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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 45377 | 22 September 2009 18:29:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Resulting from an Unlatched Door Serving as a High Energy Line Break Barrier | Based on the results of a past operability evaluation completed on 1329 (CDT), 9/22/2009, it appears that an unanalyzed condition existed intermittently for short periods of time in which a door that serves as a High Energy Line Break (HELB) barrier may have been unlatched. With the door not latched, an engineering evaluation concluded that a critical crack (HELB) in the Main Feedwater pipe traversing the south penetration room would force the door (DR-19) open, creating a harsh environment in the adjoining Emergency Feedwater (EFW) pump room. Because the EFW pump room is not evaluated for harsh conditions, it must be conservatively assumed that both pumps may fail to operate following this HELB event. Therefore, this condition is being reported in accordance with 10CFR50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. The licensee informed the Arkansas Department of Health. The licensee will be notifying the NRC Resident Inspector. | Feedwater | |
| ENS 47120 | 2 August 2011 18:46:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Vulnerability from a Potential High Energy Line Break | The following condition is being reported by Arkansas Nuclear One, Unit 2 (ANO-2) in accordance with 10CFR 50.72(b)(3)(ii)(B), 'Unanalyzed Condition' and in accordance with 10CFR 50.72(b)(3)(v)(D), 'A Condition That Could Have Prevented Fulfillment of a Safety Function.' On 08/02/2011 at 1346 CDT, the ANO Unit 2 Control Room was notified by Engineering that a postulated High Energy Line Break (HELB) could potentially cause both the Red and Green Train Emergency Safeguard Features (ESF) Rooms to exceed their environmentally qualified temperature limits. This postulated condition would be possible due to normally open room purge dampers exposing ESF equipment in these rooms to a common area impacted by HELB conditions. The ESF Rooms contain the Red and Green Trains of High Pressure Safely Injection Pumps, Low Pressure Safety Injection Pumps, Containment Spray Pumps, and Shutdown Cooling Heat Exchangers. Until further Engineering evaluation can be performed to validate this postulated scenario, ANO-2 has closed ESF room purge dampers to provide Red and Green ESF train separation during a potential HELB event. Refer to (ANO-2) Condition Report CR-ANO-2-2011-02772 for further information. The NRC Resident has been notified.
The purpose of this notification is to retract a previous report made by Arkansas Nuclear One, Unit 2 (ANO-2) on 08/02/2011 at 2127 (EDT) (EN# 47120). The initial report documented that a postulated High Energy Line Break (HELB) could potentially cause rooms containing both trains of Emergency Safeguard Features (ESF) equipment to exceed their environmentally qualified temperature limits. The ESF rooms contain the High Pressure Safely Injection Pumps, Low Pressure Safety Injection Pumps, Containment Spray Pumps, and Shutdown Cooling Heat Exchangers. Specifically, normally open ESF room purge dampers exposing both trains of ESF equipment to a common area impacted by postulated HELB conditions were not modeled in the ANO-2 HELB analysis. This condition was reported in accordance with 10CFR 50.72(b)(3)(ii)(B), 'Unanalyzed Condition' and 10CFR 50.72(b)(3)(v)(D), 'A Condition That Could Have Prevented Fulfillment of a Safety Function'. Since the initial report, Engineering has revised the ANO-2 HELB model to include the effects of the open ESF room purge dampers. The resulting analysis shows that a HELB event will not cause the required ESF equipment to exceed analyzed temperature limits with the room purge dampers in the open configuration. Therefore, the condition did not result in 'a condition that could have prevented the fulfillment of a safety function' and did not result in an 'unanalyzed condition that significantly degrades plant safety'. Based on the revised HELB analysis, the previous report (EN#47120) describes a condition that does not meet the reporting requirements of 10CFR 50.72(b)(3)(v)(D) or 10CFR 50.72(b)(3)(ii)(B) and is therefore retracted. The NRC Resident Inspector has been informed of the retraction. Notified R4DO (Hay). | Shutdown Cooling Containment Spray | |
