Semantic search
| Start date | Title | Description | Topic | |
|---|---|---|---|---|
| ENS 54979 | 3 November 2020 05:11:00 | Both Trains of Salt Water Inoperable | At 0011 EST on 11/03/20, it was discovered that BOTH trains of salt water were simultaneously INOPERABLE. While in a planned (limiting condition for operation) LCO window with the 21 salt water train INOPERABLE for post-maintenance testing, debris intrusion in the 22 salt water header rendered the redundant salt water train INOPERABLE. Due to this INOPERABILITY, the system was in a condition that could have prevented the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). One train of salt water was restored to operable at time 0026 EST. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. This event did not affect Unit 1.
Following the eight hour 10 CFR 50.72 notification made on 11/03/2020 (EN 54979), further engineering analysis determined that 22 Saltwater subsystem flow remained at levels sufficient to fulfill its safety function based on the conditions existing at the time of the event. Despite flow in 22 Saltwater subsystem falling below the short term (four hour) minimum value for approximately 15 minutes, engineering analysis was able to determine the increased heat removal capacity associated with the lower bay temperatures was sufficient to offset the reduced heat removal capacity associated with the lower 22 Saltwater subsystem flow. This demonstrated that actual heat transfer to the saltwater subsystem was sufficient to ensure all safety functions were fulfilled during the event. Therefore, this event notification is being retracted as it is not reportable pursuant to 10 CFR 50.72(b)(3)(v)(A), (B) and (D). The NRC Resident has been informed. Notified R1DO (Greives) | |
| ENS 52640 | 25 March 2017 05:42:00 | Refueling Water Tank Level Inadvertently Lowered Below Ts | While performing a purification subsystem alignment on the Unit-2 Refueling Water Tank, an inadvertent transfer of Refueling Water Tank level to the common Spent Fuel Pool occurred. This transfer resulted in lowering Unit-2 Refueling Water Tank level below the Technical Specification (TS) required limit for the current mode of operation at 0142 (EDT) on 3/25/17. Upon recognition of the inadvertent transfer, Operations secured the lineup and restored Unit-2 Refueling Water Tank level to its normal operating band at 0225 on 3/25/17. This event is reportable under 10 CFR 50.72(b)(3)(v)(D) '...any event or condition that at the time of discovery could have prevented the fulfillment of the safety function structures or systems that are needed to mitigate the consequences of an accident.' With less than the required Technical Specification volume in the Refueling Water Tank, insufficient volume existed in the Refueling Water Tank to maintain 30 minutes of full flow Safety Injection, and subsequent continued pump operation after transition to recirculation mode of operation. This level is required by Technical Specification 3.5.4.B and has a one hour action statement to restore level. The NRC Resident Inspector has been notified.
The purpose of this notification is to retract ENS notification 52640 made on March 25, 2017 for Calvert Cliffs. After further evaluation, it has been determined that the volume of water in the Unit 2 Refueling Water Tank was never below the TS required volume of 400,000 gallons. The evaluation considered the as-found condition of the level transmitter and the existing environmental conditions of the tank in determining the actual RWT water volume on the day of the event. Therefore, this event does not meet the criteria of 10 CFR 50.72(b)(3)(v)(D) and the ENS report is being retracted. The licensee will notify the NRC Resident Inspector. Notified R1DO (Cook). | Time of Discovery |
| ENS 50752 | 22 January 2015 22:45:00 | Unanalyzed Heat Exchanger Lineup Could Exceed Design Basis Temperatures | On 1/20/15, it was determined that a certain line up of component cooling heat exchangers and shutdown cooling heat exchangers could exceed the design basis temperatures for the component cooling water system following a design basis accident. Although not a safety concern at this time because of low ultimate heat sink temperatures (which cools component cooling water), in the past the ultimate heat sink temperatures have been high enough to create this condition. This particular heat exchanger line up was unanalyzed in that the ultimate heat sink temperature limits were not known until 1/22/15. This issue has been entered into the corrective action program. A review of Control Room logs for 2014 showed that in 1 instance for Unit 1 and 1 instance for Unit 2, the Units were in an unanalyzed lineup with ultimate heat sink temperature greater than the maximum now calculated. During these instances, both Units had an unanalyzed condition that had potential to significantly degrade plant safety and is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.
