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 Start dateReport dateSiteReporting criterionSystemEvent description
ENS 5589613 May 2022 16:11:00Monticello10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionSteam Jet Air EjectorThe following information was provided by the licensee via email: On 5/13/22 at 1111 CDT the station entered LCO 3.7.4 Condition B for Control Room Envelope being inoperable. This was due to results from an inspection in the Steam Jet Air Ejector room that identified steam leakage exceeding the leakage rate assumptions made in the Alternate Source Term (AST) dose analysis calculation. Therefore, this is being reported in accordance with 10CFR50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades plant safety and 10CFR50.72(b)(3)(v)(D) for any event or condition that at the time of discovery could have prevented fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. There is no impact to the health and safety of the public. NRC Resident has been notified.
ENS 5557614 November 2021 16:50:00Millstone10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionSteam Generator
Main Steam Isolation Valve
Auxiliary Feedwater
Main Steam Line
On November 14, 2021, at 1150 EST, while operating in Mode 1 at 100 percent power, the supply check valve from the Number 2 steam generator to the turbine driven auxiliary feedwater pump was determined during troubleshooting that it is not able to perform its isolation function. This failure would have resulted in the blowdown of both steam generators during a main steam line break in the Number 2 steam generator main steam line upstream of the main steam isolation valves until the operators could isolate the faulted steam generator. Previous evaluation has determined that this condition constituted an unanalyzed condition that could impact containment pressure. There has been no radioactive release to the environment. The steam lines from the steam generators to the turbine driven auxiliary feedwater pump have been isolated by use of a motor operated valve in the discharge line of the Number 2 steam generator. There has been no impact to Unit 3 which remains at 100 percent power. The NRC Senior Resident has been notified. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B) as a condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.
ENS 555656 November 2021 15:00:00Millstone10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionSteam Generator
Feedwater
Main Steam Isolation Valve
Auxiliary Feedwater
Main Steam Line
During a Unit 2 refueling outage valve overhaul activity on the steam supply check valve from the number 2 steam generator to the turbine driven auxiliary feedwater pump, 2-MS-4B, the check valve was found with its disc separated from the disc arm. This failure would have resulted in the blowdown of both steam generators during a main steam line break in the steam generator number 2 main steam line upstream of the main steam isolation valves until the operators could isolate the faulted steam generator. On November 6, at approximately 1100 EDT evaluation determined that this condition constituted an unanalyzed condition that could impact containment pressure. There has been no radioactive release to the environment. The valve has been repaired. The check valve in the steam supply from the number 1 steam generator to the turbine driven auxiliary feedwater pump was inspected and found to be satisfactory. There has been no impact to Unit 3 which remains at 100% power. The Senior Resident has been notified. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B) as a condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.
ENS 5545913 September 2021 22:22:00Surry10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionEmergency Diesel GeneratorOn September 13, 2021, at 1822 EDT, an apparent non-compliance with 10 CFR 50, Appendix R, section III.G.2 (separation of redundant fire safe shutdown equipment) was identified. Specifically, it was determined that some Emergency Diesel Generator (EDG) cables may be susceptible to a hot short/spurious operation to the close circuit. A spurious closure of the emergency bus normal supply breakers after the EDG is powering the bus could result in non-synchronous paralleling, EDG overloading, or EDG output breaker tripping due to faulted power cable from normal supply breaker. The spurious closure of the normal supply breakers is not currently addressed in the Appendix R Report or previous Multiple Spurious Operations (MSO) analysis. This condition is associated with the Appendix R safe-shutdown function of the Emergency Power System. The Emergency Power System is considered operable but not fully qualified for its safety-related design function. The following fire areas are impacted: 1) Fire Area 13, Unit 1 Normal Switchgear Room 2) Fire Area 46, Unit 1 Cable Tray Room 3) Fire Area 3, Unit 1 Emergency Switchgear and Relay Room 4) Fire Area 2, Unit 2 Cable Vault and Tunnel Until this condition is analyzed, Surry has implemented mitigating actions in the above fire areas. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B). This is also reportable as a 60-day written report pursuant to 10 CFR 50.73(a)(2)(ii)(B). This event was entered into the licensee's Corrective Action Program as CR (condition report) 1180502. The NRC Resident Inspector has been notified of this event. Mitigating actions include posting fire watches in the affected areas.
ENS 5542724 August 2021 18:06:00FitzPatrick10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionDuring an extent of condition review of DC control circuits, it was identified there are additional unprotected DC control circuits which are routed between separate Appendix R fire areas. A postulated fire in one area can cause a short circuit and potentially result in secondary fires or cable fires in other fire areas where the cables are routed. The secondary fires or cable failures degrade the degree of separation for redundant safe shutdown trains and are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B). Compensatory actions for affected fire areas have been implemented. Design modifications in the affected control circuits are being developed and will be scheduled to correct this condition.
ENS 5537522 July 2021 21:51:00North Anna10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionEmergency Diesel GeneratorOn July 20, 2021, at 1707 EDT, an apparent non-compliance with 10 CFR 50, Appendix R, section III.G.2 (separation of redundant fire safe shutdown equipment) was identified. This issue was initially categorized as not affecting train separation or the ability of the equipment to perform their Design Basis functions. The original concern was entered into the licensee's Corrective Action Program as CR1177199. Subsequently, on July 22, 2021, at 1751 EDT, a further review of the affected control circuits for the Unit 1 and Unit 2 Emergency Diesel Generator (EDG) output breakers and emergency bus feeder breakers identified a concern that breaker position interlocks routed to or through non-safety related components or spaces may affect the ability to provide emergency power on the affected unit due to impacts on the control power circuits during an Appendix R fire associated with a loss of offsite power. The following are the affected fire areas: - Unit 1 and Unit 2 Turbine Buildings - Unit 1 and Unit 2 Cable Spreading Rooms - Unit 1 and Unit 2 Normal (307) Switchgear Rooms This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B). This is also reportable as a 60-day written report pursuant to 10 CFR 50.73(a)(2)(ii)(B). This event was entered into the licensee's Corrective Action Program as CR 1177399. The NRC Resident Inspector has been notified of this event.
ENS 5532122 June 2021 16:08:00Davis Besse10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionAuxiliary FeedwaterAt 1208 (EDT) on 6/22/2021, the high-energy line break door separating Auxiliary Feedwater Train Rooms 1 and 2 was not able to be latched following normal usage. The door was able to be closed, protecting Train 1 equipment from a break in Room 2. However, it is assumed a break in Room 1 would push the unlatched door open and allow high-energy fluids to enter Room 2. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. The door was able to be latched at 1215 (EDT) on 6/22/2021 following repairs to the door latch interlocking mechanism. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). No other equipment was inoperable during this event. The NRC Resident Inspector has been notified.
ENS 552313 May 2021 13:30:00Fermi10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionAt 0930 EDT on 5/3/2021, it was determined that during entries into the Fermi 2 Reactor Building Steam Tunnel (RBST) on 4/17/2021, 4/18/2021, and 4/21/2021 that the door was not controlled according to site procedures. The RBST door is credited as a hazard barrier for various high-energy line break (HELB) scenarios. On the identified dates, the RBST door was left open for brief periods during maintenance related activities in the RBST. This condition is not bounded by existing analyses as the door is assumed to be closed throughout a HELB event. The time period that the door was open was less than one hour in each case, as stay times in the room are inherently limited by industrial and radiological conditions. Individuals remained in the area to close the door if needed, but existing analyses do not address the ability to perform those actions under all HELB scenarios. There is no impact to the health and safety of the public or plant personnel as the door is currently closed and latched and access into the area has been restricted to normal ingress and egress per site procedures, which ensures consistency with existing analyses. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). Investigation into the cause is ongoing. Preliminary review of the extent of this condition identified entries into the RBST on other occasions during the past three years where the conditions may also have not been bounded by existing analyses. The additional occasions where the door may have been held open were on 9/22/2018 (MODE 3), 10/26/2018 (MODE 1 ), 11/2/2018 (MODE 1), and 3/21/2020 (MODE 3). Each of these instances was also less than one hour with the exception of the occurrence beginning on 10/26/2018 which lasted approximately 10 hours to support packing leak repairs on a HPCI (High Pressure Coolant Injection) Outboard Isolation Valve. The licensee notified the NRC Resident Inspector.
ENS 5480630 July 2020 13:15:00Callaway10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition

