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	Discovered date	Reporting criterion	Title	Event description
ENS 56641	18 July 2023 20:14:00	10 CFR 21.21(d)(3)(i), Failure to Comply or Defect	Part 21 Report - Motor Driven Relay Failed Testing	The following information is a synopsis provided by the licensee via email: River Bend Station completed an internal Part 21 evaluation concerning a motor driven relay that failed pre-installation testing due to a buildup of corrosion between the rotor and relay core. The relay was planned for use in the Remote Shutdown System. The NRC Resident has been notified. A written notification will be provided within 30 days. Affected known plants include only River Bend at the time of the notification.
ENS 54266	11 September 2019 00:34:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition Due to Lightning Strike	A lightning strike occurred at approximately 1502 CDT on 09/10/2019, and a resulting power surge damaged some of the security door card reader system equipment. However, this did not affect access to plant areas for personnel who were already within protected area. At 1830 on 09/10/2019, it was discovered that some of the oncoming night shift personnel could not access particular areas that required the use of security card readers. Extent of condition check at 1934 on 09/10/2019 determined that access to 1A and 3A Electric Board Rooms, which contain remote shutdown panels and Fire Safe Shutdown equipment. was prohibited for the night shift personnel. This condition is reportable under 10 CFR 50.72(b)(3)(ii)(B) - Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. Access was restored to all plant areas at 2106 on 9/10/2019. No plant events occurred during the time frame that the 1A & 3A Electric Board Rooms inaccessible that would have required access to these areas. The NRC Resident Inspector has been notified.
ENS 52531	3 February 2017 19:58:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition - Appendix R Fire Analysis	In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10CFR50.48(b) (Appendix R) to 10CFR50.48(c) (NFPA 805), a weak-link and operator manual action (OMA) analysis for Information Notice (IN) 92-18 type hot shorts on motor-operated valves (MOVs) was performed to support the Plant Hatch Appendix R Safe Shutdown Analysis. As a result of the analysis, it was identified that cable impacts can bypass valve limit and torque switches, resulting in physical damage to valves required for Safe Shutdown. This would prevent the valves from being operated locally or being operated from the remote shutdown panel. These cable failures could also cause the valve motors to fail. This updated analysis has identified circuit configuration deficiencies in Fire Areas 0024 A & C (Main Control Room & Cable Spread Room), 1203F (U1 Reactor Building SE), 1205F (U1 Reactor Building NE), and 2203F (U2 Reactor Building NE). Therefore, due to the identified deficient conditions, it was determined that in the event of a postulated fire in the affected fire areas, the paths of safe shutdown on the affected unit(s) could be compromised and impact the ability to achieve safe shutdown conditions. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions in these same fire areas. The presence of the compensatory measures in addition to automatic fire detection in these fire areas ensure that the safe shutdown paths are preserved until the degraded conditions can be repaired. CRs 10326399, 10326401, 10326402, 10326404 and 10326405 The licensee notified the NRC Resident Inspector. The unanalyzed condition is applicable to 10CFR50.48(b) Appendix R and not to 10CFR50.48(c) (NFPA 805).
ENS 52329	28 October 2016 13:40:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition Due to Fire Barrier/Helb Door Inoperable	The licensee reported an unanalyzed condition under 10 CFR 50.72(b)(3)(ii)(B) due to a fire barrier/HELB (high energy line break) door being inoperable during maintenance. This resulted in two of five safe shutdown panels to be declared inoperable. The door, located between the Turbine and Administrative Buildings, was opened for approximately two minutes for 'tool pouch work'. When Operations discovered the door was opened for maintenance, they declared the door inoperable until Operations performed the surveillance required to declare the door operable. The total time the door was inoperable was approximately 1 hour and 11 minutes. The licensee has notified the NRC Resident Inspector. * * RETRACTION AT 1642 EST ON 12/27/2016 FROM DUSTIN SCURLOCK TO MARK ABRAMOVITZ * * * The condition reported in ENS 52329 pursuant to 10 CFR 50.72(b)(3)(ii)(B) has been evaluated, and determined not to be an unanalyzed condition that significantly degraded plant safety. NRC Regulatory Issue Summary (RIS) 2001-09, 'Control of Hazard Barriers,' allows breaching of HELB barriers, provided the risk associated with the applicable maintenance activity is assessed and managed in accordance with 10 CFR 50.65(a)(4) of the Maintenance Rule. The hazard barrier controls procedure at JAF (James A. Fitzpatrick) is consistent with this guidance, and includes compensatory measures for opening of the subject HELB door (76FDR-A-272-26). Per the JAF hazard barrier controls procedure the secondary HELB doors are to be verified operable, and the Alternate Shutdown Panels 25ASP-4 and 25ASP-5 declared inoperable. Based on a review of previous performances of ST-76Y, Fire Door Inspection and Operability Test, and the JAF Paperless Condition Reporting System, all applicable secondary HELB doors were operable prior to and during the 'tool pouch work' on 76FDR-A-272-26. JAF TS LCO 3.3.3.2, Remote Shutdown System (RSS), stipulates a completion time of thirty days to restore one or more required remote shutdown functions to operable. The duration of the 'tool pouch work' and inoperability of 76FDR-A-272-26 is well within this thirty day allowed outage time. In addition, the Alternate Shutdown Panels that were rendered inoperable by this condition are not required for mitigation of a HELB, and steam line break accidents are not discussed in the Technical Specification (TS) Bases for the Remote Shutdown System. The licensee notified the NRC Resident Inspector. Notified the R1DO (Lilliendahl).
ENS 51440	2 October 2015 12:25:00	10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Loose Safety Relief Valve Discharge Flange Bolts	On October 2nd, at approximately 0825 EDT, maintenance technicians were performing as-found torque checks on the discharge flange of the 'B' Safety Relief Valve (SRV). 12 of the 16 bolts were not adequately torqued. The 'B' Safety Relief Valve is credited for Remote Shutdown. The as-found condition of inadequate torque potentially impacts the seismic qualification of the 'B' SRV. An investigation and extent of condition review is ongoing. The NRC Resident Inspector has been notified. Before the outage, there were no abnormal indications of leakage as indicated by a rise in drywell temperature or pressure. The SRVs had been cycled under pressure with no abnormal indications. The four bolts that were tight were in a diagonal pattern. The looses bolts were described as "finger tight." The licensee is determining the actions to take regarding the remaining 14 SRVs. * * RETRACTION FROM STEVE WARD TO DONALD NORWOOD AT 1552 EST ON 11/20/2015 * * * As part of the event investigation and extent of condition review, the as-found torque values of the inlet and outlet Safety Relief Valve (SRV) flange connections were measured and an engineering evaluation of the as-found condition was performed. The evaluation confirmed that all 15 SRVs would have remained operable during a design basis earthquake. Any potential discharge flange connection leakage during SRV operation would be bounded by the design basis Loss of Coolant Accident analysis described in the UFSAR. Subsequent investigation activities of the as-found condition of SRV 'B' determined that the four tight bolts were not oriented in a diagonal pattern across the discharge flange as originally reported. This information is provided only to clarify previously reported information and does not affect the original basis for reporting or the current basis for retraction. The licensee notified the NRC Resident Inspector. Notified R3DO (McCraw).
