Semantic search
| Entered date | Event description | |
|---|---|---|
| ENS 47130 | 6 August 2011 10:45:00 | On August 6, 2011, Reactor Protection System (RPS) power supply 1B failed resulting in a partial loss of power to Primary Containment Isolation System (PCIS) groups and an invalid actuation of those PCIS groups. PCIS groups 1 and 2 received partial isolation signals with no subsequent system isolations, as designed. PCIS group 3, 6, and 8 received partial isolation signals with resulting system isolations, also as designed. The combination of loss of RPS 1B and PCIS group 6 isolation resulted in the isolation of the Drywell Floor Drain Sump and the Drywell Continuous Atmospheric Monitor for both particulate and gaseous activity. Thus, both means of automatic monitoring of Reactor Coolant System leakage became inoperable. Unit 1 entered Technical Specification Limiting Condition for Operation (LCO) 3.4.5.D (all required leakage detection systems inoperable) and immediately entered LCO 3.0.3 as required. At the time of occurrence, RPS 1A was being supplied from its alternate source for scheduled maintenance. Thus, the alternate source was not available to RPS 1B. Unit 1 entered LCO 3.0.3 at 0524 (CDT), 'Initiate actions within one hour to place the unit in MODE 2 within 10 hours; MODE 3 within 13 hours; and MODE 4 within 37 hours.' At 0617, Unit 1 began reducing reactor power to comply with LCO 3.0.3. This event requires a 4 hour report IAW 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The PCIS isolations which occurred at 0524 CDT are also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) 'Any event or condition that results in a valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B)(2), 'General Containment Isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs)), except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). The event time for the PCIS isolations is 0524 CDT. The NRC resident inspector has been notified. Service Request 412927 was initiated in the Corrective Action Program.
Browns Ferry restored power to the 1B Reactor Protection System power supply at 1208 CDT, reset all isolations and exited LCO 3.0.3. The licensee plans to return the unit to full power. The licensee notified the NRC Resident Inspector. Notified R2DO (Binoy Desai). |
| ENS 46801 | 29 April 2011 07:26:00 | At 2338 (CDT) on 04/28/2011, with Browns Ferry Nuclear Unit 1 and Unit 2 in Mode 4, Browns Ferry Nuclear Plant, performed a shutdown of the Unit 1/2 Emergency Diesel Generator 'C,' due to an oil leak coming from its governor causing voltage and frequency fluctuations. Following securing of the Unit 1/2 Diesel Generator 'C,' the 4kV Shutdown Board 'C,' which was being powered by DG 'C,' de-energized. This resulted in a loss of power to the 1B RPS, causing a Primary Containment Isolation System (PCIS) actuation and the automatic initiation of the three trains of Standby Gas and 1 train of CREV (Control Room Emergency Ventilation System). The PCIS isolation (Group 2) also caused a loss of Shutdown Cooling on Unit 1 which was restored at 0025 (CDT) 04/29/2011 . In addition, the loss of power to the 4kV Shutdown Board 'C.' also caused the loss of 2B RHR Pump, leading to a momentary suspension of shutdown cooling to Unit 2. Shutdown cooling was immediately restored to Unit 2 using the 2D RHR Pump at 2342 (CDT). The general containment isolation signals affecting containment isolation valves in more than one system is reportable as an 8 hour notification to the NRC IAW 10CFR50.72(b)(3)(iv)(A), as 'Any event or condition that results in a valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' There were no new Technical Specification LCO's entered as a result of this event. This is also reportable as 60 day written report lAW 10CFR50.73(a)(2)(iv). The NRC Resident (Inspector) has been notified of this event. This event was entered into the licensee's Corrective Action Program as SR# 361382. |
| ENS 45282 | 21 August 2009 15:02:00 | Safe Shutdown Instruction (SSI) 0-SSI-9 for an Appendix R fire event requires for a Unit 1 Safe Shutdown success path that 4kV Shutdown Board A be supplied by Diesel Generator 3A via 4kV Shutdown Board 3EA. During a Unit 2 simulator exercise utilizing this SSI, reactor vessel water level lowered to Level 1 (-122 inches) causing a Common Accident Signal (CAS) trip of 4 Kv Shutdown Board 3EA inter-tie breaker (1844) that de-energized 4kV Shutdown Board A. This condition will not support the Unit 1 success path for alternate decay heat removal during an Appendix R fire. The 4 Kv Shutdown Board A must be re-energized by closing breaker 1844. For this SSI (0-SSI-9) the subsequent automatic start of the 2A RHR and Core Spray pumps in response to the CAS logic has not been considered in the development of the SSIs. In response to this condition, compensatory measures have been implemented to provide equivalent shutdown capability, documented on Fire Protection Impairment Permit (FPIP) 09-1920. This event is reportable within 8 hours under 10CFR50.72 (b)(3)(v)(A) - Any event or condition that results in the condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: Shut down the reactor and maintain it in a safe shutdown condition. This event is also a reportable within 60 days under 10CFR50.73(a)(2)(v)(A) - Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: Shut down the reactor and maintain it in a safe shutdown condition. This condition has been documented in the BFN Corrective Action program as PER# 177130. Unit 1, 2 and 3 remain at 100% power and are not affected by this event. The NRC Resident Inspector has been notified. |
