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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56959 | 11 February 2024 15:11:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Primary Containment Degraded | The following information was provided by the licensee via email: At 1011 EST on 02/11/2024, during a refueling outage at 0 percent power, while performing local leakage rate testing (LLRT) of the feedwater check valves (part of the containment boundary), it was determined that the Unit 1 primary containment leakage rate did not meet 10 CFR 50 Appendix J requirements specified in Technical Specification 5.5.12. 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 Inspector has been notified. | Feedwater Primary containment | |
| ENS 56342 | 7 February 2023 22:38:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Primary Containment Degraded | The following information was provided by the licensee via email: At 1738 EST on 02/07/2023, while in mode 5 at 0 percent power, it was determined during local leak rate testing (LLRT) that the primary containment leakage rate exceeded the allowable limit defined in 10 CFR 50, Appendix J, 'Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.' Both primary containment isolation valves in a penetration failed LLRT requirements which represents a failure to maintain primary containment integrity. 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 Inspector has been notified. | Primary containment | |
| ENS 54463 | 4 January 2020 16:09:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Primary Containment Degraded | At 1109 (EST) on 01/04/2020, it was determined that the primary containment leakage rate did not meet value La, defined in 10 CFR 50, Appendix J, 'Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors'. An additional, tested valve has been closed to maintain leakage below maximum allowable leakage, La. 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 Inspector has been notified. | Primary containment | |
| ENS 52567 | 20 February 2017 04:23:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Containment Penetration Exceeded Allowable Leakage | On 2/19/2017 at 2323 EST, during LLRT (local leak rate test) testing per 42SV-TET-001-2, 2T48F320 exceeded the maximum allowable leakage limit. The companion isolation valve in the same line (2T48F319) had previously failed LLRT. The failure of the 2T48F320 represents a failure of the 2T23X26 penetration to maintain primary containment integrity. This event is reportable per 10CFR50.72(b)(3)(ii)(A) since the failure of the 2T23X26 penetration caused primary containment leakage to exceed La (allowable leakage) and thus represents a degraded principle safety barrier. CR (condition report) 10333178. NRC Resident Inspector has been notified. 2T48F319 and 2T48F320 are 18 inch dampers. This event places the licensee in a Technical Specification limit that requires the dampers to be repaired and pass LLRT prior to the plant entering Mode 3. | Primary containment | |
| ENS 51742 | 16 February 2016 11:31:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Weld Overlay Flaw Did Not Meet Acceptance Criteria | As part of the upgrade to the full structural weld overlays (FSWOL) per NRC-approved ISI (In-service Inspection) Alternative HNP-ISI-ALT-15-01, the surface of the existing weld overlay for the 1E Recirculation weld (1B31-1RC-12BR-E-5) was ground to prepare the surface for receipt of a new Alloy 52M overlay. Upon performance of the subsequent liquid penetrant testing examination, it was discovered that the as-found condition of the flaw did not meet acceptance criteria. It was determined that the flaw constituted a defect in the primary coolant system that could not be found acceptable per ASME Section XI. Therefore, this event notification is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A). Evaluation of the flaw found in the weld overlay suggests that the non-satisfactory liquid penetrant surface examination is a result of the propagation of the original flaw that was found on the 1E Recirculation Loop Piping. The flaw is axial in nature and therefore there is no impact on structure integrity degradation. No leakage is currently present or was seen during the previous operating cycle from this flaw. There is also reasonable assurance that there was not a breach in the credited RCS (Reactor Coolant System) boundary during the previous operating cycle. As part of the corrective action to fix the flaw, the 1E Recirculation weld will be upgraded to full structural weld overlay per the NRC-approved ISI Alternative. The Licensee notified the NRC Resident Inspector. | ||
| ENS 48795 | 28 February 2013 16:55:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Closed Cooling Water Isolation Valve Fails Local Leak Rate Test | On February 28, 2013, at 1155 EST, with the unit in Refueling Mode, a determination was made that reactor building closed cooling water (RBCCW) isolation valve (2P42-F051) exceeded its acceptance criteria for designed leakage when performing local leak rate testing. Diagnostic testing confirmed that all the leakage from its test boundary is going through this valve with an 'as found' leakage of >200,000 sccm at 32.87 psig. This valve is the outboard isolation barrier for that affected primary containment penetration with the inboard barrier being the RBCCW system itself as a 'closed' system. Previous practice is to conservatively include any leakage through this valve when performing as found leak rate tests as part of the primary containment leakage summary or as part of 0.6La. This is considered conservative since the RBCCW system inside containment is assumed to remain intact following a design basis accident (DBA) loss of coolant accident (LOCA). If the closed system remains intact there is no path for leakage to exit primary containment through this system. Since the past practice is to include the 'as found' leakage through this valve as part of 0.6La and since the 'as found' leakage would result in exceeding La, this condition is being considered a condition that results in the principal safety barriers being seriously degraded. This leakage would represent a loss of the containment function since the leak rate exceeded the Technical Specification limiting condition for operation (LCO) for primary containment. Further investigation is underway to determine if leakage through this single containment barrier is required to be included in the Appendix J primary containment leakage summary, since it is associated with a closed system. The licensee notified the NRC Resident Inspector.