ENS 51120	4 June 2015 14:03:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition for a Postulated Fire	In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10 CFR 50.48(c) (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Unit 1 and Unit 2 Reactor Buildings. This updated analysis has identified circuit configurations in four Fire Areas where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. These are Category 1 barrier impairments. In the Unit 1 Safe Shutdown Analysis, RCIC (1E51C001) (Path 1) components are impacted by a fire in Fire Area 1203. The postulated failure described above impacts HPCI (1E41C001) (Path 2) operation. Therefore, in the updated analysis there is no safe shutdown method for high pressure injection that remains free of fire damage for an Appendix R postulated fire in Fire Area 1203. While this represents an unanalyzed condition for Appendix R, the described scenario is only possible given a fire has occurred in Fire Area 1203. In the Unit 1 Safe Shutdown Analysis, Path 2 components are impacted by a fire in Fire Area 1205. The postulated failure described above impacts the 1E 4160 Kv (1R22S005) emergency bus power to Path 1 components. Therefore, in the updated analysis there is no safe shutdown method that remains available for an Appendix R postulated fire in Fire Area 1205. While this represents an unanalyzed condition for Appendix R, the described scenario is only possible given a fire has occurred in Fire Area 1205. In the Unit 2 Safe Shutdown Analysis, Path 2 components are impacted by a fire in Fire Area 2205. The postulated failure described above impacts the 2E 4160 Kv (2R22S005) emergency bus power to Path 1 components. Therefore, in the updated analysis there is no safe shutdown method that remains available for an Appendix R postulated fire in Fire Area 2205. While this represents an unanalyzed condition for Appendix R, the described scenario is only possible given a fire has occurred in Fire Area 2205. In the updated post-fire safe shutdown model, both safe shutdown paths include the same three options for Torus Water Temperature indication (1T48R072, 1T47R611 or 1T47R612). Only one of these three components is required to succeed, however, all are impacted by the postulated fire. Thus, there is no Unit 1 Torus Water Temperature Indication available for a fire in Fire Area 1205. While this represents an unanalyzed condition for Appendix R, the described scenario is only possible given a fire has occurred in Fire Area 1205. Based on the updated Plant Hatch Appendix R Safe Shutdown analysis recommendations and the plant's Fire Hazard Analysis (FHA), compensatory measures have been taken and will remain in place until the conditions are resolved. The presence of the compensatory measures, in addition to portable fire protection equipment and installed fire protection and detection equipment, ensures the safe shutdown paths are preserved until the conditions are resolved. CR 10079009, 10079019, 10079022, 10079025 The licensee has notified the NRC Resident Inspector. * * UPDATE FROM STANLEY STONE TO DONALD NORWOOD AT 1634 EDT ON 6/17/2015 * * * In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10 CFR 50.48(c) (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Unit 1 and Unit 2 Turbine Building. This updated analysis has identified circuit configurations in two Fire Areas where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. These are Category 1 barrier impairments. 1) In the Unit 1 Safe Shutdown Analysis, Path 1 RCIC components are impacted by a fire in Fire Area 1105. The postulated failure would impact Path 2 (HPCI) operation. Therefore, in the current analysis for the credited safe shutdown method for high pressure injection may be affected for an Appendix R postulated fire in Fire Area 1105. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1105. 2) In the updated post-fire safe shutdown model, both safe shutdown paths include the same two options for Torus Water Level Indication: 2T48-R622A and 2T48-R622B. Only one of these two components is required to succeed, however both would be impacted by a postulated fire in Fire Area 2104. Consequently, both credited paths of Unit 2 Torus Water Level Indication could potentially be affected due to a fire in Fire Area 2104. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 2104. Based on the updated Plant Hatch Appendix R Safe Shutdown analysis recommendations and the plant's Fire Hazard Analysis (FHA), compensatory measures have been taken and will remain in place until the conditions are resolved. The presence of the compensatory measures, in addition to portable fire protection equipment and installed fire protection and detection equipment, ensures the safe shutdown paths are preserved until the conditions are resolved. The analysis associated with the transition of the Plant Hatch Fire Protection Licensing Basis from Appendix R to NFPA 805 is continuing, and this and any subsequent similar conditions that meet reporting requirements will be in included in an ENS Update Report. CR 10084753, CR 10084757. The licensee notified the NRC Resident Inspector. Notified R2DO (HAAG). * * UPDATE FROM SCOTT BRITT TO VINCE KLCO ON 6/24/15 AT 2114 EDT * * * In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10 CFR 50.48(c) (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Diesel Generator Building. This updated analysis has identified circuit configurations in five Fire Areas where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. These are Category 1 barrier impairments. 1) An Appendix R postulated fire in Fire Area 1404 is assessed to impact a cable required for RHR Inboard Injection Valve A, 1E11-F015A, to open. This cable was not identified in the current Safe Shutdown Analysis Report (SSAR) for this component. This valve is normally closed and is required to open to support the operation of RHR Loop A in LPCI mode, which is the credited Low Pressure Injection system for Unit 1 in support of Inventory Control to the RPV for a fire in Fire Area 1404. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1404. RHR Loop B is not available in this fire area due to fire impacts. 2) An Appendix R postulated fire in Fire Area 1408 is assessed to impact cables required for RHR Inboard Injection Valve B, 1E11-F015B, to open. These cables were not identified in the current Safe Shutdown Analysis Report (SSAR) for this component. This valve is normally closed and is required to open to support the operation of RHR Loop B in LPCI mode, which is the credited Low Pressure Injection system for Unit 1 in support of Inventory Control to the RPV for a fire in Fire Area 1408. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1408. RHR Loop A is not available in this fire area due to fire impacts. 3) An Appendix R postulated fire in Fire Area 1412 is assessed to impact a cable required for RHR Inboard Injection Valve B, 1E11-F015B, to open. This cable was not identified in the current Safe Shutdown Analysis Report (SSAR) for this component. This valve is normally closed and is required to open to support the operation of RHR Loop B in LPCI mode, which is the credited Low Pressure Injection system for Unit 1 in support of Inventory Control to the RPV for a fire in Fire Area 1412. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1412. RHR Loop A is not available in this fire area due to fire impacts. 4) An Appendix R postulated fire in Fire Area 2404 is assessed to impact a cable required for RHR Inboard Injection Valve B, 2E11-F015B, to open. This cable was not identified in the current Safe Shutdown Analysis Report (SSAR) for this component. This valve is normally closed and is required to open to support the operation of RHR Loop B in LPCI mode, which is the credited Low Pressure Injection system for Unit 2 in support of Inventory Control to the RPV for a fire in Fire Area 2404. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 2404. RHR Loop A is not available in this fire area due to fire impacts. 5) An Appendix R postulated fire in Fire Area 2408 is assessed to impact cables required for RHR Inboard Injection Valve B, 2E11-F015B, to open. These cables were not identified in the current Safe Shutdown Analysis Report (SSAR) for this component. This valve is normally closed and is required to open to support the operation of RHR Loop B in LPCI mode, which is the credited Low Pressure Injection system for Unit 2 in support of Inventory Control to the RPV for a fire in Fire Area 2408. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 2408. RHR Loop A is not available in this fire area due to fire impacts. Based on the updated Plant Hatch Appendix R Safe Shutdown analysis recommendations and the plant's Fire Hazard Analysis (FHA), compensatory measures have been taken and will remain in place until the conditions are resolved. The presence of the compensatory measures, in addition to portable fire protection equipment and installed fire protection and detection equipment, ensures the safe shutdown paths are preserved until the conditions are resolved. The analysis associated with the transition of the Plant Hatch Fire Protection Licensing Basis from Appendix R to NFPA 805 is continuing, and this and any subsequent similar conditions that meet reporting requirements will be in included in an ENS Update Report. CR 10088142 The licensee will notify the NRC Resident Inspector. Notified the R2DO (O'Donohue). * * UPDATE AT 1739 EDT ON 08/13/15 FROM PAUL UNDERWOOD TO JEFF HERRERA * * * In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10 CFR 50.48(c) (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Control Building. This updated analysis has identified circuit configurations in a Fire Area where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. This is a Category 1 barrier impairment. 1) An Appendix R postulated fire in Fire Area 0014 is assessed to impact a cable that is required for Diesel Building MCC 1C, 1R24-S027, to remain energized. Further analysis has shown that an inter-cable hot short between two conductors could cause the feeder breaker to this MCC to trip. This MCC is required to support the operation of Diesel Generator 1C, which is a credited power source in the Safe Shutdown analysis for both Unit 1 and Unit 2 in the event of a fire in this area. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 0014. Based on the updated Plant Hatch Appendix R Safe Shutdown analysis recommendations and the plant's Fire Hazard Analysis (FHA), compensatory measures have been taken and will remain in place until the conditions are resolved. The presence of the compensatory measures, in addition to portable fire protection equipment and installed fire protection and detection equipment, ensures the safe shutdown paths are preserved until the conditions are resolved. CR 10108999. The licensee notified the NRC Resident Inspector. Notified the R2DO (Nease). * * UPDATE AT 1331 EDT ON 08/25/15 FROM JOHN MITCHELL TO JEFF HERRERA * * * In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10 CFR 50.48c (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Diesel Building. This updated analysis has identified circuit configurations in a Fire Area where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. This is Category 1 barrier impairment. 