi	10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability	Feedwater High Pressure Coolant Injection Reactor Core Isolation Cooling Primary containment Reactor Pressure Vessel Core Spray Residual Heat Removal Low Pressure Coolant Injection	On January 22, 2016, at 1923 EST, both divisions of the Residual Heat Removal (RHR) system were declared inoperable for the Low Pressure Coolant Injection (LPCI) mode of operation due to a failure of the division 1 LPCI outboard injection motor operated valve (MOV), El 1 50F017A. While performing the division 1 RHR pump and valve operability surveillance test, E1150F017A closed properly but failed to open during its required stroke time test. With this valve closed and unable to automatically open, LPCI injection into the Reactor Pressure Vessel (RPV) from both divisions of RHR would be prevented if the LPCI loop select logic selected the division 1 recirculation loop for injection; therefore, this failure rendered both divisions of RHR inoperable for the LPCI function. Technical Specification limiting condition for operation (LCO) 3.5.1, Condition K, was entered, which requires immediate entry into LCO 3.0.3. The cause of the failure was subsequently identified as a foreign material (screw) that affected the function of the MOV contactor. The root cause was determined to be less than adequate inspection procedures and susceptibility of the contactor to foreign material. Inspection of all other susceptible equipment is ongoing to tighten loose screws and a modification is planned to install Foreign Material Exclusion (FME) barriers.
05000261/LER-2016-001	19 January 2016 21 March 2016	21 March 2016	Robinson	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)	Steam Generator Service water Auxiliary Feedwater	At 1540 hours EST on 01/19/2016 with H. B. Robinson Steam Electric Plant, Unit No. 2, in Mode 1 at 100 percent power, Motor- Driven Auxiliary Feedwater (MDAFW) pump "B" failed its surveillance test due to flow switch FSL-1633B failure to indicate service water flow to the Oil Cooler and Packing Channel Cooler. It was discovered that flow was blocked due to stem and disc separation on the MDAFW pump "A" and "B" cooling flow return isolation valve, SW-115. This blockage rendered both MDAFW trains inoperable for a period longer than allowed by plant Technical Specifications (TS). TS 3.7.4, Auxiliary Feedwater (AFW) System, requires four AFW flow paths and three AFW pumps be operable in Modes 1, 2, and 3, and Mode 4 when steam generator is used for heat removal, allowing 24 hours to restore one of two inoperable MDAFW pumps to operable status. Since two MDAFW pumps were inoperable due to common cause for approximately 52 hours, this condition is reportable as a condition prohibited by TS under 10 CFR 50.73(a)(2)(i) (B), and reportable under 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability of Independent Trains or Channels. The total duration of both trains of MDAFW being unavailable was limited to approximately 3 hours. An investigation concluded that the failed valve (SW-115) was likely installed in poor condition, without documentation, and outside of design specifications prior to 1978. The valve has been replaced with a valve meeting current design specifications. The health and safety of the public was not jeopardized as a result of this condition.
05000333/LER-2015-008	18 December 2015 16 February 2016	16 February 2016	FitzPatrick	10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Steam Generator Emergency Diesel Generator Primary containment Automatic Depressurization System Control Rod	On December 18, 2015, James A. FitzPatrick Nuclear Power Plant (JAF) was operating at 100 percent power when a 10 CFR 21.21(d)(3)(ii) Notification was received from Nutherm International. It identified a defect in Moore Industries temperature transmitters. Specifically, insulation was damaged in the T2 transformer during assembly which could result in premature failure. These components were installed starting in June 2015 at 27TT-113A and 27TT-113B in the Containment Atmosphere Dilution (CAD) system. The defect caused failures in July and November which resulted in either the "A" or "B" CAD subsystem isolating. Corrective actions included replacing both temperature transmitters with ones that were confirmed to not contain this defect. Even though these defective temperature transmitters function appropriately until they fail, this defect reduced the reliability of the CAD system to perform its function for its entire mission time. Therefore, this deficiency resulted in a loss of safety function to mitigate the consequences of an accident, reportable per 10 CFR 50.73(a)(2)(v)(D). Also, a single cause affected the safety function of independent CAD trains, reportable per 10 CFR 50.73(a)(2)(vii)(D); and, this condition existed longer then allowed by Technical Specifications 3.6.3.2, reportable per 10 CFR 50.73(a)(2)(i)(B).