Further investigation revealed an Appendix J exemption that was granted for the Hatch Unit 2 local leak rate test (LLRT) program in the 1978 time frame that specifically addressed the primary containment penetration that has primary containment isolation valve (PCIV) 2P42-F051 as its outboard barrier and the 'closed' system as its inboard barrier. The exemption recognizes that this system is designed to be intact and water filled post-LOCA, allows testing of the of the PCIV with water and states that the leakage through this PCIV is not included in the 0.6 La total. Since the Hatch RBCCW system supplying components inside primary containment is a 'closed' system and remains intact and water filled post-LOCA, there is no leakage path from primary containment through this RBCCW penetration. Since no leakage from primary containment can occur through this penetration in its 'as found' state, this condition does not represent a condition that seriously degrades a principal safety barrier. As such this condition has been determined to no longer meet reporting requirement 10CFR50.72(b)(3) and is therefore not reportable. Based on this information the previous notification is being retracted. The licensee notified the NRC Resident Inspector. Notified R2DO (O'Donohue). | Primary containment | |
| ENS 48772 | 20 February 2013 10:28:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition - Containment Penetration Fails Local Leak-Rate Testing | On February 20, 2013, at 0538 EST, local leak-rate testing (LLRT) of the 'A' feedwater check valves 2B21-F010A and 2B21-F077A revealed that neither valve would pressurize. Based on this information this line would not remain water filled post-LOCA and would result in the 'as found' minimum pathway leakage exceeding the limiting condition of operation (LCO) for Technical Specification 3.6.1.1. The cause for the LLRT failures will be determined and required corrective maintenance will be performed and valves successfully tested during the current refueling outage. The licensee notified the NRC Resident Inspector.