1) An Appendix R postulated fire in Fire Area 1408 is assessed to impact a cable that is required for Station Battery Chargers 1D, 1E, and 1F to remain energized. These chargers support 125V DC Switchgear 1B which is the credited DC Switchgear for Unit 1 Path 2 Safe Shutdown in the event of a fire in this area. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1408. 2) An Appendix R postulated fire in Fire Area 2408 is assessed to impact a cable that is required for Station Battery Chargers 2D, 2E, and 2F to remain energized. These chargers support 125V DC Switchgear 2B which is the credited DC Switchgear for Unit 2 Path 2 Safe Shutdown in the event of a fire in this area. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 2408. Based on the updated Plant Hatch Appendix R Safe Shutdown analysis recommendations and the plant's Fire Hazard Analysis (FHA), compensatory measures have been taken and will remain in place until the conditions are resolved. The presence of the compensatory measures, in addition to portable fire protection equipment and installed fire protection and detection equipment, ensures the safe shutdown paths are preserved until the conditions are resolved. The analysis associated with the transition of the Plant Hatch Fire Protection Licensing Basis from Appendix R to NFPA 805 is continuing, and this and any subsequent similar conditions that meet reporting requirements will be in included in an ENS Update Report. CR 10113740, CR 10113745 The Licensee notified the NRC Resident Inspector. Notified the R2DO (Rose). * * UPDATE FROM KENNY HUNTER TO DONALD NORWOOD AT 1717 EDT ON 8/28/2015 * * * In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10 CFR 50.48(c) (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Turbine Building. This updated analysis has identified circuit configurations in a Fire Area where an Appendix R postulated fire could impact the ability to achieve safe shutdown (SSD) conditions. This is a Category 1 barrier impairment. 1) An Appendix R postulated fire in Fire Area 1105 is assessed to impact cables which are required for HPCI Steam Supply Isolation MOV, 1E41-F002, to remain open. This valve is required open in support of HPCI (SSD Path 2), which is the credited form of high pressure injection in this fire area. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1105. 2) An Appendix R postulated fire in Fire Area 1104 is assessed to impact a cable required for the RCIC Vacuum Breaker Isolation MOV, 1E51-F105, to remain open. This valve is required open to ensure operability of the RCIC turbine if RCIC is required to stop and restart. Failure of this valve to remain open could cause a siphon that would impact the operability of RCIC, and thus disable Safe Shutdown Path 1 High Pressure Injection. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1104. In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10 CFR 50.48(c) (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Reactor Building. This updated analysis has identified circuit configurations in a Fire Area where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. This is a Category 1 barrier impairment. 1) An Appendix R postulated fire in Fire Area 1203 is assessed to impact a cable required for HPCI Steam Supply Isolation MOV, 1E41-F002, to remain open. This valve is required open to ensure steam flow to the HPCI turbine. Failure of this valve to remain open would isolate steam to the HPCI turbine, which would disable HPCI, and thus disable Safe Shutdown Path 2 High Pressure Injection. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1203. 2) An Appendix R postulated fire in Fire Area 2203 is assessed to impact cables required for RHR Outboard Injection Valve B, 2E11-F017B, to remain open. This valve is required open to support RHR Loop B in LPCI mode, which is the credited lineup for Path 2 Safe Shutdown Decay Heat Removal. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 2203. 3) An Appendix R postulated fire in Fire Area 2203 is assessed to impact cables required for HPCI Vacuum Breaker Isolation Valve, 2E41-F104, to remain open. This valve is required open in support of Safe Shutdown Path 2 High Pressure Injection. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 2203. Based on the updated Plant Hatch Appendix R Safe Shutdown analysis recommendations and the plant's Fire Hazard Analysis (FHA), compensatory measures have been taken and will remain in place until the conditions are resolved. The presence of the compensatory measures, in addition to portable fire protection equipment and installed fire protection and detection equipment, ensures the safe shutdown paths are preserved until the conditions are resolved. The analysis associated with the transition of the Plant Hatch Fire Protection Licensing Basis from Appendix R to NFPA 805 is continuing, and this and any subsequent similar conditions that meet reporting requirements will be in included in an ENS Update Report. CR 10115432, CR10115473, CR10115436, CR10115446, CR10115444 The licensee will notify the NRC Resident Inspector. Notified R2DO (Rose). * * UPDATE PROVIDED BY GUY GRIFFIS TO JEFF ROTTON AT 1815 EDT ON 09/04/2015 * * * In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10CFR50.48(c) (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Control Building and Reactor Building. This updated analysis has identified circuit configurations in Fire Area's where an Appendix R postulated fire could impact the ability to achieve safe shutdown (SSD) conditions. These are Category 1 barrier impairments. 1) An Appendix R postulated fire in Fire Area 0024 is assessed to impact a cable that is required for Torus Suction Valve, 1E11-F065B to remain open. This valve is required to remain open in support of LPCI train B which is credited for Unit 1 Safe Shutdown in the event that the RPV has spuriously depressurized and low pressure inventory control is performed from the remote shutdown panel. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 0024. 2) An Appendix R postulated fire in Fire Area 0024 is assessed to impact a cable required for Torus Suction Valve, 2E11-F065B to remain open. This valve is required to remain open in support of LPCI train B which is credited for Unit 2 Safe Shutdown in the event that the RPV has spuriously depressurized and low pressure inventory control is performed from the remote shutdown panel. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 0024. 3) An Appendix R postulated fire in Fire Area 0014 is assessed to impact all three Air Handling Units; 1Z41-B003A, 1Z41-B003B, and 1Z41-B003C. The fire impacts a cable required for MCC 1C, 1R23-S003 to remain energized. This MCC supports the operation of Air Handling Unit B, 1Z41-B003B which is required in support of Main Control Room HVAC. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 0014. 4) An Appendix R postulated fire in Fire Area 0031 is assessed to impact all three Air Handling Units; 1Z41-B003A, 1Z41-B003B, and 1Z41-B003C. These AHUs are required in support of MCR HVAC. MCR HVAC was not required in the current Safe Shutdown Analysis Report, and thus these failures were not evaluated in this fire area. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 0031. 5) An Appendix R postulated fire in Fire Area 2014 is assessed to impact a cable required for Station Battery Chargers 2A (2R42-S026) 2B (2R42-S027) and 2C (2R42-S028) to remain energized. These chargers support 125 VDC Switchgear 2A (2R22-S016), which is the credited DC Switchgear for Path 1 Safe Shutdown. Path 2 Safe Shutdown is not available in this fire area due to fire impacts. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 2014. 6) An Appendix R postulated fire in Fire Area 2014 is assessed to impact a cable required for 125 VDC Switchgear 2A (2R22-S016) to remain energized. This is the credited DC Switchgear for Path 1 Safe Shutdown. Path 2 Safe Shutdown is not available in this fire area due to fire impacts. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 2014. 7) An Appendix R postulated fire in Fire Area 0014 is assessed to impact cables required for Station Battery Chargers 1D (1R42-S029), 1E (1R42-S030), and 1F (1R42-S031) to remain energized. These chargers support 125VDC Switchgear 1B (1R22-S017) which is the credited DC Switchgear for Path 2 Safe Shutdown. Path 1 Safe Shutdown is not available in this fire area due to fire impacts. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 0014. Based on the updated Plant Hatch Appendix R Safe Shutdown analysis recommendations and the plant's Fire Hazard Analysis (FHA), compensatory measures have been taken and will remain in place until the conditions are resolved. The presence of the compensatory measures, in addition to portable fire protection equipment and installed fire protection and detection equipment, ensures the safe shutdown paths are preserved until the conditions are resolved. The analysis associated with the transition of the Plant Hatch Fire Protection Licensing Basis from Appendix R to NFPA 805 is continuing, and this and any subsequent similar conditions that meet reporting requirements will be in included in an ENS Update Report. CR 10118312, CR 10118328, CR10118333, CR10118338, CR10118345 The licensee will notify the NRC Resident Inspector. Notified R2DO (Seymour)
ENS 50675	12 December 2014 21:00:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition with Reactor Core Isolation Cooling	Engineering identified fuse and breaker coordination issues with Reactor Core Isolation Cooling (RCIC) valves operated at the Remote Shutdown Panel (RSDP). The coordination issues are such that, given a fire in the main control room, it is possible that RCIC valve power supply breakers could trip prior to tripping control power fuses. Operation of RCIC from the RSDP could be impaired in this scenario without compensatory actions to reset breakers. RCIC is the single credited source of makeup to the reactor pressure vessel during this scenario. The current licensing basis (Fire Protection Report) does not identify the compensatory actions required to reset breakers prior to RCIC operation at the RSDP. This condition is applicable to Unit 1 and Unit 2. This report is being made pursuant to 10CFR50.72(b)(3)(ii)(B), 'Event or Condition that results in an unanalyzed condition that significantly degrades plant safety'. Actions are being taken to amend the appropriate operating procedures to take the required steps to ensure proper operation of RCIC in the postulated scenario. The licensee has notified the NRC Resident Inspector.
