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 Entered dateSiteRegionReactor typeEvent description
ENS 5432010 October 2019 02:17:00FarleyNRC Region 2At 2340 CDT, on October 09, 2019, a site contractor was transported offsite for treatment at an offsite medical facility. Due to the nature of the medical emergency, the individual was not thoroughly surveyed prior to being transported offsite. This is an eight-hour notification, non-emergency for the transportation of a contaminated person offsite. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xii). Following the individual being transported offsite, but prior to the individual arriving at the offsite medical facility, the individual was confirmed to not be contaminated. This occurred at approximately 2350 CDT, on October 09, 2019. The NRC Resident Inspector has been notified."
ENS 543189 October 2019 12:48:00VogtleNRC Region 2At 1023 EDT, on October 9, 2019, with Unit 2 in Mode 1 and 100 percent power, an actuation of the Emergency Diesel Generator and Auxiliary Feedwater Systems occurred. The reason for the Emergency Diesel Generator auto-start was the loss of power to the 4160V 1E electrical bus 2AA02 due to a fault at an offsite electrical switchyard. The Emergency Diesel Generator started and energized the 4160V safety bus, and Auxiliary Feedwater Systems automatically started as designed when the undervoltage condition on the safety bus was detected. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Emergency Diesel Generator and Auxiliary Feedwater Systems. There was no impact on the health and safety of the public or plant personnel. The NRC resident has been notified."
ENS 5428721 September 2019 10:28:00FarleyNRC Region 2At 0800 (CDT), with Unit 2 in Mode 1 at 100 percent (power), the reactor was manually tripped due to elevated vibration indication on the 2C reactor coolant pump exceeding annunciator response procedure trip criteria. The trip was not complex, with all systems responding normally post trip. Auxiliary Feedwater (AFW) auto actuated as expected following the manual reactor trip. Operations responded and stabilized the plant. Decay heat is being removed via the use of AFW and subsequent steaming of the steam generators to the main condenser. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b(2)(iv)(B). In addition, this event report is being reported as an eight-hour non-emergency notification per 10 CFR50.72(B)(3)(iv)(A) for a specified system actuation. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. Farley reported that there was no increase in containment unidentified leakage or fluctuations with RCP seal flow during this event.
ENS 5423723 August 2019 00:33:00HatchNRC Region 2At 2234 (EDT), on 08/22/2019, while Unit (2) was at approximately 100 percent power in MODE 1, it was discovered that Unit 2 HPCI was INOPERABLE. HPCI does not have a redundant system; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Initial evaluation is that it is a power supply issue. The system had no indications, followed by smoke from behind the circuitry, and then indications returning. This is associated with a 14-day Limiting Condition for Operation.
ENS 5417519 July 2019 13:05:00VogtleNRC Region 2At 0945 (EDT) on July 19, 2019, with Unit 2 in Mode 1 and 100 percent power, the reactor automatically tripped due to Loop 2 'B' Main Steam Isolation Valve failing shut. The Auxiliary Feedwater system (AFW) started automatically as a result of the automatic reactor trip. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam lines through the steam dumps and into the condenser. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Due to the valid AFW actuation from the reactor trip, this event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). Unit 1 was not affected. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspectors have been notified. All control rods fully inserted.
ENS 540401 May 2019 21:03:00FarleyNRC Region 2

EN Revision Text: MANUAL REACTOR TRIP DUE TO MISALIGNED CONTROL ROD At 1643 (CDT), with Unit 2 in Mode 2 during low power physics testing, the reactor was manually tripped per procedure due to a misaligned control rod. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the atmosphere using the atmospheric relief valves. Unit 1 is not affected. Due to the Reactor Protection System actuation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 05/08/2019 AT 1212 EDT FROM MIKE CONNER TO JEFFREY WHITED * * *

This Event Notification is being updated to clarify that the reactor was not critical when this event occurred. Therefore, the reporting requirement is changed from 10 CFR 50.72(b)(2)(iv)(B) to 10 CFR 50.72 (b)(3)(iv)(A). The reactor was tripped during low power physics testing. The misaligned rod was encountered during rod group insertion and the affected bank had been inserted to the extent that the reactor was subcritical when the operators tripped the reactor. The licensee notified the NRC Resident Inspector. Notified R2DO (Lopez)

ENS 5401722 April 2019 16:16:00VogtleNRC Region 2A contract supervisor tested positive for drugs on a follow-up fitness-for-duty test. The contractor's access to the facility has been revoked and his badge was confiscated. Additionally, the supervisor failed a random test administered the next day (see EN #54018). The licensee notified the NRC Resident Inspector.
ENS 5401822 April 2019 16:16:00VogtleNRC Region 2

EN Revision Text: CONTRACT SUPERVISOR TESTED POSITIVE ON A RANDOM FITNESS-FOR-DUTY TEST A contract supervisor tested positive for drugs on a random fitness-for-duty test. The contractor's access to the facility has been revoked and his badge was confiscated. Additionally, the supervisor failed a follow-up test administered the previous day (see EN #54017).

  • * * RETRACTION ON 4/30/2019 AT 1642 EDT FROM KELLI ROBERTS TO BRIAN LIN * * *

On April 16, 2019, an individual was selected for a follow-up drug test. The same individual was selected again on April 17, 2019 for a random drug test. The results for both tests were ruled by the Medical Review Officer (MRO) on the same day and ruled positive for the same drug on April 22, 2019. These FFD violations were reported to the NRC on April 22, 2019, as EN #54017 and EN #54018, respectively. As allowed by 10 CFR 26.185(o), the MRO further reviewed the quantitation of the drug in both tests and determined that no further drug use had occurred since the first positive test. Therefore, the MRO concluded that this should be considered one FFD violation, and EN #54018 is being retracted. No changes are needed to EN #54017. The NRC Resident Inspector has been notified of this retraction. Notified R2DO (Heisserer) and FFD Group (email).

ENS 5400216 April 2019 02:59:00FarleyNRC Region 2At 2355 CDT on 4/15/19, life-saving activities by offsite medical personnel for a Farley employee were terminated. The coroner declared the individual deceased at the plant site at 0130 CDT. The fatality is not believed to be work-related and the individual was inside of the Radiological Controlled Area. This is a four-hour notification, non-emergency for a notification of other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. The licensee will be notifying the Occupational Safety and Health Administration due to the on-site fatality. The licensee will perform a radiological survey of the individual prior to transportation offsite.
ENS 5396731 March 2019 00:17:00VogtleNRC Region 2At 2130 (EDT) on March 30, 2019, with Unit 2 in Mode 1 at 30 percent reactor power, the reactor was manually tripped due to a main steam isolation valve failing closed. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam lines through the steam dumps and into the condenser. The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). Unit 1 was not affected. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspectors have been notified."
ENS 5395324 March 2019 05:23:00HatchNRC Region 2At 0159 (EDT), with Unit 2 in Mode 1 at 25 percent power, the reactor was manually tripped due to degrading condenser vacuum. After the turbine was tripped, the station service electrical buses did not transfer to alternate supply resulting in loss of the condensate feedwater system and level being controlled by the RCIC system. Operators responded and stabilized the plant. Reactor water level is being maintained via the RCIC system. Pressure is being controlled and decay heat is being removed by the HPCI system in pressure control mode. Unit 1 is not affected. Additionally, an actuation of the primary containment isolation system occurred during the reactor scram. The reason for the actuation was a group II isolation signal was received on reactor water level and a group I isolation was received on decreasing vacuum. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non- emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, this event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the primary containment isolation system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."
ENS 5394419 March 2019 07:41:00HatchNRC Region 2At 0140 (EDT) on 03/19/2019, while the unit was at approximately 1% power and 154 psig pressure in MODE 2, it was discovered that Unit 2 High Pressure Coolant Injection (HPCI) was INOPERABLE. HPCI does not have a redundant system, therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Unit 2 entered LCO 3.5.1.c for the HPCI being inoperable. There is no effect on Unit 1.
ENS 5389323 February 2019 09:13:00HatchNRC Region 2At 0212 EST on February 23, 2019, with Unit 2 in Mode 5, an actuation signal for the 2C Emergency Diesel Generator (EDG) was received during the Loss of Coolant Accident / Loss of Offsite Power logic system functional test. The 2C EDG was running and tied onto the 2G 4160 emergency bus when the alternate supply breaker was closed as required per the test procedure. Immediately upon closing the alternate supply breaker, both the alternate supply breaker and 2C EDG output breaker tripped open. The 2C EDG output breaker reclosed once the 2G 4160 bus undervoltage relays sensed a deenergized bus. When the 2C EDG tied to the 2G 4160 bus, the bus voltage was noted as being high, and the 2C EDG was secured. Investigation is ongoing to determine the cause of the initial bus undervoltage and the subsequent bus excessive voltage. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the emergency AC power system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. This event puts Unit 1 in a 72 hour Limiting Condition for Operation for the 1C Startup Transformer being out of service.
ENS 5383217 January 2019 16:00:00VogtleNRC Region 2A non-licensed contractor supervisor had a confirmed positive for alcohol during a for-cause fitness-for-duty test. The contractor's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 5382715 January 2019 12:25:00FarleyNRC Region 2At 0800 CST on January 15, 2019, a non-licensed employee supervisor had a confirmed positive for alcohol during a for-cause fitness-for-duty test. The employee's access to the plant has been placed on hold. The NRC Resident Inspector has been notified.
ENS 538178 January 2019 08:12:00HatchNRC Region 2On November 12, 2018, at 1331 EST, Unit 1 secondary containment isolated and Standby Gas Treatment (SBGT) systems started on Unit 1 and Unit 2 due to a blown fuse. The blown fuse was caused by a degraded refuel floor radiation monitoring relay, causing the radiation monitor to trip and resulted in an invalid actuation of the Unit 1 Group 10 and Group 11 primary containment isolation valves, all Unit 1 secondary containment isolation valves, and auto start of the Unit 1 and Unit 2 SBGT system. The Unit 1 Fission Product Monitor isolated and tripped and both Unit 1 H202 Analyzers isolated. This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) since the containment isolation and auto-start of SBGT on both units was not part of a pre-planned sequence and the event resulted in the invalid actuation of general containment isolation valves in more than one system. All primary and secondary containment isolation valves, with the exception of the 2T41F003A, Refueling Floor Inboard Vent Supply Isolation valve, functioned successfully. The refuel floor inboard vent supply isolation valve failed to travel fully closed on the secondary containment isolation signal and was therefore declared inoperable. The 2T41F003B refuel floor outboard vent supply isolation valve was verified to go fully closed and therefore isolation of that associated penetration line was successful. After assistance from maintenance, the valve was verified to be fully closed. All SBGT systems functioned successfully. The associated fuse and relay were replaced, and secondary containment was returned to normal service. The licensee notified the NRC Resident Inspector."
ENS 5378913 December 2018 14:37:00VogtleNRC Region 2At 1700 EST on December 12, 2018, a contractor supervisor violated the licensee's Fitness-for-Duty (FFD) program by subverting a follow-up Fitness for Duty Test. The contractor's site access has been terminated. The NRC Resident Inspector was notified. No work was performed on safety related equipment. The licensee has made a PADs entry.
ENS 536434 October 2018 07:57:00VogtleNRC Region 2

EN Revision Text: MANUAL REACTOR TRIP DURING LOW POWER PHYSICS TESTING At 0544 EDT on October 4, 2018, with Unit 1 in Mode 2 with reactor power in the intermediate range performing low power physics testing, the reactor was manually tripped due to a rod control urgent failure alarm. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam system. Unit 2 was not affected. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspectors have been notified. All control rods inserted as expected. The cause of the rod control urgent failure is being investigated.