Subsequent investigation into the reported LLRT failure revealed that the initial LLRT performed on feedwater check valve 2B21-F010A was not considered an acceptable test, since that LLRT was not representative of the 'as found' condition of this check valve. The test volume for this valve had been slowly filled such that the check valve did not have the normal expected differential pressure across the valve disc to achieve normal check valve seating. After draining the test volume and refilling it by allowing the test volume to gravity fill from the reactor pressure vessel, the expected differential pressure across the valve disc occurred and seated the disc in such a way that it was more representative of the 'as found' condition for the check valve. An LLRT was then performed with a leakage of 50 accm (actual cubic centimeters per minute) against an acceptance criterion of 194 accm. No maintenance or operation of the check valve had occurred between the initial invalid test and the subsequent test performed with the disc in its 'as found' condition. An engineering evaluation was performed that documented the acceptability of using this means for establishing the test volume for feedwater check valves 2B21-F010A and 2B21-F010B for the 'A' and 'B' loops of feedwater, respectively. This engineering evaluation concluded that establishment of the required test volume in the manner described for primary containment penetration 9A satisfies the Hatch LLRT program requirements and that the leakage acceptance criterion for feedwater check valve 2B21-F010A in its 'as found' state was satisfied. The 2B21-F077A valve will be retested at a later date. Based on this information, the LLRT of this check valve in its 'as found' state was successful which actually resulted in successful minimum pathway leak rate test results for primary containment penetration 9A. These conclusive test results clearly indicated that the initial test results were incorrect and the 'as found' condition of this penetration isolation capability did not represent a significant degradation of a principal safety barrier as described in 10CFR50.72(b)(3)(ii)(A). For these reasons Notification # 48772 is being retracted. The licensee notified the NRC Resident Inspector. Notified the R2DO (McCoy). | Feedwater Primary containment Reactor Pressure Vessel | |
| ENS 44038 | 6 March 2008 09:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Pin Hole Leak Discovered in 1 Inch Instrument Line During Rpv Pressure Test | With Unit 1 in Mode 4 for a planned refueling outage, a pin hole leak was discovered on a 1 inch line between the 'A' Main Steam Line (MSL) and MSL flow instrument condensing chamber (1B21-D006B) in a weld at a 45 degree elbow. Leakage was identified as approximately 2 gallons per hour (GPH). This leakage was identified during RPV pressure test while test pressure was 1050 psig. This elbow is near the 1B21-D006B condensing chamber and is located in the Unit 1 drywell (primary containment). This item constitutes a primary coolant boundary leak discovered while shutdown. The licensee notified the NRC Resident Inspector. | Primary containment Main Steam Line | |
| ENS 44005 | 23 February 2008 21:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Unacceptable Flaw in Rpv Cap-To-Pipe Weld | At 1600 EST, it was determined by a phased array ultrasonic examination that an unacceptable flaw existed in reactor pressure vessel penetration N9. This penetration leads to a capped line and the flaw is in the cap-to-pipe weld. The flaw is 2.3 inches in length and reaches a maximum depth of 43.6 percent through-wall on a 5.75 inch OD pipe. The flaw was discovered during the routine ISI (In-service Inspection) examination of this penetration. The flaw has been found unacceptable per paragraph IWB-3514.4 of the 2003 Addenda of ASME Section XI and is therefore reportable. Upon discovery of the flaw, all penetrations of this type were examined with no further findings. A weld overlay repair is planned and should be completed by 3/6/08. The licensee notified the NRC Resident Inspector. * * * UPDATE FROM J. ANDERSON TO P. SNYDER ON 3/20/08 AT 0943 * * * On 2/23/2008 event number 44005 was made to report an unacceptable flaw in reactor pressure vessel penetration N9. The assumption made at that time was that the flaw seriously degraded the affected control rod drive (CRD) return line that was capped. This was based on the fact that the flaw was unacceptable from an ASME Section XI perspective and therefore reportable. Based on this assumption the event was reported in accordance with the following reporting requirement. '10 CFR 50.72 (b)(3)(ii)(A) Any event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded;' The depth of the flaw was initially reported to be 43.6% through wall and the thickness of the wall was measured to be 0.75 inches. Phased array ultrasonic examinations in accordance with Appendix VIII of Section XI were used to fully interrogate and characterize the circumferentially oriented flaw. The final dimensions were 2.3 inches in length on the inner diameter and a maximum depth of 60% through wall. A flaw evaluation of the Unit 1 CRD return line nozzle cap weld was subsequently performed by Structural Integrity Associates, Inc., to evaluate the 'as-found' condition and to determine the implications of the flaw on the affected system at the time of shutdown. The evaluation considered the flaw size and the appropriate stresses for operation. No