EN Revision Imported Date : 8/18/2020 EMERGENCY PROCEDURE ERROR POTENTIALLY PREVENTING TIMELY COMPLETION OF EMERGENCY CORE COOLING SYSTEM RECIRCULATION ALIGNMENT At 0815 CDT on 7/30/2020, it was determined that a procedural error in emergency procedure ES1.3, Transfer to Cold Leg Recirculation, could delay realignment from emergency core cooling system (ECCS) injection phase to recirculation phase under lower plant operational modes. It is noted this scenario is postulated to occur only when the boron dilution mitigation system is operable in lower modes of operation as per Technical Specification 3.3.9 (required operable in Mode 2 (below P-6), 3, 4 and 5). Current plant conditions require this feature nonfunctional so this issue does not impact current plant conditions. This condition is not bounded by existing design and licensing documents; however, it poses no current impact to the health and safety of the public or plant personnel. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 8/17/2020 AT 1603 EDT FROM JOSH COPELAND TO KERBY SCALES * * *

Event Notification (EN) 54806, made on 7/30/2020, is being retracted because re-evaluation performed subsequent to the notification has demonstrated that the error in Emergency Operating Procedure ES1.3 would not have resulted in a condition outside of the current licensing basis analyses of record for the Callaway Plant. This re-evaluation addressed core effects, containment pressure-temperature and radiological consequences analyses, documented in the plant's corrective action program. The re-evaluation has led to the conclusion that the procedural error in ES1.3 would not have prevented any system required to be OPERABLE by the Technical Specifications from performing its specified safety functions. With all systems capable of performing their specified safety functions, the current licensing basis analyses of record for Callaway Plant remain valid and bounding. Based on these considerations, it has been determined that the condition reported in EN 54806 did not result in the plant being in an unanalyzed condition that significantly degraded plant safety. Consequently the condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades safety. The NRC Resident Inspector has been notified of this Event Notification retraction. Notified R4DO (Taylor)

ENS 5453320 February 2020 17:40:00FitzPatrick10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition(On February 20, 2020, at 1240 EST, the Licensee determined the following information:) This notification is in reference to reports EN 54130 and LER 2019-002, which were retracted. James A. FitzPatrick Nuclear Power Plant received additional information on the technical basis for the retraction. Further review, including testing of the terminal blocks, demonstrated that the short circuit current would result in heat levels in excess of cable insulation ratings. Unprotected DC control circuits for non-safety related DC motors are routed between separate fire areas. A postulated fire in one area can cause a short circuit and potentially result in secondary fires or cable fires in other fire areas where the cables are routed. The secondary fires or cable failures degrade the degree of separation for redundant safe shutdown trains and are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B). Compensatory actions per the Technical Requirements Manual (TRM) for affected fire areas have been implemented. A modification to install fuses in the control circuits for 94P-2(M), 31P-7A(M), 31P-7B(M), and 94P-13(M) has been scheduled and shall correct this condition. The NRC Resident Inspector has been notified.
ENS 5448722 January 2020 03:18:00Sequoyah10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition

EN Revision Imported Date : 2/21/2020 CONTAINMENT RELIEF VALVES INOPERABLE At 22:18 (EST) on 1/21/20, it was discovered that all Unit 1 containment vacuum relief isolation valves were closed and all vacuum relief lines were simultaneously inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The isolation valves were opened and the vacuum relief valves were restored to operable. There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 02/20/2020 AT 1626 EST FROM FRANK SCHULTE TO BRIAN P. SMITH * * *

At 1549 (EST), February 20, 2020, a completed engineering evaluation of the condition initially reported on January 22, 2020 determined that the inoperability of the Sequoyah Unit 1 Containment Vacuum Relief System affected the ability to protect containment against an external pressure event. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. The condition was resolved when isolation valves were opened on January 21, 2020 and the vacuum relief lines were restored to an operable status. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B), "an unanalyzed condition that significantly degrades plant safety. Subsequent to the initial notification, continued evaluation of the reported condition has concluded that the isolation of the containment vacuum relief function did not prevent the fulfillment of a safety function that is needed to control the release of radioactive material; nor mitigate the consequences of an accident therefore this event is not reportable under 10 CFR 50.72(b)(3)(v), "Event or Condition that could have prevented fulfillment of a safety function. The NRC Resident has been notified. Notified R2DO (Musser)

ENS 5443711 December 2019 18:56:00Surry10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionService water
Auxiliary Feedwater
On December 11, 2019, at 1356 EST, it was concluded that certain safety-related equipment is vulnerable to design basis tornado missiles which could render the equipment inoperable and not able to perform its design function. This applies to the following Technical Specification equipment: 1. Component cooling water piping for the 'A' spent fuel cooling water system heat exchanger. This heat exchanger is vulnerable to a horizontal missile traveling through the roll-up door, which would challenge operability of the Technical Specification required component cooling system equipment. 2. All three (3) emergency service water pumps and their diesel fuel oil supply tank. The emergency service water pumps and diesel fuel oil tank are vulnerable to a horizontal missile penetrating the missile screens. 3. Certain component cooling water system pump discharge piping is vulnerable from a vertical missile penetrating the auxiliary building roof. 4. The Unit 1 auxiliary feedwater (AFW) system pumps and the pump suction and discharge piping are vulnerable to a missile traveling through the screens on the sides and roof of the main steam valve house. This vulnerability also exists for the Unit 2 AFW. This condition puts Unit 1 and 2 into Technical Specification 3.01 which requires the units to be in hot shutdown within 6 hours and in cold shutdown within the following 30 hours. The NRC Resident Inspector has been notified.
ENS 544255 December 2019 14:10:00Cooper10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
The following was received via email from Cooper Nuclear Station: At 0810 (CST), on 12/5/19, Operations personnel discovered BLDG-DOOR-R209, FIRE DOOR BETWEEN CRITICAL SWITCHGEAR ROOMS F & G, was unlatched. The door was immediately latched upon discovery. Based on door logs, the door separating the two critical switchgear rooms was inadvertently left unlatched for approximately 5 minutes. This door is a Steam Exclusion Boundary (SEB) door. It is required to be closed and latched when the Auxiliary Steam Boiler is in service due to Auxiliary Steam piping passing through Critical Switchgear Room 'G'. If a steam line break was to occur with the door unlatched, steam could render both Critical Switchgear busses inoperable. This is being reported under 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition, and 10 CFR 50.72(b)(3)(v), 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 (B) remove residual heat and to (D) mitigate consequences of an accident. There was no impact on the health and safety of the public or plant personnel. The door closes automatically and appeared to have been left unlatched by the last person passing through. The door was tested and latches as required. The licensee notified the NRC Resident Inspector.
ENS 5441730 November 2019 19:00:00Diablo Canyon10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
On November 30, 2019, at 1100 PST, with Unit 2 in Mode 4, Operations identified that both trains of containment spray had been removed from service earlier at approximately 0217 hours as part of preparations for a planned Mode 5 entry. The containment spray pumps are required to be operable (along with the containment fan cooler units) in Modes 1 through 4 in accordance with Technical Specification 3.6.6. With both containment spray pumps inoperable, TS 3.6.6 Action F requires the Unit to be shut down in accordance with TS 3.0.3. At 1125 hours, both trains of containment spray were returned to operable and the required actions of TS 3.6.6 and TS 3.0.3 were exited. The five containment fan cooler units remained operable for the duration of the occurrence. This notification is being made in accordance with the requirement of 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented fulfillment of a safety function, and 10 CFR 50.72(b)(3)(ii) as an event or condition that may have resulted in the plant being in an unanalyzed condition. The NRC Senior Resident Inspector has been notified.
ENS 543662 November 2019 19:15:00Beaver Valley10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionAt 1515 on November 2, 2019, the Refueling Water Storage Tank (RWST) was declared inoperable due to a Low Head Safety Injection relief valve discharging to the Safeguards Sump during routine surveillance testing. The leakage from the Low Head Safety Injection system in conjunction with a postulated Design Basis Accident (DBA) Loss of Coolant Accident (LOCA) with transfer to Safety Injection Recirculation may result in dose exceeding the Dose Analysis of the Exclusion Area Boundary (EAB) and the Control Room, which is common to both Unit 1 and Unit 2. This condition may not be bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. The Low Head Safety Injection relief valve has been isolated to prevent further leakage, and makeup to the RWST completed. At 1602 on November 2, 2019 the RWST was declared Operable. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(B), (C), (D) as an Unanalyzed Condition and a condition that could have prevented the Fulfillment of a Safety Function." The licensee notified the NRC resident inspector.
ENS 5426611 September 2019 00:34:00Browns Ferry10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionRemote shutdownA lightning strike occurred at approximately 1502 CDT on 09/10/2019, and a resulting power surge damaged some of the security door card reader system equipment. However, this did not affect access to plant areas for personnel who were already within protected area. At 1830 on 09/10/2019, it was discovered that some of the oncoming night shift personnel could not access particular areas that required the use of security card readers. Extent of condition check at 1934 on 09/10/2019 determined that access to 1A and 3A Electric Board Rooms, which contain remote shutdown panels and Fire Safe Shutdown equipment. was prohibited for the night shift personnel. This condition is reportable under 10 CFR 50.72(b)(3)(ii)(B) - Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. Access was restored to all plant areas at 2106 on 9/10/2019. No plant events occurred during the time frame that the 1A & 3A Electric Board Rooms inaccessible that would have required access to these areas. The NRC Resident Inspector has been notified.
ENS 542576 September 2019 02:15:00South Texas10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
Reactor Coolant System