  • * * UPDATE FROM KEVIN LOWE TO DONALD NORWOOD AT 1408 EDT ON 10/19/2018 * * *

This Event Notification is being updated to clarify that the reactor was not critical when this event occurred. Therefore, the reporting requirement is changed from 10 CFR 50.72(b)(2)(iv)(B) to 10 CFR 50.72 (b)(3)(iv)(A). During Dynamic Rod Worth Measurement testing, Control Bank Charlie was inserted approximately 153 steps when the urgent failure occurred (CBC positioned at 75 steps out). Following the scram, additional analysis concluded that the reactor was subcritical when the Reactor Protection System was actuated." The licensee notified the NRC Resident Inspector. Notified the R2DO (McCoy).

ENS 5362827 September 2018 14:42:00VogtleNRC Region 2At 1000 EDT on September 27, 2018, Southern Nuclear Operating Company (SNC) determined a contractor supervisor confirmed positive for drugs during a random Fitness-for-Duty (FFD) test. The employee's unescorted access to the plant has been suspended. The NRC Resident Inspector has been notified."
ENS 5361018 September 2018 08:28:00VogtleNRC Region 2At 1610 (EDT) on September 17, 2018, Southern Nuclear Operating Company (SNC) determined a contractor supervisor confirmed positive for drugs during a Fitness-for-Duty (FFD) test. The employee's unescorted access to the plant has been suspended. The NRC Resident Inspector has been notified."
ENS 5359912 September 2018 14:14:00VogtleNRC Region 2At 1115 EDT on September 12, 2018, Southern Nuclear Operating Company (SNC) determined a contractor supervisor confirmed positive for drugs during a Fitness-for-Duty (FFD) test. The employee's unescorted access to the plant has been suspended. The Resident Inspector has been notified."
ENS 5359511 September 2018 17:53:00VogtleNRC Region 2At 1430 EDT on September 11, 2018, Southern Nuclear Operating Company (SNC) determined a contractor supervisor confirmed positive for drugs during a Fitness-for-Duty (FFD) test. The employee's unescorted access to the plant has been suspended. The NRC Resident Inspector has been notified."
ENS 5356324 August 2018 15:14:00VogtleNRC Region 2At 1034 (EDT) on August 24, 2018, a contractor supervisor violated the licensee's Fitness-for-Duty (FFD) program by subverting a follow-up Fitness for Duty Test. The contractor's site access has been terminated. The NRC Resident Inspector was notified."
ENS 535387 August 2018 14:45:00VogtleNRC Region 2At 0909 (EDT) on August 7, 2018, Southern Nuclear Operating Company (SNC) determined a contractor supervisor confirmed positive for alcohol during a random Fitness-for-Duty (FFD) test. The employee's unescorted access to the plant has been suspended. The NRC Resident Inspector has been notified."
ENS 5352827 July 2018 13:41:00VogtleNRC Region 2A non-licensed contractor supervisor had a confirmed positive for alcohol during a random fitness-for-duty (FFD) test. The employee's unescorted access to the plant has been suspended. The licensee notified the NRC Resident Inspector.
ENS 5351719 July 2018 12:36:00VogtleNRC Region 2At 1300 (EDT) on July 18, 2018, a contractor supervisor violated the licensee's Fitness-for-Duty (FFD) program by subverting the Fitness for Duty process. The contractor's site access has been terminated. The NRC Resident Inspector was notified."
ENS 5351518 July 2018 19:27:00VogtleNRC Region 2A contractor supervisor tested positive for alcohol during an access upgrade fitness-for-duty test. The employee's access to the facility has been suspended. The licensee has notified the NRC Resident Inspector.
ENS 5351117 July 2018 17:01:00VogtleNRC Region 2At 1338 (EDT) on July 17, 2018, Southern Nuclear Operating Company (SNC) determined a contractor supervisor confirmed positive for a controlled substance during a random Fitness-for-Duty (FFD) test. The employee's unescorted access to the plant has been suspended. The Resident Inspector has been notified."
ENS 534999 July 2018 20:05:00VogtleNRC Region 2At 1515 (EDT) on July 9, 2018, Southern Nuclear Operating Company (SNC) determined a contractor supervisor confirmed positive for a controlled substance during a for cause Fitness-for-Duty (FFD) test. The employee's unescorted access to the plant has been suspended. The NRC Resident Inspector has been notified."
ENS 534843 July 2018 12:00:00VogtleNRC Region 2At 0954 (EDT) on July 3, 2018, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually tripped due to high steam generator water level. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam lines through the steam dumps and into the condenser. The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). Unit 2 was not affected. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspectors have been notified. All control rods inserted and Unit 1 is in an electrical shutdown lineup. The cause of the high steam generator water level transient is being investigated.
ENS 5347928 June 2018 13:05:00VogtleNRC Region 2At 0902 EDT on June 28, 2018, a non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 5346922 June 2018 15:12:00VogtleNRC Region 2At 0900 (EDT) on June 22, 2018, a non-licensed supervisory contractor subverted a random Fitness for Duty test. The contractor's site access has been terminated. The NRC Resident Inspector was notified. "
ENS 533938 May 2018 10:38:00FarleyNRC Region 2Westinghouse PWR 3-LoopOn May 8, 2018 at 0139 Central Daylight Time, Farley Nuclear Plant Unit 1 declared containment inoperable due to total containment leak rate greater than technical specifications. The 1B containment cooler had seat leakage of approximately 30 gallons per minute from a service water drain valve. Though the containment cooler service water supply is not tested per the Appendix J program, a loss of the containment barrier is possible under accident conditions. The service water flow path to the 1B containment cooler has been isolated to exit the condition. The licensee will notify the NRC resident inspector.
ENS 533928 May 2018 01:39:00FarleyNRC Region 2Westinghouse PWR 3-LoopOn May 7, 2018 at 1041 CDT, Unit 1 performed an RCS (reactor coolant system) leakrate procedure that calculated an unidentified RCS leakrate of 0.202 gpm. The leak source investigation concluded at 2150 that the packing for the charging flow control valve (FCV) was the source of the RCS leakage when it was bypassed, which isolated the leakage. A second RCS leakrate calculation was performed after the charging flow control valve was isolated which calculated an acceptable leakrate of 0.00 gpm. The packing leakage from the charging flow control valve represented leakage external to containment which would result in a greater that 5 Rem dose projection to control room personnel during accident conditions which does not satisfy the GDC19 criteria described in Technical Specification Bases 3.7.10. Therefore the control room emergency filtration system would not be able to fulfill its design function resulting in an unanalyzed condition. This condition is being reported pursuant to 10CFR50.72(b)(3)(ii) for a 'condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety'. The packing leak from the charging flow control valve will remain isolated until repaired under work order SNC944374. The NRC Resident Inspector has been notified.
ENS 533092 April 2018 18:33:00VogtleNRC Region 2W-AP1000A contractor employee supervisor had a confirmed positive for illegal drugs during a fitness for duty test. The employee's access to the site has been terminated. The NRC Resident Inspector has been notified.
ENS 5329026 March 2018 18:19:00FarleyNRC Region 2Westinghouse PWR 3-Loop

On March 25, 2018 at 1833 CDT, while at 100 percent power, Farley Unit 1 (FNP-1) conservatively declared a single Main Steam Isolation Valve (MSIV) inoperable on the 1C Steam Generator line due to indication of Steam Generator pressure rise with a corresponding reduction in flow of that loop. FNP-1 began a reactor shutdown at 0400 CDT on March 26, 2018 to establish plant conditions to support testing the affected main steam line MSIVs while in the required action time of Technical Specification 3.7.2. At 1338 CDT on March 26, 2018, testing confirmed that the single MSIV was inoperable and that valve disassembly will be required. The duration of the valve repair would exceed the required action time of Technical Specification 3.7.2. This report is being made in accordance with 10 CFR 50.72(b)(2)(i), as a plant shutdown required by technical specifications. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM DOUGLAS HOBSON TO KEN MOTT AT 0202 EDT ON 5/16/18 * * *

This EN (event notification) is being updated to clarify the reporting criteria as 'Voluntary'. Farley Technical Specification 3.7.2 allows continuous operation in MODE 2 with an INOPERABLE MSIV as long as the other MSIV in the affected Main Steam Line is closed. The initiation of the shutdown was performed as a prudent action to repair and restore OPERABILITY of the affected MSIV and was not a requirement of the Farley Technical Specifications. The licensee notified the NRC Resident Inspector. The R2DO (Masters) was notified.

ENS 5327722 March 2018 09:31:00VogtleNRC Region 2W-AP1000A non-licensed contract supervisor attempted to subvert a random Fitness For Duty test using a subversion kit on their person. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 531599 January 2018 23:38:00FarleyNRC Region 2Westinghouse PWR 3-Loop

On January 9, 2018, at 1759 CST, during review of NFPA 805 requirements and circuit analysis, it was determined that the NFPA 805 analysis and Fire Safe Shutdown Modeling did not consider all fire-induced failures. As such, a condition could possibly exist during a postulated fire where both safety related electrical trains could be impacted. This notification is to report a condition involving the fire safe shutdown analysis. The condition could result in an adverse impact on the ability of operators to respond to a postulated fire in these areas. Therefore, this notification is being made pursuant to 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. Compensatory fire watches have been established in the affected areas. The licensee notified the NRC Resident Inspector.

  • * * RETRACTION FROM ANTONIO BENFORD TO HOWIE CROUCH AT 1752 EST ON 2/28/18 * * *

Following additional refinements to the NFPA 805 Fire PRA Model, the circuits which initiated the initial report of an unanalyzed condition have now been evaluated and have proven that no significant degradation to plant safety existed. Therefore, EN 53159 is being retracted. The NRC Resident Inspector has been notified. Notified R2DO (Michel).

ENS 5301816 October 2017 17:59:00VogtleNRC Region 2W-AP1000A non-licensed supervisory contractor subverted a random Fitness for Duty (FFD) test. At 0851 (EDT) on October 16, 2017, a contractor supervisor was notified to report for a random FFD test and (the individual) did not report to the testing facility and left the premises. The contractor's site access has been terminated. The NRC Resident Inspector was notified.
ENS 5299226 September 2017 11:03:00VogtleNRC Region 2Westinghouse PWR 4-LoopAt approximately 0543 (EDT), while (recovering from the performance of) 2B Emergency Diesel Generator and ESFAS testing, a (subsequent) valid undervoltage actuation signal was sent to the 2B Emergency Diesel Generator (EDG). The 2B AC emergency bus (2BA03) was load shed, the 2B EDG automatically started, and tied to 2BA03. The 2BA03 bus was loaded by the automatic load sequencer. The actuation was identified by the Control Room operators and the 2B EDG was locally monitored while in service. This actuation is reportable due to the automatic actuation of a system listed in 10 CFR 50.72(b)(3)(iv)(B). The reactor was not critical at the time of the event and not challenged throughout the event. Decay heat removal and spent fuel pool cooling were not challenged throughout the event. The NRC Resident Inspector has been notified. The cause of the undervoltage condition is under investigation.
ENS 5296814 September 2017 09:10:00VogtleNRC Region 2W-AP1000On September 13, 2017, Southern Nuclear Operating Company (SNC) determined a Contractor Manager confirmed positive for alcohol during a for cause fitness-for-duty test. The employee's unescorted access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 5281822 June 2017 09:32:00HatchNRC Region 2GE-4On April 27, 2017 at 0029 EDT, Unit 2 received an invalid partial Group 2 isolation due to a failed relay (2D11-K80) on the auxiliary trip unit. Both of the U1 and U2 Standby Gas Treatment (SGT) trains started and the U2 Group II primary containment and all secondary containment inboard isolation valves closed. Also, the refuel floor isolation dampers closed, the reactor building supply and exhaust fans tripped, and the refueling floor supply and exhaust fans tripped. This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) since the containment isolation and auto-start of SGT on both units was not part of a pre-planned sequence and the event resulted in the invalid actuation of general containment isolation valves in more than one system. All primary and secondary containment inboard isolation valves and SGT systems functioned successfully. The failed relay was replaced and the systems were restored to normal alignment. The licensee will notify the NRC Resident Inspector.
ENS 5280314 June 2017 08:42:00HatchNRC Region 2GE-4On April 17, 2017 at 1120 EDT, following scheduled maintenance, the Reactor Protection System (RPS) 'A' bus was returned to its normal supply, the RPS 'A' motor generator (MG) set. The RPS MG set had been running loaded for 1 hour when the RPS 'A' bus tripped. Maintenance personnel had connected probes of a grounded oscilloscope to check for proper operation of the MG set, resulting in the RPS 'A' bus trip. The controlling procedure did not contain a caution about using only an ungrounded oscilloscope. The trip of the RPS 'A' MG set resulted in a half scram and an invalid isolation signal causing primary containment isolation valves in multiple systems to isolate. This event is reportable per 10 CFR 50.73 (a)(2)(iv)(A) since the containment isolation was not part of a pre-planned sequence and the event resulted in the invalid actuation of general containment isolation valves in more than one system. Corrective actions include revising the governing procedure with the proper precaution and limitation to require the use of an ungrounded oscilloscope. A training needs analysis will also be performed to cover the lessons learned from this event. The licensee notified the NRC Resident Inspector.
ENS 527852 June 2017 17:07:00FarleyNRC Region 2Westinghouse PWR 3-Loop