| ENS 47777 | 28 March 2012 01:20:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Spent Fuel Handling Machine Not Fully Qualified for a Seismic Event | The following condition is being reported by Arkansas Nuclear One, Unit 2 (ANO-2) in accordance with 10CFR 50.72(b)(3)(ii)(B), 'Unanalyzed Condition.' Conservative engineering analysis has determined that the Spent Fuel Handling Machine (2H-3) is not qualified in a design basis earthquake event. Current seismic analysis is indicating a lack of margin for several structural parts of the machine. The Spent Fuel Handling Machine (2H-3) is currently parked and de-energized in a safe position and administratively prohibited from being moved over any irradiated fuel assemblies in the spent fuel pool. Until further engineering evaluation or modifications can occur the Spent Fuel Handling Machine (2H-3) will remain in its current position and de-energized. The licensee has notified the NRC Resident Inspector. * * * RETRACTION FROM STEVEN COFFMAN TO PETE SNYDER AT 1104 EDT ON 5/23/12 * * * The purpose of this notification is to retract the previous Event Notification Report #47777 made by Arkansas Nuclear One, Unit 2 on 03/27/2012. The initial report indicated that the Spent Fuel Handling Machine (2H-3) may not have been qualified during a design basis earthquake event due to seismic analysis indicating a lack of margin for several structural parts of the machine. 2H3 was parked and de-energized in a safe position, and the condition was reported in accordance with 10 CFR 50.72(b)(3)(ii)(B), 'Unanalyzed Condition.' Since the initial report, Engineering has completed a new seismic analysis which demonstrates that the discovered conditions of the Spent Fuel Handling Machine structural parts were adequate to preclude structural failure of the machine in the area of the Spent Fuel Pool. References to the calculations that formulate the basis for this conclusion are documented in the Licensee's corrective action program. Based on the new seismic analysis, the condition initially reported in EN #47777 did not result in 'an unanalyzed condition that significantly degrades plant safety' and is therefore retracted. The NRC Resident Inspector has been briefed on the seismic analysis results and informed of this retraction. Notified R4DO (Spitzberg). | ||
| ENS 48644 | 2 January 2013 19:08:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Inadvertent Actuation of Safety Injection, Containment Cooling and Containment Isolation Signals | At 1308 hours (CST) on January 2, 2013, Arkansas Nuclear One Unit 2 (ANO-2) experienced an inadvertent safety injection actuation signal (SIAS), containment cooling actuation signal (CCAS) and containment isolation actuation signal (CIAS) during matrix testing on 'C' channel plant protective system (PPS). All components actuated as designed for the stated actuation signals. After verification that the actuation signal was not valid the high pressure safety injection (HPSI) pumps and low pressure safety injection (LPSI) pumps were secured and their hand switches placed in pull-to-lock as directed by the abnormal operating procedure for inadvertent SIAS at 1312 hours. The HPSI and LPSI pumps were the restored to normal standby condition and available for automatic operation at 1352 hours. The unit did down power to approximately 87% power when all three coolant charging pumps ran with suction aligned to the boric acid makeup tanks. The unit currently remains in mode 1. SIAS, CCAS, and CIAS have been reset. The cause of the inadvertent actuation signals is under investigation. The NRC Resident Inspector has been informed. Unit 2 entered Technical Specification 3.0.3 for approximately 40 minutes while the LPSI and HPSI pumps were in pull-to-lock. | 05000368/LER-2013-001 | |
| ENS 49873 | 5 March 2014 16:20:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Discovery of Individual Pathways That Could Bypass Flood Barriers | During walk downs to ensure the availability of flood protection barriers, a condition was identified which had the potential to adversely impact the ability to address external flooding conditions. Several individual pathways between both unit's Turbine Building and Auxiliary Building were identified that could bypass flood barriers. In the aggregate however, the current equipment could become overwhelmed and the flooding in the Auxiliary Building could then potentially challenge equipment necessary to remove residual heat. The identified pathways were for the most part unscheduled partially filled conduits. There were no isolation features on these pathways and no barriers to flooding were in place between the Turbine Building and Auxiliary Building thus the potential existed to bypass the existing flood barriers. Flooding of the Turbine Building conceivably could have resulted in the accumulation of water in sufficient quantities to fill the Turbine Building to the height of the external floodwaters which could enter the Auxiliary Building via one or more deficient flood barrier. These floodwaters would then potentially challenge equipment, located within the Auxiliary Building, which is required to remove residual heat. This condition has been determined to be reportable per 10CFR50.72(b)(3)(v)(B) and 10CFR50.72(b)(3)(ii)(B). This condition is a non-emergency condition. This condition has been entered into the Corrective Action Program. Barriers are being installed in these pathways as they are identified or compensatory measures implemented. The walk downs were performed in response to Fukishima lessons learned. The licensee notified the NRC Resident Inspector. | ||