05000331/LER-2015-002	31 March 2015 28 January 2016	28 January 2016	Duane Arnold	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	High Pressure Coolant Injection Reactor Core Isolation Cooling Primary containment Core Spray Residual Heat Removal Emergency Core Cooling System	On March 31, 2015, while operating at 100% power, with no structures, systems, or components inoperable, an unanalyzed condition regarding the primary containment suppression pool coating was identified. Specifically, during an inspection of suppression pool (torus) during the October 2014 refueling outage, degradation of the torus coating was discovered. Some of the coating had become delaminated. NextEra Energy Duane Arnold took immediate action to restore the coating to within design parameters during the refueling outage and the degraded condition no longer exists. Extensive analysis was performed to determine effect of the delaminated material. Upon completion of this investigation, it was determined that an unanalyzed condition, a condition prohibited by Technical Specifications, an event or condition that could have prevented fulfillment of a safety function and common cause inoperability existed and is reporting the condition under various sections of 10 CFR 50.73. The root causes of this event were less than adequate coating application specification and work instructions and less than adequate project oversight and control.
05000259/LER-2015-005	29 October 2015	28 December 2015	Browns Ferry	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability	Main Steam Isolation Valve Primary containment Automatic Depressurization System	On October 29, 2015 at 1149 Central Daylight Time, it was discovered upon receipt of a vendor report that the Main Steam Isolation Valve (MSIV) accumulators on all BFN inboard MSIVs are of insufficient size to provide the MSIV actuators adequate air volume, at the required pressure, to close the MSIV during a Loss of Coolant Accident (LOCA). Therefore, availability of Drywell Control Air (DWCA) nitrogen from the Containment Inerting system or from the Containment Atmospheric Dilution system was determined to be necessary for operability of inboard MSIVs. From December 1, 2012, to the time of discovery, there were multiple occasions where BFN Unit 1, 2, or 3 DWCA systems were aligned to receive nitrogen from the Plant Control Air system, resulting in the inoperability of multiple MSIVs for longer than allowed by BFN Technical Specification Limiting Conditions for Operation 3.6.1.3, Condition A. The causes of this event were the failure of the original design of the MSIV actuators and accumulators to account for elevated drywell pressure during the time that the MSIVs are required to stroke for a Design Basis LOCA, and the failure to incorporate internal and external operating experience into the BFN Air Operated Valve (AOV) program. Corrective actions are to issue and implement design changes to resolve negative margin issue with Units 1, 2, and 3 inboard MSIVs, to review calculations for accumulators associated with the Automatic Depressurization System relief valves to ensure they address LOCA conditions, and to ensure design basis calculations are developed for all Category 2 AOVs at BFN in addition to the Category 1 calculation reviews already in progress.
05000275/LER-2015-002	5 October 2015	3 December 2015	Diablo Canyon	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Steam Generator Reactor Coolant System Decay Heat Removal	As part of an apparent cause evaluation, Pacific Gas and Electric Company (PG&E) identified an incorrect insulation configuration, installed in 2010, on the thermal extension piping that houses the wires for the wide range (WR) reactor coolant system (RCS) resistance temperature detectors (RTDs). The insulation configuration, as installed, trapped heat inside the thermal extension piping and overheated the wires. The cause of the incorrect configuration of the insulation was insufficient guidance in the associated work package instructions. An engineering analysis completed on October 5, 2015, determined that the eight WR RCS RTDs had either failed or were operating outside the environmental qualification temperature range. As a result, on October 5, 2015, PG&E determined that the required number of WR RTDs would not have been operable and therefore a violation of Technical Specification 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," had occurred. As part of the corrective actions, PG&E replaced all eight WR RTDs, restored the insulation per the design requirements, revised the drawings for Unit 1 WR RTDs to provide adequate level of detail, and revised the work order to include the correct drawing and level of details for proper installation of all WR RTDs. This event did not adversely affect the health or safety of the public.
