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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 51837 | 30 March 2016 21:15:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Due to Discovery of Pressure Boundary Leakage | On March 30, 2016, at 1715 EDT, with the Unit shutdown and in Mode 6 for refueling, evidence of leakage was identified on a 3/4-inch flexible braided piping connection on Reactor Coolant Pump (RCP) 1-1, and this issue was determined to be reactor coolant system pressure boundary leakage. This flexible piping is for RCP 1-1 first stage seal cavity vent line, and is categorized as ASME Section III Class 2 piping. The leakage was identified due to the discovery of a small amount of boric acid (approximately 1/2 teaspoon) on the welded end connection of the flexible piping. No active leakage was identified at the time of discovery with the Reactor Coolant System depressurized and approximately 110 degrees F. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.13, 'RCS Operational Leakage,' does not apply in the current plant condition (Mode 6). The cause and resolution of the leakage are under evaluation. This event is reportable within 8 hours per 10 CFR 50.72(b)(3)(ii)(A). The NRC Resident Inspector has been notified. | Reactor Coolant System | 05000346/LER-2016-003 |
| ENS 49163 | 1 July 2013 15:03:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Due to Rcp Seal Line Pressure Boundary Leakage | On 07/01/2013 at 1103 (EDT), inspection personnel identified leakage from a 3/4 inch small-bore pipe socket weld for RCP 1-2 first stage seal cavity vent line. At current Reactor Coolant System conditions (Normal Operating Pressure and Temperature), the leak rate is approximately 8 to 9 drops per minute. The plant entered Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.13, 'RCS Operational Leakage', Condition B. Repairs to the line are being evaluated. This pressure boundary leakage is reportable per 10 CFR 50.72(b)(3)(ii)(A). The NRC Resident Inspector has been notified. The LCO requires the licensee to be in Mode 5 by 2303 EDT on 7/2/13. | Reactor Coolant System | 05000346/LER-2013-002 |
| ENS 48000 | 6 June 2012 23:56:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Due to Discovery of Pressure Boundary Leakage | On June 6, 2012, at 1956 EDT, with the Unit shutdown for refueling, leakage was identified from a 3/4-inch weld during Reactor Coolant System (RCS) walkdown inspections. The leakage amount was approximately 0.1 gpm pinhole spray. During the performance of MODE 3 engineering walkdown inspections in accordance with procedure DB-PF-03010 (ASME Section III, Class 1 and 2), with the RCS at Normal Operating Temperature and Pressure, a pressure boundary leak was identified on the Reactor Coolant Pump (RCP) 1-2 1st seal cavity vent line upstream weld of 3/4 inch small bore pipe socketweld at a 90 degree elbow between the RCP pump and valve RC-407 (1st Seal Cavity Vent Isolation). The plant was in MODE 3 at Normal Operating Pressure and Normal Operating Temperature (NOP/NOT) for the inspections. The plant entered Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.13, 'RCS Operational Leakage,' Condition B and procedure DB-OP-02522. 'Small RCS Leaks,' abnormal operating procedure. Plant cooldown to comply with LCO 3.4.13, Condition B, Required Action B.2 is in progress. The cause and resolution are under evaluation. This event is reportable within 8 hours under 10CFR50.72(b)(3)(ii)(A). The NRC Resident Inspector has been notified. This condition has been documented in the Davis-Besse Corrective Action program as Condition Report 2012-09381. The plant is required to be in MODE 5 within 36 hours. | Reactor Coolant System | 05000346/LER-2012-002 |
| ENS 45764 | 13 March 2010 02:43:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Control Rod Drive Mechanism Nozzle Indications | On March 12, 2010, during the Davis-Besse Nuclear Power Station Unit No. 1 (DBNPS) refueling outage, the documented results of planned ultrasonic (UT) examinations performed on the Control Rod Drive Mechanism (CRDM) nozzles penetrating the reactor vessel closure head (RVCH) identified that two of the nozzles inspected to date did not meet the applicable acceptance criteria. Each of these two nozzles have similar indications that appear to penetrate into the nozzle walls from a lack of fusion point at the outer diameter of the nozzle and the J-Groove weld. Leak path detection was performed on both nozzles with the results showing no leak path. Bare metal visual examinations are scheduled to be performed on all nozzles to determine if there is any pressure boundary leakage. Both indications will require repair prior to returning the vessel head to service. There are sixty-nine nozzles and all will be subject to these UT inspections. It is important to note that this notification is being provided prior to the completion of all of the required UT and VT (Visual Tests) examinations for these two nozzles and the remaining nozzles. The indications were detected with a blade probe used to detect axially-oriented indications. The remaining probe that is used to complete the UT examination in each of these two penetrations is a blade probe that is used to identify circumferentially-oriented indications. The examinations are being performed to meet the requirements of 10CFR50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1, to identify potential flaws/indications well before they grow to a size that could potentially affect the structural integrity of the reactor vessel head pressure boundary. The licensee notified the NRC Resident Inspector and will notify the State of Ohio and both Lucas and Ottawa Counties.
