Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 40092 | 20 August 2003 16:20:00 | 10 CFR 73.71(b)(1), Safeguards Event | Unauthorized Access to a Vital Area | Actual individual had been granted unescorted access to a vital area. Compensatory measures immediately taken upon discovery. NRC Resident Inspector will be notified of this event by the licensee.
Based on subsequent review, the licensee has determined that there was no security violation. This event has been retracted. For additional information, contact the Headquarters Operations Officer. NRC Resident Inspector was notified of this event by the licensee. | ||
| ENS 40213 | 1 October 2003 12:00:00 | 10 CFR 26.73, Applicability | Ffd 26.73 Report | On 10/01/03 at approximately 07:00 CDT, an access authorization (AA) group employee detected the possible odor of alcohol on a contract supervisor while speaking with the individual in the Point Beach Nuclear Plant (PBNP) Training building (outside of the protected area). AA supervision was immediately notified so corroboration of the condition would be made. The AA employee and AA supervisor were unsuccessful in locating the contractor supervisor. Security was then contacted and directed to place a hold on the individual's badge in order to prevent entry to the PBNP protected area (PA). The contract supervisor entered the PA at 07:12:56. The SAS operator placed the person's badge on hold at 07:13:30. The individual attempted to enter the plant's north service building at 07:17:00. Security personnel responded and escorted the individual to the gatehouse, at which time the contract supervisor was instructed to return to the AA group. At approximately 07:35, the FFD Manager met with the individual whereupon the odor of alcohol was corroborated. The individual was escorted to Medical and given a "for cause" test. Results of the test were positive for alcohol. A confirmatory blood draw was offered to the individual, which was accepted. Results of confirmatory testing are pending. The individual was denied access to PBNP and all other Nuclear Management Company (NMC) sites. A review of the individual's work was performed. This review concluded the individual had not been engaged in safety-related work activities. The NRC Region III safeguards inspector assigned to PBNP was informed of this event at approximately 14:35. | ||
| ENS 40626 | 30 March 2004 18:15:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Safety Injection Accumulator Level Greater than Allowed by Technical Specification | Unit 2 'A' Safety Injection Accumulator level was determined to be greater than level band allowed by technical specification 3.5.1 for Safety Injection Accumulators at 1127 (CST). A power reduction in preparation for a normal reactor shutdown was commenced at 1215 in accordance with normal plant procedures. The Safety Injection Accumulator level instruments were in question due to one indicator being within the normal range and the other being off-scale high. During troubleshooting efforts, an ultra-sonic detector was used to verify 'A' Safety Accumulator tank level. The ultra-sonic detector indicated that level was above the high scale indication, and the accumulator was declared OOS due to being above the maximum accumulator volume allowed by SR 3.5.1.2 (<1136 cubic feet). The Unit 2 'B' Safety Injection Accumulator was unaffected by the event and remains operable. The Unit 2 'A' Safety Injection Accumulator level and pressure was returned to technical specification parameters at 1403. The Unit 2 shutdown was stopped with the unit at 82% power and actions are in progress to return to full power operation. The licensee notified the NRC Resident Inspector. | 05000301/LER-2004-001 | |
| ENS 40630 | 31 March 2004 13:57:00 | Other Unspec Reqmnt 10 CFR 26.73, Applicability | Non-Licensed Employee Tested Positive for Alcohol | A non-licensed supervisor tested positive for alcohol during a "for cause" test. A review of the employee's work concluded that the individual had not been engaged in safety-related activities. The individual has been denied access to Point Beach Nuclear Plant and to all other Nuclear Management Company, LLC (NMC) sites. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. | ||
| ENS 40728 | 6 May 2004 15:46:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Vessel Head Nozzle Weld Area Flaw | During performance of NDE examinations of the Point Beach Nuclear Plant (PBNP) Unit 1 reactor pressure vessel (RPV) head required by the First Revised NRC Order (EA-03-009), flaw indications were identified on Nozzle 26. The ultrasonic examination (UT) signal for nozzle 26 identified flaw indications in the J-groove weld area that extend into the CRDM tube base material. A dye-penetrant (PT) examination of the nozzle 26 CRDM J-groove material was also performed. The PT exams showed minor surface indications that required further evaluation. Following minor excavation of the weld surface, additional examinations of the J-groove surface were performed. The results of these exams indicated the existence of flaws in the weld that do not meet accepted flaw evaluation guidance. Based upon preliminary analysis, it is expected that these indications would not be found acceptable under ASME standards. Therefore, this condition represents degradation of a principal safety barrier reportable under 10 CFR 50.72(b)(3)(ii)(A). PBNP Unit 1 Nozzle 26 is planned to be repaired prior to placing the RPV Head back into service. The condition was determined to be reportable at 1046 CDT on May 6, 2004. The licensee has notified the Resident Inspectors. | Reactor Pressure Vessel | 05000266/LER-2004-001 |
| ENS 40743 | 13 May 2004 15:29:00 | 10 CFR 26.73, Applicability | Non-Active Licensed Operator Tested Positive for Alcohol | A non-active licensed employee was determined to be under the influence of alcohol during for cause testing. The employee's access has been suspended to the Point Beach Nuclear Plant and all other Nuclear Management Company, LLC (NMC) sites. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. | ||
| ENS 40754 | 15 May 2004 16:54:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Diver Entangled in Intake Crib | A Unit 2 manual reactor trip was initiated when the control room was notified that a diver was entangled in the intake crib. Divers were being used to inspect the intake crib, install buoys, and the fish deterrent system. The diver's umbilical cord became snagged and attempts to free it were unsuccessful. The Unit 2 circulating water system was secured to aid in removing the diver from the water. The diver still had breathing air available during the transient. The diver was subsequently removed from the water unhurt. Plant systems functioned as required, including the Reactor Protection and Auxiliary Feedwater Systems. There was no Emergency Core Cooling System actuation. Note: The condenser was unavailable because circulating water was secured. This caused a loss of condenser vacuum and its use as a heat sink. The atmospheric steam dumps are currently being used for heat removal from the steam generators. The circulating water system was subsequently restored to service. This event is reportable pursuant to 50.72(b)(2)(iv)(B), any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical and 50.72(b)(3)(iv)(A), PWR auxiliary feedwater system. All control rods inserted into the core. The electrical busses are in a normal shutdown line up. The licensee notified the NRC Resident inspector. | Steam Generator Reactor Protection System Auxiliary Feedwater Circulating Water System Emergency Core Cooling System Control Rod | 05000301/LER-2004-002 |
