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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 53239 | 1 March 2018 22:43:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 Tornado Missile Vulnerabilities | During review of protection of equipment from damaging effects of tornados, Point Beach Nuclear Plant identified a potential vulnerability for the turbine driven auxiliary feedwater pumps due to steam supply piping that is not routed through a Class 1 structure. Immediate compensatory measures were taken to mitigate the potential consequences of a tornado generated missile impact. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) and per 10 CFR 50.72(b)(3)(v)(A) and (D). The identified vulnerability is being addressed in accordance with EGM 15-002 and DSS-ISG-2016-01, enforcement discretion memorandum and interim guidance document for resolution of noncompliance with tornado-generated missile protection. The NRC Resident Inspector has been notified. | Auxiliary Feedwater | |
| ENS 52976 | 18 September 2017 22:24:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 | Control Room Envelope Not Maintained | At 1724 (CDT) on 9/18/17 during Control Room Ventilation testing Door-61, South Control Room Door, became wedged against its door stop and stuck open. Door-61 is a credited High Energy Line Break (HELB) / Fire / Flood Barrier in addition to its function to maintain the Control Room envelope. The door stop was subsequently unbolted from the floor and the door was free to close. Door-61, South Control Room Door, has since been inspected, and at 1750 (CDT), was declared functional as a HELB / Fire / Flood Barrier and Operational for purposes of maintaining the Control Room Envelope. During the 26 minutes the door was stuck open, the Control Room was in an unanalyzed condition with regards to protection from a High Energy Line Break. The licensee notified the NRC Resident Inspector. | Control Room Envelope | |
| ENS 50740 | 15 January 2015 23:04:00 | 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 | Inadequately Sealed Piping Penetrations and Postulated Flooding Event | On November 19, 2014, inadequately sealed piping penetrations were discovered in the Residual Heat Removal (RHR) piping and valve gallery walls that could have allowed a postulated flooding event to potentially impact both RHR pumps. The same configuration was identified on both Units. Following analysis of the postulated flooding leakage sources and flow paths, it was discovered that there was a time when an unacceptable volume of flood water could have entered the pipe and valve gallery areas. This condition was found to have been previously corrected on 12/16/14, following changes to Operations flood mitigation strategies and installation of seals in upstream pipeway trenches. Therefore, this issue is reported as an unanalyzed past condition that had the potential to adversely impact the RHR pumps. This condition was determined to be reportable per 50.72(b)(3)(ii)(B) and 50.72(b)(3)(v)(B). The licensee has informed the NRC Resident Inspector. | Residual Heat Removal | |
| ENS 48402 | 12 October 2012 09:20:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Reactor Power Exceeded Fsar Analyzed Value | On 10/12/12 at 0420 CDT the Unit 2 Steam Generator B Atmospheric Steam Dump Valve (ADV) spuriously opened while in automatic control. This resulted in indicated reactor power exceeding the FSAR analyzed value of 1810.8 MWt. Prompt operator action was taken and reactor power was restored to within limits in approximately four minutes. The operators placed the Atmospheric Steam Dump Controller to manual and closed the ADV successfully. This event is being reported under the criteria in 10 CFR 50.72 (b)(3)(ii)(B). All other plant systems responded as expected. The licensee has notified the NRC Resident Inspector.