ENS 49607	4 December 2013 20:40:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	a Potential Electrical Hot Short in the Rhr Shutdown Cooling Control Cable Could Result in an Inter-System Loca	As a result of questions raised by inspectors as part of the 2013 Triennial Fire Protection Inspection, a vulnerability from a postulated fire in the Unit 1 Cable Spreading Room was identified. This vulnerability involved the assumption of a fire occurring in Fire Area 0024A (Cable Spreading Room) which would create an inter-cable vulnerability that could result in an inter-system LOCA (Loss of Coolant Accident). Hatch's licensing basis included credit for the use of disconnect or remote shutdown panel 'Emergency' switches located on the respective remote shutdown panels to isolate the circuits in the cable spreading room thereby eliminating this vulnerability. However, the presence of these switches does not fully address this vulnerability. The Unit 1 RHR (Residual Heat Removal) shutdown cooling isolation valves 1E11-F008 and 1E11-F009 need to be de-energized in order to preclude the opening of these valves should this vulnerability occur on Unit 1. Since the Unit 2 RHR shutdown cooling isolation valves 2E11-F008 and 2E11-F009 are already closed and deactivated, they are not presently impacted by this additional vulnerability. Immediate actions were taken to de-energize the valves in the 'closed' position which removed the vulnerability. The postulated intersystem LOCA represents an unanalyzed condition that significantly degraded plant safety. The licensee notified the NRC Resident Inspector.
ENS 49285	15 August 2013 20:10:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	a Hot Short in the Rhr Shutdown Cooling Control Cable Could Result in an Inter-System Loca	A condition was identified that resulted from an inter-cable circuit analysis as part of the safe shutdown analysis that identified a vulnerability associated with two Unit 2 valves with controls in Fire Area 2203. Specifically, during the postulated fire scenario, an inter-cable hot short could occur on the control cables for the RHR shutdown cooling suction valve 2E11-F008 valve and cause the valve to open in the event of a postulated fire in Fire Area 2203F which is in the vicinity of the Unit 2 remote shutdown panel. In addition, a spurious opening of RHR shutdown cooling suction valve 2E11-F009 valve could occur due to a hot short on the control cables. The fire is postulated while in Mode 1 which could cause both valves to open during power operation. This postulated event would expose the low pressure RHR-shutdown cooling suction line to normal operating pressures which would result in an inter-system LOCA. Immediate actions were taken to de-energize the valves in the 'closed' position which removed the vulnerability. When this condition was first discovered, the consequences of this postulated condition were evaluated and there was reasonable assurance that the condition did not represent an unanalyzed condition that significantly degrades/degraded plant safety. A review of the FSAR, design documents and regulatory requirements was performed to document the foundational logic for the engineering judgment to support the original conclusion that there was reasonable assurance that the inter-system LOCA did not represent an unanalyzed condition that significantly degraded plant safety and that this would not result in a loss of a safety function. Based on information learned in this review there was not sufficient information to make a conclusive determination. Since a conclusive determination cannot be made at this time and since there is some doubt regarding whether or not the report is needed, this report is being made in accordance with 10CFR50.72(b)(3)(ii). The licensee has notified the NRC Resident Inspector.
ENS 48676	16 January 2013 02:12:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Loss of a 4160 Volt Bus While Performing Testing	On January 15, 2013 at 2112 EST while performing testing associated with the remote shutdown system at the James A. FitzPatrick Nuclear Power Plant, an unexpected loss of the 10600 bus 'B' division AC vital power system occurred. This loss of power to the 10600 bus resulted in an automatic actuation of the 'B' and 'D' Emergency Diesel Generators. The diesel generators started as expected, but did not close in to energize the 10600 Bus due to the configuration at the time of the event. As a result of the loss of the 10600 bus, the 'B' Reactor Protection System (RPS) lost power resulting in a half scram signal and a Group II Primary Containment Isolation System (PCIS) actuation. This actuation resulted in closing containment isolation valves in multiple systems and isolating Reactor Water Clean-Up (RWCU). Based on these system actuations, the event is reportable under criterion 10 CFR 50.72(b)(3)(iv). Power to the 10600 Bus was restored January 16, 2013 at 0400 EST, and the half scram and isolation signals have been reset. Additional actions to restore systems to a normal operating line-up are on-going. Investigation into the cause of the unexpected power loss is on-going and will be addressed through the corrective action program. The NRC Resident Inspector has been notified.
ENS 48404	13 October 2012 01:00:00	10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness	Technical Support Center Removed from Service for Planned Maintenance	On October 12, 2012, at 2100 EDT, the Hope Creek Technical Support Center (TSC) ventilation system was removed from service to perform planned preventive maintenance on the system. The maintenance consists of minor maintenance to clean the intake louvers that supply outside air to the emergency filter unit (00-VH313), the supply unit fan (00-VH314) and the remote shutdown panel (RSP) supply unit fan (00-VH316). The removal of the ventilation potentially affects the TSC habitability during a declared radiological emergency requiring activation. Appropriate compensatory measures are in place while the ventilation is out of service. The Emergency Response Organization duty team has been notified of the maintenance and the possible need to activate the TSC in an alternate location. The ventilation system is scheduled to be out of service for approximately 25 hours and the TSC Ventilation will be returned to service at 2200 on October 13, 2012. The licensee has notified the NRC Resident Inspector. The local township will be notified.
ENS 48092	14 May 2012 13:28:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification for an Invalid Emergency Service Water Actuation	At 0928 EDT on May 14, 2012, the 'B' Emergency Service Water (ESW) pump started during testing from the Unit 1 remote shutdown panel (RSP). The likely cause of the start was either human error in performing continuity checks or inadvertent contact with the manual start circuit in the RSP. Based on the likely cause, this was an invalid actuation of a system listed in 10 CFR 50.73(a)(2)(iv). As indicated in 10 CFR 50.73(a)(1), in the case of an invalid actuation reported under 10 CFR 50.73(a)(2)(iv) other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. This 60-day telephone notification is being made to meet the reporting requirements instead of submitting an LER since the actuation was invalid and was not an RPS actuation with the reactor critical. The following additional information is being provided as specified in NUREG-1022: The specific train(s) and system(s) that were actuated: The 'B' ESW pump inadvertently started during testing from the Unit 1 remote shutdown panel. Whether each train actuation was complete or partial: This was a partial actuation (one of four ESW pumps). Whether or not the system started and functioned successfully: The 'B' ESW pump started successfully, operated properly, and continued running until manually secured via normal controls. At the time of the event, the licensee was performing a surveillance where control was shifted from the control room to the remote shutdown panel. The licensee notified the NRC Resident Inspector.