05000247/LER-2015-002	18 August 2015	19 October 2015	Indian Point	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability	Reactor Coolant System Residual Heat Removal	On August 18, 2015, the operations shift manager entered Technical Specification (TS) 3.0.3 upon determination that the Residual Heat Removal (RHR) heat exchanger outlet valves (MOV-746 and MOV-747) would not remain operable during a degraded voltage (DV) condition. RHR heat exchanger outlet valves MOV-746 and MOV-747 are normally closed therefore their failure would result in both trains of RHR becoming inoperable.' The RHR outlet valves are required to open during a Design Basis Accident for the RHR system to perform its safety function. As a result of NRC inspector questioning, an evaluation of the electrical coordination calculations associated with fuses for MOV-746 and MOV-747 determined the fuses would not support continued operability during a DV condition. The fuses were replaced with fuses that would remain operable under DV conditions and the RHR trains restored to operable status. Direct cause was the electrical coordination calculations for MOV-746 and MOV-747 did not support operability during a DV condition. The apparent cause was that the Industry Operating Experience (OE) (prior NRC violations and findings) was not properly acted upon during the Focused Self-Assessment on CDBI Preparations due to an incorrect assumption. Corrective actions included replacement of fuses, and communication to engineering personnel of the lessons learned from the event. Initiated CR-IP2-2015-03725 recording NRC 2015 CBDI Item 103 and CR-IP2-2015-03702 recording NRC 2015 CBDI Item 104 to addresses. EOC findings and calculation updates which will be processed via Engineering Changes. The event had no significant effect on public health and safety. NRC FORM 366AU.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) Indian Point Unit 2 05000-247 Note: The Energy Industry Identification System Codes are identified within the brackets (1. Contents 1 DESCRIPTION OF EVENT 2 Cause of Event 3 Corrective Actions 4 Event Analysis 5 Past Similar Events 6 Safety Significance DESCRIPTION OF EVENT On August 18, 2015, while at 100% steady state reactor power, the operations shift manager entered Technical Specification (TS) 3.0.3 at 13:31 hours, upon determination that the Residual Heat Removal (RHR) (BP) heat exchanger fHX1 outlet valves (MOV-746 and MOV-747) (ISV) would not remain operable during a degraded voltage (DV) condition. RHR heat exchanger outlet valves MOV-746 and MOV-747 are normally closed therefore their failure due to inadequate fuses for DV conditions would result in both trains of RHR being inoperable. These valves are required to open during a Design Basis Accident for the RHR system to perform its safety function. During a scheduled NRC Component Design Basis Inspection (CDBI), an inspector review of the 480 Volt Motor Control Center (MCC) coordination calculation for MCC 26B resulted in a question concerning the survivability of safety related loads following a degraded voltage (DV) condition. Specifically, the NRC inspector questioned the ability of the electrical protective devices (fuses) (FU) to hold during locked rotor currents at degraded voltages for the duration of the degraded voltage time delay. The NRC inspector review of protective device coordination plots for Motor Operated Valves (MOVs) 746 and 747 (RHR heat exchanger outlet valves) questioned whether the Shawmut Type A4J30 (S156) fast acting fuses in the supply circuit would coordinate with the MOV locked rotor currents expected under degraded, normal and higher than normal voltage conditions. As a result of inspector questioning, the fuses for MOV-746 and MOV-747 were evaluated in accordance with the guidance contained in Nuclear Energy Institute (NEI) 15-01 (An analytical Approach for Establishing Degraded Voltage Relay (DVR) Settings). Engineering concluded after review of the NEI 15-01 guidelines that the electrical coordination calculations associated with MOV-746 and MOV-747 did not support continued operability during a DV condition. The condition was recorded in the Indian Point Energy Center (IPEC) Corrective Action Program (CAP) as Condition Report CR-IP2-2015- 03688. On August 18, 2015, Operations entered TS 3.5.2 (ECCS Operating) for two trains of RHR and Recirculation inoperable due to valves MOV-746 and MOV-747 being determined to be inoperable. Entered TS 3.5.2 Condition C for less than 100 percent equivalent flow of one RHR pump and one Recirculation pump available. Required action C.1 is to enter TS Limiting Condition for Operation (LCO) 3.0.3 immediately. Entered TS 3.0.3 at 13:31 hours, whose actions are to place the unit in a mode or other specified condition in which the LCO is not applicable with actions to be initiated within 1-hour. The installed fuses were replaced with Shawmut Type AJT30 time delayed fuses which have no coordination issues. During change out of fuses with replacement fuses that would remain operable under DV conditions, operators changed out one set of fuses affecting one valve at a time. At 14:19 hours the fuses for MOV-746 were installed and TS 3.0.3 exited. Entered TS 3.5.2 Condition A for one or more trains inoperable with required action A.1 to restore train to operable status with completion time of 72 hours. At 14:31 hours, the fuses for MOV-747 were installed and TS 3.5.2 actions exited. The RHR system is a subsystem of the Emergency Core Cooling System (ECCS) divided into two 100 percent capacity subsystems. Each RHR subsystem consists of one RHR pump and one RHR heat exchanger as well as associated piping and valves to transfer water from the suction source to the reactor core. ECCS analysis assumes RHR injection into all four reactor coolant system (RCS) cold legs. During the injection phase of a LOCA recovery, a suction header supplies water from the Refueling Water Storage Tank (RWST) to the High Head Safety Injection (HHSI) and RHR pumps. The discharge from the HHSI and RHR pumps divides and feeds an injection line to each of the RCS cold legs. During the recirculation phase of LOCA recovery, the Containment Recirculation pumps take suction from the containment recirculation sump and direct flow through the RHR heat exchangers to the cold legs. NRC FORM 366AU.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) Indian Point Unit 2 05000-247 2015 002 00 The RHR pumps can also be used to provide a backup method of recirculation. MOV-746 and MOV-747 are RHR heat exchanger outlet valves that are normally closed during power operation but are required to open for a DBA (LOCA). This issue has been an NRC concern regarding the adequacy of power plant electrical distribution systems voltages as a result of previous events with degraded voltage protection for power plant Class lE electrical safety buses for degraded transmission network (grid) voltage conditions. Previous electrical grid events demonstrated that when Class lE buses are supplied by the offsite power system, sustained degraded voltage conditions on the grid can cause adverse effects on the operation of class lE loads. The degraded voltage conditions will not be detected by the Loss-of-Voltage Relays (LVRs) which are designed to detect loss of power to the bus from offsite circuits. The NRC issued actions to licensees followed up with Generic Letter 79-36, Branch Technical Position (BTP) PSB-1 and Regulatory Issues Summary (RIS) 2011-12. Indian Point used the NRC guidance to evaluate the plant and developed calculations to address the issue. In 1997, IEEE Standard 741 was issued providing guidance on the use of degraded voltage relays in protective schemes but the guidance was not endorsed by the NRC. IEEE Standard 741 was not part of the Unit 2 protective setting and coordination criteria which was used to assess the Unit 2 coordination adequacy. As a result, the methodology provided in IEEE Standard 741 was not used to develop the coordination calculations. The current Entergy Engineering Standard (ES) invokes only certain portions of the IEEE Standard 741 criteria. Specifically, the one second - margin guideline for prevention of spurious tripping is not included in the ES. RIS 2011-12 was issued to clarify regulatory requirements but did not detail any specific analytical approach to meet the requirements. In March 2015, the Nuclear Electric Institute (NEI) developed technical guidance document NEI 15-01 to provide an analytical approach that could be used to establish the settings for degraded voltage protection schemes. The NRC used NEI 15-01 and RIS 2011-12 in the CDBI to evaluate the current design basis of the electrical protective devices for MOV-746 and MOV-747. In preparation for the CDBI, a Focused Self-Assessment (FSA) was performed and the issued identified but a detailed review was not performed. During the FSA the results of recent NRC CDBI inspections were reviewed and the FSA determined that further review of design basis electrical calculations should be performed to determine if Indian Point was susceptible to similar calculation weaknesses that resulted in NRC findings at other plants. The failure to properly act upon the Industry OE review per the FSA finding resulted in a missed opportunity. An