This notification is being made as required by 10 CFR 50.72(b)(3)(v) due to both trains of Penetration Room Filtration (PRF) being inoperable due to an inoperable PRF Boundary. At 0920 (CDT) on 6/2/2017, a gap was discovered between an electrical penetration room ceiling and the containment wall where seismic gap material was noted to be missing. The gap was subsequently closed and PRF testing completed sat. The condition was exited at 1345 (CDT). The licensee notified the NRC Resident Inspector.

  • * * RETRACTION ON 7/25/17 AT 1725 EDT FROM MATT STANLEY TO DONG PARK * * *

On 6/2/17 at 1707 CDT Farley Nuclear Plant notified the NRC Operations Center of an entry into Technical Specification 3.7.12 Condition B for Unit 1 loss of two trains of Penetration Room Filtration (PRF). At 0920 (CDT) on 6/2/2017, a gap had been discovered between an electrical penetration and containment where seismic gap material was noted to be missing. The report was made pursuant to 10 CFR 50.72(b)(3)(v) under Event Notification 52785. Upon further engineering review and satisfactory testing to support operability, Farley has determined that the configuration did not meet the criteria for a condition that could have prevented fulfillment of a safety function, and is retracting the notification. The NRC Resident Inspector has been notified. Notified R2DO (Blamey).

ENS 5276318 May 2017 15:14:00VogtleNRC Region 2W-AP1000On May 17, 2017, at 1852 (EDT), Southern Nuclear Operating Company (SNC) contractor informed SNC that a contractor Fitness-for-Duty (FFD) collector failed to follow the process defined in both NRC Regulation 10 CFR 26.105(b) and contractor FFD procedures. The collector was fully trained, qualified, and knowledgeable of the process requirements. This determination was made based on initial department investigations and corroborated via independent investigations performed by the contractor. SNC has determined that this is reportable under 10 CFR 26.719(b)(3). The licensee informed the NRC Resident Inspector and NRC Region II.
ENS 5270021 April 2017 05:41:00HatchNRC Region 2GE-4At 2345 (EDT) on 04/20/2017, the Unit 1 Reactor Mode Switch was taken to the Shutdown position to comply with Technical Specification 3.10.4 due to having no operable IRM's (Intermediate Range Monitors) in one quadrant of the reactor vessel as a result of maintenance activities. Placing the mode switch to Shutdown inserts a valid scram signal into the Reactor Protection System (RPS). All control rods had been previously inserted and no rod movement occurred when the mode switch was positioned to Shutdown. Due to this valid RPS scram, and not being a part of a preplanned evolution, this condition is reportable under criteria 50.72(b)(3)(iv)(A) as an event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. The licensee notified the NRC Resident Inspector.
ENS 5269620 April 2017 05:57:00HatchNRC Region 2GE-4On 04/20/2017 at 0302 EST during a reactor startup, a reactor scram resulted from upscale spike on two Intermediate Range Monitors (IRMs), 1C51K601A and 1C51K601B. IRM A, 1C51K601A is in Reactor Protection System Channel A and IRM B, 1C51K601B is in Reactor Protection System Channel B. All control rods fully inserted. No PCIS (Primary Containment Isolation System) actuations occurred and none were expected to occur based upon plant condition following the reactor scram. Investigation is in progress. Condition was not due to a true flux event. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B) as an event or condition that resulted in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. CR 10356172 The NRC Resident has been notified. The reactor was at 0.5% (percent) power at the time of the event and will remain in Hot Shutdown pending the results of the root cause investigation.
ENS 5268517 April 2017 13:35:00HatchNRC Region 2GE-4On February 17, 2017 at 1021 EST, secondary containment isolated and Standby Gas Treatment (SBGT) systems started on Unit 1 and Unit 2 during a maintenance activity to replace a relay in the Unit 2 primary containment isolation system. The work was being conducted as part of planned maintenance during the Unit 2 refueling outage. Poor work instruction led to a jumper not being installed as required, thus causing relays to de-energize, resulting in an invalid actuation of the Unit 1 and Unit 2 outboard primary and secondary containment isolation valves and auto-start of the Unit 1 and Unit 2 SBGT system. This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) since the containment isolation and auto-start of SBGT on both units was not part of a pre-planned sequence and the event resulted in the invalid actuation of general containment isolation valves in more than one system. All primary and secondary containment isolation valves and SBGT systems functioned successfully. The associated wires were re-landed and secondary containment was returned to normal service. The licensee notified the NRC Resident Inspector.
ENS 5265030 March 2017 12:58:00HatchNRC Region 2GE-4During the evaluation of tornado missile vulnerabilities and the potential impacts to safety-related plant equipment, it was concluded that the following structures are vulnerable to tornado generated missiles: All five of the Unit 1 and Unit 2 Emergency Diesel Generator (EDG) fuel oil storage tanks have ventilation pipe extending approximately 5 feet above grade. In the event that a tornado missile impact occurs on the aforementioned ventilation piping, there is a possibility that the vent lines could crimp. This could prevent the tanks from venting and inhibit the transfer of fuel oil from the main fuel oil storage tanks to the day tank. Ultimately, this causes the emergency diesels to be inoperable. These conditions are reportable per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(D) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety and also an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequence of an accident. This issue is being addressed in accordance with EGM-15-002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance. The NRC Resident Inspector has also been notified.
ENS 5264127 March 2017 16:02:00VogtleNRC Region 2W-AP1000A non-licensed contractor employee had a confirmed positive for a controlled substance during a random fitness-for-duty test. The employee's access to the site has been terminated. The licensee notified the NRC Resident Inspector.
ENS 5261917 March 2017 17:12:00VogtleNRC Region 2Westinghouse PWR 4-LoopAt approximately 1517 (EDT), while restoring protective relay power to the 1B Reserve Auxiliary Transformer, a valid undervoltage actuation signal was sent to the 1B Emergency Diesel Generator (EDG). The 1B EDG automatically started and tied to the safety bus (1BA03). The 1BA03 bus was loaded by the automatic load sequencer. This actuation was identified by the Control Room Operators and the 1B EDG was locally monitored while in service. This actuation is reportable due to the automatic actuation of a system listed in 10 CFR 50.72(b)(3)(iv)(B). The reactor was not critical at the time of the event and not challenged throughout the event. Containment Cooler Number 8 did not automatically start in 'Fast Speed' as expected. Containment Cooler Number 8 was successfully started in 'Fast Speed' manually by the Control Room Operators. The NRC Resident Inspector has been notified.
ENS 5256720 February 2017 01:14:00HatchNRC Region 2GE-4On 2/19/2017 at 2323 EST, during LLRT (local leak rate test) testing per 42SV-TET-001-2, 2T48F320 exceeded the maximum allowable leakage limit. The companion isolation valve in the same line (2T48F319) had previously failed LLRT. The failure of the 2T48F320 represents a failure of the 2T23X26 penetration to maintain primary containment integrity. This event is reportable per 10CFR50.72(b)(3)(ii)(A) since the failure of the 2T23X26 penetration caused primary containment leakage to exceed La (allowable leakage) and thus represents a degraded principle safety barrier. CR (condition report) 10333178. NRC Resident Inspector has been notified. 2T48F319 and 2T48F320 are 18 inch dampers. This event places the licensee in a Technical Specification limit that requires the dampers to be repaired and pass LLRT prior to the plant entering Mode 3.
ENS 5256317 February 2017 17:46:00HatchNRC Region 2GE-4On 2/17/2017 at 1414 EST, secondary containment was declared inoperable due to the discovery of an 18-inch open pipe penetration in the secondary containment boundary. During walkdown activities, it was discovered that a blind flange installed to support removal of a Unit 2 secondary containment isolation valve had been installed on the wrong flange to provide isolation for secondary containment. At 1503 EST, the blind flange was moved to the correct side of the flange and secondary containment was declared operable. This event is reportable per 10 CFR 50.72(b)(3)(v)(C) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system needed to control the release of radioactive material. In conjunction with operation of the Standby Gas Treatment (SGT) subsystems, secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations. Therefore, the lack of a qualified isolation device to limit the release of radioactive material constitutes a loss of safety function due to a loss of secondary containment integrity. CR 10332592 The NRC Resident has been notified.
ENS 5255816 February 2017 17:28:00HatchNRC Region 2GE-4On February 16, 2017 at 1320 EST, the 2A Emergency Diesel Generator (EDG) started in response to a valid actuation signal due to the momentary loss of the 2C Startup Transformer (SAT). While performing maintenance activities on the 2D SAT, the alternate supply breaker tripped and reclosed, allowing the 4160 2E Emergency Bus to be momentarily de-energized. When the 4160 2E Emergency Bus de-energized, the 2A EDG received a valid autostart signal due to emergency bus low voltage. Although, the 2A EDG did autostart, it did not tie to the 4160 2E Emergency Bus as the 4160 2E Emergency Bus was re-energized from the 2C SAT. This event is reportable per 10 CFR 50.72(b)(3)(iv)(A) since the autostart of the 2A EDG was not part of a pre-planned sequence and the event resulted in the valid actuation of an emergency ac electrical power system. CR 10332134 The NRC Resident has been notified.
ENS 525438 February 2017 17:31:00HatchNRC Region 2GE-4During a control room panel walk down by an on-shift Reactor Operator at approximately 1151 (EST) on 2/8/2017, Unit 1 High Pressure Coolant Injection (HPCI) suction and discharge pressure indicators were noted to be downscale. I & C investigated and found the output of 1E41K603, DC to AC inverter, degraded. This inverter also powers the HPCI flow controller. Without the flow controller HPCI would not auto-start to mitigate the consequences of an accident, thus HPCI was declared inoperable. All other emergency core cooling systems and the Reactor Core Isolation Cooling (RCIC) system remain operable. HPCI is a single train system with no redundant equipment in the same system, thus this failure is reportable as an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident, 10CFR50.72(b)(3)(v)(D). Inverter 1E41K603 was replaced and functionally tested satisfactorily at 1630 on 2/8/2017, restoring HPCI to operable status. The NRC Resident Inspector was notified.
ENS 525343 February 2017 18:56:00VogtleNRC Region 2Westinghouse PWR 4-LoopAt 1545 EST on 2/3/17, Vogtle Unit 1 was manually tripped from 100% power when loop 1 Main Steam Isolation Valve (MSIV) started to fail closed. Non-Safety Related 4160V bus 1NA01 failed to transfer to alternate incoming power supply automatically and was successfully manually energized. All control rods fully inserted and AFW (Auxiliary Feedwater) and FWI (Feedwater Isolation) actuated as expected. Unit 1 is in Mode 3 and stable with decay heat being removed by AFW. The licensee informed the NRC Resident Inspector.
ENS 525313 February 2017 16:33:00HatchNRC Region 2GE-4In 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 525261 February 2017 16:46:00HatchNRC Region 2GE-4

Plant Hatch declared a notification of unusual event. Subsequent investigation determined the paraphernalia was related to plant drills. The Unusual Event was terminated at 1727 EST. The licensee notified the NRC Resident Inspector. Notified the DHS SWO, FEMA, NICC, and NNSA (via e-mail).