ENS 47967	25 May 2012 17:22:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	De-Energization of Both Divisions of Rhr Suction Line Primary Containment Isolations Valves During Testing	On Friday, May 25th 2012 at 1322 EDT, Nine Mile Point Unit 2 experienced a loss of power to 600V 2EJSUS1 emergency load center while performing scheduled surveillance testing of the Division 1 Remote Shutdown System disconnect switches. Disconnect switch SW 1-2CESA20 was taken to the actuate position which isolated main control room control, bypassed the housing limit switches and aligned the trip test switch for local breaker control of 2EJSUS1 supply breaker 1-3B. Contacts in the trip test switch for 2EJSUS1 supply breaker 1-3B were found to be closed which energized its trip coil. This resulted in a loss of motive power to Division 1 Residual Heat Removal (RHR) system primary containment isolation valve 2RHSMOV113 on the shutdown cooling suction line from the reactor vessel. At the time of the event, the Division 1 RHR shutdown cooling system was in-service with the Division 2 shutdown cooling suction line primary containment isolation valve 2RHSMOV112 de-energized open to prevent inadvertent or spurious closure, which would interrupt the shutdown cooling decay heat removal function. The result of the event was that both the Division 1 and Division 2 isolation valves on the common RHR shutdown cooling suction line (2RHSMOV112 and 2RHSMOV113) were open with no motive power. Thus, neither valve was capable of automatically closing in the event of a reactor level low (level 3) signal due to a leak in the RHR shutdown cooling system. The loss of this isolation function is being reported in accordance with 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented fulfillment of a safety function of a system that is needed to (D) mitigate the consequences of an accident. Technical Specification 3.6.1.3 Condition G was entered and actions to restore the valves to operable status were immediately initiated in accordance with Required Action G.2. Power to 2RHSMOV113 was restored at 1824 hrs, re-enabling its automatic isolation capability. The licenses has notified the NRC Resident Inspector.
ENS 47515	9 December 2011 23:35:00	10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat	Part 21 Issue with Seismic Clips on Rcic System Controllers Results in System Inoperability	On December 7, 2011, a 10 CFR 21 report (reference NRC EN No. 47498) was received from a vendor for a defect with NUS Controllers. The defect involves spring clips that form part of the seismic restraints for the controllers. The controllers referenced in the report are installed for the Reactor Core Isolation Cooling (RCIC) system in the control room and remote shutdown panel. Based on initial information provided by the vendor, it was determined that the RCIC system remained operable. On December 9, 2011, additional information provided by the vendor did not support the immediate operability determination and the RCIC system was declared inoperable for Technical Specification (TS) Limiting Condition for Operation (LCO) 3.5.3 Condition A at 1835 hours (EST). At 1932 hours (EST), the High Pressure Core Spray system was verified operable per TS LCO 3.5.3 Required Action A.1. TS LCO 3.5.3 Required Action A2 requires restoration of the RCIC system to operable status within 14 days. Qualified spring clips have been obtained and will be installed on the controllers. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(v)(B) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system needed to remove residual heat. The NRC Resident Inspector has been notified. * * UPDATE FROM CHARLES ELBERFELD TO JOHN KNOKE AT 1415 EST ON 12/10/11 * * * As a follow-up to the condition reported above, we have replaced the affected seismic clips on the controllers and the Reactor Core Isolation Cooling system is now operable as of 0734 on December 10, 2011. The NRC Resident Inspector has been notified." R3DO (Skokowski) notified. * * RETRACTION FROM LLOYD ZERR TO CHARLES TEAL ON 2/6/12 AT 1504 EST * * * The vendor provided a seismic report to the station. This report showed that the seismic clips holding the Reactor Core Isolation Cooling (RCIC) controller meet the Operating Basis Earthquake (OBE) test requirements and design requirements for a Safe Shutdown Earthquake (SSE) for Perry. Based on this review, it was determined that the spring clips would function properly during and OBE and SSE. Because the condition reported in Event Number 47515 would not have prevented the fulfillment of the safety function of a system needed to remove residual heat, the condition is not reportable, and this notification is being retracted. The evaluation for this condition is documented in condition report 2011-06531. The NRC Resident Inspector has been informed." Notified R3DO (Giessner) and Part 21 Group via email.
ENS 46782	25 April 2011 13:50:00	10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor	Single Train Charging Pump Declared Inoperable	During normal operation with 23 Charging pump in service, an unusual sound was noticed by Operations. All other parameters were normal for the pump, but 21 Charging pump was placed in service as a precaution. Subsequent investigation was unable to conclusively identify the source of the unusual sound and a lower-than-expected crankcase oil pressure was observed on a maintenance run. Therefore, 23 Charging Pump was declared inoperable. This is the alternate safe shutdown pump responsible for inventory control for remote shutdown. NRC guidance indicates that a loss of a single train system not credited in accident analysis is reportable when identified in LCD 3.3.4 'Remote Shutdown.' The licensee notified the NRC Resident Inspector and the New York Public Service Commission. * * UPDATE AT 1408 EDT ON 04/29/11 FROM MIKE BURNEY TO S. SANDIN * * * The licensee is retracting this report based on the following: Indian Point Unit 2 is retracting the 8-hour non-emergency notification made on April 25, 2011, at 1606 EDT (EN #46782). The notification on April 25, 2011, reported a loss of a single train system identified in LCO 3.3.4 'Remote Shutdown' (SSFF (safety system functional failure)) as a result of declaring the 23 Charging Pump (CP) inoperable. The 23 CP exhibited unusual sound during operation. Because the 23 CP is credited in Technical Specification (TS) 3.3.4 for Remote Shutdown, the inoperable condition was determined to be a loss of safety function. The TS 3.3.4 Allowed Outage Time (AOT) for inoperable conditions is 30 days. In accordance with recent NRC guidance provided to Indian Point for loss of single train systems, although not credited in the accident analysis and specified in Technical Specification 3.3.4, this condition is reportable as a SSFF. Subsequent investigations determined that the documented condition does not pose a challenge to the operability of the 23 CP. The noise was associated with the normal operation of the charging pump internal check valves. This condition is a long term issue that does not affect the current operation of the pump. The pump was in operation with no abnormal parameters noted at the time the noise was noted. No SSFF occurred. The licensee will inform the NRC Resident Inspector. Notified R1DO (Schmidt).
ENS 46603	8 February 2011 14:00:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Event Could Initiate High Pressure Core Spray and Overfill the Reactor Pressure Vessel	During the performance of a Fire Protection self assessment, it was discovered that a calculation for the safe shutdown analysis has an assumed action for an operator to locally depress the internal breaker trip plunger to trip the High Pressure Core Spray (HPCS) pump in response to a fire in the main control room. However, due to personnel safety concerns related to potential arc flashing events associated with this action, the remote shutdown procedure was revised to locally close the HPCS injection valve (1E22FOO4) in lieu of depressing the internal breaker trip plunger. During engineering's review of this procedure and supporting calculation, it was determined that the HPCS system could be initiated due to concurrent fire induced hot short cable damage to the two automatic initiation logic instrument cables routed in the same raceway in the area. In this event, even if the HPCS breaker could be tripped or the HPCS injection valve could be closed locally, HPCS would continue to fill the reactor pressure vessel (RPV) and flood the main steam lines. Once pressure reaches the setpoint for the Main Steam Safety Relief Valves (MSSRVs), they would lift and discharge mixed-phase water through the discharge line to the suppression pool. This conservatively postulated scenario would place the MSSRVs and their associated tailpipes in an unanalyzed condition for the stresses expected during the two-phase flow event. While it is not expected that a failure of the MSSRV discharge line will occur, a confirmatory analysis will be performed. Compensatory measures for Multiple Spurious Operations have been determined to be adequate until the analysis is complete. The licensee added additional fire zone surveillance to operator plant walk downs and will investigate to determine further corrective actions. The license has notified the NRC Senior Resident Inspector.