| ENS 40822 | 16 June 2004 19:59:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Emergency Sirens Inadvertently Activated | During a weekly growl test of the Point Beach Nuclear Plant (PBNP) emergency sirens on June 16, 2004, the Manitowoc County sirens were inadvertently activated at 1459 CDT. PBNP is investigating the cause of this event. The sirens were activated for approximately one minute. As a result of the inadvertent actuation, the Village of Mishicot, the City of Two Rivers, Manitowoc County, and Kewaunee County, Wisconsin were notified. There was no impact to the plant as a result of the activation of the sirens. The alert notification system remains operable. This report is being submitted in accordance with 10 CFR 50.72(b)(2)(xi) as a result of the notifications made. The Resident Inspector was notified | ||
| ENS 40940 | 11 August 2004 13:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification of State Agency Regarding Potentially Oil Contaminated Soil | On August 11, 2004, at approximately 0830 AM CDT, a non-Emergency notification was made to the Wisconsin Department of Natural Resources regarding a discovery, on August 4, 2004, of diesel fuel odor in soil at the Point Beach Nuclear Plant (PBNP). The source of the odor is believed to be oil contaminated soil caused by a past accidental environmental release of fuel oil. This condition was identified while digging holes for a new fence on the northwest side of the plant as part of the PBNP security upgrade. During the excavation, an odor reminiscent of diesel fuel was noted to be emanating from the dirt that was dug up in one area. The dirt from the holes was dry with no obvious fuel oil present. No diesel fuel odor was noted from water and soil at the outfall of a drain in the vicinity of this area. The quantity of diesel contaminated soil is unknown. NMC plans to conduct an investigation to determine the nature and extent of contamination and to ultimately conduct remediation as appropriate. The licensee notified the NRC Resident Inspector. | ||
| ENS 41212 | 19 November 2004 22:22:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to a Steam Leak on an Instrument Isolation Valve Inside Containment | The following information was obtained from the licensee via facsimile: On November 9, 2004, NMC personnel performed a containment entry for Point Beach Nuclear Plant Unit 2 to locate and identify the source of a secondary steam leak. A leak was suspected due to increased pumping of the Unit 2 containment sump 'A'. A steam leak was identified on the body of valve 2MS-465D. This valve is an isolation valve for main steam flow transmitter 2FT-465. For containment isolation considerations, the main steam system is considered to be a closed system inside containment. The steam leak discovered on 2MS-465D cannot be isolated from containment penetration P-1, the main steam line penetration. Accordingly, we declared Technical Specification Condition 3.6.3.C not met. (This condition is applicable to penetration flow paths with only one containment isolation valve and a closed system) and entered TSAC (Technical Specification Action Condition) C.1. The required action for this condition is to isolate the affected penetration flow path with a completion time of 72 hours. In order to repair this valve wall leak, it will be necessary to shutdown and cooldown PBNP Unit 2. Accordingly, at 1622 CST we initiated a reactor shutdown of PBNP Unit 2. Since we anticipate that we would be unable to (meet) the allowed completion time for this TSAC, we have determined that this reactor shutdown is required by the Technical Specifications and is reportable in accordance with 10 CFR 50.72.(b)(2)(i). The licensee has notified the NRC Resident Inspector. | Main Steam Line Main Steam | 05000301/LER-2004-003 |
| ENS 41447 | 27 February 2005 17:58:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to Loss of Dc Battery Chargers | The following information was provided by the licensee via facsimile: A Technical Specification (TS)-required plant shutdown was initiated on Point Beach Unit 2 at 11:58 a.m. CST on February 27, 2005. The initiation of the unit shutdown was required by TS 3.8.4, Condition 13, due to the Completion Time of TS 3.8.4 Condition A not being met. Both units were operating at 100% power prior to the event. At 09:48 CST, offsite transmission network line 151 was lost. The associated electrical transient resulted in the loss of all four required battery chargers. This condition resulted in TS 3.8.4 not being met; however, no associated Actions are provided in TS 3.8.4 for the Condition of four inoperable DC power subsystems, requiring entry of both operating units into TS LCO 3.0.3. LCO 3.0.3 specifies that action be initiated within 1 hour to shut down both units. Three of the four required battery chargers were subsequently restored to operable status by 10:00 a.m. CST. As a result, LCO 3.0.3 was exited prior to initiation of power reduction on either unit. During the 12 minutes while less than 3 battery chargers were available, the associated station DC loads were being provided by the station batteries. The DC battery drain during this time was minimal and the DC electrical subsystems remained within their analyzed condition. At 10:00 a.m. CST, the only remaining inoperable battery charger was swing battery charger D09. D09 had been powering the D02 electrical subsystem at the time due to previously ongoing maintenance on normal battery charger D08. The continued unavailability of D09 resulted in TS 3.8.4 not being met for both Unit 1 and Unit 2 due to D02 electrical power subsystem becoming inoperable, Required Action 3.8.4 A.1 directs restoring the electrical power subsystem to operable status with a Completion Time of 2 hours. At 11:48 a.m. CST, Required Actions 3.8.4 B.1 and B.2 were entered for both units due to the Completion Time of TS 3.8.4 Condition A not being met. This Required Action directs that the units be in Mode 3 in 6 hours and Mode 5 in 36 hours. Plant operators planned an orderly sequential shutdown of both units and initiated the process by beginning a shutdown of Unit 2 at 11:58 a.m. CST. This condition is reportable per 10 CFR 50.72(b)(2)(i). At 12:02 p.m. CST, battery charger D09 and electrical subsystem D02 were restored to operable status. This resulted in TS 3.8.4 being met and Conditions 3.8.4 A and B were exited. Reactor power on Unit 2 had only been reduced to approximately 99% power at that time. No shutdown was initiated on Unit 1, which remained at 100% power. The power reduction on Unit 2 continued in accordance with normal plant procedures governing loss of offsite transmission network line 151. Plant procedures specify a reduction to less than 50% power as a precautionary measure associated with a potential for grid instability under certain postulated conditions. Both reactors continue to operate in a safe and stable manner. The licensee stated that the loss of offsite power line 151 (345KV) was due to an insulator problem on a high voltage tower about 7 miles from the plant. The estimated repair and restoration time for line 151 is approximately 2100 CST tonight. The licensee has notified the NRC Resident Inspector. | ||
| ENS 41538 | 28 March 2005 19:43:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Siren Capability Due to Power Outage | The purpose of this 8 Hour notification is to inform the Nuclear Regulatory Commission of a Loss of Emergency Preparedness Capabilities per NUREG 1022 and 10CFR50.72(b)(3)(xiii). Specifically, at 1455 central standard time on 3/28/05, Point Beach Nuclear Plant was notified that Emergency Plan Sirens, P012, P013, and P014 were out of service. These three sirens constitute 59.32% loss of population coverage. The time that the emergency sirens went out of service was 1343 CST, 3/28/05. These sirens are fed from Two Rivers Water/Light (WIPPI). Point Beach Nuclear Plant has contacted Manitowoc County Dispatch at 1517 of the loss of population coverage, and subsequent power restoration. Two Rivers Municipal Utilities has restored power to the Emergency Plan Sirens at 1533 on 3/28/05. The licensee informed the NRC Resident Inspector.