| ENS 50519 | 8 October 2014 15:35:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Potential Fire Induced Cable Failure in Multiple Fire Zones | INPO-IER-L4-14-33 (Direct Current Circuits Challenge Appendix R Fire Analysis) was reviewed to determine applicability to ANO Unit 1 & 2. It was determined that 2P-21, Turbine Generator Emergency Seal Oil Pump, control cables are not fused and are routed through multiple fire zones containing safe shutdown equipment. A potential fire induced cable failure in any of these fire zones could result in a secondary fire or damage adjacent cables along the path of the unprotected cable. The concern is that under fire safe shutdown conditions, it is postulated that a fire in one zone can cause short circuits potentially resulting in secondary fires or cable failures in other fire zones 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. Compensatory measures (fire watches) have been implemented for affected zones of the plant. The NRC Resident Inspector has been notified. | ||
| ENS 50641 | 25 November 2014 18:11:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Deficiency in Methodology Used for Emergency Core Cooling System Performance Requirements | On Tuesday November 25, 2014, at 1211 CST, Arkansas Nuclear One, Unit 1 (ANO-1) reviewed AREVA 10 CFR 50.46 Notification Letter FAB14-00632. This letter indicates that a deficiency was discovered in the uranium thermal conductivity models used in the ANO-1 Loss of Coolant Accident (LOCA) analysis of record. When the deficiency is corrected, the LOCA Peak Cladding Temperature (PCT) limits may be in excess of 2200 degrees Fahrenheit (F). 10 CFR 50.46 paragraph (b) defines the acceptance criteria for the LOCA analysis process. The ANO-1 licensing basis PCT is evaluated for compliance with the criterion 10 CFR 50.46(b)(1) and must not exceed a PCT of 2200 degrees F. During AREVA's review of the issue, AREVA had provided compensatory measures in the form of reductions in LOCA linear heat rates as a contingency in case the errors were found to be substantiated, which were then translated into reduced axial imbalance limits so that ANO-1 would operate within 10 CFR 50.46 limits. As a precautionary measure pending the completed analysis, ANO-1 implemented the compensatory measures on October 20, 2014, and as a result, the errors reported have no impact on current plant operation or public health and safety. This event is being conservatively reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Based on 50.46(a)(3)(ii) criteria, ANO-1 will submit a written report within 30 days. ANO-1 has notified the NRC Senior Resident Inspector. | Emergency Core Cooling System | 05000313/LER-2014-002 |
| ENS 51808 | 19 March 2016 22:10:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition That Could Challenge Rhr Equipment During Flood Conditions | Two (2) potentially degraded flood barriers at penetrations 0073-01-0034 and 0073-01-0063 were identified in the area between the Unit 1 Turbine Building and Auxiliary Building. The deficient barriers are a 'blockout section' of the floor designed to house multiple penetrations that transition from the Turbine Building to the Auxiliary Building. Attempts have been made to investigate the status of the flood barrier with no definitive results. Investigations and additional evaluations are continuing, however, it is currently unknown if the aggregate of these two flood barriers could potentially overwhelm and flood the Auxiliary Building which would challenge equipment necessary to remove residual heat and constitute an unanalyzed condition. Based on current conditions (i.e., no forecast flooding conditions), this condition does not present an immediate safety concern. This condition has been determined to be reportable per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(B). This condition is a non-emergency condition. This condition has been entered into the Corrective Action Program. Compensatory measures have been prepared to allow placement of a seal over the identified deficient barriers. If required these seals can be installed well in advance of forecast flood conditions. Permanent repairs are currently being designed for installation. The NRC Resident Inspector has been notified.