EN Revision Text: CONTAINMENT PENETRATION DISCOVERED NOT ISOLATED At 2115 CDT on 9/5/2019, an inside containment test connection and inoperable outside containment isolation valve were discovered to be open for a containment air sample penetration. This resulted in the containment penetration not being isolated. The inside containment test connection was closed at 2322 CDT on 9/5/2019.

This event is being reported under 10 CFR 50.72(b)(3)(v)(C) and (D) and 10 CFR 50.72(b)(3)(ii)(B).

There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM PAUL BURTON TO HOWIE CROUCH AT 1342 EST ON 11/7/19 * * *

This event was originally reported on September 6, 2019 under 10 CFR 50.72(b)(3)(v)(C) and (D) and 10 CFR 50.72(b)(3)(ii)(B). Upon completion of the investigation of the event, it was determined that the event had insignificant safety consequences because the containment breach was disconnected from the Reactor Coolant System by a series of closed valves for the duration of the event. Additionally, the lines to the inside containment connection and the outside inoperable containment isolation valve that was found to be open as well as the main line connecting and passing through the penetration were one-inch diameter lines. Analysis determined that containment breaches that are less than a three-inch diameter do not lead to a large radiation release. The event did not place the plant in an unanalyzed condition that significantly degrades plant safety. Therefore, 10 CFR 50.72(b)(3)(ii)(B) did not apply to this event and this notification is to retract reporting under that criterion. The licensee notified the NRC Resident Inspector. Notified R4DO (Drake).

ENS 5413024 June 2019 22:15:00FitzPatrick10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition

EN Revision Text: POTENTIAL UNANALYZED CONDITION DUE TO UNPROTECTED CONTROL CIRCUITS RUNNING THROUGH MUTILPLE FIRE AREAS During a review of industry Operating Experience it was identified that there were unprotected DC control circuits for non safety-related DC motors which are routed from the Battery Charger Rooms to other separate fire areas. Circuit Breakers used to protect the motor power conductors appear to be inadequate to protect the control conductors. The concern is that under fire safe shutdown conditions, it is postulated that a fire in one area can cause short circuits potentially resulting in secondary fires or cable fires in other fire areas where the cables are routed. The secondary fires or cable failures are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis. This condition is reportable as an 8-hour ENS report in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Requirements of the Technical Requirements Manual (TRM) for the affected fire areas will be implemented." The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION FROM ROBERT GRAHAM TO HOWIE CROUCH AT 2045 EDT ON 9/30/19 * * *

In accordance with NUREG-1022, Sections 2.8 and 5.1.2, James A. FitzPatrick Nuclear Power Plant is retracting (formally withdrawing) Licensee Event Report (LER) Number 2019-002. LER 2019-002 was transmitted to the NRC via letter JAFP-19-0080 dated August 23, 2019. The LER reported, under 10 CFR 50.73(a)(2)(ii)(B), the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety. Subsequent to submittal of LER 2019-002, FitzPatrick Engineering completed analyses using more accurate input conditions. This analysis has determined no credible hot short scenario will result in damage to adjacent cables in other fire zones, showing that the postulated condition would not degrade plant safety. Therefore, James A. FitzPatrick Nuclear Power Plant is retracting LER 2019-002 (and this event notification). The licensee will notify the NRC Resident Inspector and the New York State Public Service Commission. Notified R1DO (DeFrancisco).

ENS 5411111 June 2019 16:32:00Monticello10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
Primary containmentAt 1132 CDT on 6/11/2019, both manual primary containment isolation valves in a one-inch service air line were found open. This resulted in an open primary containment penetration. Both valves are required to be closed for Primary Containment Isolation Valve Operability. Both valves were closed and independently verified closed at 1149 CDT on 6/11/2019. This is being reported under 10 CFR 50.72(b)(3)(v)(C) and (D), and 10 CFR 50.72(b)(3)(ii)(B). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The licensee also notified the State of Minnesota State Duty Officer.
ENS 5399712 April 2019 23:15:00Monticello10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
Core Spray

EN Revision Text: HIGH ENERGY LINE BREAK DOOR FOUND IN INCORRECT POSITION RESULTING IN LPCI AND CORE SPRAY BEING INOPERABLE At approximately 1815 CDT on April 12, 2019, High Energy Line Break (HELB) Door-410A in the Reactor Building was discovered in the closed position. HELB Door-410B was previously closed for maintenance. Either Door-410A or Door-410B must be open to support the current HELB analyses. With both doors closed, this is considered an unanalyzed condition resulting in the loss of a post-HELB safe shutdown path. With Door-410A and Door-410B closed, LPCI (Low Pressure Coolant Injection) and Core Spray injection valves in both divisions are no longer considered available. This condition is being reported under 10 CFR 50.72(b)(3)(ii) as an unanalyzed condition that significantly degrades plant safety and 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented the fulfillment of a safety function. The condition was resolved at approximately 1845 CDT on April 12, 2019 when Door-410A was blocked open. The health and safety of the public was not affected by this condition. The NRC Resident has been notified.

  • * * RETRACTION FROM JESSE TYGUM TO HOWIE CROUCH AT 1330 EDT ON 5/24/19 * * *

Event Notification (EN) #53997, made on 4/13/2019, is being retracted. An engineering evaluation completed subsequent to this event analyzed the discovered condition with both Door-410A and Door-410B being closed. The engineering evaluation determined that the environmental conditions present with both Door-410A and Door-410B closed would not have impacted the availability of both divisions of the LPCI (Low Pressure Coolant Injection) and Core Spray injection valves nor would it have resulted in the loss of a post-HELB safe shutdown path. Therefore, this condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii) as an unanalyzed condition that significantly degrades plant safety or per 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified. The licensee also notified the Minnesota State Duty Officer. Notified R3DO (Cameron).