extent of condition (EOC) review determined the condition is unique to Design Engineering and design basis electrical calculations because the condition specifically concerns survivability of electrical protective devices during a DV condition. As such, this condition does not extend to other equipment, calculations, processes or organizations not already evaluated. As a result of High Risk 1 EOC, an EOC review was performed on the effects of degraded grid voltage on electrical protective devices for safety related motors, motor operated valves and static loads. Loads on the Unit 2 and Unit 3 480 volt safeguards buses and MCCs were reviewed. This review identified two valves, MOV-746 and MOV-747, in need of fuse replacement. Calculation updates associated with this EOC review are tracked by CR-IP2-2015-03725. As a result of High Risk 2 EOC, a review was performed that included an evaluation of all safety related loads at degraded grid voltages. Operating voltage (running and starting) to all Unit 2 and Unit 3 motors, MOVs, static loads and MCC contactors on Unit 2 and 3 480 volt safeguards buses and MCCs were evaluated at degraded voltage conditions. This review determined that the required loads would successfully cope with a DV condition. Calculation updates associated with this EOC review are tracked by CR-IP2-2015-03702. The calculation updates will be processed via Engineering Changes (ECs) and all affected documents will be evaluated for update. The calculation updates will be processed via Engineering Changes and all affected documents, including Engineering Standard ENN-EE-S-003-IP and the 480 Volt Electrical System Design Basis Document will be reviewed and any necessary changes incorporated into the appropriate documents. Cause of Event Direct cause was electrical coordination calculations for MOV-746 and MOV-747 did not support continued operability during a degraded voltage condition. The apparent cause was that the Industry Operating Experience (OE) (prior NRC violations and findings) was not properly acted upon during the Focused Self-Assessment on CDBI preparation due to an incorrect assumption. Corrective Actions The following corrective actions have been or will be performed under Entergy's Corrective Action Program to address the cause and prevent recurrence: Evaluated and replaced the fuses in the supply circuit for MOV-746 and MOV-747 in accordance with Engineering Change (EC) 59435. Issued EC mark-ups for the affected calculations as part of EC-59435. Communicated to Design and Programs Engineering Personnel the lessons learned from this event and to raise the level of awareness regarding recent guidance on the evaluation of degraded voltage conditions and to reinforce importance of verifying assumptions. Initiated CR-IP2-2015-03725 recording NRC 2015 CBDI Item 103 and CR-IP2-2015-03702 recording NRC 2015 CBDI Item 104 to addresses EOC findings and calculation updates which will be processed via ECs. All affected documents, including Engineering Standard ENN-EE-S-003-IP and the Design Basis Document for the 480 Volt Electrical System (IP2-480V DBD) will be evaluated for update as part of the ECs for this issue. Performed and documented EOC review for CBDI Item 103 and Item 104 per EC-59116 and EC-59123. Event Analysis The event is reportable under 10CFR50.73(a)(2)(v) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) shut down the reactor and maintain it in a safe shutdown condition, (B) remove residual heat, (D) mitigate the consequences of an accident (safety system functional failure), and is also reportable under 10CFR50.73(a)(2)(ii)(B) for any event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety and under 10CFR50.73(a)(2)(vii) for any event where a single cause or condition caused at least two independent trains or channels to become inoperable in a single system designed to: (A) shut down the reactor and maintain it in a safe shutdown condition, (B) remove residual heat, (D) mitigate the consequences of an accident. The condition recorded in CR-IP2-2015-03688 meets these reporting criteria since the possible failure of the fuses under degraded voltage conditions for RHR heat exchanger outlet valves MOV-746 and MOV-747 could result in the loss on both RHR trains. Past Similar Events A review was performed of the past three years of Licensee Event Reports (LERs) for events that involved a SSFF, Unanalyzed Condition or common mode failure due to loss of redundant ECCS trains as a result of inadequate engineering. No LERs were identified reporting a loss of ECCS function. FACILITY NAME (1) DOCKET (2) LER NUMBER 6) PAGE (3) Safety Significance