  • * * RETRACTION FROM KENNETH HUNTER TO MARK ABRAMOVITZ AT 1829 EST ON 2/1/2017 * * *

The event was retracted. The licensee notified the NRC Resident Inspector. Notified the R2DO (Shakur), NSIR (Holian), IRD MOC (Gott), DHS SWO, FEMA, NICC, and NNSA (via e-mail).

ENS 5251830 January 2017 14:38:00VogtleNRC Region 2Westinghouse PWR 4-LoopAt approximately 1049 EST, an individual was discovered to be unresponsive at a site training facility outside of the Protected Area. The individual was transported via ambulance to the Burke County Hospital and was declared deceased at 1143 EST. The licensee has notified the NRC Resident Inspector. The licensee notified the State of Georgia Department of Labor - OSHA at 1414 EST.
ENS 5248211 January 2017 13:52:00VogtleNRC Region 2W-AP1000A non-licensed supervisor tested positive for alcohol during a for-cause test. The individual's access to the site has been terminated pending the results of an investigation. The licensee notified RII (Patterson) and the NRC Resident Inspector.
ENS 524239 December 2016 23:46:00VogtleNRC Region 2Westinghouse PWR 4-LoopOn December 09, 2016 at 1734 EST, U2 Train-A NSCW (Nuclear Service Cooling Water) Transfer Pump #8 tripped during Return To Service Surveillance testing for Train-B NSCW Transfer pump #7. Technical Spec 3.7.9 Condition E entered at 1734 with Required Actions to be in M3 (mode 3, Hot Standby) in 6 hours AND M4 (mode 4, Hot Shutdown) in 12 hours. A unit shutdown was commenced at 2042 EST (as a conservative measure) to comply with TS 3.7.9 Condition E. At 1937, U2 B-train NSCW Transfer Pump #7 was declared operable and TS 3.7.9 Condition E was exited. The plant is currently raising power to 100%. The licensee notified the NRC Resident Inspector.
ENS 524147 December 2016 17:24:00FarleyNRC Region 2Westinghouse PWR 3-LoopDuring the evaluation of tornado missile vulnerabilities and the potential impacts to Technical Specification (TS) plant equipment, it was concluded that the following SSCs (systems, structures, and components) were vulnerable to tornado generated missiles. The Service Water Intake Structure (SWIS) intake and exhaust ventilation hoods, located on the roof of the SWIS, are not adequately protected from missiles generated by a tornado. Should a tornado-generated missile strike the SWIS intake and exhaust ventilation hoods, the hoods could crimp thus reducing air flow and challenging the performance of their heating and cooling safety functions. If the intake hoods were damaged or removed due to a missile strike, entry of rainwater could occur due to severe weather high wind velocity, and could affect safety related electrical equipment in the rooms directly below the hoods. These potential conditions could render Service Water trains inoperable on either or both units. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v)(D) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. This condition is being addressed in accordance with EGM-15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents). Required actions have been taken. Corrective actions will be documented in a follow up licensee event report. The NRC Resident Inspector will be notified. The licensee is evaluating the operability of the service water system. Should one train be declared inoperable, the licensee would be in a 72 hr. LCO action statement. If both trains are inoperable, then the licensee would enter T.S. 3.0.3.
ENS 524127 December 2016 13:12:00VogtleNRC Region 2W-AP1000In accordance with 10 CFR 52.99(c)(2), Vogtle Units 3 and 4 Construction is making this notification to NRC for determining that Inspection, Test, Analysis, and Acceptance Criteria (ITAAC) 2.6.01.02.ii (ECS System Seismic Category I Equipment Design Basis Loads) for both units requires additional actions to restore its completed status. The Closure Notification for this ITAAC (NRC Index No. 580) was originally submitted on May 17, 2016 (reference ML 16138A080 and ML 16166A030). On November 1, 2016, it was determined by the Vogtle 3&4 Contractor that modifications to the Reactor Coolant Pump (RCP) breaker (i.e., switchgear) cabinet design were required to ensure compliance with the applicable portions of IEEE 384, 'Standard Criteria for Independence of Class 1E Equipment and Circuits.' The modification involves an engineering change which adds different equipment to the RCP Breaker cabinet that function to trip the RCP. The new components were not previously qualified for use in the RCP breaker cabinet assembly. Additional seismic qualification type testing and analysis of components are being performed for the added components in the RCP breaker. Update of the Equipment Qualification Data Package and Equipment Qualification Summary Report for the RCP breaker to confirm the breaker withstands seismic design basis loads and Licensee's acceptance is in progress. The revised ITAAC Completion Notice will be submitted to the NRC once all related ITAAC activities have been completed. The licensee has notified the NRC Resident Inspector.
ENS 5239527 November 2016 03:08:00FarleyNRC Region 2Westinghouse PWR 3-LoopAt 0026 (CST) on November 27, 2016, Farley Unit 1 was manually tripped from 100% reactor power due to voltage swings suspected to be caused by the Auto Voltage Regulator. All control rods fully inserted and Auxiliary Feedwater (AFW) auto-started as expected. All systems responded as expected. The plant is currently stable in Mode 3 (Hot Standby). The cause of the main generator voltage oscillations is under investigation. The NRC Resident Inspector has been notified. The trip was uncomplicated. Decay heat is being removed via the steam dumps to condenser. The plant is at normal operating pressure and temperature with auxiliary feedwater supplying the steam generators. The electrical grid is stable and supplying plant loads. All safety equipment is available, if needed. Unit 2 was unaffected by the event and remains at 100% power.
ENS 5237417 November 2016 20:51:00FarleyNRC Region 2Westinghouse PWR 3-LoopAt 1859 CST on November 17, 2016, Farley Nuclear Plant Unit 1 initiated a shutdown from approximately 99 percent reactor power. The shutdown was initiated per Technical Specification LCO 3.0.3. This LCO entry was based on having no operable steam flow channels on the C loop for Farley Nuclear Plant Unit 1. Unit 2 as not affected. The NRC Resident Inspector has been notified.
ENS 523568 November 2016 17:36:00FarleyNRC Region 2Westinghouse PWR 3-LoopAt 1331 (CST) on November 8, 2016, Farley Nuclear Plant Unit 1 manually tripped from 32% reactor power. The plant was ramping down to remove the main generator from service due to an unrelated issue. 1A SGFP did not respond to control Steam Generator (SG) level as expected when the miniflow was opened per procedure. SG levels lowered due to lower feed flow and the reactor was manually tripped in accordance with plant procedures. All control rods fully inserted and Auxiliary feedwater (AFW) auto started as expected. The Main Steam Isolation Valves were closed to minimize the cool-down. Decay heat is being removed through the Atmospheric Relief Valves. All other systems responded as expected. The plant is currently stable in Mode 3 (Hot Standby). The failure of the 1A SGFP control is under investigation. Unit 2 was not affected. The NRC Resident Inspector has been notified. There is no primary to secondary leakage.
ENS 523401 November 2016 19:01:00FarleyNRC Region 2Westinghouse PWR 3-Loop

At 1743 CDT on 11/1/16, readings were obtained that indicated ammonia above IDLH (Immediately Dangerous to Life and Health) values in the Unit 1 Auxiliary Building and an Alert was declared by the Emergency Director. The plant is stable with Unit 1 at 2% power. Unit 2 is at 100% power (and stable). The site is currently investigating to determine the source and ventilate the area." The licensee has informed the NRC Resident Inspector. Notified DHS SWO, FEMA Ops Center, USDA Ops Center, HHS Ops Center, DOE Ops Center, DHS NICC Watch Officer, EPA EOC, FEMA National Watch Center (email), FDA EOC (email), Nuclear SSA (email).

  • * * UPDATE FROM BLAKE MITCHELL TO STEVEN VITTO ON 11/02/2016 AT 0443 EDT * * *

The following is the Emergency Preparedness 8-hour report for the Plant Farley ALERT declaration. An ALERT was declared at 1743 CDT on November 1, 2016, based on Emergency Action Level (EAL) HA3 (Release of toxic, asphyxiant, or flammable gases within vital areas which jeopardizes operation of systems required to maintain safe operations or establish or maintain safe shutdown) due to an ammonia discharge into the Auxiliary Building. Farley Unit 1 was in Mode 2 at 2 percent power and Unit 2 was in Mode 1 at 100 percent power throughout the event. Actions completed during and leading up to termination from the event included walkdown of the area to locate the ammonia source, and isolating the ammonia source while atmospheric monitoring was performed in all affected and adjacent areas. Installation of additional ventilation fans was completed. This event has been entered into the Farley Corrective Action Program under Condition Report 10293519 (Ammonia levels in 100 Feet Radside). A review of the Alert classification has determined that the classification was timely and accurate. All required notifications were completed accurately and in a timely manner. The ERO (Emergency Response Organization) notification system functioned as expected and all emergency response facilities were activated in a timely manner. There were no missing or injured personnel during the event. There is no evidence of tampering or sabotage of plant equipment leading to this event. At 2202 CDT on 11/1/2016, the ammonia source was identified as Valve N1P20V913. At 2228 CDT, the ammonia source was isolated. Final assessment of the Auxiliary Building showed that the ammonia levels were less than 5 ppm. It was determined that a deficiency in the valve component was the cause of the ammonia leak. The Alert was terminated on 11/1/2016 at 2340 CDT. The Licensee has notified the NRC Resident Inspector. Notified R2DO(Ernstes), IRD (Stapleton), NRR EO (Miller). Notified DHS SWO, FEMA Ops Center, USDA Ops Center, HHS Ops Center, DOE Ops Center, DHS NICC Watch Officer, EPA EOC, FEMA National Watch Center (email), FDA EOC (email), Nuclear SSA (email).

ENS 5231926 October 2016 16:35:00VogtleNRC Region 2Westinghouse PWR 4-LoopDuring the evaluation of tornado missile vulnerabilities and the potential impacts to Technical Specification (TS) plant equipment, it was concluded that the following SSCs (Structures, Systems, and Components) were vulnerable to tornado generated missiles: The Turbine Driving Auxiliary Feedwater (TDAFW) pump exhaust vent is not adequately protected from tornado generated missiles. A tornado could generate missiles capable of striking the exhaust piping on the Turbine Driven Auxiliary Feedwater Pump, crimping the piping resulting in reduced flow capacity and rendering the TDAFW pump inoperable. The Condensate Storage Tank (CST) vents are not adequately protected from tornado generated missiles. A tornado generated missile could strike the vent piping rendering it unable to relieve vacuum on the top of the CST and allow adequate water flow to the Auxiliary Feedwater Pumps and rendering the CST inoperable. These conditions are reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v)(D) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. This condition is being addressed in accordance with EGM-15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents). The NRC Resident lnspector has been notified.
ENS 522741 October 2016 09:42:00FarleyNRC Region 2Westinghouse PWR 3-LoopAt 0512 (CDT) on October 1, 2016, Farley Nuclear Plant Unit 1 automatically tripped from 99 percent reactor power due to the inadvertent closure of a main steam isolation valve (MSIV). The closure of the MSIV caused a turbine trip resulting in an automatic reactor trip. Concurrent with the reactor trip, a safety injection (SI) occurred. The plant is stable in Mode 3 (Hot Standby) and auxiliary feedwater (AFW) autostarted as expected. The cause of MSIV closure and SI actuation is under investigation. Cooldown will continue to Mode 5 (Cold Shutdown) as planned for entry into a scheduled refueling outage. Restart is not planned until the completion of the refueling outage. Unit 2 was not affected. The NRC Resident Inspector has been notified. The MSIVs are open with the steam generators discharging steam to the main condenser using the turbine bypass valves. SI was from high head injection which has been secured.
ENS 5220830 August 2016 16:05:00VogtleNRC Region 2W-AP1000A non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated.
ENS 521575 August 2016 20:55:00FarleyNRC Region 2Westinghouse PWR 3-Loop

At 1209 CDT on 8/5/16, during testing of Unit 2 Auxiliary Building Pre-action Sprinkler Systems, all zones on the Unit 2 Pyrotronics Fire Detection Panel went into an alarm state and were unable to be reset. This condition is reportable per 10 CFR 50.72(b)(3)(xiii) as a Major Loss of Assessment capability. The NRC Resident has been notified. The licensee has initiated all necessary compensatory and corrective actions.