Further engineering analysis has refined the ultimate heat sink temperature that provides an acceptable safety system response with the component cooling water and shutdown cooling heat exchanger lineups in question. The revised information demonstrates that the system lineups that occurred in the last 12 months did not result in an unanalyzed condition that significantly degrades plant safety. This event notification is retracted. The NRC Resident Inspector will be notified. The R1DO (Burritt) was notified.
The retraction statement provided on 3/3/15 incorrectly addressed system lineups limited to the last 12 months. The original withdrawal was prematurely submitted and therefore the original notification (Event report 50752) is not retracted and remains valid. The NRC Resident Inspector has been notified. The R1DO (Cook) was notified. | Unanalyzed Condition Ultimate heat sink |
| ENS 50502 | 1 October 2014 05:24:00 | Partial Loss of Communications in Emergency Operations Facility and Joint Information Center | At 0750 EDT on October 1, 2014, the Shift Manager was notified that site Information Technology (IT) personnel were being mobilized to investigate a potential voice and network loss at the Emergency Operations Facility (EOF) and Joint Information Center (JIC). Site IT personnel were notified by offsite IT resources at 0727 EDT on October 1, 2014 of the issue that was first identified by IT monitoring software at 0124 on October 1, 2014. The site IT personnel that responded to the EOF and JIC reported to site Control Room and Emergency Preparedness (EP) personnel at 0845 that connectivity to the Exelon network and the internet was unavailable at both the EOF and the JIC. This loss of connectivity would prevent the ability of the EOF Emergency Response Organization (ERO) personnel to directly monitor key plant parameters via the site's Plant Process Computer (including the Site Parameter Display System) and other network-based plant parameter display systems. Site IT and EP personnel determined that the following communications equipment was not impacted by the connectivity issue: - Dedicated Offsite Agency Phones (primary method for contacting state and local agencies) - Commercial Phones and dedicated bridge line (primary method for contacting other site Emergency Response Facilities) - FTS-2001 Phones (e.g., ENS and HPN lines) - ERDS Additionally, EP personnel verified with Dose Assessment Office personnel that dose assessment and dose monitoring functions from the EOF could still be performed without delay. Site IT personnel reported to the Control Room at 1135 that connectivity to the Exelon network and the internet had been restored to a fully functional status. While site and fleet IT personnel continue to address and verify all appropriate corrective actions have been taken to prevent recurrence of the connectivity issue, the site has employed appropriate compensatory measures to ensure that the verbal transmission of key plant parameters from the site (Technical Support Center or Control Room) to the EOF is recognized and maintained. The NRC Resident Inspector has been notified.
This update retracts Event Report #50502, which reported that a loss of connectivity to the Exelon network and internet at the Emergency Operations Facility (EOF) and Joint Information Center (JIC) had impacted the ability of staff in these facilities to directly monitor key plant parameters via the site's Plant Process Computer and other network-based plant parameter display systems. Subsequent to the identification of this event, further investigation by site and fleet staff determined that adequate direction was included in applicable Emergency Response Organization (ERO) procedures to respond to data display system failures of this type. Specifically, the checklist (procedure) for the Operations Communicator in the EOF provided adequate direction for this ERO member to obtain required plant data from the Operations Communicator located in the Control Room via alternate methods (e.g., over the phone - phone lines remained functional throughout the time that the loss of computer connectivity condition existed). Therefore, this event did not result in a major loss of emergency assessment capability and was not reportable to the NRC under 10CFR50.72(b)(3)(xiii). The NRC Resident Inspector has been notified. Notified R1DO (Bickett) and Cyber Assessment Team via email. | |
| ENS 48367 | 3 October 2012 04:10:00 | Unanalyzed Condition Due to High Energy Line Break Barrier Being Partially Open | At 0010 (EDT) on 10/3/12, it was determined that an unanalyzed condition existed for the switchgear rooms on Unit 1. A degraded condition of a high energy line break (HELB) barrier was created during security surveillance. A HELB barrier door for the 27' elevation switchgear room was partially opened for a 14 minute period and could have allowed steam from a HELB on the 27' elevation of the Turbine Building to potentially impact equipment in both the 27' and 45' elevation switchgear rooms because these rooms are connected by a common ventilation system. The switchgear rooms are not analyzed for a steam environment. The degree of the impact could not be readily determined, but could likely affect the safety related equipment in the switchgear rooms. At 0024 (EDT) on 10/3/12, the HELB barrier door for the 27' elevation switchgear room was closed and the barrier restored which eliminated the potential unanalyzed condition. An 8 hour report to the NRC is required 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' since there was not a reasonable expectation that the switchgear room environment could support operation of safety related equipment with the 27' elevation switchgear room door partially open. Further analysis is underway. The licensee has notified the NRC Resident Inspector.