This event had no significant effect on the health and safety of the public. There were no actual safety consequences for the event because there were no accidents or transients during the time of the event. A risk assessment was performed by the NRC as discussed in Inspection Report 05000247/2015-007 dated October 5, 2015. The NRC reviewed the Entergy operability assessment and determined it was adequate. However, for the degraded voltage issue, the originally installed fast-acting Shawmut Type A4J30 fuses would likely actuate prior to the RHR MOVs successfully stroking open. That condition results in the potential inoperability of both trains of low pressure injection/recirculation for longer than the TS LCO allowed outage time. A detailed risk evaluation was performed for that condition. The calculated cumulative conditional core damage probability (CCDP) was determined to be 5.5E-6 with an exposure time of one year. To account for the consequential degraded grid voltage condition, the CCDP value was multiplied by 2E-2 to approximate the probability of a LOOP given a LOCA has occurred. The evaluation determined that the estimated increase in core damage frequency (CDF) associated with this performance deficiency is 1E-7/year or very low safety significance. The risk significance approximation is overly conservative and is considered a worst case bounding evaluation. The risk assessment included a review for potential LERF and external events contributions. Based on Unit 2 being a pressurized water reactor with a large dry containment, the finding screens out for LERF consideration. Because the conditional event sequences of interest involve loss of coolant accidents, external events coincident with or contributing to these accidents would be of extremely low probability and considered beyond the plants design basis. Accordingly, there is no external event contribution to core damage risk for this issue.
05000400/LER-2015-005	16 June 2015	17 August 2015	Harris	10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)	Feedwater Main Steam	On June 16, 2015, while Harris Nuclear Plant (HNP) was operating at 100% in mode 1, two doors between the Reactor Auxiliary Building (RAB) and Main Steam Tunnel (MST) were opened to support a maintenance activity. These doors are credited in the high energy line break (HELB) equipment qualification and internal flooding analyses for HNP; however the opening of these doors is not addressed in these analyses. If a HELB occurred in the MST during a time in which these doors are open, the Essential Services Chilled Water system could become inoperable. The root cause of this event was determined to be that HNP Engineering failed to develop and implement control measures for hazard barriers credited for mitigating HELB events. Immediate corrective action was taken to close the doors and issue a Standing Instruction that prohibits these doors from being blocked open during modes 1 through 4. Corrective action is planned to develop and implement an engineering change that evaluates the required passive design features needed to support HELB analysis and overall licensing bases. This engineering change will identify required passive design features and establish the necessary process to ensure barriers are appropriately identified and controlled.
05000266/LER-2015-004	4 June 2015	3 August 2015	Point Beach	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Emergency Diesel Generator Decay Heat Removal	On June 4, 2015 with both units at full power, it was identified that removal of the W-185A(B), G-03(04) Emergency Diesel Generator (EDG) Switchgear Room Exhaust Fan from service may result in the inability to maintain switchgear room temperatures below that required to maintain equipment operable. Subsequent engineering evaluation has determined that room temperature could have exceeded environmental conditions at which the components would be reasonably expected to function for their respective mission times with the G-03(04) EDG Switchgear Room Exhaust Fan(s) out of service. This condition resulted in a condition prohibited by Technical Specifications, Technical Specification 3.8.9. Distribution Systems- Operating. Additionally, Technical Specification 3.8.9 required action A.1 would have required immediate declaration of the associated supported required feature(s) inoperable. No opposite train plant systems were affected by this condition. This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B), Operation or Condition Prohibited by Technical Sf.J~:dfi .... auuns, and 10 CFR 50.73(a)(2)(vii), Common Cause lnoperability of Independent Trains or Channels.