This is an 8 hour non-emergency supplemental notification to previously issued Event Notification number 51808. In EN 51808, two non-functional barriers were identified and reported, and during an extent of condition review, a third barrier has been identified that does not conform to expected flood barrier standards A potentially degraded flood barrier at 'blockout' penetration 0073-01-9018 was identified in the area between the Unit 1 Turbine Building and Auxiliary Building. The deficient barrier is a 'blockout section' of the floor designed to house multiple penetrations that transition from the Turbine Building to the Auxiliary Building. Attempts have been made to investigate the status of the flood barrier with no definitive results. Investigations and additional evaluations are continuing; however, it is currently unknown if this flood barrier could potentially be overwhelmed and flood the Auxiliary Building which would challenge equipment necessary to remove residual heat and constitute an unanalyzed condition. Based on current conditions (i.e., no forecast flooding conditions), this condition does not present an immediate safety concern. This condition has been determined to be reportable per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(B). This condition is a non-emergency condition. This condition has been entered into the Corrective Action Program. Compensatory measures have been prepared to allow placement of a seal over the identified deficient barrier. If required this seal can be installed well in advance of forecast flood conditions. Permanent repairs are currently being designed for installation. The NRC Resident Inspector was notified earlier in the evening that this event would be updated. Notified the R4DO (Haire). | ||
| ENS 52195 | 24 August 2016 13:41:00 | 10 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 | Unanalyzed Condition Associated with Damaging Effects of Tornados | During performance of an extent of condition evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Arkansas Nuclear One, Unit 1, identified non-conforming conditions in the plant design such that specific TS equipment on Unit 1 is considered to not be adequately protected from tornado missiles. The reportable condition is postulated by tornado missiles entering the Unit 1 Cable Spreading Room through penetrating the hollow metal door or potentially from spalling of the block wall separating Room 96 and 97. A tornado could generate multiple missiles capable of striking the Unit 1 Cable Spreading Room and rendering both safety related electrical trains 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. Licensee Event Report (LER) 50-313/2016-002-00 was recently submitted addressing previously identified tornado missile vulnerabilities at the Unit 1 plant. 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 52234 | 11 September 2016 19:51:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Inadequate Design Protection of Safety-Related Equipment Identified During Tornado Missile Impact Evaluation | During performance of an extent of condition evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Arkansas Nuclear One, Unit 1, identified non-conforming conditions in the plant design such that specific TS equipment on Unit 1 is considered not be adequately protected from tornado missiles. The reportable condition is postulated by tornado missiles entering the Unit 1 Controlled Access area, elevation 386', Upper North Electrical Penetration Room (UNEPR) through penetrating a hollow metal door and then striking safety related cables. A tornado could generate multiple missiles capable of striking the Unit 1 UNEPR and rendering both safety related emergency feedwater trains 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)(B) and (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 (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. Licensee Event Report (LER) 50-313/2016-002-00 was recently submitted addressing previously identified tornado missile vulnerabilities at the Unit 1 plant. 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. A similar evaluation is on going for Unit 2. | Feedwater | |
| ENS 52242 | 15 September 2016 14:08:00 | 10 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 | Potential Unanalyzed Condition Based on Identified Non-Conforming Conditions | During performance of an extent of condition evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornadoes, Arkansas Nuclear One, Unit 1, identified non-conforming conditions in the plant design such that specific TS equipment on Unit 1 is considered not (to) be adequately protected from tornado missiles. The reportable condition is postulated by tornado missiles entering vital switchgear rooms 99 and 100 and striking vital switchgears in the rooms. A tornado could generate multiple missiles capable of striking the Unit 1 vital switchgear and rendering both safety related AC electrical trains 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) safe shutdown capability, (B) residual heat removal capability, or (D) accident mitigation. This condition was identified as part of an on-going extent of condition review of potential tornado missile related site impacts. Licensee Event Report (LER) 50-313/2016-002-00 was recently submitted addressing previously identified tornado missile vulnerabilities at the Unit 1 plant. 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. | Residual Heat Removal | |