Subsequent engineering analysis conducted for this condition demonstrated that having the roll-up door stuck open six inches would not, should a HELB occur in the vicinity of the 27' switchgear room roll-up door, result in conditions within both switchgear rooms that would challenge the design limits and which would render the Class 1E electrical busses inoperable. Therefore the unit was not in a degraded or unanalyzed condition. As a result this condition is not reportable under 50.72(b)(3)(ii)(B). Based on this information, Calvert Cliffs is retracting notification message EN #48367. The licensee has notified the NRC Resident Inspector. Notified R1DO (Dwyer). | Unanalyzed Condition |
| ENS 47619 | 23 January 2012 18:00:00 | Breach in High Energy Line Break Barrier | At 1300 EST on 1/23/12, it was determined that an unanalyzed condition existed for the Unit 1 Cable Spreading Room. A high energy line break (HELB) barrier issue was discovered while performing a HELB inspection and the condition is believed to have existed from initial plant construction. A HELB barrier was found to have a breach in it that could allow steam from a high energy line break in the Unit 1 Turbine Building (into the cable spreading room and thus into the control room). The Control Room is not analyzed for a steam environment. The degree of the impact could not be readily determined, but could likely affect the safety related equipment in the Cable Spreading Room. Therefore, an 8-hour report to the NRC is required 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.' The licensee will notify the NRC Resident Inspector.
Engineering performed an evaluation to address the impact of the degraded condition on the barrier's design functions. The evaluation concluded that the barrier remained capable of performing its design function with the degraded seal present. Therefore, this condition does not represent an unanalyzed condition that significantly degrades plant safety. The NRC Resident Inspector has been notified. Notified the R1DO (Dentel). | Unanalyzed Condition |
| ENS 47512 | 8 December 2011 22:55:00 | Unanalyzed Condition Potentially Could Affect the Common Control Room | At 1755 on 12/8/11, it was determined that an unanalyzed condition existed for the common Control Room for both Units. A high energy line break (HELB) barrier issue was discovered while performing a fire barrier surveillance and the condition is believed to have existed from initial plant construction. A HELB barrier was found to have a significant breach in it that could allow steam from a HELB in the Unit 2 Steam Generator Blowdown system to potentially impact equipment in the Control Room. The Control Room is not analyzed for a steam environment. The degree of the impact could not be readily determined, but could likely affect the safety related equipment in the Control Room. At 1803 on 12/8/11, Unit 2 Steam Generator Blowdown was secured to eliminate the potential for a HELB in the affected area which eliminated the potential unanalyzed condition. Therefore, an 8 hour report to the NRC is required 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' since there was not a reasonable expectation that the Control Room environment could support operation of safety related equipment with Unit 2 Steam Generator Blowdown in service. Further analysis is underway. The licensee will notify the NRC Resident Inspector
Engineering performed an evaluation to address the impact of the degraded condition on the barrier's design functions. The evaluation concluded that the barrier remained capable of performing its design function with the degraded seal present. Therefore, this condition does not represent an unanalyzed condition that significantly degrades plant safety. The licensee will notify the NRC Resident Inspector. R1DO (Burritt) notified. | Unanalyzed Condition Fire Barrier |
| ENS 47413 | 3 November 2011 22:00:00 | Diesel Generator Starting Air Not Compliant with Appendix R Requirements | During the conduct of a system vulnerability assessment by Engineering on the emergency diesel system. a discussion was held regarding the impact of a fire on the DG starting air system. Subsequently it was recognized that components of the system were vulnerable to damage during a fire. A review of the Appendix R analysis was conducted and it was determined that this vulnerability was not analyzed in the evaluation. This is reportable to Unit 2, as Unit 1 has a separate diesel building for the 1-Alpha emergency diesel. The NRC Resident Inspector has been notified.
