Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 55122 | 4 March 2021 08:23:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Notification of Degraded Condition | At time 0323 (EST) on March 04, 2021, it was determined that the Reactor Coolant System (RCS) pressure boundary did not meet the acceptance criteria under ASME, Section XI IWB-3600, "Analytical Evaluation of Flaws." This condition will be resolved prior to plant start up. This event is being reported as an eight hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident has been notified. | Reactor Coolant System | |
| ENS 51747 | 20 February 2016 08:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Discovered on Pressurizer Safety Relief Valve Weld | During the CCNPP (Calvert Cliff Nuclear Power Plant) Unit 1 Refueling, (an ultrasonic test) examination of the Unit 1 dissimilar metal weld (DMW) of the pressurizer safety relief valve RV-201 safe end weld found an indication that is approximately 80 percent though-wall, inside diameter connected axial flaw in weld #4-SR-1006-1. This is an ASME Class 1 component and Unit 1 has entered the requirements of Technical Requirements Manual (TRM) 15.4.3.A. CCNPP Event #4254. The licensee has notified the NRC Resident Inspector. | Safety Relief Valve | 05000317/LER-2016-002 |
| ENS 50307 | 25 July 2014 03:15:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Technical Specification Required Shutdown Due to Reactor Coolant System Pressure Boundary Leakage | At 2315 EDT on July 24th, 2014, CCNPP (Calvert Cliffs Nuclear Power Plant) U1 identified RCS (Reactor Coolant System) pressure boundary leakage from the instrument line to 1-PDT-123A, 11A reactor coolant pump differential pressure transmitter. Technical Specification 3.4.13, Action B was entered and requires that the unit be placed in Mode 3 within 6 hours and Mode 5 within 36 hours. (CCNPP has) initiated plant shutdown in accordance with this Technical Specification. Therefore, this is reportable under 10 CFR 50.72(b)(2)(i) Plant Shutdown Required by Technical Specifications. This is also reportable under 50.72(B)(3)(ii)(A) as a material defect in the primary coolant system that cannot be found acceptable under ASME Section XI, IWB-3600 or ASME Section XI, Table IWB-3410-1. The licensee notified the NRC Resident Inspector.
The licensee reduced reactor power to 10 percent and entered containment. The pressure boundary leak was isolated and the licensee exited the Technical Specification 3.4.13, Action B. The Technical Specification required shutdown was terminated. The licensee is making preparation to increase power to full load. The licensee will notify the NRC Resident Inspector. Notified the R1DO (Gray). | Reactor Coolant System | |
| ENS 48763 | 17 February 2013 18:45:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Pressure Boundary Leakage Discovered During Containment Walkdown | CCNPP (Calvert Cliffs Nuclear Power Plant) U-2 discovered during the Mode 3 NOP/NOT containment walkdown that 2CV-100F Pressurizer Spray valve had a pin hole leak on the leak-off welded cap. This indicates that the bellows has a leak and in addition, the packing is leaking, pressurizing the leak-off line thus the leak-off line is the pressure boundary making this pressure boundary leakage. Technical Specification 3.4.13 (Action B) requires the unit be in mode 3 in 6 hrs. Unit was in mode 3 when discovered and unit will be placed in mode 5 within 36 hrs. Currently, the plant is at 400 degrees and 950 PSIA. The licensee has notified the NRC Resident Inspector. | 05000318/LER-2013-001 | |
| ENS 48124 | 21 July 2012 16:10:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Technical Specification Required Shutdown Due to Rcs Pressure Boundary Instrument Line Leakage | (Calvert Cliffs Nuclear Power Plant) CCNPP U1 identified RCS Pressure Boundary Leakage from the instrument line to 1-PDT-123A, 11A reactor coolant pump differential pressure transmitter. Technical Specification 3.4.13, Action B was entered and requires that the Unit be placed in Mode 3 within 6 hours and Mode 5 within 36 hours. (The licensee has) initiated plant shutdown in accordance with this Technical Specification. Therefore, this is reportable under 10CFR50.72(b)(2)(i) Plant Shutdown Required by Technical Specifications. This is also reportable under 50.72(B)(3)(ii)(A) as a material defect in the primary coolant system that cannot be found acceptable under ASME Section XI, IWB-3600 or ASME Section XI, Table IWB-3410-1. The leak is from an instrument line which monitors reactor coolant pump differential pressure. It