ENS 539681 April 2019 03:06:00Palo Verde10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material

At 2006 (MST), on 3/31/2019, the Palo Verde Nuclear Generating Station Unit 1 Shift Manager was informed that leakage was measured from the Train A Emergency Core Cooling System (ECCS) piping at approximately 100 ml/minute through a High Pressure Safety Injection (HPSI) A drain valve. This value exceeds the assumed 3000 ml/hour ECCS leakage for a large break loss of coolant accident analysis. At 0230 (MST) on April 1, 2019, the valve was flushed and the leakage reduced to 10 ml/minute (600 ml/hour) and was no longer above the limit of the safety analysis. This condition is being reported as an unanalyzed condition per 10 CFR 50.72(b)3)(ii)(B) and a condition that could have prevented the fulfillment of a safety function to the control the release of radioactive material per 10 CFR 50.72(b)(3)(v)(C). This event did not result in an abnormal release of radioactive material. Notification received by Caty Nolan and emailed to HOO.HOC@NRC.GOV The NRC asked a followup question: Why was the criterion for Control of Radioactive Material selected? per the PVNGS Unit 1 Shift Manager, this criterion was selected due to the potential of exceeding offsite dose projections, post recirculation, following a Design Basis Accident. The resident inspector has been notified.

  • * * UPDATE ON 05/15/19 AT 1417 EDT FROM SEAN DORNSEIF TO BETHANY CECERE * * *

An engineering evaluation concluded that the as-found ECCS leakage would not have degraded the performance of the Pump Room Exhaust Air Cleanup system; therefore, it remained operable. The evaluation also concluded that the as-found leakage was within the analysis margins for HPSI pump hydraulic performance and containment flood level following a Large Break Loss of Coolant Accident; therefore, the ECCS also remained operable. Based on the above information, the condition identified on March 31, 2019, was an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B), but did not prevent the fulfillment of the safety function of the structures or systems that are needed to control the release of radioactive material per 10 CFR 50.72(b)(3)(v)(C). The NRC resident inspectors have been informed. Notified R4DO (Proulx).

ENS 5380120 December 2018 05:00:00Watts Bar10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionAt 1642 Eastern Standard Time (EST) on December 20, 2018, it was determined that both trains of Containment Air Return Fan (CARF) were simultaneously INOPERABLE from 0817 (EST) to 1129 (EST) on November 20, 2018. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.
ENS 537121 November 2018 04:00:00Fermi10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition

EN Revision Text: UNANALYZED CONDITION DUE TO MODIFICATION NOT ADDED TO PROCEDURE On November 1, 2018, at approximately 1300 EDT, Fermi 2 identified that a Station Blackout (SBO) procedure was deficient as a result of a modification installed during a recent refueling outage. A review identified that the performance of the SBO procedure could have resulted in a challenge to having an alternate AC source available within one hour as outlined in the Updated Final Safety Analysis Report (UFSAR) 8.4.2. The alternate AC source was always available to be manually aligned in accordance with other standard operating procedures. The modification did not affect the function for Appendix R alternative shutdown. Immediate actions are underway to revise the impacted procedure. The health and safety of the public was not affected as offsite power has remained available since the modification was installed. Investigation into the cause and corrective actions is ongoing. Fermi 2 is reporting this event as an unanalyzed condition pursuant to the requirements of 10 CFR 50.72(b)(3)(ii)(B). The licensee notified the NRC Resident Inspector.

  • * * RETRACTION ON 12/28/18 AT 1228 EST FROM JEFFREY MYERS TO JEFFREY WHITED * * *

The purpose of this notification is to retract a previous report made on November 1, 2018 (EN 53712) under 10 CFR 50.72(b)(3)(ii)(B). Subsequent to the initial notification, the event, site procedures, and the NRC guidance in NUREG-1022 pertaining to 10 CFR 50.72(b)(3)(ii)(B) were reviewed further. The evaluation determined that at the time of the event, there were multiple methods defined in existing station procedures to establish an available alternate AC source within one hour as outlined in the Updated Final Safety Analysis Report (UFSAR) 8.4.2. Under these circumstances, the event does not represent an unanalyzed condition under 10 CFR 50.72(b)(3)(ii)(B). Therefore, EN 53712 can be retracted and no Licensee Event Report (LER) under 10 CFR 50.73(a)(2)(ii)(B) is required to be submitted. The licensee has notified the NRC Resident Inspector. Notified R3DO (Riemer).

ENS 5367419 October 2018 04:00:00Fermi10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionOn 10/19/2018, at approximately 0400 EDT, during an investigation into a failed surveillance test for a Loss of Offsite Power (LOP) coincident with a Loss of Coolant Accident (LOCA), it was identified that the Engineered Safety System Bus degraded voltage relay scheme contained a time delay setting that could inhibit all Low Pressure Core Injection (LPCI) pumps from automatically starting and operating during a LOP/LOCA, thus making LPCI incapable of meeting its functional requirement of automatic startup and operation regardless of the availability of offsite power supply (UFSAR Section 6.3.1.4 and Tech. Spec. Surveillance Requirement 3.8.1.17). The condition was identified during the first-time performance of a revised surveillance procedure for a LOP coincident with a LOCA signal. Fermi is currently in Mode 4 (Cold Shutdown) and LPCI auto start on a LOP/LOCA signal is not required. However, the initial investigation identified the condition likely existed in the past during modes of operation where LPCI auto start on LOP/LOCA was required. Investigation into the cause and corrective actions is ongoing. Since LPCI auto start is not required at the time of discovery (Mode 4), this event is being reported pursuant to 50.72(b)(3)(ii)(b). The NRC Resident Inspector has been notified.
ENS 5367519 October 2018 04:00:00Perry10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition

During extent of condition review of a previously identified fire induced hot-short (Ref. EN#53644) an unfused circuit associated with the 0M23C0002A, Miscellaneous Switchgear Recirculation Fan was discovered. This condition is not bounded by existing design and licensing documents. This results in an unanalyzed condition due to the possibility for a postulated fire induced hot-short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in the Appendix R Evaluation, Safe Shutdown Capabilities Report, due to an unfused circuit. Without overcurrent protection for this circuit, the potential exists that an initial fire event affecting this circuit could cause a short circuit that would cause excessive current through the circuit beyond the capacity rating of the conductors. This could lead to a secondary fire in another plant area where this circuit is routed, challenging the ability to achieve and maintain safe shutdown. The postulated event would affect multiple fire zones in the control complex. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE ON 10/19/2018 AT 1454 EDT FROM EDWARD CONDO TO ANDREW WAUGH * * *

Further extent of condition reviews have discovered another unfused circuit. The circuitry is related to 0M24C001A, Battery Room Exhaust Fan. This condition is not bounded by existing design and licensing documents. This results in an unanalyzed condition due to the possibility for a postulated fire induced hot-short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in the Appendix R Evaluation, Safe Shutdown Capabilities Report, due to an unfused circuit. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Senior Resident Inspector. Notified R3DO (Hills).