  • * * UPDATE FROM BLAKE MITCHELL TO VINCE KLCO AT 1718 EDT ON 8/6/2016 * * *

At 1600 CDT on 8/6/16, the Unit 2 Pyrotronics Fire Detection Panel was declared functional following repair of master override reset test switch and supply fuse. The Pyrotronics Fire Detection Panel was successfully tested following maintenance. The emergency assessment capability for the site's Emergency Plan has been fully restored. The NRC Resident has been notified Notified the R2DO (Suggs).

ENS 521472 August 2016 16:22:00VogtleNRC Region 2W-AP1000A member of the Fitness-For-Duty (FFD) staff violated the FFD policy in December, 2014 by responding to a call-in for unscheduled work after having consumed alcohol within the prohibited 5 hour period. The licensee informed the NRC Resident Inspector.
ENS 5211621 July 2016 18:28:00FarleyNRC Region 2Westinghouse PWR 3-LoopSouthern Nuclear Operating Company had a non-licensed supervisory contractor employee confirmed positive result for alcohol during a follow-up fitness-for-duty test. The contractor employee's unescorted access to the plant has been revoked. The NRC Senior Resident Inspector has been informed.
ENS 520611 July 2016 17:00:00FarleyNRC Region 2Westinghouse PWR 3-LoopOn 7/1/16, the Houston County Dispatch (911 Center) was conducting a monthly weather siren test. At approximately 1205 CDT, there was an inadvertent actuation of 41 emergency sirens in Houston County. The emergency sirens were deactivated at approximately 1208 CDT. Houston Country EMA issued a press release to notify the public of the inadvertent actuation. This is being reported under 10 CFR 50.72(b)(2)(xi) due to the inadvertent actuation and subsequent press release. There was no impact to the health and safety of the public as a result of this event as the offsite response capabilities remain functional. The site is operating normally with no emergency conditions present. The NRC Resident Inspector has been notified.
ENS 519867 June 2016 14:10:00VogtleNRC Region 2W-AP1000A non-licensed contract supervisor had a confirmed positive for a controlled substance during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector.
ENS 5195625 May 2016 04:54:00VogtleNRC Region 2Westinghouse PWR 4-LoopAt 0206 EDT 5/25/16, Vogtle Unit 2 tripped from 100% when SG (Steam Generator) #1 Level began to lower for an unknown reason. Cause for level issue is under investigation. All control rods fully inserted and AFW (Auxiliary Feedwater) and FWI (Feedwater Isolation) actuated as expected. Unit 2 is in Mode 3 and stable with decay heat being removed by Aux Feedwater. Prior to the trip, I & C (Instrumentation & Calibration) was performing a loop #1 narrow range instrument calibration. Unit 2 is in a normal post trip electrical lineup with all source of offsite power available. The licensee informed the NRC Resident Inspector.
ENS 5195023 May 2016 17:02:00HatchNRC Region 2GE-4On May 23, 2016, at 1009 EDT, while personnel were performing turbine testing with Unit 2 offline for planned maintenance, an event resulted in the actuation of containment isolation valves in more than one system. In response to this unexpected signal, 2B21F016 (Steam Line Drain Line Inboard Isolation Valve), 2B21F019 (Steam Line Drain Line Outboard Isolation Valve), and 2B31F019 (Reactor Water Sample Inboard Isolation Valve) went closed, all of which are primary containment isolation valves actuated by Group 1 Isolation. The Group 1 Isolation signal initiated based on low condenser vacuum during the turbine testing procedure, a valid condition that was expected to have been bypassed in the logic during the performance of this procedure. Human performance is believed to be the cause of these systems having actuated in a way that was not part of the planned evolution. Although the Unit was shut down when this signal was received, and primary containment isolation was not required to mitigate the consequences of an event, this Isolation signal has been determined to have been valid due to the initiation in response to actual plant conditions or parameters which satisfy the requirements for initiation of the system. The event is reportable as required by 10 CFR 50.72(b)(3)(iv)(A)(2): (A) Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam Isolation valves (MSIVs). The licensee has notified the NRC Resident Inspector.
ENS 5191811 May 2016 10:56:00FarleyNRC Region 2Westinghouse PWR 3-LoopAt 0653 (CDT) on 5/11/16, Farley Unit 2 reactor was manually tripped from 29 (percent) power. The initiating event was hi-hi Steam Generator level. Steam Generator levels began to rise following the start of a second condensate pump. The hi-hi steam generator level setpoint was reached causing the only running main feedwater pump to trip, a main feedwater isolation, and an automatic turbine trip. Auxiliary feedwater automatically started as expected. The reactor was manually tripped per procedure. All other systems responded properly for the event and there were no complications. The plant is currently stable in Mode 3. The NRC Resident Inspector has been notified.
ENS 5187521 April 2016 18:18:00VogtleNRC Region 2W-AP1000A non-licensed supervisory contractor admitted to attempting to subvert a Fitness for Duty test. The contractor's site access has been terminated. The NRC Resident Inspector was notified.
ENS 5174218 February 2016 15:48:00HatchNRC Region 2GE-4As part of the upgrade to the full structural weld overlays (FSWOL) per NRC-approved ISI (In-service Inspection) Alternative HNP-ISI-ALT-15-01, the surface of the existing weld overlay for the 1E Recirculation weld (1B31-1RC-12BR-E-5) was ground to prepare the surface for receipt of a new Alloy 52M overlay. Upon performance of the subsequent liquid penetrant testing examination, it was discovered that the as-found condition of the flaw did not meet acceptance criteria. It was determined that the flaw constituted a defect in the primary coolant system that could not be found acceptable per ASME Section XI. Therefore, this event notification is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A). Evaluation of the flaw found in the weld overlay suggests that the non-satisfactory liquid penetrant surface examination is a result of the propagation of the original flaw that was found on the 1E Recirculation Loop Piping. The flaw is axial in nature and therefore there is no impact on structure integrity degradation. No leakage is currently present or was seen during the previous operating cycle from this flaw. There is also reasonable assurance that there was not a breach in the credited RCS (Reactor Coolant System) boundary during the previous operating cycle. As part of the corrective action to fix the flaw, the 1E Recirculation weld will be upgraded to full structural weld overlay per the NRC-approved ISI Alternative. The Licensee notified the NRC Resident Inspector.
ENS 5134325 August 2015 16:01:00VogtleNRC Region 2Westinghouse PWR 4-LoopOn August 25, 2015, at approximately 12:52 EDT, an unplanned loss of power to the Vogtle Meteorological Towers occurred. The loss of power was the result of an electrical disturbance in the Plant Wilson switchyard. Power was restored 14:45 EDT. The electrical disturbance impacted meteorological instrumentation only and did not impact operation of either Unit 1 or Unit 2. The TSC and EOF remained functional and compensatory measures existed within emergency plan implementing procedures to obtain meteorological tower data locally and with the National Weather Service. Meteorological information could have been provided via the emergency notification system to the NRC Operations Center. This event is reportable per 10 CFR 50. 72(b)(3)(xiii) for 'any event that results in a major loss of emergency assessment capability.' The Emergency Response Organization was notified and necessary compensatory actions were established. The licensee notified the NRC Resident Inspector.
ENS 5125623 July 2015 15:52:00HatchNRC Region 2GE-4At 1120 EDT on 07/23/2015, while performing Chemistry rounds, a Chemistry Technician discovered an underground leak near the Plant Service Water bleach moat. The Plant Service Water bleach injection pumps were secured at 1130 EDT on 07/23/2015. The leak was later confirmed to be bleach. Approximately 200-300 gallons of 12.5% sodium hypochlorite are estimated to have leaked out of a broken underground pipe. This amount exceeded the reportable limit (100 lbs. or 80 gallons) as specified by the National Response Center. The bleach leaked into the surrounding soil and into a Plant Hatch storm drain where it pooled before reaching the Altamaha river. No bleach discharged into the river. Cleanup began immediately by neutralizing the bleach with sodium bisulfite and is still on-going. Plant Management authorized Southern Nuclear Company Environmental Affairs to report to the US Coast Guard National Response Center at 1305 EDT on 07/23/2015 in accordance with corporate procedures. The event was reported to the National Response Center at 1316 EDT. The Georgia Environmental Protection Division was also contacted. This event is reportable under 10 CFR 50.72(b)(2)(xi) as an event or situation related to the protection of the environment for which a notification to other government agencies have been made. NRC Resident Inspector was notified at 1530 at 07/23/2015. CR (Condition Report) 10099779
ENS 5122313 July 2015 07:50:00VogtleNRC Region 2Westinghouse PWR 4-Loop

On July 13, 2015, planned activities will be performed to upgrade the Vogtle Electric Generating Plant Meteorological Towers and its communication equipment. The emergency response data system will be affected. The work will be completed within approximately 5 days. This activity requires the meteorological tower to be out of service greater than 72 hours rendering equipment used for emergency assessment non-functional. The TSC and EOF remain functional during the outage. Compensatory measures exist within emergency plan implementing procedures to obtain meteorological tower data via Savannah River Site and with the National Weather Service. Meteorological information will be relayed via the emergency notification system to the NRC Operations Center. This event is reportable per 10 CFR 50.72(b)(3)(xiii) for 'any event that results in a major loss of emergency assessment capability.' The Emergency Response Organization has been notified of the upgrade and necessary compensatory actions. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM DANIEL MONAHON TO DONALD NORWOOD AT 2013 EDT ON 7/18/2015 * * *

Vogtle primary and backup Meteorological Towers have been returned to service and functionality of equipment has been restored. The NRC Resident Inspector will be notified. Notified R2DO (King).

ENS 512199 July 2015 18:05:00VogtleNRC Region 2Westinghouse PWR 4-LoopThe Vogtle Unit-1 control room received a fire alarm from inside the containment building. Per plant requirements, verification of the fire was not obtained within 15 minutes and an Unusual Event was declared. Plant personnel were able to enter containment and verify that there was no fire present. The reactor remained at 100% power throughout the event. Unusual Event entered: 1732 EDT Unusual Event exited: 1848 EDT The licensee notified the NRC Resident Inspector, State, and local governments. Notified DHS, FEMA, and NICC. NuclearSSA (email) and FEMA NWC (email).
ENS 5114811 June 2015 13:15:00FarleyNRC Region 2Westinghouse PWR 3-LoopA licensed employee supervisor had a confirmed positive test for a controlled substance during a random fitness-for-duty test. The employee's plant access has been placed on administrative hold. The NRC Resident Inspector has been notified.
ENS 511204 June 2015 12:56:00HatchNRC Region 2GE-4

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 5109126 May 2015 17:45:00HatchNRC Region 2GE-4
  • * * UPDATE ON 05/26/15 AT 1745 EDT FROM SCOTT BRITT TO JEFF HERRERA * * *

During an expanded scope inspection, deficiencies in the Control Building 147(foot) elevation were observed that caused the affected barrier to be considered nonfunctional: - Gaps were identified around cables in the foam cable tray penetration seal for penetration 1Z43H006F in the floor of the Cable Spreading Room (separating Fire Areas 0024A and 1104). The nonconforming condition observed for the affected fire barrier was identified as affecting both safe shutdown paths for Units 1 and 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10074859 The licensee will be notifying the NRC Resident Inspector.