The licensee faxed the following update: The purpose of this update is to provide additional clarification to EN#41538. At 1343 CST, Emergency Plan Siren P012 went out of service due to a power outage. The loss of P012 represents a 3.38% loss in population coverage in and of itself. At 1502 CST, Emergency Plan Sirens P013, and P014 went out of service due to a power outage. When all three sirens are out, as they were from 1502 to 1533 CST, the loss of coverage to the population is 59.32%. Power was restored to Emergency Plan Sirens P012, P013, and P014, and they were returned to service at 1533 CST. The remainder of EN#41538 remains accurate as written. The licensee informed the NRC Resident Inspector. Notified the R3DO( Riemer). | ||
| ENS 41683 | 10 May 2005 08:23:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Notification System (Ens) Communication Lines | Licensee discovered that the ENS Communication Lines from both the Control Room and the Technical Support Center were unavailable. This constitutes a loss of emergency preparedness capability. The licensee notified the NRC Resident Inspector. | ||
| ENS 41705 | 17 May 2005 14:05:00 | 10 CFR 26.73, Applicability | Fitness for Duty | A non-licensed contract employee found a can of beer in his lunch box. The licensee's investigation determined that the introduction of alcohol into the Protected Area was not an intentional act. The contract employee's access was restored. The licensee notified the NRC Resident Inspector. Contact the Headquarters Operations Officer for additional details.
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| 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. | ||
| 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 41836 | 17 May 2005 05:42:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of the "B" Residual Heat Removal (Rhr) Pump. | July 12, 2005 Telephone Report in Accordance with 10 CFR 50.73(a)(2)(iv)(A) Invalid Actuation of the "B" Residual Heat Removal (RHR) Pump. SPECIFIC TRAINS AND SYSTEMS THAT WERE ACTUATED During a maintenance activity involving replacement of a safeguards relay, "B" Train RHR Pump 2P-10B was started inadvertently. The RHR pump is part of the emergency core cooling system (ECCS). DESCRIPTION OF WHETHER EACH TRAIN ACTUATION WAS COMPLETE OR PARTIAL On May 17, 2005, Unit 2 was in a shutdown condition (MODE 6) for routine refueling outage. At approximately 0042, Unit 2 RHR Pump 2P-10B was inadvertently started. Since RHR Pump 2P-10A had already been running for normal shutdown decay heat removal, the inadvertent start of the B RHR pump resulted in both RHR pumps running. Operations observed an increase in RHR flow but initially attributed the flow change to the RHR Heat Exchanger (HX) Bypass Flow Control Valve, which was operating in automatic. At the time of the inadvertent pump start, technicians were landing a wire as part of performance of a procedure to replace a safeguards relay. During this activity, the technicians heard a breaker close in the 480 VAC safeguards bus. Primary Plant Computer System (PPCS) computer data indicates that this breaker closure corresponded to the 2P-10B RHR Pump start. The technicians were not aware that this breaker closure was caused by their activity. The inadvertent RHR pump start was identified at 0525. The investigation of this event determined that a current path was created during the relay replacement, which caused starting of 2P-10B. No other ECCS components were actuated during this event. A review of the activity could not determine the specific wire that was lifted landed to cause the pump start. Rather several wire combinations were noted, each of which could have caused what is commonly referred to as a 'sneak' current path. These current paths can result in the RHR pump starting circuit being energized. DESCRIPTION OF WHETHER OR NOT THE SYSTEM STARTED AND FUNCTIONED SUCCESSFULLY This event was not safety significant. The RHR pump started and functioned successfully. The RHR pump was the only ECCS component in the system affected. Since the pump was aligned in the standby mode to provide normal decay heat removal cooling, its start appropriately resulted in additional cooling flow being pumped to the reactor core. The RHR pump was secured and returned to standby mode following discovery of this condition. As corrective action, a procedure change was initiated to add a precaution statement to ensure the associated safeguards relay cabinet is deenergized to the extent possible prior to similar maintenance activities to prevent inadvertent safeguards actuations. This item is documented in the PBNP corrective action process system (CAP 064616). The NRC Resident Inspector was notified of this event report by the licensee. | Residual Heat Removal Emergency Core Cooling System Decay Heat Removal | |
| 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 41885 | 2 August 2005 01:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Potentially Inoperable Safety Injection Pumps | As part of an ongoing calculation review project, design data is being evaluated on safety related motors. On 07/15/2005 at 2245, design data was accepted for 1P-15A Safety Injection motor. This data determined that during a design basis accident with degraded safeguards bus voltage, the 1P-15A Safety Injection Pump could trip and lockout on over current prior to the safeguards bus stripping on under voltage. The Safety Injection Pump lockout would then prevent an auto start of the Safety Injection Pump during Emergency Diesel Generator load sequencing. 1P-15A Safety Injection Pump was declared out of service at that time. No other safety related motors were affected by this data. The 1P-15A Safety Injection Pump time over current set point was adjusted, based on the revised motor data, returning 1P-15A to service on July 17 at 2230. 2P-15A Safety Injection Pump time over current set point was also conservatively reset. On 08/01/2005 at 2000, additional design motor data was accepted, which impacted the time over current set point for 1P-15B and 2P-15B Safety lnjection motors. Review of this data determined that a similar potential exists that during a design basis accident with degraded safeguards bus voltage, 1P-15B or 2P-15B Safety Injection Pump could trip and lockout on over current prior to the respective safeguards bus stripping on under voltage. The Safety Injection Pump lockout would then prevent an auto start of the Safety Injection Pump during Emergency Diesel Generator load sequencing. 1P-15B and 2P-15B Safety Injection Pumps have been declared out of service as a result of accepting this new data. Based on the combined effect of all received design motor data, a condition existed prior to 07/15/2005, which could have impacted the design function of Unit 1 or Unit 2 Safety Injection Pumps during a design basis accident with degraded safeguards bus voltage. Both Unit 1 and both Unit 2 Safety Injection Pumps had the potential for this effect to prevent auto start on the Emergency Diesel Generator Loading Sequence per design. Therefore, this condition is being reported under 10 CFR 50.72(b)(3)(v)(D) due to the potential to prevent fulfillment of a safety function to mitigate an accident. Previous corrective action has been completed for Unit 1 and Unit 2 P-15A Safety Injection motors therefore, they remain Operable and capable of fulfilling design safety function. The licensee is currently in a 72 hour Tech. Spec. LCO 3.5.2 (A) action statement for ECCS train B on both Units. Repairs are expected to take six to eight hours. Safety Injection train A equipment is being protected by administrative requirements put in place by the licensee. Operations personnel have been briefed of the potential impacts. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator | 05000266/LER-2005-004 05000266/LER-2005-003 |