ENS 5366311 October 2018 04:00:00Calvert Cliffs10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
During a post maintenance start of the 1B diesel generator, the air start solenoid valves did not close as expected. This resulted in lowering air pressure in the common air start headers causing inoperability of the 2A and 2B diesel generators at time 23:03. The 1B diesel generator was isolated from the common air start header, which restored the air start header pressure to the 2A and 2B diesel generators. The 2A and 2B diesel generators were declared operable at 23:34. The NRC Resident Inspector was notified.
ENS 536444 October 2018 04:00:00Perry10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionDegraded or unanalyzed condition due to the possibility for a postulated fire induced hot short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in calculation SSC-001 due to an unfused circuit associated with the 1M43C0001A, Diesel Generator Building Ventilation Fan. This condition is not bounded by existing design and licensing documents. Without overcurrent protection for this circuit, the potential exists that an initial fire event affecting this circuit could cause a short circuit without protection that would cause excessive current through the circuit beyond the capacity rating of the conductors. This could lead to a secondary fire in another plant area where this circuit is routed challenging the ability to achieve and maintain safe shutdown. The postulated event would affect the following fire zones: 1CC-3c (Unit 1, Division 1 4160V and 480V Switchgear Room, 620 feet 6 inch elevation), 1CC-3e (Unit 1 West Corridor North of Elevator, 620 feet 6 inch elevation), DG-1d (Hallway Diesel Generator Building 620 feet 6 inch elevation), and 1DG-1c (Unit 1, Division 1 Diesel Generator Building 620 feet 6 inch elevation). This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Senior Resident Inspector.
ENS 5362225 September 2018 05:00:00River Bend10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
At 1200 CDT on September 25, 2018, while the plant was in MODE 1 at 90 percent power, it was identified that an additional condition existed which had not previously been considered in developing the compensatory measures implemented for design flaws and single point vulnerabilities associated with the Control Building Chilled Water System. Specifically, a 20 minute 'quick restart timer' on Control Building Chillers that have analog control systems (HVK-CHL1A & 1B) would prevent the chillers from starting in specific scenarios. The recommended compensatory actions to address the new condition were implemented at 1235 CDT on September 25, 2018. Currently the Chilled Water System is otherwise operating as designed. Operator actions are in place to ensure the plant meets all required design safety system functions. Work is currently underway to identify and correct all design vulnerabilities. The (NRC) Senior Resident Inspector has been notified. This was identified by engineering during an extended condition search.
ENS 5359711 September 2018 05:00:00Monticello10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
Core SprayOn 9/10/2018, the 11 Core Spray (CSP) loop was placed in service to support quarterly surveillance testing. With the 11 CSP pump in service it was identified that the check valves isolating the 11 CSP system from the keep fill supply were leaking by. At 1129 CDT on 9/11/2018, it was identified that this leakage may have exceeded the leakage rate assumptions made in the dose analysis calculation for emergency core cooling system (ECCS) leakage outside containment following a loss of coolant accident (LOCA). Therefore, this is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades plant safety and 10 CFR 50.72(b)(3)(v) 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: (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. The potential ECCS leak pathway has been isolated. There is no impact to health and safety of the public. The NRC Resident Inspector has been notified.
ENS 5354812 August 2018 04:00:00Beaver Valley10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown
Emergency Diesel Generator

EN Revision Text: TECHNICAL SPECIFICATION REQUIRED SHUTDOWN - LOSS OF 480 VOLTAGE EMERGENCY BUS On 8-12-18 at 0158 EDT, Beaver Valley Unit 2 experienced a loss of 480 Volt 2P Emergency Bus. This resulted in a Loss of Safety Function due to the 2-2 Emergency Diesel Generator (EDG) being Inoperable coincident with the Residual Heat Release Valve (2SVS-HCV104). A Technical Specification shutdown is required per LCO 3.0.3. The Licensee also stated they were in an unanalyzed condition due to the EDG and Residual Heat Release Valve being inoperable at the same time. The Licensee is shutting down to Mode 5 (Cold Shutdown). The Licensee is notifying the Resident Inspector. The Licensee will be making a Press Release about the unplanned shutdown.

  • * * UPDATE ON 08/16/2018 AT 1424 EDT FROM BLASE BARTKO TO KEN MOTT * * *

On 8-12-18 at 0158 (EDT) Beaver Valley Unit 2 experienced a loss of 480 Volt 2P Emergency Bus. Per operational guidance, this was determined to be a Loss of Safety Function due to the Unit 2 Emergency Diesel Generator (EDG) being INOPERABLE coincident with the Residual Heat Release Valve (2SVS-HCV104) 10 CFR 50.72(b)(3)(v)(B) and (D). This was also reported as an Unanalyzed Condition 10 CFR 50.72(b)(3)(ii)(b). No Press Release was performed for this event. The NRC Resident Inspector was notified. At 0410 (EDT) a Technical Specification Shutdown was commenced 10 CFR 50.72(b)(2)(i). At 2011 (EDT) the 480 Volt 2P Emergency Bus was restored and energized. Further evaluation of the event has determined that this event was not an Unanalyzed Condition and did not result in a Loss of Safety Function. The classifications of Unanalyzed Condition and Loss of Safety Function are being retracted. The accuracy of the existing guidance relative to Safety Function has been entered in the Corrective Action Program and interim actions have been taken to provide accurate guidance. Notified R1DO (Young) via email.

ENS 534853 July 2018 05:00:00Callaway10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionService water
Auxiliary Feedwater

EN Revision Text: DISCOVERY OF AN UNANALYZED CONDITION THAT SIGNIFICANTLY DEGRADES PLANT SAFETY On July 3, 2018, while performing a review of Emergency Operating Procedures, a concern was identified regarding the potential for excessive loss of ultimate heat sink inventory (UHS) through the auxiliary feedwater (AFW) system mini-flow recirculation pathway. This condition would have the potential to prevent the ultimate heat sink from providing an adequate inventory of water for a 30-day mission time.

The normal water supply for the Callaway AFW system is the condensate storage tank (CST). The CST is a non-safety grade component. The safety-grade supply for AFW is the essential service water (ESW) system. The ESW system is supplied by the UHS. The UHS thermal performance analysis accounts for a loss of UHS inventory to the AFW system up until the point of the accident sequence that the AFW pumps would be secured. The analysis did not include an allowance for loss of UHS inventory through the AFW mini-flow recirculation pathway following the AFW pumps being secured. The EOP guidance that secures the AFW pumps does not isolate the mini-flow recirculation pathway.

Initial estimates indicate that loss of UHS inventory through the mini-flow recirculation pathway, if not isolated, would preclude the UHS from completing its 30-day mission time. This potential for depletion of the UHS placed the plant in an unanalyzed condition that significantly degraded safety.

Callaway has issued interim guidance to the on-shift personnel regarding this concern to ensure that the ultimate heat sink water level is maintained at a level that will be adequate to mitigate the potential loss of inventory.

This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades safety. The NRC Resident Inspectors have been notified of this condition.

  • * * RETRACTION ON 07/31/2018 AT 1430 EDT FROM LEE YOUNG TO ANDREW WAUGH * * *

Event Notification (EN) 53485, made on July 3, 2018, is being retracted because re-evaluation performed subsequent to the notification has demonstrated, based on actual plant equipment and environmental conditions, that the unanalyzed inventory losses previously reported by EN 53485 would not have depleted the UHS inventory to an unacceptable level during its 30-day mission time. The re-evaluation has led to the conclusion that the previously unanalyzed losses of UHS inventory would not have prevented the UHS from performing its specified safety functions and meeting its 30-day mission time requirements. With the UHS capable of performing its specified safety functions and meeting its 30-day mission time requirements, the systems supported by the UHS would have remained capable of performing their specified safety functions. Based on these considerations, it has been determined that the condition reported in EN 53485 did not result in the plant being in an unanalyzed condition that significantly degraded safety. Consequently, the condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades safety. The NRC Resident Inspector has been notified of the Event Notification retraction. Notified R4DO (Gaddy).

ENS 534381 June 2018 04:00:00Salem10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
Service water
Auxiliary Feedwater

During the period of evaluation of tornado missile vulnerabilities and the potential impacts to technical specification (TS) plant equipment, it was determined that the power cables to a safety related motor control center (MCC) in the service water (SW) intake structure are not adequately protected from tornado generated missiles. During walk downs, it was identified that the installed SW pipe tunnel barrier is not adequate. A tornado could generate missiles capable of striking the power cables and rendering a SW MCC inoperable. These conditions are reportable per 10 CFR 50.72(b)(3)(ii)(B) and per 10 CFR 50.72(b)(3)(v)(D). This condition is being addressed in accordance with NRC enforcement guidance provided in Enforcement Guidance Memorandum (EGM) 15-002, 'Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' Revision 1, and DSS-ISG-2016-01, 'Clarification of Licensee Actions in Receipt of Enforcement Discretion per Enforcement Guidance Memorandum EGM 15-002, Enforcement Discretion for Tornado- Generated Missile Protection Noncompliance,' Revision 1. Compensatory measures have been implemented in accordance with these documents. The NRC Resident Inspector has been notified. The licensee will be notifying the Lower Alloways Creek Township.