is for monitoring purposes only. The leak rate is currently about .08 gpm. The licensee has begun power reduction of the unit and is currently at approximately 73% rated thermal power. The recovery plan is to complete the unit shutdown and then enter containment and repair the leak. The leak is believed to be coming from the same instrument line that was reported to be leaking on July 17, 2012 (see EN #48116). At that time, power was reduced to approximately 10% and a containment entry was made. At that time it was believed that the leak had been isolated and full power operations were resumed. Based on containment sump pump run times and another containment entry, it was determined that the leak apparently had not been isolated. The licensee notified the NRC Resident Inspector. | ||
| ENS 48116 | 17 July 2012 04:00:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Technical Specification Required Shutdown Due to Small Rcs Pressure Boundary Instrument Line Leak | At 1715 EDT on July 17th, 2012, Calvert Cliffs Nuclear Power Plant Unit 1 identified RCS Pressure Boundary Leakage from the instrument line to 1-PDT-123A, 11A reactor coolant pump differential pressure transmitter. Technical Specification 3.4.13, Action B was entered and requires that the Unit be placed in Mode 3 within 6 hours and Mode 5 within 36 hours. The licensee has initiated plant shutdown in accordance with this Technical Specification. Therefore, this is reportable under 10 CFR 50.72(b)(2)(i) Plant Shutdown Required by Technical Specifications. This is also reportable under 50.72(B)(3)(ii)(A) as a material defect in the primary coolant system that cannot be found acceptable under ASME Section XI, IWB-3600 or ASME Section XI, Table IWB-3410-1. The licensee initially discovered the leak during an RCS leak rate surveillance that indicated RCS leakage on the order of 0.08 gpm. A video camera was used to look for the leak and was found to be on the identified instrument line. The licensee has initiated a power reduction and is currently at approximately 50%. The recovery plan at this time is to reduce power to about 10% which will reduce radiation levels in the area of the leak and isolate the instrument line and isolate the leakage. The instrument line is for a reactor coolant pump differential pressure transmitter which is for monitoring purposes only and does not provide any safety related functions. If the leak can be secured, the licensee intends to return to power. If the leak cannot be secured, the plant will shutdown in compliance with the technical specification action statement. The licensee has notified the NRC Resident Inspector. | ||
| ENS 47316 | 3 October 2011 14:35:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Technical Specification (Ts) Required Shutdown Due to Both 'A' Train Edg'S Inoperable | Dual Unit Shutdown due to T.S. 3.0.3. Unit 1 and Unit 2 'A' Train Emergency Diesel Generators (EDGs) OOS (Out-of-Service). '2A' DG OOS due to '21' Salt Water Header scheduled maintenance. '1A' DG OOS due to emergent loss of '16' Battery charger for the '1A' DG DC bus. Neither '11' nor '22' 125 VDC busses have an Operable charger. (Chargers have power, but do not have an Operable DG). " The licensee entered TS 3.0.3 at 1035 EDT and is expected to be in Mode 3 on both units at 1735 EDT unless the licensee exits T.S. 3.0.3 upon restoration of the battery charger. '0C' EDG is functional but not credited towards T.S. The salt water header is not expected to be restored until 10/4/11. There are no issues with offsite power. The NRC Resident Inspector has been notified. The licensee will notify the Calvert Control Center and plans to make a press release. CR No.: CR-2011-009860; Calvert Cliffs internal event notification number 4194. | Emergency Diesel Generator | |
| ENS 46623 | 18 February 2011 05:51:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Inactive Leak Identified on Presurizer Heater Well | During a scheduled VT-2 boric acid examination on the Unit 2 Pressurizer lower level heater penetrations a boric acid leak was identified on heater penetration N-3." The licensee characterized the leak as inactive. The approximate sizes of the round boric acid deposits are 1/32 inch and 1/64 inch. The licensee continues to investigate and plan recovery actions. The licensee notified the NRC Resident Inspector. | 05000318/LER-2011-001 | |
| ENS 45737 | 2 March 2010 12:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | 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) | Safety Relief Valve | |