| 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 42030 | 30 September 2005 11:05:00 | Other Unspec Reqmnt 10 CFR 26.73, Applicability | Fitness for Duty | A non-licensed employee supervisor had a confirmed positive for alcohol during a "for-cause" test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. | ||
| 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 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 42199 | 13 December 2005 09:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Loss of Condenser Vacuum | Point Beach Unit 1 was manually tripped at 0339 CST on 12/13/05 due to a loss of condenser vacuum caused by a mechanical failure of the running circulating water pump. All plant systems responded normally, including an auxiliary feedwater actuation. The trip was uncomplicated. All rods fully inserted. MSIVs were isolated due to the loss of condenser vacuum so decay heat is being removed by the atmospheric dump valves. The licensee indicated that there are no known steam generator tube leak issues. All systems functioned as required. The licensee was not in any significant LCO at the time of the trip. The trip had no impact on the electrical lineup or on Unit 2 operations. The cause of the circ water pump failure is still under investigation but there is evidence of sheared bolts on the pump coupling. The licensee noted the turbine condenser rupture disks blew due to high pressure. The licensee notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater | 05000266/LER-2005-008 |
| ENS 42426 | 17 March 2006 13:58:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification - Local Sheriff Contacted | On March 17 at approximately 0758 Central Standard Time (CST), the remains of a human body were discovered to have washed ashore immediately North of the Unit 2 circulating water discharge canal. The Manitowoc County Sheriff's Department was contacted to investigate. The body is believed to have randomly washed ashore from Lake Michigan. The county is presently on-site to pick up the human remains. The licensee notified the NRC Resident Inspector and expects to issue a press release at a later date.
NMC previously reported that on March 17 at approximately 0758 Central Standard Time (CST), the remains of a human body were discovered to have washed ashore immediately North of the Unit 2 circulating water discharge canal. The Manitowoc County Sheriff's Department was contacted to investigate. The body is believed to have randomly washed ashore from Lake Michigan. The following article has subsequently appeared on the web site of the local newspaper (Herald Times Reporter). MANITOWOC - Manitowoc County Sheriff's Department investigators responded to the area of the Point Beach Nuclear Plant before 8 a.m. today after a body washed up onto the shore of Lake Michigan there, according to Inspector (Deleted). (Deleted) said the body is that of a 46-year-old man, but said further details are not immediately available. Sheriff (Deleted) said early indications point to suicide, adding that is still under investigation. Department personnel were still on the scene, near the Two Creeks boat launch, as of 10 a.m. today. The County Sheriff and Coroner removed the discovered body and left site at 1008 on 3/17/06. The NRC Resident Inspector and local officials were notified of this event by the licensee. NRC R3DO (Dave Passehl) notified. | ||
| ENS 42435 | 20 March 2006 23:11:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Oil Spill Onsite | A notification was made to the Wisconsin Department of Natural Resources Spill Response Center by We Energies at 1711 hours on 03/20/06 for a small lube oil spill that occurred on 3/19/06. The spill was initially believed to be very small (less than 5 gallons) and there was no evidence of a release to Lake Michigan. Following a maintenance inspection of the water treatment deaerator vacuum pump, it was conservatively assumed that a maximum quantity of 25 gallons of oil could have been spilled. The NRC resident inspector has been notified of this event. | ||
| ENS 42611 | 30 May 2006 14:23:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Control Room Emergency Filtration System Inoperable | Control Room Emergency Filtration System (CREFS) was declared Inoperable at 0923 on 5/30/2006 per TS 5.5.10.c 'Ventilation Filter Testing Program'. Laboratory testing results of a sample of the charcoal adsorber taken on 5/2/2006 did not meet the methyl iodide penetration percentage acceptance criteria of less than or equal to 1.0%. This condition is covered by TS 3.7.9 and both Unit 1 and Unit 2 entered Condition A, 'CREFS Inoperable' with a Required Action of 'Restore CREFS to OPERABLE status' with a Completion Time of 7 Days. CREFS is a single train system. Based on the guidance in NUREG-1022 for single train systems that perform safety functions, this condition was determined to be reportable under 10 CFR 50.72(b)(3)(v), 'Event or Condition That Could Have Prevented Fulfillment of a Safety Function'. Technical Specifications allow this system to be inoperable for a period of seven days. This condition was not reported within the 8 Hour Non-Emergency reporting requirements. The charcoal has been replaced and the licensee estimates testing at approximately 17:00. The licensee notified the NRC Resident Inspector. | Control Room Emergency Filtration System | |
| ENS 42912 | 17 October 2006 02:26:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event - Identified Leakage Greater than 25 Gpm | At 2126, PBNP Unit 2 had a inadvertent PORV (431C) lift. This was a momentary event, however, a PRT level change of 60 gallons was noticed. This event is classifiable per C.U.1.2 identified leakage greater than 25 GPM on Unit 2. This event has been terminated at 2139. This event was caused by a momentary loss of power to the PORV controller. Power was immediately restored. The event was caused when a manually operated breaker (MOB) was inadvertently operated while the licensee was working inside the panels on the "B" train RHR system. The licensee notified the NRC Resident Inspector, State and local officials.