05000298/LER-2015-003	30 May 2015	28 July 2015	Cooper	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Reactor Protection System Main Steam Isolation Valve Primary containment Main Turbine Main Steam Line Main Steam	In January 2015, during Quarterly Surveillance Testing on the Main Steam Isolation Valves (MSIVs), inboard MSIV C failed to actuate its associated Reactor Protection System (RPS) relay. The limit switch and associated RPS relay were declared inoperable and the associated RPS channel was placed in trip to satisfy Technical Specifications requirements. In May 2015, during Quarterly Surveillance Testing on the MSIVs, the inboard MSIV A and inboard MSIV B also failed to actuate their associated RPS relay. The limit switches and associated RPS relay were declared inoperable and the associated RPS channel was placed in trip to satisfy Technical Specifications requirements. As a result, the plant was in an increased risk of an inadvertent full scram. A decision was made to shut the plant down and replace the limit switches. The limit switches were removed and are being evaluated for cause. The event is currently under investigation. CNS will provide a supplement to this Licensee Event Report.
05000400/LER-2015-002	4 April 2015	3 June 2015	Harris	10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	HVAC Control Room Emergency Filtration System Remote shutdown	On April 4, 2015, Harris Nuclear Plant was shut down for a scheduled refueling outage in mode 5 and was performing the Remote Shutdown System Operability test. Following transfer back to the Main Control Board, the supply breakers to the normal air intake isolation dampers' motor actuators both independently tripped due to high instantaneous current from the attempted direction reversal of their respective motor actuators. These trips caused both dampers to be in the partially open position, rendering the Control Room Envelope (CRE) boundary inoperable. The apparent cause of this event is that the HMCP model breaker/starter combination installed by a Design Change is more sensitive to peak current spikes than the original EF3 model breakers. The contributing cause associated with this event was that industry operating experience (OE) was not adequately reviewed to identify existing OE on the need to raise the trip setting on HMCP model breakers. Immediate corrective action was taken to manually close the dampers and restore integrity of the CRE boundary. The corrective action taken to address the breaker sensitivity observed was that the trip settings for the impacted HMCP model breakers installed by the Design Change were revised to add margin to the trip settings.
05000298/LER-2015-002	19 February 2015	16 April 2015	Cooper	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Reactor Protection System Main Steam Isolation Valve Reactor Core Isolation Cooling Primary containment Emergency Core Cooling System Main Steam Line Control Rod Main Steam	On February 19, 2015, during performance of surveillance procedures, three of eight Division 2 Main Steam Differential Pressure Indicating Switches failed to trip prior to exceeding limits set in Technical Specifications (TS). Investigation revealed that the setpoint change calculations and surveillance procedures had been inappropriately revised during implementation of TSTF-493. The applicable setpoint change calculations and surveillance procedures have been revised to pre-TSTF-493 values. This event is being reported as an operation or condition prohibited by TS, and also as a common cause inoperability of independent trains or channels. The event has negligible safety significance based on the safety function associated with the main steam high flow isolation signal was preserved through successful testing of five of the eight Division 2 switches.
05000261/LER-2015-001	28 January 2015	30 March 2015	Robinson	10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Feedwater Reactor Protection System Main Steam Line	At 1957 hours EST on 01/28/2015 with H. B. Robinson Steam Electric Plant, Unit No. 2 in Mode 1 at 100 percent power, the plant entered Improved Technical Specifications (ITS) Limiting Condition for Operation (LCO) 3.0.3 due to inoperability of both trains of the Reactor Protection System (RPS). It was discovered that modification work performed during the Fall 2013 refueling outage inadvertently connected in parallel both safety trains of the RPS and both trains of the DC Electrical Distribution System (DC-EDS). This parallel connection rendered both of these systems inoperable because the required independence and redundancy of systems was eliminated. This condition was present for a time greater than allowed by technical specifications (TS) and is reportable as a condition prohibited by TS under 10 CFR 50.73(a)(2)(i)(B). This condition also rendered inoperable one train of multiple systems designed to mitigate the consequences of an accident and is reportable under 10 CFR 50.73(a)(2)(vii). Immediate actions taken to restore compliance with regulations included completion of an emergency Work Order that restored the independence of safety trains of both systems and permitted exit of ITS LCO 3.0.3 at 2048 on 01/28/2015. The condition was entered into the site corrective action program and a root cause investigation is in progress.