  • * * UPDATE ON 6/18/2018 AT 1604 EDT FROM JUSTIN HARGRAVE TO RICHARD SMITH * * *

During subsequent walk downs, PSEG (Public Service Enterprise Group) identified that both the Unit 1 and Unit 2 turbine driven auxiliary feedwater pumps are also not adequately protected from tornado generated missiles. The steam exhaust pipe could be potentially impacted and cause crimping that could reduce steam exhaust flow and pump capacity. EN 53438 is updated to include both Salem units and these additional components. This condition is being addressed in accordance with NRC enforcement guidance provided in enforcement Guidance Memorandum (EGM) 15-002, 'Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' Revision 1, and DSS-ISG-2016-01, 'Clarification of Licensee Actions in Receipt of Enforcement Discretion per Enforcement Guidance Memorandum EGM 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' Revision 1. Compensatory measures have been implemented in accordance with these documents." The NRC Resident Inspector has been notified. The licensee will be notifying the Lower Alloways Creek Township. Notified R1DO (Burritt).

ENS 5341922 May 2018 04:00:00Beaver Valley10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown

EN Revision Text: GAS VOIDS DISCOVERED IN BOTH TRAINS OF LOW HEAD SAFETY INJECTION On 5/22/2018, while operating at approximately 100 percent power, Ultrasonic Testing of the Beaver Valley Power Station (BVPS) Unit 1 Low Head Safety Injection (LHSI) pump suction piping identified gas voids in excess of the acceptable limit for void volume. Both trains of LHSI were declared inoperable. Technical Specification (TS) 3.5.2 for both trains of the LHSI system was entered along with TS 3.0.3 which requires the initiation of a plant shutdown. Time of TS entry was 12:56 (EDT). Plant shutdown was commenced at 15:56 (EDT) in accordance with plant procedures. At 15:59 (EDT) Train 'A' LHSI was restored to operable status, TS 3.0.3 Action was exited and the power reduction was stopped at approximately 99 percent. At 17:43 (EDT) Train 'B' LHSI was restored to operable status, TS 3.5.2 Actions were exited. This is reportable per 10 CFR 50.72(b)(3)(ii) Unanalyzed Condition, 10 CFR 50.72(b)(3)(v) Event or Condition that Could Have Prevented the Fulfillment of a Safety Function and 10 CFR 50. 72(b )(2)(i) TS Required Shutdown. The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 6/21/18 AT 1535 EDT FROM SHAWN KEENER TO RICHARD SMITH * * *

Further engineering evaluation has determined that the gas voids that existed at the time of discovery would not have rendered the LHSI (Low Head Safety Injection) system inoperable if it were required to actuate. The engineering evaluation concluded that filling of the containment sump during a Design Basis Accident would result in a void volume reduction such that the void in the LHSI suction piping would not be large enough to significantly impact the operability of the system. Therefore, the system remained operable but degraded. No TSs (Technical Specifications) were required to be entered and no shutdown was required. As such, all three reporting criteria do not apply and are being retracted. The NRC Resident Inspector has been notified. Notified R1DO (Burritt).

ENS 5340114 May 2018 17:05:00Calvert Cliffs10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
On May 14, 2018, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of a tornado generated missile, Calvert Cliffs identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from a tornado generated missile. A tornado could generate a missile that could strike the Unit 1 Saltwater system header and associated piping. This could result in damage to the unit 1 Saltwater system header which could affect the ability of the Unit 1 Saltwater subsystems to perform their design function if such a tornado would occur. 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 NRC enforcement guidance provided in EGM 15-002 and DSS-ISG-2016-01. Compensatory measures have been implemented in accordance with these documents. The NRC Resident Inspector has been informed of this notification.
ENS 533928 May 2018 04:00:00Farley10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionReactor Coolant System
Control Room Emergency Filtration System
On 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 533824 May 2018 16:29:00River Bend10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
During performance of an extent of condition evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, River Bend Station identified non-conforming conditions in the plant design such that specific TS equipment is considered to not be adequately protected from tornado missiles. The reportable condition is postulated by tornado missiles entering the Diesel Generator Building through conduit and pipe penetrations. A tornado could generate multiple missiles capable of striking Division 1, Division 2, and Division 3 Diesel Generator support equipment rendering all Safety Related Diesel Generators inoperable. 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) 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 (A) Shut down the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) Mitigate the consequences of an accident. This condition was identified as part of an on-going extent of condition review of potential tornado missile related site impacts. Enforcement discretion per Enforcement Guidance Memorandum EGM 15-002 has been implemented and required actions taken. Corrective actions will be documented in a follow-on licensee event report. The licensee has notified the NRC Resident Inspector.
ENS 5334919 April 2018 23:44:00Watts Bar10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionResidual Heat RemovalOn April 19, 2018 at 1944 EDT, Watts Bar Nuclear Plant (WBN) determined that a preliminary analysis shows current acceptance criteria for gas accumulation in the WBN Unit 1 and Unit 2 Safety Injection System (SIS) and Residual Heat Removal System (RHRS) discharge piping may be non-conservative. The surveillances that check void values and allow venting of the systems are to be performed utilizing conservative criteria at more frequent intervals to ensure gas void volumes remain under acceptable limits. Additional analysis is being performed to determine final actions. The NRC Resident Inspector has been notified.
ENS 5332411 April 2018 06:50:00River Bend10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
At time 0150 CDT on April 11, 2018, a condition was identified that could impair the ability of the Control Building Air Conditioning System to perform its design function. Engineering determined that the time delay relays HVKA11-80YB or HVKA11-80YD (Division II chilled water LOW FLOW relays) could fail in a manner that challenges the design safety function of the Control Building Chilled Water System during a Loss of Offsite Power (LOP) Event. A failure of the time delay relay HVKA11-80YB or HVKA11-80YD (Division II chilled water LOW FLOW relays) to provide the time delay function would cause both the Division I and Division II HVK chilled water pumps to start after a LOP, which in turn could hinder the auto start of either Division I or Division II chillers. Currently the Chilled Water System is otherwise operating as designed. All operator actions are in place to ensure the plant meets all required designed safety system functions. Work is currently underway to correct this design vulnerability. The NRC Resident Inspector has been notified of this condition.
ENS 5330029 March 2018 18:44:00Browns Ferry10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionEmergency Equipment Cooling WaterAt 1344 on March 29, 2018, it was determined (engineering evaluation) that an unanalyzed condition that significantly degraded plant safety previously existed. During a postulated control room abandonment due to a fire, and concurrent with a Loss of Offsite Power (LOOP), the required number of Emergency Equipment Cooling Water (EECW) pumps would not have been available from 10/28/2015 to 3/10/2018. On March 8, 2018, during relay functional testing it was discovered that the C3 Emergency Equipment Cooling Water (EECW) pump closing springs did not recharge with the breaker transfer switch in emergency. On August 23, 2012, a wire modification was performed that contained a drawing error resulting in wire placement on the incorrect connection points for the C3 EECW pump. On March 10, 2018, the C3 EECW pump breaker wiring was corrected and subsequent testing was completed satisfactorily. Prior to 10/28/2015, Brown's Ferry Nuclear Plant (BFN) adhered to Appendix R fire protection requirements which did not credit the C3 EECW pump for fire protection from the backup control location. On 10/28/2015, BFN transitioned to National Fire Protection Association (NFPA) 805 fire protection requirements which takes credit for the C3 EECW pump from the backup control location. This condition is being reported 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'. This is also reportable as a 60-day written report in accordance with 10 CFR 50.73(a)(2)(ii)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified of this event.
ENS 5328725 March 2018 20:16:00Pilgrim10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded
Primary containment

On March 25, 2018 at 1616 hours (EDT), with the reactor in cold shutdown condition, two control rod drive piping lines were determined to be potentially inoperable in the event of a design basis earthquake due to support defects. The control rod drive piping forms a portion of the reactor coolant pressure boundary and primary containment boundary. The supports will be repaired prior to plant startup. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. The licensee will notify the Commonwealth of Massachusetts.