Unit 2 was in Mode 5 at ~150 psig, ~112 degrees F and the pressurizer (PZR) was in a water-solid condition per plant procedures. The following is a timeline of the event for clarification: 20:47:49 A manually operated breaker (MOB-281) inadvertently opened while removing a caution tag resulting in an increase of flow in the RHR system. 20:48:43 MOB-281 was reclosed correcting the RHR flow perturbation, however the adjacent breaker (MOB-282) inadvertently opened due to the vibration of MOB-281 being manipulated. This resulted in the loss of power to the RCS pressure instrumentation providing an input to a power-operated relief valve (PORV) causing the affected PORV to open. The PORV opening provided a flow path from the PZR to the Pressurizer Relief Tank (PRT). 20:49:06 MOB-282 was reclosed resulting In the PORV shutting and isolating the flow path from the PZR to the PRT. 20:55 The PRT level indication was checked on the Main Control Board, no change in level was observed as compared to prior to the event. While additional actions to stabilize the plant were being taken, the Shift Manager directed additional data to be obtained from the Plant Process Computer System (PPCS) regarding PRT parameters as the amount of water transferred was not large enough to be identified on the control board level Indication due to meter resolution. 21:26 Analysis of the PPCS data indicated that 62.5 gallons of RCS inventory had been transferred from the PZR to the PRT while the PORV was open. It was identified that this exceeded allowable RCS leakage limits, and the Emergency Plan was implemented. 21:32 An Unusual Event was declared per EAL CU 1.2, identified leakage greater than 25 gpm. It was recognized that this was a transient event and would be promptly terminated as the leakage flow path had been secured. 21:39 The Unusual Event was terminated. 21:43 State and County agencies were notified. 21:50 NRC Resident notified. 22:15 1 hour ENS Notification for Unusual Event and subsequent termination completed. The licensee notified the NRC Resident Inspector. Notified R3DO (Skokowski). | ||
| ENS 42913 | 17 October 2006 13:32:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification - Press Release | On 10/17/06 at 0832 Nuclear Management Company, LLC issued press release concerning the Unusual Event at Point Beach Nuclear Plant that was declared on 10/16/06 at 21:26 CDT. The Unusual Event was previously reported via EN# 42912 and occurred as a result of a momentary loss of power to the power-operated relief valve (PORV) instrumentation which caused the valve to open. The valve opening resulted in reactor coolant leakage into the pressurizer relief tank. The tank is located in the Unit 2 containment building and is designed to collect water flow from the PORV's. At no time was reactor coolant released outside of the piping systems. The NRC Resident Inspector has been informed of the press release and this subsequent notification. | ||
| ENS 43040 | 9 December 2006 00:12:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Crefs System Inoperable | Control Room Emergency Filtration System (CREFS) was declared inoperable at 1812 on 12/8/06 because the W-14A, F-16 Control Room Charcoal Filter Fan tripped during performance of the monthly technical specification surveillance test, TS-9. This fan is required to be operable for operability of the CREFS System. This condition is covered by TS 3.7.9, Control Room Emergency Filtration System and both units have entered action condition A, 'CREFS Inoperable' with a required action to 'Restore CREFS to OPERABLE Status' with a completion time of 7 days. CREFS is a single train system. Based on the guidance in NUREG-1022 for single train systems that perform safety functions, this condition was determined to be reportable under 10CFR50.72(b)(3)(v), 'Event or Condition That Could Have Prevented Fulfillment of a Safety Function'.
On December 8, 2006, at 22.42 EST, EN 43040 was made by the Point Beach Nuclear Plant in accordance with 10 CFR 50.72(b)(3)(v)(D) as a condition at the time of discovery could have prevented the fulfillment of a system function that is needed to mitigate the consequences of an accident. Control Room Emergency Filtration System (CREFS) fan W-1 4A, F-16 Control Room Charcoal Filter Fan tripped during performance of the monthly Technical Specification surveillance test. CREFS was declared inoperable. After further evaluation, it was determined that the safety function of CREFS was not lost. Redundant fan W-14B was operable for the duration of the time fan W-14A was out of service. Single train portions of the system were not affected by the W-1 4A fan trip. Accordingly, this event is not reportable. EN 43040 is, therefore, retracted. Notified R2DO (Burgess) | Control Room Emergency Filtration System | |
| ENS 43101 | 15 January 2007 19:21:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification - Oil Spill | At 1321 on January 15, 2007, the Point Beach Nuclear Plant asset owner notified the Wisconsin Department of Natural Resources (WDNR) that at about 10:00 a.m. on Sunday, January 14, 2007, approximately 20-25 gallons of vacuum pump hydraulic oil was released to the roof of the Point Beach Nuclear Plant heating boiler room. This event was caused by a failed seal in a vacuum pump. A small amount of this oil entered the rooftop drain system, and from there, it entered the plant's storm water drain system. The plant staff contained and removed the oil from the roof system. Absorbent pads were placed in the storm water system and at the exit of the system near the riprap above the beach. Approximately one gallon of the hydraulic oil was collected in the absorbent material at the exit of the storm water system. No oil discharged to the beach or the surface water. The licensee notified the NRC Resident Inspector. | ||
| ENS 43149 | 6 February 2007 20:15:00 | 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)(A), Seriously Degraded | Control Room Emergency Filtration System Declared Inoperable | The Control Room Emergency Filtration System (CREFS) was declared inoperable at 1957 on 02/03/07 due to W-14B, 'F-16 Control Room Charcoal Filter Fan,' being declared inoperable during monthly Technical Specification surveillance testing. Upon subsequent investigation of the inoperability of the W-14B fan, the W-14A fan was declared inoperable at 1415 CST on 02/06/07. The cause of the failure of the fans is under investigation. These fans are required to be operable to support operability of the CREFS System. This condition is covered by TS 3.7.9 'Control Room Emergency Filtration System' and both units have entered Action Condition A 'CREFS Inoperable,' with a Required Action to 'Restore CREFS to an OPERABLE Status' by 1957 CST on 02/10/07. Although the W-14 fans are redundant, CREFS is a single train system. Based on the guidance in NUREG-1022 for single train systems that perform safety functions, this condition was determined to be reportable under 10 CFR 50.72(b)(3)(v)(D), 'Event or Condition That Could Have Prevented Fulfillment of a Safety Function.' Additionally, the failure places PBNP CREFS in a degraded condition that significantly affects plant safety under 50.72(b)(3)(ii)(A), 'Degraded or Unanalyzed Condition.' The licensee notified the NRC Resident Inspector. | Control Room Emergency Filtration System | 05000266/LER-2007-001 |