| ENS 52271 | 30 September 2016 02:00:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Unisolable Leak on Decay Heat Removal Piping Due to Weld Failure on a 1" Common Pipe | At 2100 CDT on 09/29/16, while in Mode 6, both trains of Decay Heat (Residual Heat Removal) were declared inoperable due to a cracked weld on a 1" common pipe. The leak developed in a USAS B31.7, Class1 pipe at a weld upstream of pressure indication isolation valve DH-1037. The leak is not isolable from the common 8-inch Decay Heat piping and encompasses approximately 1/3 (one third) of the pipe circumference. At the time of discovery, the unit was in Lowered Inventory with both Loops of Decay Heat in service. Subsequently, one train of Decay Heat has been secured to reduce the likelihood of crack propagation. One Train of Decay Heat remains in service providing the function of removing Decay Heat and the other train is readily available. The leakage impacts redundant equipment required to fulfill a safety function. In the current condition, both trains are required to be operable to meet Technical Specification LCO 3.9.5, Decay Heat Removal (DHR) and Coolant Circulation-Low water Level. This condition is reportable per 10 CFR 50.72(b)(3)(v)(B) for any event or condition that results in a loss of Safety Function associated with the Decay Heat System (Residual Heat Removal System). The licensee has notified the NRC Resident Inspector. The leak is approximately 0.25 gallons per minute and pipe pressure is 140 psi. Compensatory measures are in place and include an individual posted to watch the pipe in case plugging is necessary. Repairs to the pipe will be completed once pipe is able to be drained. | Residual Heat Removal Decay Heat Removal | 05000313/LER-2016-004 |
| ENS 53456 | 20 May 2024 20:53:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Pressure Boundary Leakage Identified While Shutdown | On June 12, 2018, at 1500 CDT, a Reactor Coolant System (RCS) Pressure Boundary leak was identified during a Mode 3, hot shutdown walkdown on a High Pressure Injection Line (HPI) to Reactor Coolant Pump (P32C) drain line weld near MU-1066A HPI Line Drain Valve and MU-1066B HPI Line Drain Valve. The 3/4 inch drain line containing drain valves MU-1066A and MU-1066B on the 'C' HPI header (CCA-5 pipe class) has a through-wall defect on the pipe stub or welds between the sockolet and valve MU-1066A. The leak location is in the ASME Class I RCS Pressure Boundary. The hot shutdown walkdown was being performed as part of a planned outage to investigate excessive Reactor Building Sump inleakage. Total unidentified RCS leakage prior to the investigation was determined to be at 0.165 gpm. After the initial investigation of the leakage, the following Tech Specs (TS) were determined be applicable: TS 3.4.5 - RCS Loops Mode 3, TS 3.4.13 - RCS Leakage, TS 3.5.2 - ECCS. Unit 1 is currently in Mode 3 and in progress of an RCS cooldown to comply with Tech Spec requirements. The licensee notified the NRC Resident Inspector. | Reactor Coolant System | |
| ENS 54339 | 20 October 2019 15:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Envelope Breach | At 1030 CDT, it was discovered that the loop seal on the condensate drain was empty for VUC-9 Control Room AC Unit. This creates a breach in the Control Room envelope. Unit 2 entered (Technical Specification) T.S. 3.7.6.1 Action D. Unit 1 is in Mode 6; therefore, not in a mode of applicability. Compensatory action were being performed and the licensee was in the process of sealing the loop. The licensee will notify the NRC Resident Inspector.
This report is being retracted. The Control Room Envelope (CRE) provides a safety function which limits radiological dose to occupants to no more than 5 rem for 30 days post-accident. The dose limitation assumes the occupants are stationed within the CRE 24 hours a day for the entire 30-day period. The CRE also functions to protect occupants from potential hazards such as smoke or toxic chemicals. The CRE is declared inoperable when a potential breach is identified, regardless of the ability to seal the breach. With respect to the event of October 20, 2019, the water level in a loop seal could not be maintained at the desired level. Subsequent evaluation determined that sufficient water was maintained in the loop seal to prevent a breach of the CRE. The subject reporting criterion is based on the assumption that safety-related systems, structures, and components (SSCs) may no longer be capable of mitigating the consequences of an accident. In accordance with NUREG 1022, 'Event Report Guidelines 10 CFR 50.72 and 50.73,' a report may be retracted based on a revised operability determination. The CRE remained operable; therefore, this report may be retracted. The licensee notified the NRC Resident Inspector. Notified R4DO (Young). | Control Room Envelope | |
| ENS 54954 | 19 October 2020 04:13:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Envelope Inoperable | On October 18, 2020 at 2313 CDT, Arkansas Nuclear One (ANO) discovered that 2VRA-1B (2VSF-9 outside air damper reserve bottle) was below 600 psig. This condition caused the control room envelope to be inoperable in accordance with OP-2104.007 Attachment L. ANO Unit 1 entered TS 3.7.9 Condition B for inoperable control room boundary. ANO Unit 2 entered TS 3.7.6.1 Action D for inoperable control room boundary. A procedurally controlled temporary modification was implemented to install a blank flange on the 2VSF-9 outside air damper. Both Units declared the control room boundary operable at 2358 CDT. The associated control room emergency recirculation fan remains inoperable with the blank flange installed. This is a 7-day shutdown-LCO for both units. The licensee informed the NRC Resident Inspector.
Previously, Entergy notified the NRC that ANO control room envelope was inoperable due to 2VRA-1B (2VSF-9 outside air damper reserve bottle) was below required pressure of 600 psig. After additional engineering evaluation, it was determined the control room boundary remained intact for this condition. As documented in version 2 operability determination for condition report ANO-C-2020-2818, the control room ventilation boundary remained intact for the condition identified and was able to fulfill its function for the required 30-day mission time. In accordance with NUREG-1022, 'Event Report Guidelines 10 CFR 50.72 and 50.73,' a report may be retracted based on a revised operability determination. The CRE remained operable; therefore, this report may be retracted. The NRC Resident Inspector has been informed. Notified R4DO (Pick). | Control Room Envelope |