  • * * RETRACTION FROM JOE FRATTASIO TO HOWIE CROUCH AT 1500 EDT ON 4/13/18 * * *

The purpose of the notification is to retract ENS notification 53287 made on 03/25/18 for Pilgrim Nuclear Power Station. The previous notification reported that control rod drive (CRD) piping could be potentially inoperable in the event of a design basis earthquake, at the time of discovery, due to piping support defects. Subsequent evaluation has demonstrated that the piping was not inoperable. Specifically, after an engineering evaluation, it has been determined that the CRD Hydraulic System operability was never lost and the system was operable, although non-conforming, based on the support configuration not conforming to the pipe support drawings. The affected pipe supports have been restored or reworked to the proper design condition in accordance with the design drawings. The CRD System has subsequently been restored to a fully operable status. Notified R1DO (Jackson) and IRD MOC (Pham).

ENS 5326716 March 2018 05:00:00Browns Ferry10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionResidual Heat Removal
Residual Heat Removal Service Water

At 1604 (CDT) on March 16, 2018, Browns Ferry Nuclear Plant (BFN) Engineering reported an unanalyzed condition affecting the Residual Heat Removal (RHR) heat exchangers in a postulated fire event. It was discovered that the Residual Heat Removal Service Water (RHRSW) heat exchanger piping associated (with) the credited heat exchangers in the NFPA 805 Nuclear Safety Capability Analysis (NSCA) could experience water hammer damage. Fire damage to the cables for the RHRSW outlet motor operated valves could cause the valves to spuriously open and drain the RHRSW piping. Subsequent starting of the RHRSW pumps on the affected header could cause water hammer loads and damage the piping. Review of NFPA 805 analyses show the cables associated with this condition are routed in Fire Areas 01-03, 02-03, 02-04, 03-03, 16 and 23. There are 11 cases where the deterministically credited heat exchanger could be affected. Compensatory fire watch measures have been established. This event requires an 8 hour report in accordance with 10CFR50.72(b)(3)(ii)(B), 'Any event or condition that results in: (B) The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. CR 1139620 documents this condition in the Corrective Action Program. The NRC Resident Inspector has been notified.

  • * * RETRACTION AT 2215 EST ON 11/29/2018 FROM NEEL SHUKLA TO MARK ABRAMOVITZ * * *

NRC notification 53267 was made to ensure that the eight-hour non-emergency reporting requirements of 10 CFR 50.72 were met when the licensee discovered an unanalyzed condition with the potential to significantly degrade plant safety. On August 22, 2018, an independent analysis was completed which determined that the RHRSW system would remain functional during the postulated scenario. Based on this analysis, a revised functional evaluation was performed by BFN which determined that the condition did not constitute an unanalyzed condition that significantly degraded plant safety. For credited RHR heat exchangers for fire events in Fire Areas 01-03, 02-03, 02-04, 03-03, 16, and 23, the RHRSW piping will remain intact and the valves will operate manually after a water hammer event. This condition did not significantly degrade plant safety and is therefore not reportable under 10 CFR 50.72(a)(2)(ii)(B). On November 16, 2018, TVA canceled the 60 day report which had been submitted for this condition. TVA's evaluation of this event notification is documented in the corrective action program. The licensee has notified the NRC Resident Inspector. Notified the R2DO (Shaeffer).

ENS 5326515 March 2018 19:24:00Peach Bottom10 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionFeedwater
Reactor Core Isolation Cooling
Reactor Water Cleanup
At 1524 (EDT) on Thursday, March 15, 2018, Operations was notified of a failure to meet Appendix R requirements for Peach Bottom Atomic Power Station (PBAPS) Unit 2 and Unit 3. Valves associated with the feedwater system for both units were not properly considered as Hi-Lo Pressure interface valves as required by the Appendix R program. This results in the susceptibility to a hot short condition that could open valves, diverting flow from the reactor, damage piping and prevent injection. U3 (Unit 3) Fire Safe Shutdown Credited Reactor Core Isolation Cooling (RCIC) System is affected. U2 (Unit 2) is affected by a potential leak path through the Reactor Water Cleanup system. This event is being reported as an occurrence of an event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety under 10 CFR 50.72(b)(3)(ii). The Station (PBAPS) is performing hourly fire watches for the impacted areas and is also evaluating this condition for corrective action. The licensee notified the NRC Resident Inspector.
ENS 532391 March 2018 22:43:00Point Beach10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
Auxiliary FeedwaterDuring review of protection of equipment from damaging effects of tornados, Point Beach Nuclear Plant identified a potential vulnerability for the turbine driven auxiliary feedwater pumps due to steam supply piping that is not routed through a Class 1 structure. Immediate compensatory measures were taken to mitigate the potential consequences of a tornado generated missile impact. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) and per 10 CFR 50.72(b)(3)(v)(A) and (D). The identified vulnerability is being addressed in accordance with EGM 15-002 and DSS-ISG-2016-01, enforcement discretion memorandum and interim guidance document for resolution of noncompliance with tornado-generated missile protection. The NRC Resident Inspector has been notified.
ENS 532351 March 2018 18:10:00Quad Cities10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
Emergency Diesel Generator

EN Revision Text: UNANALYZED CONDITION FOR TORNADO GENERATED MISSILES On March 1, 2018, during evaluation of protection for Technical Specifications (TS) equipment from the damaging effects of tornado generated missiles, Quad Cities Station identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from tornado generated missiles. Tornado generated missiles could strike the Unit l, Unit 1/2, and Unit 2 Emergency Diesel Generator intake and exhaust stacks, day tank vents, and main fuel oil tank vents. This could result in crimping of the intake/exhaust stacks and vents, which would affect the ability of the Emergency Diesel Generators to perform their function if a tornado would occur. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B) as a condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and 10 CFR 50.72(b)(3)(v)(D) as a 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 NRC enforcement guidance provided in Enforcement Guidance Memorandum (EGM) 15-002, 'Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' Revision 1, and DSS-ISG-2016-01, 'Clarification of Licensee Actions in Receipt of Enforcement Discretion per Enforcement Guidance Memorandum EGM 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' Revision 1. Compensatory measures have been implemented in accordance with these documents. The NRC Resident Inspector has been informed of this notification.

  • * * RETRACTION FROM NICK JOHNSON TO HOWIE CROUCH AT 1900 EST ON 11/7/19 * * *

The purpose of this notification today (11/07/19) is to retract event notification #53235 made on March 1, 2018, for Quad Cities Station. Additional review determined that the current design of all three Emergency Diesel Generators and associated Day Tanks and Main Fuel Oil Storage Tanks Vents is consistent with the licensing basis for Quad Cities Station. There was no non-conformance of Quad Cities' tornado missile protection design, and the EDGs were operable at the time the event notification was made. Therefore, this event does not meet the criteria of 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(D) and ENS report #53235 is being retracted. The NRC Resident Inspector has been informed of this notification. Notified R3DO (Peterson).

ENS 5322320 February 2018 18:25:00Callaway10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
Service water
Auxiliary Feedwater

At 1225 CST, all three Auxiliary Feedwater (AFW) pumps were declared inoperable at the Callaway Plant upon discovery that a door (DSK13311) credited for protection of equipment from the effects of a high-energy line break (HELB) hazard had come partially open due to vibration harmonics from the running turbine-driven Auxiliary Feedwater pump (TDAFP). Immediate investigation identified that play in the mechanism that holds the door closed had rendered it susceptible to movement from the vibration harmonics. The affected HELB door specifically protects safety-related instruments that provide a swap-over signal upon detection of a low suction pressure condition for the AFW pumps and thereby automatically effect a suction transfer for the AFW pumps from the condensate storage tank (normal/standby source) to the Essential Service Water (ESW) system (credited safety-related source). All of the AFW pump suction transfer instrument channels were declared inoperable. Per Technical Specifications (TS) 3.3.2, 'Engineered Safety Feature Actuation System (ESFAS) Instrumentation,' the applicable Condition(s) and Required Action(s) for inoperable AFW pump suction transfer instrumentation only addresses a single channel being inoperable. Thus, the condition of having all three instrument channels inoperable required entry into TS Limiting Condition for Operation (LCO) 3.0.3. At the same time, however, with the automatic suction transfer capability rendered inoperable, all three AFW pumps, i.e., the TDAFP and the 'A' and 'B' motor-driven AFW pumps, were declared inoperable. Although LCO 3.0.3 was applicable, entry into the Required Actions of LCO 3.0.3 was suspended per the Note attached to Required Action E.1 of TS 3.7.5, 'Auxiliary Feedwater (AFW) System,' which states, 'LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.' At 1336 CST, Operations took actions to prevent the TDAFP from running, so the remaining (AFW Pumps) could be returned to Operable status. Operations then declared the affected instrumentation and the 'A' and 'B' motor-driven AFW pumps Operable. This allowed LCO 3.0.3 and Conditions A, B, D, and E under TS 3.7.5 to be exited. With only the TDAFP inoperable, TS 3.7.5 Condition C and its Required Actions remain in effect. Due to the degraded HELB door rendering all three AFW pumps inoperable, the unidentified condition is being reported as an unanalyzed condition that significantly degraded plant safety (per 10 CFR 50.72(b)(3)(ii)(B)) as well as a condition that could have prevented the fulfillment of the safety functions of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, and mitigate the consequences of an accident (per 10 CFR 50.72(b)(3)(v)(A), (B), and (D), respectively). The NRC Senior Resident Inspector has been notified.