  • * * UPDATE FROM STANLEY STONE TO JOHN SHOEMAKER AT 2017 EDT ON 6/1/15 * * *

During an expanded scope inspection for penetration seals, using more intrusive tools and methods, fire barriers in the Control Building El. 112 (foot) were found not to meet acceptance criteria. The fire protection engineering staff has examined the situations and recommends that these conditions be considered NON-FUNCTIONAL. - An issue was identified with the wall separating the el. 112 (foot) Control Building Working Floor, Fire Area (FA) 0001 from the Station Battery Room 1B, FA 1005. - An issue was identified with the wall separating the Station Battery Room 2A, Fire Area (FA) 2004 from the Station Battery Room 2B, FA 2005, on el. 112 (foot). The nonconforming conditions observed for the affected fire barriers were identified as affecting both safe shutdown paths for Units 1 and 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10077573 & 10077574. The licensee has notified the NRC Resident Inspector. Notified the R2DO(O'Donohue).

  • * * UPDATE FROM PAUL UNDERWOOD TO DONG PARK AT 1704 EDT ON 6/2/15 * * *

During an expanded scope inspection for penetration seals, using more intrusive tools and methods, fire barriers in the Control Building El. 112 (foot) were found not to meet acceptance criteria. The fire protection engineering staff has examined the situations and recommends that these conditions be considered NON-FUNCTIONAL: - Issues were identified with the wall separating the el. 112 (foot) Control Building Working Floor, Fire Area (FA) 0001 from the Unit 1 AC Inverter Room, FA 1008. The nonconforming conditions observed for the affected fire barriers were identified as affecting both safe shutdown paths for Unit 1. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10078011. The licensee has notified the NRC Resident Inspector. Notified the R2DO (Rose).

  • * * UPDATE FROM PAUL UNDERWOOD TO DONG PARK AT 1501 EDT ON 6/3/15 * * *

During an expanded scope inspection for penetration seals, using more intrusive tools and methods, fire barriers in the Control Building El. 112 (foot) were found not to meet acceptance criteria. The fire protection engineering staff has examined the situations and recommends that these conditions be considered NON-FUNCTIONAL: - Issues were identified with the wall separating the el. 112 (foot) Control Building U2 Water Analysis Room, Fire Area (FA) 2006 from the Control Building East Corridor, FA 0007 The nonconforming conditions observed for the affected fire barriers were identified as affecting both safe shutdown paths for Unit 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10078561. The licensee has notified the NRC Resident Inspector. Notified the R2DO (Rose).

ENS 5107014 May 2015 23:02:00HatchNRC Region 2GE-4

On May 14, 2015, it was determined that the number of 55 gallon drums of 2-Butoxyethanol analyzed to be transported and stored within the owner-controlled area (OCA) at any one time had been exceeded. It was discovered that contrary to the toxic gas analysis performed, fourteen 55-gallon drums of 2-Butoxyethanol were transported and stored in the OCA to support the construction of a dome being built to provide storage for FLEX equipment. The number of drums exceeded the limitation specified in the toxic gas analysis performed as part of the design project for transportation and storage could potentially impact Control Room habitability, emergency diesel generator air intake and have an adverse impact on security personnel. Upon determination that an unanalyzed condition existed, Operations placed the control room ventilation system in the 'isolation mode' until the number of drums on-site was reduced within the analyzed number. A substantial covering had been placed over the drums which also decreased the likelihood that any of the drums would fail and would also limit the potential dispersion of chemicals should a breach occur. The excess number of drums of 2-Butoxyethanol being transported and stored on-site is considered an unanalyzed condition that significantly degraded plant safety and is reportable in accordance with 10CFR50.72(b)(3)(ii)(B). The allowed number of 55 gallon drums of 2-butoxyethanol allowed per the current toxic gas analysis is 4 drums. The number of drums has been reduced to 2 as of 2024 EDT. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM PAUL UNDERWOOD TO VINCE KLCO ON 7/8/2015 AT 1657 EDT * * *

Further investigation into the chemicals transported on-site (F183M, of which 2-Butoxyethanol is a component) revealed that an on-site spill of all fourteen 55-gallon drums will not adversely affect Main Control Room Habitability, Security Personnel, or the Emergency Diesel Generators (EDGs). The Control Room Habitability Determination flowchart in Attachment 3 of NMP-CH-002-002 establishes a 10 mmHg vapor pressure threshold for determining if chemicals need to be evaluated for impact on the Main Control Room. Chemicals with a vapor pressure less than 10 mmHg do not need to be evaluated for control room habitability. The basis for the threshold is found in Reg Guide 1.78 Revision 0 section C.5.a, 'For chemicals that are not gases at 100F and normal atmospheric pressure but are liquids with vapor pressures in excess of 10 torr (10 mmHg), consideration should be given to the rate of flashing and boil off to determine the rate of release to the atmosphere and the appropriate time duration of the release.' The individual chemical component (including 2-Butoxyethanol) vapor pressures are less than 10 mmHg. As their vaporization rate is too low to adversely affect Control Room Habitability, it is also too low to create a hazard for Security personnel or to adversely affect the Emergency Diesel Generators. Based on this information, the transportation and storage of these chemical barrels did NOT represent a condition that significantly degraded plant safety. As such, this condition has been determined to no longer meet reporting requirement 10CFR50.72(b)(3)(ii)(B) and is therefore NOT reportable. Based on this information the previous notification for Event 51070 is being retracted. The licensee notified the NRC Resident Inspector Notified the R2DO (McCoy).

ENS 510477 May 2015 09:45:00FarleyNRC Region 2Westinghouse PWR 3-LoopThis notification is being made as required by 10 CFR 50.72(b)(2)(i) due to a Farley Nuclear Plant Unit 1 shutdown required by Technical Specifications. At 0509 CDT on 5/7/2015, 1B Reactor Coolant Pump (RCP) tripped during transfer of 1B 4160V bus to 1B unit auxiliary transformer. Technical Specification LCO 3.4.4 Condition A was entered for loss of a Reactor Coolant System (RCS) loop. Unit 1 reactor was shut down per operating procedures and entered Mode 3 at 0740 CDT. The NRC Resident Inspector has been notified.
ENS 510435 May 2015 13:05:00FarleyNRC Region 2Westinghouse PWR 3-LoopAt 0422 CDT on 5/5/2015, with Farley Nuclear Plant Unit 1 in Mode 2, and the 1A Steam Generator Feedwater Pump (SGFP) in the tripped condition, the 1B SGFP was manually tripped during troubleshooting. The trip of the second SGFP initiated the auto start signal for the MDAFWPs (motor driven aux feedwater pumps) due to the auto start signal not being defeated. Both MDAFW pumps were in service supplying AFW to the steam generators (SG) when the actuation signal was received. The effects of the auto start signal were to fully open the AFW Flow Control Valves and isolate SG blowdown and SG blowdown sample valves. These actions occurred successfully and the auto start signal was reset. There was no adverse impact to the plant and decay heat continued to be removed through the condenser throughout the event. The NRC Resident Inspector has been notified.
ENS 5100824 April 2015 00:15:00FarleyNRC Region 2Westinghouse PWR 3-LoopThis notification is being reported to NRC in accordance with 10 CFR 50.72(b)(2)(xi) for notification of an on-site fatality of a contract employee. In addition, the contracting company plans to notify the Occupational Safety and Health Administration (OSHA) of a fatality per 29 CFR 1904.39. At approximately 1717 CDT on 4/23/15, a 911 call was received in the Control Room regarding a contract employee who was found unresponsive and unattended in a temporary break room set up on the Turbine Deck during the Unit 1 refueling outage. Resuscitation by first responders and paramedics from a nearby town was unsuccessful. Resuscitation efforts were suspended at 1750. The Houston County Sheriff's Office was notified at approximately 1800 and they responded to the site at 1822. The county coroner was notified and arrived on site at 1850. (Farley Nuclear Plant) received notification at approximately 2035 that the contractor company intended to notify OSHA. A press release is not planned at this time. The NRC Resident Inspector has been notified. Unit 1 remains in Mode 6 and Unit 2 remains in Mode 1 at 100% power.
ENS 5099820 April 2015 21:26:00HatchNRC Region 2GE-4
  • * * UPDATE FROM STEVE BRUNSON TO CHARLES TEAL ON 4/20/15 AT 2126 EDT * * *

During an expanded scope inspection, deficiencies in the following areas were observed that caused the affected barriers to be considered nonfunctional: - A gap 1/4" wide, 1" tall, and 6" deep was found at penetration 1Z43H594D on the west wall of the U1 East Cableway (separating Fire Area 1104 and Fire Area 1020) - Near penetration 1Z43J837D, and approximately 12" south and above 1Z43H837D, gaps were observed in the mortar joint between CMU on the west wall of the U1 East Cableway (separating Fire Area 1104 and Fire Area 1020) - A triangular gap 1" wide, 1" tall and 6" deep was found at penetration 1Z43H592D on the west wall of the U1 East Cableway (separating Fire Area 1104 and Fire Area 1020) - A gap 4" tall and 3" wide was found behind Turn Box TB1-1272 which covers penetrations 1Z43H590D, 1Z43H589D, 1Z43H588, and 1Z43H587D on the west wall of the U1 East Cableway (separating Fire Area 1104 and Fire Area 1020) - At the architectural joint between the vertical wall to the horizontal floor/ceiling assembly above door 1C-22, above and to the south of 1Z43H1105D, a gap was observed approximately 1/4" tall, 3" wide, and 6" deep on the west wall of the U1 East Cableway Foyer (separating Fire Area 1105 and Fire Area 0014K) - Gap between the grout and the conduit of penetration 1Z43H778D approximately 1/4" tall x 1.5" wide x 6" deep on the east wall of the Unit 1 130' Elevation Control Building Working Floor Hallway (separating Fire Area 0014K and Fire Area 1105) The nonconforming conditions observed for the affected fire penetrations and fire barriers were identified as affecting both safe shutdown paths for Unit 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensure the safe shutdown paths are preserved until the degraded conditions are repaired. CR 10058276; CR 10058278 The following deficiencies were also observed causing the affected penetrations to be considered nonfunctional: - A gap 1/4" wide, 1" tall, and 6" deep was located at penetration 1Z43H532D on the west wall of the U1 East Cableway (separating Fire Area 1104 and Fire Area 0014M) - A gap 1/8" wide, 1" tall and 6" deep was located at penetration 1Z43H780D on the west wall of the U1 East Cableway (separating Fire Area 1104 and Fire 0014M) - A gap 1/2" wide, 1" tall, and 6" deep was located at penetration 1Z43H781D on the west wall of the U1 East Cableway (separating Fire Area 1104 and Fire 0014M). A 1/4" x 1/2" defect was also identified at penetration 1Z43H781D on the east wall of the Men's Restroom in the Control Building (separating Fire Area 0014M and Fire Area 1104) The nonconforming conditions observed for the affected penetrations were identified as affecting both safe shutdown paths for Unit 1 and Unit 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until degraded conditions are repaired. CR 10058277 The expanded scope inspection activity is continuing and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. The licensee has notified the NRC Resident Inspector. Notified R2DO (Blamey).