The Appendix R analysis was updated and no changes to the plant or procedures resulted. Therefore, no unanalyzed condition that significantly degraded plant safety existed. This event is being retracted. The NRC Resident Inspector has been notified. R1DO (Joustra) notified. | Unanalyzed Condition |
| ENS 45737 | 2 March 2010 12:30:00 | Pressurizer Safety Nozzle Weld Degradation | A Unit-1 Pressurizer Safety Relief Valve Nozzle to Safe-end Weld, (#4SR-1006-1) was discovered to have an indication not acceptable under ASME Section XI, IWB-3600. The weld is on a 4 inch diameter line to the pressurizer safety relief valve (RV-201). The weld is in our dissimilar metal weld inspection program and was scheduled to be examined this RFO (refueling outage) in accordance with MRP-139 requirements. Initial inspections on 3/1/10 found an indication and the indication was confirmed by a second NDE level 3 examiner. Indication is circumferential, initiated at ID and propagates approximately 1.8 inches circumferentially to about approximately 70% through wall. ASME IWB-3500 allows up to 12.5% through wall for class 1 welds. The defect was found using phased array Ultrasonic Testing (UT). The NDE report is being reviewed and characterized at this time. At this time, we believe the most probable cause of the indication is primary water stress corrosion cracking. No active leaks have been identified, the defect is not through wall. Unit 1 is in Mode 6. A team has been established and repair options are being investigated. The licensee notified the NRC Resident Inspector.
On March 7, 2010, manually indexed phased-array ultrasonic examination of the U1 Pressurizer Safety Relief Valve Nozzle to Safe-end Weld was performed. Based on these inspections, it was determined on March 13, 2010 that the indication does not exhibit stress corrosion cracking characteristics and is not consistent with ultrasonic responses associated with inside diameter (ID) connected geometry. Therefore, the U1 Pressurizer Safety Relief Valve Nozzle to Safe-end Weld is acceptable from an ASME Code Section XI perspective. On this basis, the indication did not result in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded and the issue is not reportable under 10 CFR 50.72(b)(3)(ii)(A). The NRC resident was notified of this retraction. Notified the R1DO (Chris Cahill) | Dissimilar Metal Weld |
| ENS 43830 | 8 December 2007 05:50:00 | Reactor Power Inadvertently Increased Above Accident Analysis Limit | On December 8, 2007 at 0050, during plant maneuvers to return Unit Two from 84% to 100% reactor power following main turbine control valve testing, the two-minute instantaneous thermal power was inadvertently increased above the accident analysis thermal power limit of 2735.1 thermal megawatts while adjusting main turbine load to maintain Reactor Coolant temperature on program. The maximum power level attained was 2741 thermal megawatts. This is a potentially unanalyzed condition. Prompt operator action was taken to restore reactor power to within limits. Reactor thermal power was restored to within limits (below 2735.1 thermal megawatts) eight minutes later at 0058. The two-minute instantaneous thermal power was restored to below 2700 thermal megawatts (100% rated thermal power) at 0105. The licensee notified the NRC Resident Inspector.
Based on the results of an evaluation of the event, the licensee concluded that Unit 2 operated at a maximum power level of 2741 MWth (101.5 percent rated thermal power (RTP)). The maximum analyzed steady-state reactor core power levels, including uncertainties, are 102 percent of RTP or 2754 MWth. The unit was not in an unanalyzed condition. Therefore, this event is not reportable. Based on this evaluation, the associated non-emergency notification made on December 8, 2007, is retracted. Licensee has notified the NRC Resident Inspector. Notified R1DO (Perry). | Unanalyzed Condition |