| 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 43407 | 5 June 2007 20:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip After Feedwater Valve Failure | On 06/05/07 at 1512 hours CDT, operators observed that the Unit 1 main feedwater regulating valve (1 FD-476B) was going full open to full shut and entered abnormal operating procedure (AOP) 2B for a feedwater system malfunction. An immediate inspection of the valve determined that the valve positioner arm was disconnected, with the positioner arm locknut found on the floor adjacent to the valve. Operators manually tripped the Unit 1 reactor at 1517 hours CDT in response to the loss of 'B' train main feedwater control. During the trip, the auxiliary feedwater system actuated due to low level in the 'B' steam generator and an actuation of the ATWS mitigating system (AMSAC). Plant systems and equipment functioned properly following the manual trip with the following exceptions: A switchyard bus section 2 lockout occurred, resulting in loss of 345 Kv Line 121; 1FD-2603 bleeder trip valve (1 HX-22A moisture reheater drain) stuck open; and 1 P-129A turbine bearing oil lift pump did not automatically start, but was successfully started manually. Troubleshooting and additional investigations are continuing. The affected equipment has been quarantined. Unit 1 is in MODE 3 and is stable. Unit 2 was unaffected by the Unit 1 manual trip. All control rods fully inserted on the trip. No safety or relief valves lifted from the trip. Reactor pressure and temperature are being maintained with main feedwater and steaming to the main condenser. Emergency Diesel Generator GO-1 (two EDGs per train) and one Service Water Pump are out of service for replacement. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Service water Emergency Diesel Generator Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 43424 | 14 June 2007 23:19:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Tech Spec. Required Shutdown Due to 72 Hour Completion Time Not Met | At 1819 hours, an orderly shutdown on Unit 1 commenced as the result of Technical Specification Action Condition (TSAC) 3.7.5.B.1 completion time of 72 hours not being met for the 1P-29 turbine-driven auxiliary feedwater pump. The pump was declared inoperable on June 12, 2007, at 0131 hours as a result of high pump bearing temperatures. Repairs and testing performed to date have not satisfactorily resolved the problem. This non-emergency notification is being made in accordance with 10CFR50.72(b)(2)(i). The PBNP resident inspector has been notified. The licensee plans to conduct an extended duration test of the 1P-29 AFW pump. This test is scheduled to commence shortly and will last approximately 4 to 6 hours. The licensee will have two different hold points in the test to take pump bearing temperature data. The licensee has one emergency diesel generator (EDG) out of service (OOS), but an alternate EDG is aligned to ensure all four vital buses have vital power in the event of an loss of offsite power. No other safety related systems are OOS at this time. The licensee notified the NRC Resident Inspector.
This is an update to EN#43424: Regarding Unit 1 turbine driven auxiliary feedwater pump inoperability and Technical Specification required shutdown. PBNP Unit 1 entered MODE 3 on 6/15/07 at 0407. PBNP Unit 1 entered MODE 4 on 6/15/07 at 1712. Technical Specification Action Condition 3.7.5.D.1 required Unit 1 to be in MODE 3 by 0731 and MODE 4 by 1931 on 6/15/07. All Technical Specification Required Actions for the AFW pump OOS have been met within the required times. The licensee informed the NRC Resident Inspector. Notified R3DO(Stone). | Emergency Diesel Generator Auxiliary Feedwater | |
| 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 43649 | 18 September 2007 13:43:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Refueling Water Storage Tank Temperature Above Operating Limits | At 0043 on 9/18/07, during performance of Technical Specification (TS) Surveillance Requirement (SR) 3.5.4.1, the 2T-13 Refueling Water Storage Tank (RWST) was found to be at 105 degrees F. Limiting Condition for Operation (LCO) 3.5.4 A was therefore not met due to temperature being above the parametric value limit of 97 degrees F (100F technical specification value). The cause of the elevated temperatures of 2T-13 is currently under investigation. At 0843 an orderly shutdown of Unit 2 was commenced because of the continued elevated temperature of the Refueling Water Storage Tank. Troubleshooting to identify the cause of the problem and take remedial actions continues. Unit 1 RWST temperature is 77 F. Licensee is taking action to lower Unit 2 RWST temperature. The licensee notified the NRC Resident Inspector.
Unit 2 RWST temperature was verified at 97 degrees F at 1421 CDT. The 2T-13 RWST is operable and LCO 3.5.4.is now met. The Unit 2 power reduction was terminated at 20% reactor power. Current plans are to continue to cool down the RWST using station procedures that place the RWST on recirculation through the residual heat removal (RHR) heat exchanger with component cooling water (CCW) to cool the heat exchanger. Plans are in progress to return Unit 2 to full power on 09/18/07. A root cause evaluation is in progress to determine the specific cause of the event; however, preliminary investigations have determined that the heaters were on with an improper setpoint. The power supply to the heaters have been turned off and temperatures are being monitored on an hourly basis while the RWST is on recirculation. The licensee notified the NRC Resident Inspector. Notified R3DO (Peterson) | Residual Heat Removal | |
| ENS 43750 | 26 October 2007 00:30:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Both Low Temperature Overpressure Protection Systems Out of Service | This 8 hour report is being made pursuant to 10 CFR 50.72(b)(3)(v). On 10/25/2007 at 1930 CDT Point Beach Nuclear Plant (PBNP) Unit 1 and Unit 2 low temperature overpressure protection systems (LTOP) were declared inoperable as a result of the determination that the current LTOP actuation setpoint was non-conservative based on updated calculations. Specifically, 1) The mass input from the Safety Injection Pumps has significantly increased based on the use of a Point Beach Nuclear Plant specific system flow model. 2) The setpoint calculation does not consider instrument delay times during (PORV) Pilot Operated Relief Valve actuation. 3) The updated Calculation changes instrument uncertainties. LCO 3.4.12 for the LTOP system is not applicable at this time for either unit (both units in Mode 1/100% power). LCO 3.4.12 is only applicable in Mode 5, Mode 6 when the reactor vessel head is on, and Mode 4 when any cold leg temperature is at or below the temperature specified in the Pressure Temperature Limits Report (270 deg F). Changes to operating procedures to delineate operation of reactor coolant pumps and charging pumps during low temperature conditions are in progress. Implementation of these procedure changes will permit LTOP to be returned to service. The licensee stated that this issue was discovered as part of the on going calculation reconstitution initiative at the plant. The licensee notified the NRC Resident Inspector.