ENS 46392	4 November 2010 05:23:00	10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor	Charging Pump '23' Declared Inoperable During Surveillance Testing	On November 4, 2010 at 0123 (EDT), '23' Charging Pump was declared inoperable due to not meeting required test parameters during surveillance testing. '23' Charging Pump is required for remote shutdown per Technical Specification 3.3.4, Remote Shutdown, and is a single component system intended to perform the inventory control safety function. Since '23' Charging Pump has been declared inoperable, a loss of safely function has occurred and is reportable per 10CFR50.72 (b)(3)(v)(A). The cause of the inoperability is under investigation. Per Technical Specification 3.3.4, Remote Shutdown, '23' Charging Pump must be restored to operable status within 30 days. The licensee will inform the NY State PSC (Public Service Commission) and the NRC Resident Inspector. * * RETRACTION FROM CHARLES ROKES TO DONG PARK AT 1028 EST ON 12/20/10 * * * Indian Point Unit 2 is withdrawing the 8-hour non-emergency notification made on November 4, 2010, at 0305 EST. The notification on November 4, 2010, reported a safety system functional failure (SSFF) as a result of declaring the 23 Charging Pump (CP) inoperable. The 23 CP failed to meet test parameters during testing. Because the 23 CP is credited in Technical Specification (TS) 3.3.4 for Remote Shutdown, the inoperable condition was determined to be a loss of safety function. The TS 3.3.4 Allowed Outage Time (AOT) for inoperable conditions is 30 days. The remote shutdown function is not credited in the accident analysis. Subsequent investigations determined that during a two year comprehensive test (IST) the 23 CP failed the flow test due to low pump discharge flow (79.3 gpm) at the prescribed test speed (85% of rated speed). Troubleshooting determined that the 23 CP recirculation line valves 1279 and 4902 had leak-by of approximately 6.7 gpm. The leak-by results in a reduction of available discharge flow from the 23 CP to the Reactor Coolant System (RCS) due to a small amount of the pump discharge being diverted through the recirculation line. Engineering determined that the CP capacity for emergency boration with the recirculation line valves leak-by was within the capability of the 23 CP to maintain overall system function. The 23 CP is also credited for RCS inventory maintenance per TS 3.3.4 and Station Administrative Order (SAO) 703 for fire protection. Engineering concluded that with a normal CP output of approximately 87-90 gpm and a nominal rated capacity of 98 gpm, the reduction by approximately 6.7 gpm due to leak-by did not significantly impact normal operation of the CP. Therefore, the 23 CP continues to meet the established IST requirement when the 6.7 gpm recirculation line valve leak-by is accounted for. The degraded output is associated with the system rather than the 23 CP therefore the IST requirement for the 23 CP is satisfied. Additionally, the 23 CP test was performed at approximately 85% of rated speed in accordance with the IST requirements. At 100% rated speed, the remaining available capability to the RCS would be approximately 92 gpm. Engineering concludes that although degraded the 23 CP and system can perform their intended function and no SSFF occurred." The licensee will notify the NRC Resident Inspector. The R1DO (Perry) was notified.
ENS 46222	1 September 2010 20:00:00	10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor	Loss of Temperature Instrumentation	On September 1, 2010 at 1600 hrs. (it was) identified that the 21 and 22 Hot Leg Remote Shutdown Temperature Instruments are inoperable. This constitutes a safety system functional failure which is reportable. The cause of the inoperability is under investigation. Per Technical Specification 3.3.4, these instruments must be restored to operable status within 30 days. The NRC Resident Inspector has been notified.
ENS 45882	29 April 2010 22:30:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Identified Wiring Discrepancy in "B" Edg	While performing a walkdown of the 'B' emergency diesel generator remote shutdown circuitry, a wiring discrepancy was identified that could have disabled the Appendix R function of the circuitry. The wiring discrepancy could have prevented complete isolation of the 'B' emergency diesel generator breaker closure circuitry from a postulated fire in the Control Building. This configuration causes the Appendix R function of the 'B' emergency diesel generator to be inoperable. A roving firewatch has been established and a 7 day action statement has been entered in accordance with the applicable fire protection procedure. Corrective actions are being evaluated to resolve this issue. The licensee has notified the NRC Resident Inspector.
ENS 45432	14 October 2009 19:00:00	10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor	Safety System Functional Failure - Alternate Safe Shutdown Source Range Neutron Monitor Inoperable	At 1500 hours on October 14, 2009, the power supply for neutron source range detector N-38 was determined to be unable to provide reliable power to detector N-38 and the detector was declared inoperable. Technical Specification (TS) 3.3.4 (Remote Shutdown) Basis Table 3.3.4-1, Function 1.a requires one channel. Technical Requirements Manual (TRM) 3.3.D (Appendix R Alternate Safe Shutdown Instrumentation) requires entry into the condition referenced in TRM Table 3.3.0-1 and entry into the applicable related TS referenced in TRM Table 3.3.D-1 when one or more required functions with one or more required instruments in Table 3.3.D-1 (are) inoperable. TRM Table 3.3.D-1, TRO 3.3.D1, neutron flux (source range only) lists and specifies detector N-38 only and references TS 3.3.4 Basis Table 3.3.4-1, Item 1a. Although there is an operable redundant source range neutron detector N-39, the only source range detector which has indication remote from the control room is N-38. Therefore, the safety function for safe shutdown remote from the control room for reactivity control in accordance with TS Basis Table 3.3.4-1, Function 1a can not be met with N-38 inoperable. The inability to meet the TS condition is a safety system functional failure. Actions are in progress to provide a reliable power supply for N-38. Unit 2 is at 100% and is not affected by the condition. The licensee notified the NRC Resident Inspector.
ENS 44236	23 May 2008 07:30:00	10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor	Charging Pump Declared Inoperable Due to Relief Valve Lifting Below Setpoint	During normal operation with 23 Charging Pump in service, 21 Charging Pump was placed in service for quarterly testing. Shortly after starting 21 Charging Pump, 23 Charging Pump relief valve lifted below its set point value. CVCS system pressure was approximately 2520 psig. This is the alternate safe shutdown pump responsible for inventory control for remote shutdown. 23 Charging Pump was declared inoperable. Recent NRC guidance indicates that a loss of a single train system not credited in accident analysis is reportable when identified in LCO 3.3.4, Remote Shutdown. The licensee will notify the NRC Resident Inspector. * * RETRACTION AT 1446 EDT ON 7/21/08 FROM ROKES TO HUFFMAN * * * Indian Point Unit 2 is withdrawing the 8-hour non-emergency notification made on May 23, 2008, at 08:11 hours (EN #44236). The notification reported a safety system functional failure (SSFF) as a result of the relief valve for the 23 charging pump lifting below its setpoint value. The condition was discovered during Chemical Volume & Control System (CVCS) operation with the 23 charging pump in operation. Shortly after the 21 charging pump was placed in service for quarterly testing, the 23 charging pump relief valve lifted. The operating speed of the charging pumps was reduced and the 23 charging pump relief valve re-seated. The 23 charging pump is the alternate safe shutdown pump responsible for inventory control for remote shutdown from the control room. The relief valve has an active safety function in the open position to provide overpressure protection for the 23 charging pump and associated piping. The relief valve has an active safety function in the closed position to prevent charging pump discharge flow from continuously re-circulating to the pump suction. At the time of the event Operations believed the 23 charging pump relief valve lifted prematurely and declared the 23 charging pump inoperable and made the 8-hour non-emergency notification for loss of remote shutdown capability for a component specified in Technical Specification 3.3.4, 'Remote Shutdown.' The relief valve was subsequently changes out with a spare. Subsequent investigations determined that the CVCS pressure was set high in the operating band for two pump operation. Engineering determined that when the 21 charging pump was started the pulsation at the discharge of the positive displacement charging pump exceeded the limit of the relief valve pressure setting. Initial header pressure with the 23 charging pump in operation was approximately 2420 psig with typical charging header pressure at approximately 2300 psig. When the 21 charging pump was placed in service, the resultant pressure with consideration for pulsation immediately at the discharge of the charging pump would have been approximately 2701 psig. The charging pump relief valve lift pressure setting is 2735 +/- 3% (2653 psig to 2817 psig). Therefore, the resultant pressure from two pump operation was within the relief valve lifting pressure. An inspection of the removed relief valve showed only minor anticipated seat degradation of the disc ring insert. Further assessment indicated that the 23 charging pump relief valve did lift and also reseated when aligned for parallel pump operation when the CVCS pressure was adjusted. Engineering and Operations concluded that the 23 charging pump and associated relief valve were operable and capable of performing this safety function. The initial condition was determined to have been attributed to elevated initial CVCS pressure that resulted in challenging the relief valve during two pump operation. The licensee will notify the State and the Resident Inspector. R1D0 (Dentel) notified.