| ENS 45719 | 23 February 2010 20:12:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition - Weld Leak Due to Vibrating Valve Attachment Wire | On 2/22/10, we determined that there was pressure boundary leakage on Unit 2 from the leak off line associated with a pressurizer spray valve (2-RC-220). The source of leakage associated with a boron deposit discovered earlier on 2/18/10. The boron deposit formed based on a leak in a Class 1 manual valve packing leak-off line. The leak location was determined to be in the packing leak-off pipe fillet weld to the stem retaining structure of the valve in the packing gland area. The valve (2-RC-220) is a Class 1 pressure boundary. The packing leak-off line is considered an auxiliary connection in the stem retaining structure of the valve. At this time, we believe the most probable cause of the 2-RC-220 leak-off line socket weld leak was based on two factors, a small pore in the original socket weld metal and the location of a valve tag attachment wire. The leak location coincided with the location where a valve tag attachment wire laid across the weld. It is possible this wire, vibrating against the weld, opened up a subsurface pore in the weld metal which began to leak sometime after startup from the 2009 RFO (Refueling Outage) (March 2009). Since the failure may have occurred due to a material problem that resulted in abnormal degradation of a principal safety barrier (i.e., it is necessary to take corrective actions to restore the weld's integrity), this event is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A). The valve was overhauled in the 2009 RFO and subsequently passed VT-2 inspection. The overhaul did not include any welding on the affected joint. Unit 2 is in Mode 5 to allow repairs to be made to this weld. The licensee will notify the NRC Resident Inspector. | 05000318/LER-2010-002 | |
| ENS 44006 | 25 February 2008 06:40:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Due to Leakage on a Pressurizer Heater Sleeve | During in-service inspection VT-2 visual inspection, the licensee discovered dried boric acid on pressurizer heater sleeve C-3. All other pressurizer sleeves were inspected with no additional findings. The licensee will not increase coolant temperature until the repair is made to the sleeve. The licensee has notified the NRC Resident Inspector. | ||
| ENS 41448 | 27 February 2005 20:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition of Reactor Coolant Boundary Weld | The licensee provided the following information, via facsimile, about its ongoing Alloy 600 inspection: IRE-003-574 has identified an axial flaw on line 2-LD-2004-1, 22A cold leg letdown nozzle. Engineering has evaluated the flaws using the procedures of ASME Code Section XI, IWB-3600. The calculated growth rates of the postulated axial flaws are large and do not meet the ASME Section XI Code requirements. This is reportable per NUREG-1022, Rev. 2, Sect. 3.2.4 (degraded or unanalyzed condition). 'Any event or condition resulting in the condition of the nuclear power plant, including its principal safety barriers, being serious degraded.' Unit two is in Mode 6 for refueling and a repair plan is currently being developed. The discovery of this flaw will likely result in the expansion of the inspection scope. Generic concerns for both units have not yet been fully developed by the licensee. The licensee has notified the NRC Resident Inspector. See similar event #41445 | ||
| ENS 41445 | 26 February 2005 20:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition of Reactor Coolant Boundary Weld | IRE-003-469 identified flaws (one circumferential and two axial indications) on 2-DR-2007 weld of hot leg drain line. Engineering has evaluated the flaws using the procedures of ASME Code Section XI, IWB-3600. Although the circumferential flaw was found to be acceptable in accordance with ASME section XI, IWB-3600, the calculated growth rates of the postulated axial flaws are large and do not meet the ASME Section XI Code requirements per NUREG 1022, Rev. 2 sect. 3.2.4 (degraded or unanalyzed condition). 'Any event or condition resulting in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' Currently a repair plan is being developed. The licensee indicated that they will not be able to enter mode (4) until the repair is completed (using weld overlay as the repair method) and an inspection of the repair is conducted. The licensee will perform a design engineering evaluation of extent of condition to determine the impact of this on the operating unit (Unit 1). The NRC Resident Inspector will be notified. See similar event #41448 |