The following is an update to the 8 hr report made to the NRC via EN#43750: On 10/26/07 at 17:51 procedure changes which identify requirements for operation of reactor coolant pumps and charging pumps during low temperature conditions have been made. These procedures provide the guidance required to ensure that the current LTOP setpoints remain conservative. Based on the issuance of these procedures with the required guidance, LTOP is returned to service for both Unit-1 and Unit-2. The licensee notified the NRC Resident Inspector. Notified R3DO (Hills). | 05000266/LER-2007-008 | |
| ENS 43907 | 15 January 2008 20:15:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Unusual Event - Loss of Offsite Power | At 1404 1/15/2008, Central time, Point Beach experienced a loss of Unit 1X-04, Low Voltage Station Auxiliary Transformer. This loss has caused Point Beach to declare an Unusual Event (SU 1.1, Loss of all Offsite Power to essential busses for greater than 15 minutes.) The loss of Unit 1X-04 transformer is under investigation. All equipment operated properly with the exception of 1B-04, 'B' train 480 VAC Safeguards Bus, supply breaker opened for unknown reasons. The licensee notified the NRC Resident Inspector. Preparations are underway for a Unit 1 shutdown. Both Unit 1 EDGs are running and supplying Unit 1 safeguard buses. Non-safety buses are powered from an auto transfer that occurred after the loss of 1X-04. No impact on Unit 2 due to the Unit 1 electrical transient. Licensee notified state and local emergency management agencies.
On 1/15/08 at 1941 central time, FPL Energy Point Beach, LLC issued a press release concerning the Unusual Event at Point Beach that was declared on 1/15/08 at 1415 central time. The Unusual Event was reported under EN# 43907 and occurred due to the loss of all offsite power to essential busses for greater than 15 minutes from the failure of the Unit 1 X-04, Low Voltage Station Auxiliary Transformer. The NRC Resident Inspector has been informed of the press release and this subsequent notification.
At 1404 on 1/15/08 central time, Point Beach experienced a loss of Unit 1 X-04, Low Voltage Station Auxiliary Transformer. This loss caused Point Beach to declare an Unusual Event and that was reported under EN# 43907. Technical Specification 3.8.1.B, Associated unit's 13.8/4.16 kv (X04) transformer inoperable, was not met as of 1404 on 1/15/08 and this LCO has a Required Action to restore the (X04) transformer to Operable status with a Completion time of 24 hours. At 1404 1/16/08 the LCO was still not met and Unit 1 entered Technical Specification 3.8.1.H Required Action and associated Completion Time Not met, with a required action to be in Mode 3 in 6 hours and Mode 5 in 36 hours. This Event Notification Worksheet is for the initiation of the Technical Specification Required reactor shutdown that commenced at 1549 on 1/16/08, per 10 CFR 50.72(b)(2)(i). The licensee notified the NRC Resident Inspector. Notified R3DO(Stone), NRR EO(Evans), and IRD(Blount).
Update: At 1947 CST on 01/16/08, the Unit 1 reactor reached MODE 3, completing the required Technical Specification Shutdown. Offsite power has been restored to the essential busses by paralleling from Unit 2 4160 V buses and the Emergency Diesel Generators G-01 and G-03 have been placed back to a normal standby alignment. The unit is being taken to MODE 5 in accordance with plant TS requirements. This update to EN #43907 also reports the termination of the Unusual Event (AT 2035 CST) SU1.1, Loss of all offsite power to essential buses for greater than 15 minutes initially reported via EN #43907. The State of Wisconsin and the counties of Kewaunee and Manitowoc were notified of the termination via the Nuclear Accident Reporting System (NARS) at 2041 on 1/16/08. The licensee notified the NRC Resident Inspector. Notified R3DO(Stone), NRR EO(Evans), NRR ET(Dyer), DHS(Gray), FEMA(Dunker), and IRD(Blount).
Update: At 0702 CST on 01/17/08, FPL Energy Point Beach, LLC issued a press release regarding the termination of the Unusual Event and the shutdown of Unit 1 for repair of 1- X04 transformer. The NRC resident inspector was notified of the press release." R3DO (Stone) notified. | Emergency Diesel Generator | 05000266/LER-2008-001 |
| ENS 44319 | 26 June 2008 16:33:00 | 10 CFR 26.719, FFD Reporting requirements | 10 Cfr 26.719 Fitness-For-Duty - Site Access Suspended | A random FFD drug and alcohol test on a licensee supervisor indicated the presence of alcohol, although the blood alcohol content was below the limit of a positive test. As a prudent measure the Medical Review Officer recommended that the individual be further evaluated. The individual's site access to the plant has been suspended pending completion of that evaluation. Contact the Headquarters Operations Officer for additional details. The licensee informed the NRC Resident Inspector. | ||
| ENS 44336 | 4 July 2008 00:45:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Sodium Hypochlorite Leak | At 1945 Point Beach declared a UE based on HU 3.1, 'Report or detection of toxic or flammable gases that has or could enter the site area boundary in amounts that can affect NORMAL PLANT OPERATIONS.' This Emergency Action Level (EAL) was entered based on a confirmed Sodium Hypochlorite tank leak to the tanks diked area. The leak is located below the current tank level and will continue to drain to the dike until equilibrium level is reached. The dike is designed to contain the entire tank contents. There are no signs of any leakage outside of the dike. The vapors in the immediate area are not significant at this time and extra ventilation has been provided by opening an overhead roll up door in the area. The NRC Senior Resident Inspector has been notified. The sodium hypochlorite is used in the screen house to chlorinate the service water and circulating water systems to eliminate Zebra Mussels. The concrete diked area is designed to contain the entire contents of the tank.