This notification is being made to retract Event Notification EN #48402 which reported power exceeding the FSAR analyzed value of 1810.8 MWt due to a spurious opening of an Atmospheric Steam Dump Valve. While 1810.8 MWt is the normal PBNP (Point Beach Nuclear Plant) full power value without consideration to uncertainties, subsequent reviews show that the peak power level reached during the event was bounded by FSAR analyzed events. Additionally, the opening of the valve is an Anticipated Operational Occurrence, described and analyzed in the FSAR. The power excursion during the event was below excursions evaluated in the FSAR. The event was analyzed and plant safety was not significantly degraded. Therefore, the event does not meet the criteria of 10CFR50.72(b)(3) and NextEra Energy Point Beach retracts Event Notification EN #48402. The licensee notified the NRC Resident Inspector. Notified the R3DO (Pelke). | Steam Generator | |
| ENS 44351 | 16 July 2008 20:16:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Potential Fire Propagation Between Rooms Could Affect Appendix R Safe Shutdown | A potential exists for a fire in the South Area of the Auxiliary Feedwater (AFW) room to propagate to the Vital Switchgear (VSG) room. A fire 4 in the South Area of the AFW room could cause a short circuit in a cable that traverses the AFW room and the VSG room, causing ignition of the cable. The Point Beach Safe Shutdown Analysis assumes afire in a single fire area. A fire in South Area of the AFW Room, credits AFW pumps P-38B & 2P-29 for providing AFW to both unit's steam generators. A fire in the VSG Room, credits AFW pumps 1 P-29 & 2P-29 for providing AFW to both unit's steam generators. A fire in both the South Area of the AFW room and VSG room could potentially cause three of the four AFW pumps to be unavailable, which does not meet the requirements for Appendix R safe shutdown. The potential for a fire affecting two fire areas is being reported as an unanalyzed condition as defined by 10 CFR 50.72(b)(3)(ii)(B). Compensatory measures have been implemented. Licensee investigations are continuing. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater | 05000266/LER-2008-003 |
| ENS 43487 | 12 July 2007 15:15:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Fire Inspection Analysis of Pressurizer Porvs and Block Valves | During a review of abnormal operating procedure (AOP) 10A, Safe Shutdown-Local Control, by the NRC triennial fire inspection team, it was identified that fire damage to the reactor coolant system (RCS) power-operated relief valve (PORV) and block valve circuits as a result of a fire in the cable spreading room could also result in simultaneous damage to a block valve circuit and spurious actuation of a PORV. While the actions included in abnormal operating procedure (AOP)-10A provide reasonable assurance that positive control of RCS Inventory is maintained, these steps do not ensure that simultaneous failure of the block valve circuit and spurious operation of a PORV will not result in RCS depressurization. Therefore, a postulated fire may potentially remove the ability to fully implement the Safe Shutdown Strategy. Compensatory measures in the form of twice-per-shift fire rounds in the cable spreading room have been implemented. The licensee notified the NRC Resident Inspector. | Reactor Coolant System | 05000266/LER-2007-006 |
| ENS 43353 | 9 May 2007 21:23:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Identified Non Compliant Fire Protection Manual Operator Actions | 05000266/LER-2007-002 | ||
| ENS 42129 | 8 November 2005 14:44:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Design Basis for Long Term Cooling Not Correctly Modeled | While investigating an issue related to containment coatings and their potential to clog the containment sump strainers, errors were discovered in the calculations that were used as the basis for responding to GL98-04. The errors involved the improper application of a correlation that was used to derive head loss across a screen that was assessed to be partially fouled with debris and the incorrect application of the results to a partially submerged screen that would be susceptible to air intrusion. Further investigation revealed that the flow path for a partially blocked strainer was not correctly modeled for the containment sump strainer and containment sump valve (SI-850A&B). The SI-850 valves have a rising disk and are located inside and at the bottom of the containment sump strainer. These errors in the modeling fidelity potentially impact the analytical basis for demonstrating compliance with the acceptance criteria in 10 CFR 50.46 (b)(5), Long-term Cooling. Immediate actions - Operability analyses were performed. These operability analyses demonstrated that adequate NPSH would be available to the ECCS pumps to ensure long-term cooling. Long-term action - The containment sump strainer will be modified, as committed to in Point Beach letter NRC 2005-0109, "Nuclear Management Company Response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors, for Point Beach Nuclear Plant," dated September 1, 2005. This modification will result in a larger strainer surface area and a greater clearance in the vicinity of the SI-850 valves. These modifications will be supported by design analysis and testing that will demonstrate the strainers comply with the long-term cooling capability requirement of 10 CFR 50.46 (b)(5). The licensee notified the NRC Resident Inspector. | 05000266/LER-2005-006 | |