  • * * RETRACTION ON 4/4/2018 at 1109 EDT FROM JONATHAN LAUF TO DAVID AIRD * * *

Event Notification (EN) # 53223, made on 2/20/2018, is being retracted because new information has been obtained that negates the original basis for reporting the unanalyzed condition. Specifically, an evaluation of the HELB that is postulated to occur in the TDAFP room has determined that without crediting door DSK13311 for protection, the affected safety-related instruments would not be exposed to environmental conditions beyond their analyzed capability. This resulted in a conclusion that the unanalyzed condition of door DSK13311 being open did not prevent the affected safety-related instruments or their supported AFW pumps from performing their required safety functions to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, and/or mitigate the consequences of an accident, nor did it significantly degrade plant safety. Consequently, the condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(A), (B), or (D). The licensee notified the NRC Resident Inspector. Notified R4DO (Drake).

ENS 5321315 February 2018 17:00:00LaSalle10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
On February 15, 2018, during evaluation of protection for Technical Specifications (TS) equipment from the damaging effects of tornado generated missiles, LaSalle Station identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from tornado generated missiles. Tornado generated missiles could strike the components supporting the operation of Control Room (VC) and Auxiliary Electric Room (VE) ventilation. This could result in inoperable VC/VE systems, which provide a protected environment for occupants to control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke if a tornado were to occur. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B) as a condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety and 10 CFR 50.72(b)(3)(v)(D) as a 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 NRC enforcement guidance provided in Enforcement Guidance Memorandum (EGM) 15-002, Revision 1, 'Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' and DSS-ISG-2016-01, Revision 1, 'Clarification of Licensee Actions in Receipt of Enforcement Discretion' per Enforcement Guidance Memorandum EGM 15-002, 'Enforcement Discretion for Tornado Generated Missile Protection Noncompliance.' Compensatory measures have been implemented in accordance with these documents. The NRC Resident Inspector has been informed of this notification.
ENS 5320412 February 2018 18:00:00Dresden10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
Emergency Diesel Generator

EN Revision Text: UNANALYZED CONDITION FOR TORNADO GENERATED MISSILES On February 12, 2018, during evaluation of protection for Technical Specifications (TS) equipment from the damaging effects of tornado generated missiles, Dresden Station identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from tornado generated missiles. Tornado generated missiles could strike the Unit 2, Unit 2/3, and Unit 3 Emergency Diesel Generator main fuel oil tank vents. This could result in crimping of the vents, which would affect the ability of the main fuel oil tanks to perform their function if a tornado would occur. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B) as a condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety and 10 CFR 50.72(b)(3)(v)(D) as a 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 NRC enforcement guidance provided in Enforcement Guidance Memorandum (EGM) 15-002, 'Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' and DSS-ISG-2016-01, 'Clarification of Licensee Actions in Receipt of Enforcement Discretion' per Enforcement Guidance Memorandum EGM 15-002, 'Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance,' Revision 1. Compensatory measures have been implemented in accordance with these documents. The NRC Resident Inspector has been informed of this notification.

  • * * RETRACTION ON 10/11/19 AT 1031 EDT FROM SAMANTHA COSENZA TO BETHANY CECERE * * *

The purpose of this notification is to retract event notification 53204 made on February 12, 2018, for Dresden Station. Additional review determined that the current design of all three Emergency Diesel Generators and associated Main Fuel Oil Storage Tanks Vents is consistent with the licensing basis for Dresden Station. There was no non-conformance of Dresden's tornado missile protection design, and the EDGs were operable at the time the event notification was made. Therefore, this event does not meet the criteria of 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(D). The ENS 53204 report is being retracted. The NRC Resident Inspector has been informed of this notification." Notified R3DO (Hills).

ENS 531599 January 2018 23:59:00Farley10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition

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 531109 December 2017 19:48:00Clinton10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
Secondary containment
Reactor Protection System
Primary containment
High Pressure Core Spray
Core Spray
Standby Gas Treatment System
Control Rod

At approximately 1347 (CST) on 12/09/17, the Main Control Room received annunciators that indicated a trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1 breaker. Numerous Division 1 components lost power (powered from unit subs 1A and A1). The Division 1 containment Instrument Air isolation valves had failed closed by design due to the loss of power. Due to the loss of containment instrument air, several control rods began to drift into the core as expected and, by procedure, the reactor mode switch was placed in the shutdown position at 1353 (CST). All control rods fully inserted. Also due to the loss of power, the Fuel Building ventilation dampers failed closed by design. With the normal ventilation system secured, secondary containment differential pressure rose to slightly greater than 0 inches water gauge which exceeded the Technical Specification requirement of greater than 0.25 inches vacuum water gauge at 1348 (CST). The Control Room entered EOP-8, Secondary Containment Control. Secondary Containment differential pressure was restored within Technical Specification requirements at 1351 (CST) by starting the Division 2 Standby Gas Treatment system. This event is being reported as a manual actuation of the Reactor Protection System (RPS) and as a Condition that Could Have Prevented Fulfillment of a Safety Function.

The cause is currently under investigation. The NRC Resident has been notified. The licensee informed the NRC Resident Inspector.

  • * * UPDATE FROM DALE SHELTON TO VINCE KLCO AT 1658 EST ON 12/10/2017 * * *

During a review of plant logs it was identified that the primary to secondary containment differential pressure was identified to be outside of Technical Specification 3.6.1.4 limits of 0 plus or minus 0.25 psid at 2009 on 12/9/17 due to the primary containment ventilation system dampers closing as a result of the loss of power. This parameter is an initial safety analysis assumption to ensure that primary containment pressures remain within the design values during a Loss of Coolant Accident (LOCA). As a result, this condition is reportable as an unanalyzed condition that significantly degrades plant safety. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone).

  • * * UPDATE FROM MICHAEL ANTONELLI TO VINCE KLCO ON 12/11/17 AT 1805 EST * * *

During the post transient review of the trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1, it was identified that the unplanned INOPERABILITY of the Low Pressure Core Spray (LPCS) system due to the loss of power to the injection valve constitutes an event or condition that could have prevented fulfillment of a safety function and is reportable under 10CFR50.72(b)(3)(v)(D) for Accident Mitigation. The High Pressure Core Spray (HPCS) remained available to perform the core spray function, if necessary, during a design basis Loss of Coolant Accident (LOCA), however HPCS and LPCS are each considered single train safety systems. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone).

ENS 531066 December 2017 14:58:00Three Mile Island10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
Emergency Diesel GeneratorOn December 6, 2017, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornado generated missiles, Three Mile Island Nuclear Station identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from tornado generated missiles. A tornado could generate a missile that could strike the emergency diesel generator (EDG) fuel oil supply tank (DFT) vent stack. This could result in crimping of the stack, which could affect the ability of the DFT to perform its design function if such a tornado would occur. 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 NRC enforcement guidance provided in EGM 15-002 and DSS-ISG-2016-01. Compensatory measures have been implemented in accordance with these documents. The NRC Resident Inspector has been informed of this notification.