crack growth was required to be assumed since the 'as-found' flaw was compared against the allowable flaw size which was used to determine the acceptability of the flaw during plant operation. Since the weld is associated with the capped CRD return line nozzle, the applied stress at the affected location was due to pressure loading only. The flaw evaluation concluded that the requirements of Section IWB-3640 were satisfied for a flaw less than 75% through-wall. The 'as-found' depth of 60% through wall is less than the allowable depth of 75%. Based on this information, the Structural Integrity Associates, Inc. evaluation demonstrated that the flaw was acceptable per the ASME Code Section Xl, 2001 Edition through 2003 Addenda requirements. Since the as- found flaw depth at the N9 weld is less than the allowable flaw depth, the required safety factors were met at all times during plant operation. Based on this updated information the conclusion has been reached that this flaw did not seriously degrade the plant or its principal safety barriers. Since the condition does not meet a reporting requirement, this notification serves to retract Notification # 44005 made on 2/23/2008. The licensee notified the NRC Resident Inspector. | Reactor Pressure Vessel Control Rod | |
| ENS 43224 | 9 March 2007 05:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Main Steam Line Flow Instrument Line Pressure Boundary Leakage | With Unit 2 in Mode 4 for a planned refueling outage, a one-inch socket weld elbow on an instrument line off the 'D' main steam line was identified as leaking water. Leakage was quantified as less than 20 drops per minute (with the main steam lines full of water). This elbow is on a steam line flow instrument (just below its condensing chamber) and is located in the Unit 2 drywell (primary containment). This item constitutes a primary coolant boundary leak discovered while shutdown. A pressure test is planned to be performed after the repair of the instrument line elbow is completed. The licensee notified the NRC Resident Inspector. | Primary containment Main Steam Line | |
| ENS 42372 | 23 February 2006 23:50:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition of Shroud Tie Rods | While performing Unit One (In-Vessel Visual Inspection) IVVI examination of the four Shroud Tie Rods (Upper Support Horizontal Support Surface) the following results were reported: Tie Rod at 135 degree location: Crack-like indications beginning at the inner corner on both sides of the left support and extend to two thirds of the way to the outer corner with full penetration. Tie Rod at 225 degree location: Crack-like indication beginning at the inner corner on one side of the left support and extending a small portion of the way toward the outer corner. The indication is similar to that described for the shroud tie rod in the 135 degree location except that it is much less pronounced and is only on one side. Tie Rod at 45 degree location: No apparent indications present. Tie Rod at 315 degree location: No apparent indications present. One of the design criteria of the Shroud Tie Rods is to maintain zero separation between the shroud horizontal welds at 100% uprated power, assuming all of the horizontal welds (H1 thru H8) are fully cracked. The Shroud Tie Rods are also designed to maintain structural integrity of the shroud during all design basis accidents and transients. These findings bring into question the ability of these shroud tie rods to have performed their design function with the reactor in operation. This condition constitutes a serious degradation of a principal safety barrier had the unit been operating. The reactor is presently shutdown and the condition discovered does not represent an immediate safety concern for Unit 1. The extent of this condition is believed to be limited to Unit 1, since Unit 2 core shroud tie rods are made of different materials and installed in a different configuration. The licensee notified the NRC Resident Inspector.
Retraction of NRC Event # 42372: After further review and evaluation it has been determined that the eight hour call made February 23, 2006 per the guidance of 50.72(b)(3)(ii)(A) should be retracted. On 02/23/2006 at approximately 2135 EST, Unit 1 was in the refuel mode. During that time routine inspection of the Reactor Vessel Shroud Restraint Tie Rod Assemblies was in progress. The inspections revealed two cracks on the 135 degree assembly and one crack on the 225 degree assembly. In addition the mechanical preload on the 315 degree assembly was found to be below the design value. These assemblies were originally installed as a mechanical replacement of the horizontal shroud welds. An evaluation was performed of the as-found condition that considered the effects of the cracks on the upper supports and the reduced mechanical preload on the 315 degree assembly. The results of the analysis showed that sufficient compression existed for the tie rod assemblies to be considered operable in the as found condition. Inspection and evaluation of the horizontal shroud welds (original plant design) further determined that sufficient intact weld ligament existed to ensure the shroud design function was maintained without relying on the tie rod assemblies. Therefore, the structural integrity of the shroud was and is maintained for normal operation as well as all design basis accidents and transients. The licensee notified the NRC Resident Inspector. Notified R2 DO (M. Lesser) |