| ENS 42109 | 2 November 2005 05:00:00 | 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 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Tech Spec Required Shutdown Due to Degradation of Containment Coatings | On November 2, 2005 at approximately 00:00 Central Standard Time (CST), Point Beach Nuclear Plant (PBNP) Unit 2 commenced a reactor shutdown required by Technical Specification 3.0.3. During a review of the containment coatings in both Unit 1 & 2 containments, it was discovered that the containments have not been maintained with the analysis of record performed by Sergeant and Lundy (S&L). The S&L analysis performed for Unit 2 was based on the known condition of coatings when the analysis was performed. There was no explicit margin for further degradation. Subsequent discoveries of degraded or unqualified coatings cannot be accommodated by the existing analysis as written. An Operability Recommendation (OPR) was performed for Unit 2 and approved on 10/30/05 at 2000. Following this OPR, a further review of containment coatings in the Unit 2 containment was performed and showed a potential for approximately 11 square feet of unqualified coatings (in) the Zone of Influence (ZOI) for the containment sump. The OPR allowed for a maximum of 5.68 square feet of loose material in the ZOI. A Unit 2 containment walk-down was performed on the evening of November 1, 2005. This revealed that the unqualified coatings in the ZOI were approximately 11 square feet. This information placed Unit 2 in an unanalyzed condition, which lead the operators to enter Technical Specification 3.0.3 at 2300 on November 1 due to both trains of Emergency Core Cooling System (ECCS) being declared inoperable for sump recirculation capability. Actions are currently underway to remove enough unqualified coatings to be within the assumptions made in the OPR and restore Containment Sump recirculation capability. When this is completed, the technical specification shutdown will be terminated, and Unit 2 will make preparations to return to full power. Unit 1 is currently in Mode 5 and ECCS is not required. However, the condition is also applicable to Unit 1 containment. Actions have been underway since the identification of the original issue to remove unqualified containment coatings. The Plant Manager has placed a hold on entering Mode 4 on Unit 1 pending completion of corrective actions. Presently there are 2 workers and a Radiation Protection technician inside containment. The licensee said that workers will go inside containment and remove the degraded coating. This will take approximately 45 minutes and have a total exposure to personnel of 85 millirem. The licensee notified the NRC Resident Inspector.
At 01:06 CST the licensee exited from Technical Specification 3.0.3. requirements and plans to hold power on Unit 2 at 97% power pending further assessment and evaluation. The licensee will notify the NRC Resident Inspector. Notified the R3DO (Kozak).
On November 2, 2005, at 01:13 (ET) PBNP submitted Emergency Notification #42109, to report a TS required shutdown due to potential of an unanalyzed condition that significantly degrades plant safety and an event or condition that potentially could have prevented the fulfillment of a safety function. The condition related to the discovery of degraded containment coatings in the zone of influence (ZOI) for the containment sump. A subsequent evaluation concluded that the degraded coatings would not have significantly affected sump recirculation flow capability. Additionally, the zone of influence was identified to be approximately one-third that assumed for the design-basis calculation. Based on the conservatism in the sump blockage analysis, the degraded coatings in the Unit 2 containment within the original ZOI did not affect the conclusion that equipment needed for accident mitigation would have operated as designed. Therefore, the Emergency Notification made on November 2, 2005, documenting that this condition created the potential of an unanalyzed condition that significantly degrades plant safety, and potentially could have prevented the fulfillment of a safety function, is retracted. The Technical Specification required shutdown is also retracted. The licensee notified the NRC Resident Inspector. Notified R3DO (H. Peterson). | Emergency Core Cooling System | |
| ENS 42020 | 27 September 2005 15:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Faults Have Electrical Current in Excess of the Maximum Listed Interrupting Ratings. | NMC (Nuclear Management Company) has identified certain equipment in the PBNP electrical distribution system that will not assure, under certain conditions, interruption of a three phase bolted fault short circuit. These postulated faults have electrical current in excess of the maximum listed interrupting ratings for designated circuit breakers and associated bus bar bracing. This condition affects the 13.8 Kv, 4.16 Kv, and 480 V power panels, motor control centers (MCCs), and switchgear. Although the probability of bolted faults is considered low, the Point Beach bolted fault analysis is based on the worst case assumption of three phases firmly tied together and grounded. A postulated bolted fault itself would only impact equipment in a single safety train. However, the PBNP Appendix R analysis relies on breaker coordination and fault current interruption to prevent loss of safe shutdown equipment due to common enclosure/power supply associated circuit concerns. The degraded breaker coordination resulting from a bolted fault condition does not satisfy the requirements of the Appendix R safe shutdown analysis. This condition is reportable because the PBNP Appendix R analysis is based on the occurrence of a single fire in a single fire area. The postulated condition could result in a loss of safe shutdown equipment functionality beyond that previously analyzed. Compensatory measures (i.e., fire rounds - 6 times per day) have been implemented for cases where the unprotected cable length was routed beyond the original fire area. As part of the long-term corrective action, transformer tap setting changes to reduce bus voltages are being evaluated. The NRC Resident Inspector was notified of this event by the licensee.