  • * * UPDATE FROM SCOTT BRITT TO DONG PARK ON 4/23/15 AT 1654 EDT * * *

During an expanded scope inspection, deficiencies in the following areas were observed that caused the affected barriers to be considered nonfunctional: - A gap 1/4" wide, 1" tall and 7" deep was found at penetration 1Z43H1138D on the east wall of the U1 RPS MG Set Room (separating Fire Area 1013 and Fire Area 0040). No seal material was seen between the sleeve and the cinderblock on the north side of penetration. - A void 1" tall, 1" wide, and 7" deep was found in the south upper corner under a concrete beam at column line T12 above penetration 1Z43H941D on the east wall of the U1 RPS MG Set Room (separating Fire Area 1013 and Fire Area 0040). - At penetration 1Z43H1139D, it appears that combustible neoprene insulation is used around the pipe within the plane of the west wall of the Vertical Cable Chase Room (separating Fire Area 0040 and Fire Area 1013). Combustible materials would not be part of a rated pen seal. - A gap 1" wide, 1" tall and 7" deep was observed at penetration 1Z43H1138D on the west wall of the Vertical Cable Chase Room (separating Fire Area 0040 and Fire Area 1013). The nonconforming conditions observed for the affected fire penetrations and fire barriers were identified as affecting both safe shutdown paths for Unit 1. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10060228 The licensee will notify the NRC Resident Inspector. Notified R2DO (Blamey).

  • * * UPDATE FROM STANLEY STONE TO DONG PARK ON 4/27/15 AT 2047 EDT * * *

During an expanded scope inspection, deficiencies in the following areas were observed that caused the affected barriers to be considered nonfunctional: -An opening in the grout 1/4" wide, 1/2" tall and over 7" deep was found between the wall and the outside sleeve for penetration 2Z43H028D on the west wall of the U2 Transformer Room (separating Fire Area 2019 and Fire Area 2016). -A 1/4" diameter hole in the grout approximately 2.5" deep was found above conduit 2MI2128 on the west wall of the U2 Transformer Room (separating Fire Area 2019 and Fire Area 2016). The nonconforming conditions observed for the affected penetration and fire barrier were identified as affecting both safe shutdown paths for Unit 1 and Unit 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10061830 The licensee notified the NRC Resident Inspector. Notified R2DO (Blamey).

  • * * UPDATE FROM PAUL UNDERWOOD TO DONG PARK ON 4/28/15 AT 1640 EDT * * *

During an expanded scope inspection, deficiencies in the following areas were observed that caused the affected barriers in the Unit 2 Control Building 130' elevation to be considered nonfunctional and represented degraded conditions of penetrations through the wall separating the Unit 2 Switchgear Access Hallway - Fire Area 2014, from the U2 West 600 V Switchgear Room - Fire Area 2016. The following conditions were located on the south wall of the Unit 2 Switchgear Access Hallway (Fire Area 2014). 1. An opening between the conduit and the wall 1/4" wide, 2" long and probed to be at least 2 1/2" deep was identified for penetration 2Z43H668D. A similar condition exists for this penetration on the opposite side of the wall in Fire Area 2016. 2. An opening between the conduit and the wall 1/4" wide, 1/2" long and probed to be at least 3" deep was identified for penetration 2Z43H667D. A similar condition exists for this penetration on the opposite side of the wall in Fire Area 2016. The following conditions were located on the opposite side of the same wall. This is the north wall of the U2 West 600V Switchgear Room (Fire Area 2016): 3. An opening between the conduit and the wall 1/8" wide, 1" long and probed to be at least 4" deep was identified for penetration 2Z43H668D. 4. An opening between the conduit and the wall 1/8" wide, 1/2" long and probed to be at least 3" deep was identified for penetration 2Z43H667D. 5. An opening between the conduit and the wall 2 1/2" wide, 2 1/2" long and probed to be at least 4" deep was identified around the 2" continuous run conduit located above cable tray penetration 2Z43H031D. 6. An opening between the conduit and the wall 1/4" wide, 3/4" long and probed to be at least 6" deep was identified above a 3/4" continuous run conduit (first of three) located at the ceiling near column line TE. 7. An opening between the conduit and the wall 1/4" wide, 1/2" long and probed to be at least 6" deep was identified above a 3/4" continuous run conduit (second of three) located at the ceiling near column line TE. 8. An opening between the conduit and the wall 1/4" wide, 3/4" long and probed to be at least 5" deep was identified above a 3/4" continuous run conduit (third of three) located at the ceiling near column line TE. The nonconforming conditions observed for the affected penetration and fire barrier were identified as affecting both safe shutdown paths for Unit 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10062254 The licensee notified the NRC Resident Inspector. Notified R2DO (Ehrhardt).

  • * * UPDATE FROM PAUL UNDERWOOD TO DANIEL MILLS ON 4/29/15 AT 1804 EDT * * *

During an expanded scope inspection, deficiencies in the Unit 2 Control Building 130 foot elevation were observed that caused the affected barriers to be considered nonfunctional and represented degraded conditions of penetrations through the wall separating the Unit 2 West DC Switchgear Room 2A (Fire Area 2018) and the Unit 2 Switchgear Access Hallway (Fire Area 2014). The following conditions were located on the west wall of the Unit 2 Switchgear Access Hallway (Fire Area 2014). 1. An opening between the conduit and the wall 1/4 inch wide, 1 inch long and probed to be greater than 2 inch deep, was identified for penetration 2Z43H673D. 2. There is insufficient masonry material to fill the full depth of the wall above the ductwork that passes through penetration 2Z43H032D. This deficiency affects a small area on the south side of the ductwork and penetrations 2Z43H789D, 2Z43H790D, and 2Z43H791D. 3. An opening between the conduit and the wall 1/4 inch wide, 1 inch long and probed to be 4 inch deep was identified for penetration 2Z43H671D. A similar condition exists for this penetration on the opposite side of the wall (see Item 5 below). The following conditions were located on the east wall of the Unit 2 West DC Switchgear Room 2A (Fire Area 2018). 4. There are openings between the conduits and the wall 1/2 inch wide and 1 inch long for penetrations 2Z43H789D, 2Z43H790D, and 2Z43H791D. These penetrations are affected in Item 2 above. 5. An opening between the conduit and the wall 1 inch wide, 1 inch long and probed to be greater than 6 inch deep, was identified for penetration 2Z43H671D. 6. An opening between the conduit and the wall 1/4 inch wide, 1 inch long and probed to be at least 2 inch deep, was identified for penetration 2Z43H673D. 7. An opening between the conduit and the wall 1/4 inch wide, 1 inch long and probed to be at least 2 1/2 inch deep, was identified for penetration 2Z43H676D. The nonconforming conditions observed for the affected penetrations were identified as affecting both safe shutdown paths for Unit 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR10062955 The licensee notified the NRC Resident Inspector. Notified R2DO (Ehrhardt).

  • * * UPDATE FROM JOHN MITCHELL TO HOWIE CROUCH AT 2137 EDT * * *

During an expanded scope inspection, deficiencies in the Control Building 130' elevation were observed that caused the affected barriers to be considered nonfunctional and represented degraded conditions of the following penetrations through the wall separating the Unit 2 East Cableway (Fire Area 2104) and the Health Physics Hallway and Counting Room (Fire Areas 0014B and 0014G). - Penetration 2Z43H783D terminates open within a foot of the east wall of the Health Physics Counting Room (Fire Area 0014G) - Penetration 2Z43H603D contains no visible seal material and is located on the east wall of the Health Physics Hallway (Fire Area 0014B). The nonconforming conditions observed for the affected penetrations were identified as affecting both safe shutdown paths for Unit 1 and Unit 2. Deficiencies were also observed that caused the affected barriers to be considered nonfunctional and represented degraded condition of the wall separating the Unit 2 East Cableway (Fire Area 2104) from the common East Cableway Foyer (Fire Area 1105). - Gap near penetration 2Z43H170D between a conduit and the concrete masonry unit (CMU) wall located on the south wall of the Unit 2 East Cableway (Fire Area 2104).

The nonconforming conditions observed for the affected fire barriers were identified as affecting both safe shutdown paths for Unit 1. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas but were modified based on the nature of the degradations noted in the condition report and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR10063642 Notified R2DO (Ehrhardt).

  • * * UPDATE FROM JOHN MITCHELL TO JOHN SHOEMAKER AT 1638 EDT ON 5/7/15 * * *

During an expanded scope inspection, deficiencies in the Control Building 147' elevation were observed that caused the affected barriers to be considered nonfunctional. These deficiencies represented degraded conditions of the following penetrations through the wall separating the Unit 1 CO2 Tank Room (Fire Area 0025) and the Computer Room (Fire Areas 0024B) as well as a discrepancy in the affected wall.

   - In Fire Area 0024B, a small gap in the foam, approximately 6 (inch) deep was identified in Penetration 1Z43H592F.  The adjacent Fire Area is FA 0025.
   - In Fire Area 0024B, penetration 1Z43H325F was identified with no sealant for the penetration sleeve.  The adjacent Fire Area is FA 0025.
   - In Fire Area 0024B, foam sealant was missing in cable-tray, 1Z43H061F. The adjacent Fire Area is FA 0025
   - In Fire Area 0024B, a gap was identified in a concrete masonry unit (CMU) wall joint, directly above 1Z43H062F.

The nonconforming conditions observed for the affected fire barriers were identified as affecting both safe shutdown paths for Unit 1 and Unit 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR10066678 The licensee has notified the NRC Resident Inspector. Notified the R2DO (Sandal).

  • * * UPDATE AT 2151 EDT ON 05/07/15 FROM SCOTT BRITT TO S. SANDIN * * *

During an expanded scope inspection, a deficiency in the Control Building 147 ft. elevation was observed that caused the affected barrier to be considered nonfunctional. This deficiency represented degraded conditions of the following fire barrier separating the Unit 1 CO2 Tank Room (Fire Area 0025) and the Cable Spreading Room (Fire Areas 0024A). - A 1/4 inch x 2 inch x approximately 4 inch deep gap in the east CMU wall of Unit 1 CO2 Tank Room above penetration 1Z43H046F. The nonconforming condition observed for the affected fire barrier was identified as affecting both safe shutdown paths for Unit 1 and Unit 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR10066844 The licensee will inform the NRC Resident Inspector. Notified R2DO (Sandal).

  • * * UPDATE AT 2029 EDT ON 05/08/15 FROM SCOTT A. BRITT TO S. SANDIN * * *

During an expanded scope inspection, deficiencies in the Control Building 147 ft. elevation were observed that caused the affected barrier to be considered nonfunctional. These deficiencies represent degraded conditions of the following fire barrier separating the Cable Spreading Room (FA 0024A) and the CO2 Tank Room (FA 0025). - Multiple gaps in the caulk at the top of the ceiling of the west wall of the Cable Spreading Room (separating FA 0024A and FA 0025). The nonconforming condition observed for the affected fire barrier was identified as affecting both safe shutdown paths for Unit 1 and Unit 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10067163 The licensee will inform the NRC Resident Inspector. Notified R2DO (Sandal).

  • * * UPDATE FROM PAUL UNDERWOOD TO DANIEL MILLS ON 5/11/15 AT 1711 EDT * * *

During an expanded scope inspection activity, multiple fire penetrations on the Control Building El. 130' elevation were identified that resulted in the affected barriers being considered NON-FUNCTIONAL. An issue was identified with the wall separating the Vertical Cable Chase, Fire Area 0040, from the Unit 2 RPS MG Set Room, Fire Area 2013. - A 1/4" wide x 1/2" long x approximately 6" deep gap in the grout of a 2" continuous run conduit, 6" away from 2Z43H581D was identified. - A 1/4" wide x 3" long x approximately 6" deep gap in the grout of penetration 2Z43H581D was identified. - A 1/2" wide x 2" long x approximately 6" deep gap in the grout of penetration 2Z43H580D was identified. The nonconforming condition observed for the affected fire barrier was identified as affecting both safe shutdown paths for Unit 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10068138 The NRC Resident Inspector has been notified. Notified R2DO (Bonser).

