Semantic search
| Entered date | Event description | |
|---|---|---|
| ENS 51642 | 8 January 2016 00:03:00 | Prairie Island's Appendix R calculations credit a procedurally established repair instruction to the Train B Pressurizer Vent valves for a postulated fire in Fire Area 59 (Unit 1) and Fire Area 74 (Unit 2) to obtain Mode 5 during a postulated fire in the affected areas. At 1900 (CST) on 1/7/2016, during a review of corrective actions associated with Prairie Island's NFPA 805 transition, it was identified that the required procedures are not in place to make the analyzed repairs. It has been determined that this condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. As compensatory measures, hourly fire watches are in place in the affected areas of the Auxiliary Building. The operating crew and Fire Brigade have been briefed on the impact of a fire in the affected area. This brief will continue to future operating shifts via a standing instruction. Fire detection equipment for the affected zones has been protected to ensure availability and operating crews are walking down the affected areas to verify any required transient combustibles in the affected areas are controlled in accordance with plant procedure. These compensatory measures, in addition to automatic fire detection and suppression capability in these fire areas, ensure protection of the potentially affected equipment until corrective actions can be completed. This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified. |
| ENS 50801 | 10 February 2015 10:41:00 | On February 10, 2015, Prairie Island Unit 1 was shutdown in Mode 3 during a planned outage. Ultrasonic testing in support of Unit 1 Emergency Core Cooling System (ECCS) void verifications identified existing voids with calculated volumes in excess of the OPERABILITY limits specified by the procedure. This rendered both trains of Residual Heat Removal (RHR) systems inoperable requiring entry into Technical Specification 3.0.3 at 0250 (CST). The station took prompt actions to vent the identified voids. The void at 1 RH-12 was vented to within acceptable limits allowing LCO 3.0.3 to be exited at 0538 on February 10. Venting at 1RH-11 is in progress. Voiding was identified at location 1-RH-11 with a calculated volume of 62.21 cubic inches with an OPERABILITY limit of 11.62 cubic inches. Voiding was identified at location 1-RH-12 with a calculated volume of 350 cubic inches with an OPERABILITY limit of 22.84 cubic inches. There was no impact to the health and safety of the public as Safety Injection was available and the time both trains of RHR were INOPERABLE was limited. This event is being reported as a unanalyzed condition that significantly degrades plant safety and a condition that could have prevented the fulfillment of a safety function (i.e., remove residual heat) under 10CFR50.72(b)(3)(ii) and 10CFR50.72(b)(3)(v). The licensee notified the NRC Resident Inspector.
Further analysis was performed on the two void locations, 1RH-11 and 1RH-12. Based on this additional analysis from AREVA, it was determined that the void located at 1RH-11 (RHR Train A) was operable. The calculations for past operability at inspection location 1RH-11 provides reasonable assurance that a void of 65 cubic inches will not generate forces that will fault any piping and supports. The void location at 1RH-12 (RHR Train B) was considered inoperable due to exceeding current procedural operability limits. The void located at 1RH-11 was determined to be nonconforming due to exceeding procedural design basis limits. Therefore, with RHR Train A determined to be operable, this event was not an 8-hour notification for an unanalyzed condition that significantly degrades plant safety, nor a condition that could have prevented the fulfillment of a safety function (i.e., remove residual heat) under 10CFR50.72(b)(3)(ii) and 10CFR50.72(b)(3)(v). The licensee has notified the NRC Resident Inspector. Notified the R3DO (Skokowski). |
| ENS 48619 | 21 December 2012 15:17:00 | 21 Motor Driven and 22 Turbine Driven Auxiliary Feedwater Pumps (AFWP) were declared inoperable at 0900 CST on 12/21/2012 due to the Condensate Storage Tank (CST) temperature exceeding 92 degrees F. In accordance with procedure C28.6, 'Condensate Storage Tank Freeze Protection System', the maximum CST temperature shall not exceed 92 degrees F. This is to ensure the maximum AFWP discharge temperature is less than 100 degrees F when an AFWP is at design flow per USAR Table 11.9-2, 'Summary of Assumptions,' used in the AFW system design verification analyses. LCO 3.7.5 Condition D was entered for two AFW trains inoperable in Modes 1, 2 or 3. The AFWP's could start and run if required at the time of entry. Immediate action was taken to reduce the CST temperature. At 1315 CST temperature were lowered below 92 degrees F and LCO 3.7.5 condition D was exited. This condition is reportable per 10 CFR 50.72(b) (3) (v) (D) as an event or condition that could have prevented the fulfillment of a safety function. The licensee has notified the NRC Resident Inspector.
This notification is being made to retract Event Notification (EN) #48619, which reported 21 Motor Driven and 22 Turbine Driven Auxiliary Feedwater Pumps (AFWP) were declared inoperable due to the Condensate Storage Tank (CST) temperature exceeding 92 degrees F. Based on engineering analysis the 92 degree F was a margin value with no formal basis or evaluation. This value was added to the C28.6 procedure in the early 1990's as a precaution to provide Operations with additional operating information. After the 22 CST tank was found above the 92 degrees F temperature on 12/21/12, a formal engineering evaluation (EC 21354) was performed to determine CST heat values and provide an accurate number for temperature margin. The purpose of the CST tank temperature being at a specific value is to ensure that the AFWP discharge temperature stays at or under 100 degrees F to ensure the AF system can provide adequate decay heat removal if called upon during an accident. Engineering evaluation EC 21354 determined that the average CST tank temperatures could reach 96.5 degrees F before the AFWP discharge temperature could have reached 100 degrees F. Therefore, there was no loss of safety function or past operability concerns. The NRC Resident Inspector has been informed. Notified R3DO (Bloomer). |
| ENS 47017 | 1 July 2011 18:07:00 | With Unit 1 at 100% power Unit 1 was manually tripped at 1552. The manual reactor trip was in response to the right main turbine stop valve failing closed as the result of an electro-hydraulic oil leak located at the stop valve. Procedure 1E-0 'Reactor Trip or Safety Injection' was completed at 1600. No SI (safety injection) required. 1ES-0.l 'Reactor Trip Recovery' is in progress. Offsite power remains on all safeguards buses for both units. 11 and 12 AFW pumps auto started on SG (steam generator) low level and are supplying Unit 1 Steam Generators. After the trip, power was lost to non-safety related 4160 VAC buses 11 and 14 as expected due to the electrical lineup. The loss of power to 4160 VAC bus 11 upon the reactor trip resulted in a loss of power to 11 RCP. 12 RCP continues to operate on offsite power. Unit 2 remains at 100% power/Mode 1. Reportable actuations are: Unit 1 reactor protection (scram), and Unit 1 AFW pumps auto start. The NRC Resident Inspector has been notified. |
| ENS 45966 | 1 June 2010 12:10:00 | At approximately 2115 CDT on May 30, 2010, a Pierce County dispatcher was notified of inadvertent siren activations, momentarily activating for approximately 30 minutes. The dispatcher also indicated there were lightning strikes in the area about the same time. The siren vendor was dispatched and the issue was tracked down to a single siren; no physical damage was observed, however, the siren was declared out of service by the siren vendor. The licensee notified the NRC Resident Inspector. |