On March 13, 2010, additional Control Rod Drive Mechanism nozzles (4 and 59) were identified that did not meet the applicable acceptance criteria. The indications on these nozzles are similar in nature and location to the indications previously reported for nozzles 28 and 33. These indications will also require repair prior to returning the vessel head to service. Like the two previously reported nozzles, this notification is being provided prior to the completion of all of the required examinations for the nozzle and the remaining nozzles. The indications were detected with a blade probe used to detect axially-oriented indications. The remaining probe that will be used to complete the examination in this penetration is a probe sensitive to circumferentially-oriented indications. Additionally, during the bare metal visual examination of the outer surface of the reactor vessel closure head, boric acid deposits were found at CRDM nozzles 4 and 33 that are indicative of primary water leakage. The visual examination of the RVCH is continuing. The licensee notified the NRC Resident Inspector, State of Ohio, and local government. Notified the R3DO (Phillips).
Update to Davis-Besse event #45764 initially reported 3/13/2010 at 0445 and adding an additional 10 CFR 50.72 reporting criteria. On March 15, 2010, the FirstEnergy Nuclear Operating Company is issuing a press release regarding the Davis-Besse Nuclear Power Station Unit No. 1 reactor vessel head (RVCH) Control Rod Drive Mechanism nozzles that have indications that do not meet the applicable acceptance regulatory criteria. The results of the ongoing planned ultrasonic (UT) examinations performed on these nozzles penetrating the RVCH have identified indications that will require repair prior to returning the vessel head to service. Currently, there are twelve nozzles which will be repaired, two of which have evidence of primary water leakage found during the bare metal visual examination of the outer surface of the reactor vessel closure head. There are sixty-nine nozzles and all are subject to these UT inspections. It is important to note that this media release is being provided prior to the completion of all of the required examinations for these and the remaining nozzles. The licensee notified the NRC Resident Inspector and will be notifying the State of Ohio and both Lucas and Ottawa Counties. Notified R3DO (Monte Phillips) | Control Rod | |
| ENS 43880 | 4 January 2008 07:50:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Pressure Boundary Leakage Discovered During Refueling | While in Mode 6 with the Reactor Head in place and fuel in the Reactor Vessel, weld overlay was in progress on the common Decay Heat Suction line from the Reactor Coolant System. The Reactor Coolant System is in a drained condition with RCS level approximately 24 inches above the Hot Leg Centerline. During the first weld pass on the suction line weld located in Containment, the welding operator noticed water seeping from the weld. Visual inspection revealed a small leak. The leakage is too small to quantify the leakrate. Surface examinations prior to welding revealed no abnormal conditions or leakage. Welding operations on this weld have been suspended. Planned Refueling activities continue to defuel the reactor vessel. One train of Decay Heat Removal remains in service providing core cooling. The second train is aligned in standby for Decay Heat Removal. Both Trains of Decay Heat Removal have been declared inoperable due to this leak, however both trains of Decay Heat Removal remain functional. The licensee notified the NRC Resident Inspector. | Reactor Coolant System Decay Heat Removal | 05000346/LER-2008-001 |
| ENS 42437 | 21 March 2006 21:08:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition - Indication on Primary Coolant System Pressure Boundary Weld | On March 18, 2006, with the Davis-Besse Nuclear Power Station in Mode 6 (Refueling), an axial indication was discovered on the Reactor Coolant Pump 1-1 Cold Leg Drain Line Nozzle to Elbow Weld. At the time of discovery, the indication was determined to be at least 0.25 inches in length, but the through-wall depth of this indication could not be determined. Further evaluation of this indication is in progress by an offsite engineering organization. Based upon preliminary evaluation, it has been determined that this indication cannot be found acceptable under ASME Section XI. 10CFR50.72(b)(3)(ii)(A) requires reporting of any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. NUREG-1022, 'Event Reporting Guidelines - 10CFR50.72 and 10CFR50.73' lists as an example of such degradation 'Welding or material defects in the primary coolant system which cannot be found acceptable under ASME Section XI, IWB-3600, 'Analytical Evaluation of Flaws' or ASME Section XI, Table IWB-3410-1, 'Acceptance Standards.' Based upon the preliminary evaluation by the offsite engineering organization, the axial indication is being conservatively reported per 10CFR50.72(b)(3)(ii)(A) as a serious degradation of a principle safety barrier. The plant is currently defueled, and plans are underway to determine the necessary repair (to) this indication. The licensee will notify the NRC Resident Inspector. | 05000346/LER-2006-002 |