  • * * UPDATE FROM GUY GRIFFIS TO DANIEL MILLS ON 5/12/15 AT 2151 EDT * * *

During an expanded scope inspection activity, a fire barrier on the Control Building El. 164' elevation was identified as being NON-FUNCTIONAL as follows; - A discrepancy was identified with the fire barrier separating the Unit 1 Turbine Building Main Floor Area, Fire Area 0101A from the Main Control Room, Fire Area 0024C. The condition consists of a small gap 1/4" wide, 3" long and probed to be greater than 6" deep between the wall and conduit at penetration 1Z43H605J on the Turbine Building side of the wall. The nonconforming condition observed for the affected fire barrier was identified as affecting both safe shutdown paths for Units 1 and 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10068842 The NRC Resident Inspector has been notified. Notified R2DO (Bonser).

  • * * UPDATE FROM GUY GRIFFIS TO VINCE KLCO ON 5/14/15 AT 2121 EDT * * *

During an expanded scope inspection, deficiencies in the Control Building 164' elevation were observed that caused the affected barrier to be considered nonfunctional: - A 1/4 inch x 1/2 inch x approximately 6 inch deep gap in the grout of the annulus of penetration 1Z43H602J was identified in the east wall of the Main Control Room (separating Fire Areas 0024C and 0101A). - A 1/4 inch wide x 8 inch long vertical crack, approximately 6 inch deep was identified in the CMU below penetration 1Z43H602J was identified in the east wall of the Main Control Room (separating Fire Areas 0024C and 0101A). - Three abandoned anchor holes, 1/2 inch in diameter and approximately 4 inch deep, were identified below penetration 1Z43H604J in the east wall of the Main Control Room (separating Fire Areas 0024C and 0101A). - A 1 inch diameter abandoned anchor hole, approximately 6 inch deep, was identified directly above a 1/4 inch pipe penetration in the east wall of the Main Control Room (separating Fire Areas 0024C and 0101A). - A 1/2 inch to 3/4 inch gap exists between the top of each of the 3 concrete block (CMU) walls enclosing the HVAC chase and the underside of the floor/ceiling assembly separating the Main Control Room (Fire Area 0024C) and the HVAC Room Chase (Fire Area 0014L). - A 1/2 inch diameter hole exists in the CMU at the upper right corner of penetration 1Z43H1184J separating the Main Control Room (Fire Area 0024C) and the HVAC Room Chase (Fire Area 0014L). The nonconforming condition observed for the affected fire barrier was identified as affecting both safe shutdown paths for Units 1 and 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10069898; CR 10069995 The licensee will notify NRC Resident Inspector. Notified the R2DO (Bonser).

  • * * UPDATE FROM R.S. STONE TO VINCE KLCO ON 5/15/15 AT 1807 EDT * * *

During an expanded scope inspection for penetration seals, the following discrepancies were identified with the wall separating the Unit 1 Working Floor, Fire Area 0001, from the Unit 1 AC Inverter Room, Fire Area 1008 that caused the affected barriers to be considered nonfunctional:

A 1/4 inch x 1 inch x approximately 2 inch deep gap in the grout of the annulus of penetration 1Z43H553C.

A 1/4 inch x 1 inch x approximately 2 inch deep gap in the grout of the annulus of penetration 1Z43H546C.

A 1/8 inch wide x 1/2 inch tall x approximately 6 inch deep gap in the foam block out, below penetration 1Z43H546C.

A 3 inch x 3 inch x 10 inch deep gap in the grout around a 2-1/2 inch continuous run conduit.

A 1/4 inch x 1 inch x 10 inch deep gap in the grout around 1-1/2 inch continuous run and 1-1/4 inch continuous run conduits.

2 inch deep gaps in the grout around 1-1/2 inch and 2-1/2 inch continuous run conduits.

A 1/4 inch hole x 1 inch deep gap in the grout around penetration 1Z43H060C.

A 1/4 inch x 1/4 inch x 2 inch deep gap around the annulus of a 1-1/4 inch continuous run and 2 inch continuous run conduits. The nonconforming condition observed for the affected fire barrier was identified as affecting both safe shutdown paths for Unit 1. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10070439 The licensee notified the NRC Resident Inspector. Notified the R2DO (Bonser).

  • * * UPDATE ON 1638 EDT ON 05/21/15 FROM GUY S. GRIFFIS TO JEFF HERRERA * * *

During an expanded scope inspection, the following discrepancies were identified in the Unit 1 Control Building 130(foot) elevation that caused the affected fire barrier to be considered nonfunctional: - Six 3(inch) x 3(inch) holes in the wall of the Men's Rest Room (separating Fire Areas 0014M and 1104). The nonconforming conditions observed for the affected barrier were identified as affecting both safe shutdown paths for Unit 1 and Unit 2. Deficiencies were also observed in the Diesel Generator Building 130(foot) elevation that caused the affected fire barrier to be considered nonfunctional: - Through-wall gap around the conduit that passes through penetration 2Y43H511D on the south wall of the U2 Diesel Generator Switchgear Room 2F (separating Fire Areas 2408 and 2409). The nonconforming condition observed for the affected fire barrier was identified as affecting both safe shutdown paths for Unit 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR10073041; CR10073187 The licensee notified the NRC Resident Inspector. Notified the R2DO (Ernestes).

ENS 5089414 March 2015 19:19:00VogtleNRC Region 2Westinghouse PWR 4-LoopOn March 14, 2015 at 1207 EDT, operators were performing steps to disable an automatic actuation signal to the B-train Auxiliary Feedwater system (AFW), when a valid actuation signal was received on B-train only. Both motor driven Auxiliary Feedwater pumps were already running and feeding forward. The B-train discharge valves went from throttled to fully open. The AFW discharge valves were restored to their previous positions without any adverse impacts on the plant. Decay heat removal was still being removed through the Atmospheric Relief Valves. The licensee notified the NRC Resident Inspector.
ENS 5089314 March 2015 08:07:00VogtleNRC Region 2Westinghouse PWR 4-LoopVogtle Unit 2 was operating in Mode 1 at 100% rated thermal power. At 0429 EDT a Unit 2 automatic Reactor Trip and Safety Injection / Steamline Isolation occurred. All systems operated as expected and all control rods fully inserted. The Safety Injection was terminated at 0447 EDT and the emergency operating procedures were exited at 0522 EDT. Unit 2 is stable in Mode 3 with decay heat removal via the Auxiliary Feedwater system and the atmospheric relief valves. A response team is investigating the cause of the event. Unit 1 was unaffected by the event. NRC Senior Resident Inspector was notified and is at plant site for investigation.
ENS 508572 March 2015 18:31:00VogtleNRC Region 2W-AP1000On March 2, 2015, Southern Nuclear Operating Company (SNC) had a non-licensed supervisory contractor employee confirmed positive result for alcohol during a fitness-for-duty test. The contractor employee's unescorted access to the plant has been terminated. The NRC Senior Resident Inspector has been informed.
ENS 507989 February 2015 11:36:00VogtleNRC Region 2W-AP1000This is a 10 CFR 50.55 initial notification for a significant breakdown in the Quality Assurance (QA) Program of Chicago Bridge & Iron (CB&I), an agent for the Licensee of the Vogtle 3&4 Construction Project. In August 2014, deviations were found in sub-modules CA03-06, -08, and -09, which initiated the discovery and evaluation processes for both Part 21 and 10 CFR 50.55 . The conditions were determined to be not reportable by CB&I under 10 CFR 21, but an evaluation of the root cause analysis results concluded that a significant QA program breakdown had occurred that could have produced a defect. No defect has been identified. This initial notification is being made in accordance with 10 CFR 50.55 (4)(iii) and 10 CFR 50.55 (5)(i). The NRC Resident Inspector will be notified.
ENS 5067713 December 2014 10:00:00FarleyNRC Region 2Westinghouse PWR 3-LoopAt 0152 CST on December 13, 2014, both strings of the Unit 2 containment smoke detection system were declared non-functional due to a non-radioactive steam leak inside containment. The steam leak was causing spurious alarms to the smoke detection system. This condition prevents identification and assessment of a fire in containment. Required compensatory measures have been established. Since a fire in the containment building is an entry condition for the site's emergency plan, this is considered a loss of emergency assessment capability and is being reported per 10CFR50.72(b)(3)(xiii). The NRC Resident Inspector has been informed. The steam leak was caused by a faulty main steam flow detector. The licensee has shut down the reactor to effect repairs.
ENS 506629 December 2014 20:02:00HatchNRC Region 2GE-4

During an inspection of a fire penetration between Fire Area 1404, Switchgear Room 1G and Fire Area 1408, Switchgear Room 1F in the diesel generator building, the penetration was determined to be non-functional as a 3 hour fire barrier. In the event of a postulated fire in the affected areas, both safe shutdown paths on Unit 1 could be compromised. Given this information, the determination was made that this condition meets the reporting criteria of 10 CFR 50.72(b)(3)(ii)(B). Compensatory measures were established in accordance with the plant's Fire Hazard Analysis (FHA). The presence of the compensatory measures in addition to automatic fire detection in the fire areas ensure that the safe shutdown paths are preserved until the degraded condition can be repaired. (Condition report No.) CR904013 The licensee notified the NRC Resident Inspector

  • * * RETRACTION PROVIDED BY KENNY HUNTER TO JEFF ROTTON AT 0915 EST ON 12/30/2014 * * *

Further investigation revealed that after removing the outermost two inches of loose silicone foam material, and taking additional measurements there remained adequate silicone foam material to provide 10 inches of silicone foam sealing the penetration. Plant design shows that the wall in question is 18 inches thick and also that 9 inches of silicone foam is required in the penetration in order maintain a 3 hour fire rating for the wall/penetration. Since there is adequate foam in the penetration to maintain the 3 hour fire rating the penetration is fully functional. Based on this information, this penetration in its 'as found' state does NOT represent a condition that seriously degrades a principal safety barrier. As such this condition has been determined to no longer meet reporting requirement 10CFR50.72(b)(3) and is therefore NOT reportable. Based on this information the previous notification for Event 50662 is being retracted. The licensee notified the NRC Resident Inspector. Notified R2DO (Bartley)

ENS 5062919 November 2014 21:44:00HatchNRC Region 2GE-4During a fire inspection activity involving inspection of fire penetrations that serve as Appendix R barriers, degradation of a fire penetration was identified that was sufficient to prevent this penetration from meeting Appendix R requirements as a 3 hour fire barrier. In the event of a postulated fire in the affected areas both safe shutdown paths on Unit 1 and 2 could be compromised. Given this information the determination was made that this condition meets the reporting criteria of 10CFR50.72(b)(3)(ii)(B). Compensatory measures were put in place in accordance with the plant's Fire Hazard Analysis (FHA) for the degraded penetration in the affected 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 corrected. The degraded fire penetration is located between the turbine deck and the main control room. The licensee has notified the NRC Resident Inspector.
ENS 5061815 November 2014 08:14:00FarleyNRC Region 2Westinghouse PWR 3-LoopThis notification is being made as required by 10CFR 50.72(b)(3)(iv)(A) due to control room operators opening the Farley Nuclear Plant Unit 2 reactor trip breakers via the main control board hand switch during reactor startup procedures. This is a valid actuation of the reactor protection system. The reactor was not critical at the time the reactor trip breakers were opened. While withdrawing control rods for reactor startup and low power physics testing with control bank C at approximately 50 steps, Digital Rod Position indication for one of the control rods (M12) changed to 90 steps. The control room operators stopped withdrawing rods and entered FNP-2-AOP-19 (Malfunction of the Rod Control System). The reactor trip breakers were opened at 0353 CST. All rods inserted as expected. All other systems functioned normally. The plant is stable at normal operating pressure and temperature. The NRC Resident Inspector has been notified.