This is a supplemental emergency notification based on additional information identified regarding degraded voltage conditions at PBNP. On 09/27/2005 NMC reported a condition where certain equipment in the PBNP electrical distribution system would not assure, under certain conditions, interruption of a three-phase bolted fault short circuit. Licensee Event Report (LER) 266 & 301/2005-005-00 was subsequently submitted on November 18, 2005. The original Event Notification Report was associated with bolted fault conditions that potentially resulted in additional unanalyzed fire losses due to direct fire damage or uncleared faults on associated circuits. The synopsis of the LER addressed these issues, and also identified: 1. A non-conservative Technical Specification for degraded voltage time delay relay settings and their setting tolerance range in calibration procedures that could have resulted in certain safety system motors and switchgear tripping on overcurrent. Such an event could have prevented the fulfillment of the motors' safety function to mitigate the consequences of an accident. 2. Under a design basis loss of coolant accident concurrent with a reduced voltage condition, safety-related motors and switchgear may trip their protective devices on overcurrent without the degraded voltage relays being actuated. Affected equipment included certain safeguards 480V AC switchgear, 480 V AC motor control centers, both auxiliary feedwater pump motors, and one component cooling water pump motor. Corrective actions for the above issues included placing calibration procedures on administrative hold, implementation of compensatory measures consisting of fire rounds for affected zones, and administrative controls to assure that a more restrictive limit for the degraded voltage allowable value was in place for the affected Technical Specification (s), as well as implementing administrative controls on the management of 480 V loads. Long-term corrective actions are evaluation and implementation of analytical changes resulting from the completed analysis, plant modification changes as needed to address minimum bus voltage and submittal of a license amendment request. Additional reviews into the extent of condition of this issue have revealed additional potential concerns associated if a station battery charger load test is conducted under reduced or degraded grid voltage conditions. If a battery charger load test is conducted during a degraded grid voltage condition and a loss-of-coolant accident occurs with a coincident safety injection signal but a loss of off-site power does not occur, the battery chargers are not stripped from their alternating current supply. The additional potential electrical load on the AC supply has not been analyzed. Compensatory measures, in the form of administrative controls associated with battery charger testing, are being implemented. A supplement to LER 266(301)/2005-005-00 will be submitted. The senior resident inspector has been informed of this supplemental report. Subsequent conversations between the Headquarters Operations Officer and the Shift Manager and Shift Technical Advisor at Point Beach have confirmed that this is not an "emergency" notification as quoted in the first paragraph of the event update. This is an event notification only. R3DO (Stone) notified.
On April 6, 2006, NMC provided a supplemental notification to EN# 42020 regarding a condition where certain equipment in the PBNP electrical distribution system would not assure, under certain conditions, interruption of a three-phase bolted fault short circuit. The supplemental notification resulted from additional reviews into the extent of condition of this issue, which indicated additional potential concerns associated with a station battery charger load test being conducted under reduced or degraded grid voltage conditions. A subsequent evaluation of this supplemental condition concluded that its significance is minor. The April 6, 2006, supplemental notification concerned 480 VAC vital bus loading due to battery charger load during charger testing and battery charger load while recovering a battery after battery discharge testing conducted under reduced or degraded grid voltage condition. The potential impact of this testing was evaluated to be associated only with the D-109 station battery charger and not several chargers as originally determined. An evaluation of the risk significance of this issue indicates that because of the combination of simultaneous events that would need to occur, the significance is minor. Because of the minor risk significance and that the issue concerns only a single piece of equipment (D-109 station battery charger), the guidance in NUREG-1022 indicates that this condition does not meet the reportability criteria in 10 CFR 50.72(b)(3)(ii)(B). Therefore, the supplemental notification made to EN #42020 on April 6, 2006, is hereby retracted. The underlying condition will continue to be addressed through the plant's corrective action process. The licensee notified the NRC Resident Inspector. Notified R3DO (M. Phillips). | Auxiliary Feedwater | |