At 0646 (CDT) on 7/4/08 Point Beach terminated the UE that was declared at 1945 CDT on 7/3/08, based on HU 3.1. The leaking Sodium Hypochlorite tank and leak control dike have been drained and the over flow leak control area cleaned. The hazard no longer exists and the area has been released to normal access. The licensee informed State and local agencies and the NRC Resident Inspector. Notified R3DO (Kozak), EO (Brown), IRD (Gott), FEMA (Canupp) and DEHS (Inzer). | Service water Circulating Water System | |
| 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 45329 | 4 September 2009 13:15:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty - Licensed Employee Confirmed Positive for Alcohol | A licensed employee had a confirmed positive for alcohol during for-cause fitness-for-duty test. The employee's access to the plant has been suspended. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. | ||
| ENS 45337 | 9 September 2009 20:23:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Soil Sample Results in Excess of State Requirements | At 1523 CDT, on September 9, 2009, the Wisconsin Department of Natural Resources (WDNR) was notified via electronic mail that laboratory analyses of soil samples collected from the catch basins surrounding the Unit 1 and 2 Main Power Transformers, 1X01, and 2X01, revealed polycyclic aromatic hydrocarbon (PAH) levels in excess of Wisconsin Administrative Code requirements. Remediation will be performed in accordance with Wisconsin Administrative Code requirements. The licensee notified the NRC Resident Inspector. | ||
| ENS 45553 | 9 December 2009 18:31:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Notification of Local Law Enforcement Due to a Single Emergency Siren Actuation | At 1231 emergency siren P-007 located in the Mishicot, WI area inadvertently actuated. The siren covers 0.9% of total EPZ population. Severe weather, snow, ice and wind is occurring at this time. At 1340 siren actuation was verified and Manitowoc County Sheriff was notified of the sounding siren. At 1350, Point Beach performed a siren test, reset the siren and it is no longer alarming. Repair team has been dispatched to the siren location to troubleshoot and determine the cause of the actuation. The population coverage for siren P-007 is 0.9% and the siren malfunction is not reportable due to loss of population coverage. However, based on actuation and notification of the Manitowoc County Sheriffs Department, the event is reportable. The NRC resident inspector has been notified. | ||
| ENS 46032 | 19 June 2010 11:36:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Generator Lockout | At 0636 hours, control room personnel initiated a manual reactor trip of Unit 2 from MODE 2 at 0% power. Just prior to the initiation of the manual reactor trip, an automatic turbine trip had occurred as a result of the receipt of a generator lockout signal. Prior to the automatic turbine trip, an orderly power reduction to 44% power had been completed and the unit was being maintained in a stable condition in MODE 1. At the time of the turbine trip, two sets of condenser steam dump valves were isolated, in preparation for scheduled condenser waterbox tube cleaning. Following the manual reactor trip, all safety systems and equipment operated as expected. The cause of the turbine trip, including receipt of the generator lockout signal, is under investigation. The unit is stable in MODE 3 at normal RCS temperature and pressure. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) t and 10 CFR 50.72(b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified. Just prior to the trip, the unit was critical in Mode 2. All rods are fully inserted into the core. | ||
| ENS 46053 | 29 June 2010 14:55:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness for Duty - Licensed Employee Confirmed Positive for Alcohol | A non-supervisory licensed employee had a confirmed positive test for alcohol during a random fitness-for-duty test. The employee's access to the plant has been denied. The individual had not entered the protected area and had not performed any licensed duties prior to the event. The licensee notified the NRC Resident Inspector at 1053 hours. | ||
| ENS 46080 | 9 July 2010 11:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to a Feedwater Regulating Valve Failure | On 07/09/10 at 0647 (CDT) hours, control room personnel initiated a manual reactor trip of Unit 2 from approximately 64% power as a result of a failure of the "A" feedwater regulating valve (FRV). All (other) systems functioned as expected. All rods fully inserted into the core. The unit is stable in MODE 3 at normal RCS (Reactor Coolant System) pressure and temperature. The cause of the FRV failure is being investigated. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The unit electrical power is lined up to offsite power in a normal configuration. Decay heat is being removed from the steam generator through the steam dumps to the main condenser. The FRV failed open and the valve controller was unable to place the FRV into the correct position. The steam generator HI-Hi level provided a feedwater isolation signal and a high level lockout of the FRV. Currently, the feedwater bypass valve is controlling steam generator level. There was no impact on Unit 1. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Main Condenser | 05000301/LER-2010-002 |
| ENS 46129 | 27 July 2010 01:01:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Loss of Condenser Vacuum | On 7/26/2010 at 2001 (hrs. CDT), Control Room personnel initiated a manual reactor trip from approximately 19% reactor power. Unit 1 was in the process of coming off-line to support Main Generator repair. The generator breaker had just been opened and load transferred to condenser steam dumps when a loss of condenser vacuum occurred. The reactor was manually tripped due to a loss of main condenser vacuum with reactor power above P-10 permissive. All systems functioned as expected. All control rods fully inserted. Main Steam Isolation valves were manually shut. All reactor coolant system parameters are as expected, with reactor coolant temperature being maintained by atmospheric steam dumps. Currently, the plant is at normal operating temperature and pressure with the steam generators being fed by the main feed pumps. Feedwater is being supplied via the condenser and condensate storage tank. There were no lifts of safeties or reliefs during the transient. The plant is in its normal shutdown electrical line-up with no effect on Unit 2. There is no known primary-to-secondary leakage. The cause of the loss of vacuum is under investigation. The licensee has notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Main Steam Isolation Valve Main Condenser Control Rod | |
| ENS 46240 | 9 September 2010 20:20:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Emergency Sirens Inadvertently Actuate | At 1520 (CDT), Point Beach was notified by Manitowoc County Sheriff's Department (MCSD) and citizens of audible EP Siren actuation in the City of Two Rivers and the Town of Two Creeks, WI. System troubleshooting was in progress at the time of actuation. Alert and Notification System Siren configuration was restored to normal at approximately 1630 (CDT). All required ANS sirens were 'Poll tested' and are fully functional. Further reviews indicated that Sirens P-001 through P-013 (13 sirens) had each simultaneously received a 180 second activation. At this time the cause of the activation signal is unknown. All troubleshooting and testing are currently suspended; event investigation has commenced. The licensee notified the NRC Resident Inspector.
This notification is an update to ENS notification #46240 made by Point Beach. Notification #46240 was submitted September 9 at 2025 EDT regarding Point Beach notification by Manitowoc County Sheriff's Department (MCSD) and citizens of audible EP Siren actuation in the City of Two Rivers and the town of Two Creeks, WI. System troubleshooting was in progress at the time of actuation. This notification is to inform (the NRC) of the planned press release to inform the citizens of the affected communities of the event. At 0834 CDT 9/12/2010 Point Beach issued a letter to the Editor of Manitowoc Times Herald Times Reporter Newspaper regarding inadvertent Siren Activation. Corrective actions have been taken in the (Point Beach) EP siren maintenance program to avoid a recurrence and the Site Resident NRC Inspector has been notified. Notified the R3DO (Lipa). |