ENS 43379	21 May 2007 17:31:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Potential Division 1 Edg Damage Due to Loss of Service Water Cooling During Control Room Fire	During a review of an Operating Experience issue (LO-GLO-2006-0090 CA5), a condition was found which is not consistent with the assumptions of the River Bend post-fire safe shutdown analysis. 10CFR50 Appendix R states that for alternate shutdown capability (i.e shutdown from outside the main control room) support systems (service water cooling) for critical post-fire safe shutdown components must remain free from fire damage. Generic Letter 86-10, 'Implementation of Fire Protection Requirements' state that the following assumptions are required for evaluation of a control room fire: 1) fire induced spurious operation of safe shutdown components has occurred; 2) offsite power is lost and; 3) loss of automatic starting of the onsite AC generators as well as the automatic function of valves and pumps whose circuits could be affected by a control room fire. In addition to loss of automatic start of the emergency diesel generators, the post-fire safe shutdown analysis must also evaluate the consequences if the diesel generators do start concurrent with fire induced multiple spurious actuations. Since control circuits for motor operated valves for the standby service water system are routed in the control room, fire induced shorts could place these valves in a position that would prevent service water from cooling the Division 1 emergency diesel generator. In the time required for Operations personnel to evacuate the control room and re-establish control of the standby service water system at the Division 1 Remote Shutdown panel, thermal damage to the diesel generators could render the Division 1 generator incapable from performing its post-fire function. The RBS (River Bend Station) post-fire safe shutdown analysis is based on the assumption that the diesel generator high temperature trip function would remain functional based on the fact that the trip logic is located outside of the main control room and therefore would remain free from fire damage. The investigation performed during the OE review uncovered the fact that at RBS when the emergency diesel generator is started in the emergency mode the non-safety trips (such as high temperature) are by-passed. The loss of off-site power starts the diesel generator in the emergency mode; therefore the high temperature trip is by-passed. With the non-safety trips by-passed, the diesel generator will continue to run even without sufficient cooling. This condition involves compliance with 10CFR50, Appendix R. Plant equipment remains capable of performing the remaining design functions. The scope of this analysis deficiency is limited to the Main Control room fire scenario, with multiple concurrent failures. The Control Room is continuously manned. The affected cables in the MCR under-floor area are protected by automatic fire detection and automatic suppression systems, which would rapidly detect and smother a fire. Introduction of ignition sources, such as work involving welding or grinding is strictly controlled by station procedures. Furthermore Standing Order #193 Revision 3 limits hot work in the main control room during Modes 1,2 and 3. The licensee notified the NRC Resident Inspector.
ENS 42327	10 February 2006 17:00:00	10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Potential for Loss of Safe Shutdown Capability	As a result of Wolf Creek's 2005 Triennial Fire Protection Inspection, a re-evaluation of concerns described in NRC Information Notice (IN 92-18), 'POTENTIAL FOR LOSS OF REMOTE SHUTDOWN CAPABILITY DURING A CONTROL ROOM FIRE' was performed. During that re-evaluation it was identified that in the event of a fire in the control room, 39 motor operated valves (MOVs) credited for post-fire safe shutdown could potentially fail in an unanalyzed condition. Of those 39 MOVs, failure of 4 of them could potentially prevent achieving and maintaining safe shutdown (SSD) conditions. In the 39 MOV circuits identified above, an intra-cable hot short between one conductor on the hot side of the indication circuit and another conductor on the load side of the control room hand switch could bypass the torque switch and energize either the open or close coil. If this occurs, the open or close contactor will close and the motor will operate in either the open or close direction until the motor stalls, possibly resulting in damage to the valve such that it cannot be manually operated. Present system operability is not affected as there has been no occurrence of a fire in the Control Room and compensatory actions are in place to detect and mitigate the effects of a fire in the Control Room. Actions taken or planned: An hourly fire watch was established in the Control Room due to a previous condition identified on 11/16/2005 in accordance with AP 10-104, Breech procedure. The condition identified today will be referenced on the active Breach Request that was implemented on 11/16/2005. The licensee notified the NRC Resident Inspector.
ENS 42272	18 January 2006 15:00:00	Other Unspec Reqmnt	24-Hour Condition of License Report Involving Fire Protection Program Non-Compliance	This notification is being made pursuant to the Perry Nuclear Operating License section 2.C.6 (violation of the Fire Protection Program). Incorrect configuration of Division 1 remote shutdown panel wiring was identified during performance of surveillance testing. This incorrect configuration is associated with the Reactor Core Isolation Cooling motor operated exhaust valve. In the event of a fire in the control room, the motor operated exhaust valve would have the potential for spurious operation caused by fire induced shorts prior to isolation from the control room. Repairs and operator actions could have been taken to restore the valve. However, these repairs and operator actions are not currently identified in the Fire Protection Safe Shutdown analysis or associated operating procedures. Therefore, for this issue Perry does not comply with the Perry Fire Protection Program. Repairs have been completed and configuration has been restored. The licensee will notify the NRC Resident Inspector.
ENS 42243	5 January 2006 15:30:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Unanalyzed Condition Involving Rcic Operation During Mcr Evacuation	During an engineering assessment, a condition was found which does not meet the assumptions of the River Bend post-fire safe shutdown analysis. 10CFR50 Appendix R states that for alternate shutdown capability (i.e. shutdown from outside the main control room), reactor coolant system process variables shall be maintained within those predicted for a loss of normal AC power, and the fission product boundary integrity shall not be affected. Generic Letter (GL) 86-10, 'Implementation of Fire Protection Requirements,' states that the following assumptions are required for evaluation of a control room fire: 1) fire-induced spurious operation of safe shutdown components has occurred; 2) offsite power is lost; and, 3) the emergency diesel generators (DGs) do not automatically start. Based on the conservative assumptions imposed by GL 86-10, the following control room fire scenario must be addressed. A fire is assumed to cause motor-operated valve E51-MOVF063, the inboard steam supply to Reactor Core Isolation Cooling (RCIC) turbine to close. The same fire requires the main control room (MCR) to be evacuated, and during relocation to the Division 1 Remote Shutdown panel, offsite power is lost. The post-fire safe shutdown analysis has evaluated RCIC to be available from the Remote Shutdown Panel in order to maintain reactor water level, and that the Division 1 and 3 DGs are started locally. The Division 2 DG is not analyzed to remain free of damage caused by the MCR fire. Since valve E51-MOVF063 is powered from Division 2, and there is no Division 2 power available to re-open the valve, steam would not be available to power the RCIC turbine. E51-MOVF063 is located in the drywell, making manual operation of the valve impractical. Therefore, RCIC is postulated to not be available to maintain reactor level. Establishing reactor level control is a time-critical function that is required to occur within ten minutes of MCR evacuation in order to meet one of the Appendix R safe shutdown performance goals. This condition involves compliance with 10CFR50, Appendix R. Plant equipment remains operable. The scope of this analysis deficiency is limited to the MCR fire scenario, with three concurrent failures. The MCR Is continuously manned. The affected cables in the MCR under-floor area are protected by fire detection and automatic suppression systems, which would rapidly detect and smother a fire. Introduction of ignition sources, such as work involving welding or grinding, is strictly controlled by station procedures. While the assumptions of the post-fire safe shutdown analysis are not met for this scenario, it has been verified that the components required to properly align the Division 1 Residual Heat Removal system in the low pressure coolant injection mode would be available at the Division 1 Remote Shutdown Panel. Control of three safety-relief valves is also available at the Division 1 Remote Shutdown Panel to depressurize the reactor vessel for low pressure injection. An analysis is under way to determine the response of reactor water level, given these conditions. The licensee notified the NRC Resident Inspector.