| ENS 41856 | 20 July 2005 09:51:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Minimum Recirculation Valves Will Not Automatically Open in Local Operating Mode | While performing a test start of P-38A Motor Driven Auxiliary Feedwater Pump utilizing the local control station, it was discovered that AF-4007, the mini-recirculation valve for P-38A would not automatically open in the local mode of operation. The mini-recirculation valve provides a minimum flow path for pump operation to prevent pump damage. After this condition was discovered, a review of procedures associated with initiating safe shutdown via local operation disclosed that the procedure did not address manually opening the mini-recirculation valve prior to local pump start. As a result, pump damage could occur due to no flow though the pump prior to aligning a flow path into a steam generator. This condition is also applicable to AF-4014, the mini-recirculation valve for P-38B Motor Driven Auxiliary Feedwater Pump. P-38A and P-38B Motor Driven Auxiliary Feedwater pumps would only be used in a safe shutdown local control condition if steam generator level could not be maintained using the normal means of 1P-29 and 2P-29 Turbine Driven Auxiliary Feedwater Pumps. The station has taken compensatory actions to brief the operators about the condition. Additionally, temporary procedure changes are in progress to direct the operator to manually open each Motor Driven Auxiliary Feedwater Pump's associated mini-recirculation valve prior to a local start of the P-38A or P-38B Motor Driven Auxiliary Feedwater Pump. The associated mini-recirculation valve will operate when P-38A or P-38B Motor Driven Auxiliary Feedwater Pump is started from the control room. Follow up testing has verified P-38A Auxiliary Feedwater Pump operability. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater | |
| ENS 41758 | 8 June 2005 22:24:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition - Fire Organizational Plan No Longer Aligned with Safe Shutdown Analysis | It has been identified that during a revision change between Revision 5 of FOP 1.2 (July 20, 2004), and Revision 6 of FOP 1.2 (November 1, 2004), a number of omissions of safe shutdown equipment occurred. Because of these omissions, FOP 1.2 is no longer aligned with the Safe Shutdown Analysis. As a result, some of the manual actions that would be necessary to accomplish safe shutdown are no longer identified in FOP 1.2. FOP 1.2, Fire Organizational Plan is used by Operations to provide guidance on plant operation for fires within safe shutdown areas. Included in this guidance is a list of safe shutdown equipment affected by a postulated fire and the manual actions that may be required to compensate for the fire damage. It has been determined that this condition is reportable because the missing procedural guidance may result in safety significant operator actions not being performed which are credited in the Safe Shutdown Analysis. Corrective actions have been entered into the Corrective Actions program, and a procedure revision to correct FOP 1.2 is currently in progress. The licensee notified the NRC Resident Inspector. | ||
| ENS 41754 | 7 June 2005 23:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Appendix R Safe Shutdown Strategy Deficiency | This report is a result of an ongoing evaluation of a previously identified deficiency with the Appendix R Safe Shutdown Strategy with respect to use of charging pumps for a fire in Fire Area A06, 1B-32 480V MCC area. This issue was originally identified on April 8, 2005 during work on the Fire Probabilistic Risk Assessment Project. This was entered into the Point Beach Corrective Action Program, and compensatory fire rounds were initiated. A postulated fire in the east side of the MCC 1 B-32 could damage both the power and the control cables for charging pumps 1 P-2A and 1 P-2B, and the control cables for redundant charging pump 1 P-2C. The resultant condition of the 1 P-2C charging pump control circuit could prevent operation of this pump as directed in the Safe Shutdown Analysis and FOP 1.2, Potential Fire Effected Safe Shutdown Components. The condition exists as the result of a lack of physical cable separation for power and control cables for the Unit 1 charging pumps. An Operability recommendation was performed for this issue and determined that the condition was Operable but Non-Conforming. Based on a continuing review of further information related to this condition, it has been determined that this condition is reportable based on the resultant effect on available charging pump capability. The condition could have resulted in losing the availability of all but a single charging pump operating at slow speed, which would not provide sufficient reactor coolant pump seal cooling, and thereby degrade the level of plant safety. The identified condition requires a revision to the Safe Shutdown Analysis. Plant fire mitigation procedure changes to ensure adequate charging pump capability, and reactor coolant pump seal cooling have been made. The Safe Shutdown Analysis will be revised. The Resident Inspector will be notified. |