05000286/LER-2015-001, Safety System Functional Failure Due to Inoperable Refueling Water Storage Tank Level Alarms Due to Freezing of the Level Instrument Sensing Lines Caused by a Failed Strip Heater	8 January 2015	3 March 2015	Indian Point	10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
05000286/LER-2015-001	8 January 2015	3 March 2015	Indian Point	10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability	Auxiliary Feedwater Residual Heat Removal Control Rod Containment Spray	On January 8, 2015, the Refueling Water Storage Tank (RWST) level sensing instrumentation lines (LT-920 and LIC-921) were discovered frozen resulting in inoperable low-low level alarms in the Control Room. Entered Technical Specification (TS) 3.5.4 (RWST) Condition C due to both RWST low-low level alarms disabled in the CR. TS Condition C requires at least one channel of RWST low-low level to be restored to operable in one hour. Actions initiated to return one RWST level channel to operable.. Entered TS 3.5.4 Condition D (Required Action and associated Completion Time not met) D.1 be in. Mode 3 in 6 hours and D.2 be in Mode 4 in 12 hours. Commenced unit shutdown per TS for inoperable RWST level alarms. Repairs and calibrations completed returning the RWST level alarms to operable. TS 3.5.4 exited and power ascension commenced. Loss . of-both LT alarms is a safety system functional failure as the alarms are credited for operator manual switchover for recirculation. Direct cause was failure of the RWST instrument level alarm strip heater to maintain the temperature in the instrument enclosure. - Due to the failure of the heat trace circuit EHT34-1 strip heater to . function combined with a period of severe cold weather resulted in the sensing lines for the RWST.to freeze. The apparent cause was a high resistance electrical connection at the strip heater wire lug due to thermal cycling and age. Corrective actions included _repair of ring lug to strip heater and calibration of level instrumentation. Maintenance procedure 0-ELC-419-EHT will be revised to include inspection/repair of -. strip heater and ring lug connections within instrument enclosures. An action request (AR) will be initiated for a new model Work Order/PM to inspect strip heater connections and-operation. The event had no significant effect on public health and safety.
05000275/LER-2015-001, Both Trains of Residual Heat Removal Inoperable Due to Circumferential Crack on a Socket Weld	31 December 2014	2 March 2015	Diablo Canyon	10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability	Residual Heat Removal	On December 31, 2014, while performing a walkdown as part of a surveillance test procedure, plant personnel identified through-wall seepage in a Diablo Canyon Power Plant Unit 1 socket weld inside residual heat removal system. Subsequent cleanup of the boric acid accumulation revealed active seepage of 30 drops per minute. A visual inspection identified that the source of the seepage was a circumferential crack on the socket weld. This is the initial Licensee Event Report (LER) for this event. Pacific Gas & Electric will submit a supplemental LER describing event cause and corrective actions no later than May 8, 2015. This condition did not have an adverse effect on the health and safety of the public.
05000382/LER-2014-004	22 October 2014	3 June 2015	Waterford	10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications	Steam Generator Emergency Diesel Generator	During a walkdown of the Emergency Diesel Generator Feed Tank A and B vent lines on October 22, 2014, an NRC Component Design Basis Inspection inspector identified corrosion on the Emergency Diesel Generator Feed Tank A and B vent lines where the vent lines pass through the roof. A visual inspection was performed and revealed that the corrosion had created through wall holes that could allow water into both the train A and B Emergency Diesel Generator Feed Tanks. Follow up analysis has determined that some rainfall amount less than the postulated Probable Maximum Precipitation event could have resulted in water intrusion into the Emergency Diesel Generator A and B Feed Tanks that exceeds the 0.1 percent water content allowed by the vendor technical manual. This could have potentially affected the operability of both the A and B Train Emergency Diesel Generator Feed Tanks and subsequently both trains of the Emergency Diesel Generators. It is unknown how long this corrosion has existed. Compensatory measures were put in place to prevent water ingress should a large rainfall event occur. This condition is reportable under the following criteria: 10 CFR 50.73(a)(2)(i)(B), 10 CFR 50.73(a)(2)(v)(D), and 10 CFR 50.73(a)(vii).

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