ENS 41844	14 July 2005 21:10:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Station Blackout Temperature Analysis Higher than Rcic Governor Documentation	During fact gathering in response to an NRC inspection inquiry, it was determined that documentation does not exist that demonstrates that the Reactor Core Isolation Cooling (RCIC) Electronic Governor Module (EGM) would be able to operate during the required Station Blackout (SBO) coping mission time at the postulated post SBO RCIC room temperature of 206.4F. Current documentation supports operation up to 150F. The EGM is a skid-mounted module that provides speed control signals for the RCIC Woodward Governor. Failure of the EGM would result in a loss of speed control for the RCIC turbine. This could result in an overspeed, underspeed or no change condition. Overspeed of the turbine would result in a mechanical overspeed trip. This device is not in the EQ program but is Augmented Quality. RCIC continues to perform its Technical Specification required functions as defined in the Bases of Technical Specification (TS) 3.5.3. The TS function is to respond to transient events by providing makeup coolant to the reactor. The RCIC Room temperatures for the postulated TS transient events is less than the currently documented component qualification temperature. The RCIC is not an ESF system and no credit is taken in the safety analysis for RCIC system operation but is retained in the TS based on its contribution to the reduction of overall plant risk per Criterion 4 of 10 CFR 50.36. The RCIC system design requirements ensure that the criteria of 10CFR50 Appendix A, GDC 33, are satisfied. Due to the lack of supporting documentation for the EGM, the beyond design basis regulatory SBO rule requirements of 10 CFR 50.63 may not be met. This condition could potentially result in an unanalyzed condition that could significantly degrade plant safety and is therefore reportable under 10 CFR 50.72(b)(3)(ii). An analysis of the RCIC Room Heat Up Rate calculation is being performed as there are conservatisms built into the calculation that when removed will result in a lower temperature than 206.4F. Additional actions in progress include, establishing appropriate protected pathways to minimize the potential for a Loss Of Off-Site Power which could result in a SBO, performance of temperature qualification testing at SBO temperatures for the EGM, and performance of an extent of condition review for remaining RCIC components to ensure temperature qualification is met for the SBO rule. In parallel with temperature qualification testing, a modification to relocate the EGM to an area outside the RCIC room that has a lower SBO profile temperature is being pursued in the event that temperature qualification is not successful. The licensee notified the NRC Resident Inspector. * * RETRACTION FROM D. COVEYOU TO W. GOTT AT 1427 EDT ON 8/16/05 * * * A 8-hour notification was made on July 14, 2005, in accordance with 10 50.72(b)(3)(ii)(B), Unanalyzed condition. The report was made because documentation did not support the continued operation of Reactor Core Isolation Cooling (RCIC) Electronic Governor Module (EGM) during the required Station Blackout (SBO) coping mission. Since the initial report, the post SBO room heatup calculation was evaluated and determined that the decay heat removal function during the SBO coping mission was met. The decay heat removal function during SBO coping period is achieved by either High Pressure Core Spray (HPCS) or RCIC systems. In addition, the other RCIC functions (i.e., Remote Shutdown, and Safe Shutdown Fire) were evaluated and determined to be met. Since the RCIC functions and the decay heat removal and vessel inventory functions during the SBO coping mission were maintained, the plant was not in an unanalyzed condition and this issue is not reportable. Since the condition is not reportable EN 41844 is retracted. The licensee notified the NRC Resident Notified R3DO (K. O'Brien)
ENS 41742	2 June 2005 23:49:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Appendix R Assessment on Remote Shutdown Panel	The licensee provided the following information via facsimile: On 6/2/05, while performing an Appendix 'R' assessment, it was determined that the capability to achieve and maintain safe shutdown conditions, as required by Appendix 'R', has not been adequately demonstrated. Specifically, a postulated fire requiring the evacuation of the Control Room could also cause a spurious hot short in the protective relaying circuitry for the credited division of essential electrical power and subsequent loss of that division. It was recently determined that the hot short could occur at anytime during the transient, whereas previous evaluations had assumed any such hot short would be associated with the initiating event. Consequently, existing compensatory measures would only be implemented if the adverse consequences were apparent, and therefore might not be implemented within the timeframe required to assure satisfactory results. The result is that Appendix 'R' requirements are not met, and an unanalyzed condition that significantly degrades plant safety may exist. A 30 day LCO was entered at 1749 per Tech Spec LCO 3.3.3.2 Function 4.e for the Remote Shutdown Panel 'B' Essential 4160 Instrumentation transfer and control. The licensee notified the NRC Resident Inspector.
ENS 41663	3 May 2005 17:10:00	10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition	Appendix R Fire Scenario Involving Multiple High Impedance Faults	TMI Issue Report # 329440 identifies an issue associated with a previously unidentified/unanalyzed Appendix R fire scenario involving multiple high impedance faults. An engineering evaluation has determined that certain safety related power circuits are not protected against multiple high impedance faults, which in combination with a fire in the 305' elevation of the Control Building, could cause a loss of safe shutdown functions from the control room and the remote shutdown panel. An hourly fire-watch has been established in the affected fire zone in the 305' elevation of the Control Building as an interim compensatory measure. The NRC Resident Inspector will be notified.
ENS 41214	8 October 2004 01:55:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Report Due to an Invalid Actuation of Eccs and Edgs During Surveillance Activities	The purpose of this report is to provide a telephone notification under 10CFR50.73(a)(2)(iv)(A) for an invalid actuation of the Emergency Core Cooling Systems and the Emergency AC Electrical Power Systems. As allowed by 10CFR50.73(a)(1) this notification is being made via an ENS phone call. On October 7, 2004, James A. Fitzpatrick (JAF) was in a refueling outage with the reactor cavity flooded and the spent fuel pool gates removed. At approximately 2055, during the performance of an instrument surveillance procedure (ISP-5-7) for 'Remote Shutdown Reactor Vessel Pressure Indicator Calibration' a pressure perturbation was introduced to the sensing lines while returning 02-3PI-60A to service. The pressure perturbation resulted in an invalid actuation of the Emergency Core Cooling Systems (ECCS) and Emergency Diesel Generator (EDG) initiation logic. The initiation logic functioned as designed, 'A' and 'B' Core Spray Pumps and 'A', 'B', 'C', and 'D' RHR pumps started and injected into core, and EDGs 'A', 'B', 'C' and 'D' started but did not close to the emergency buses since there was no loss of off site power. Operations personnel verified plant conditions, secured ECCS injection and secured the EDGs. The event was entered in the plant corrective action program via Condition Report CR-JAF-04-04457. Preliminary investigation has identified valve manipulation to restore pressure indicator 02-3PI-60A to service as the cause of a pressure perturbation in a shared sensing line, which caused the level transmitters sharing the same line to detect a low reactor water level condition. All systems that actuated operated per their design. As a result of the injection, water overflowed to the skimmer surge tanks and into the reactor building drain system as designed, resulting in contamination of sections of the reactor building. No personnel contaminations or unexpected exposures occurred as a result. A recovery plan was developed and implemented to minimize the impact and to recover the affected plant areas. In addition, signals were received for HPCI and RCIC initiation, SBGT initiation, Recirc pump trip and Alternate Rod Insertion (ARI). Due to outage activities, these systems were removed from service and protective tagged, and thus the systems did not respond. Although not required by 10 CFR 50.73 JAF will submit a voluntary LER to address this event in greater detail. The NRC Sr. Resident Inspector was notified. The licensee also informed the State of New York.

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