| ENS 45107 | 31 May 2009 17:39:00 | |
| ENS 44680 | 23 November 2008 18:13:00 | During performance of Unit 1 vessel leak test in accordance with 1-SI-3.3.1.A - ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping, a pressure boundary leak was discovered on an instrument line connected to the reactor pressure vessel. This instrument line is a ASME Code Class 1 equivalent component (nozzle safe end) on pressure vessel nozzle N11B and connecting upstream of primary containment penetration X-29B. This caused entry into Technical Requirements Manual (TRM) 3.4.3 - Structural Integrity - Condition A - in which the applicability is at all times and the required action is to immediately restore the structural integrity of the affected component to within its limit or maintain the reactor in MODE 4 or 5 or the reactor coolant system less than 50?F above the minimum temperature required by NDT considerations, until each indication of a defect has been investigated and evaluated. The plant is currently in MODE 4. This event is reportable within 8 hours under 10CFR50.72 (b)(3)(ii)(A) Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. This event is also a reportable within 60 days under 10CFR50.73(a)(2)(ii)(A) any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC Resident Inspector has been notified. This condition has been documented in the BFN Corrective Action program as PER# 157918. Unit 2 and 3 remain at 100% power and are not affected by this event. |
| ENS 44540 | 5 October 2008 01:16:00 | On 10/04/08 at 2008 (CDT) the Unit 2 reactor scrammed due to turbine generator reverse power signal on the Main Generator. The cause of the reverse power signal is unknown and the investigation is continuing. All systems responded as expected to the generator reverse power signal. Reactor pressure was automatically controlled by the Main Turbine bypass valves. No Emergency Core Cooling System (ECCS), nor Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached, and reactor water level is being automatically controlled by the feedwater system. This report is being made as required by 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The licensee characterized the scram as uncomplicated. All control rods fully inserted. No safety valves lifted during the transient. All safety systems were available at the time of the scram. There were no impacts on Units 1 or 3. The licensee has notified the NRC Resident Inspector. |
| ENS 43083 | 8 January 2007 15:47:00 | This 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of any of the equipment specified in paragraph 50.73(a)(2)(iv)(B). In this case, the equipment actuated was the four Unit 1 and 2 emergency diesel generators (EDG) A, B, C, and D. These EDGs are common equipment for Browns Ferry Units 1 and 2. On November 10, 2006, personnel were performing surveillance requirement (SR) test 0-SR-3.8.1.6, Common Accident Signal Logic. At this time Browns Ferry Unit 1 was shutdown and defueled, and Unit 2 was operating at 100% thermal power. During conduct of this testing, at 0010 hours CST, all four of the common U1/2 EDGs were inadvertently started. This occurred when test personnel (non-licensed) inadvertently placed a keylock switch in the wrong position during the equipment restoration portion of the testing. The SR test step being performed was correct, but it was incorrectly implemented. All affected equipment operated per the plant design in response to the switch manipulation, with each of the four EDGs properly starting and running in response to the invalid start signal. No loss of normal plant electrical power occurred, and none of the EDGs connected to its associated 4-Kv shutdown board. The EDGs were shutdown in accordance with plant operating procedures approximately 13 minutes after their start. Since no actual plant condition existed which required the EDGs to start, and since the starts occurred inadvertently as a result of a human error during a test performance, this event is classified as invalid. There were no safety consequences or impacts on the health and safety of the public. The event was entered into TVA's corrective action program for evaluation and resolution. (Reference BFN corrective action document PER 114498). The NRC Senior Resident Inspector has been notified.
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| ENS 41404 | 11 February 2005 20:29:00 | The following text was obtained from the licensee via facsimile: At 1629 (hrs. CST) on 02/11/05, Unit 3 reactor scrammed from 100% power when the output breaker tripped causing a load reject. The breaker tripped due to a corresponding switchyard breaker 5268 tripping when a PK block was re-installed before the trip cutout (TCO) switches were placed in TCO. All rods inserted. Unit 2 was also at 100% power and was unaffected by this event. Water level lowered to +1" as expected and was recovered by normal feed water flow. All expected PCIS (Primary Containment Isolation System) isolations, Group 2 (RHR S/D (Residual Heat Removal) cooling), Group 3 (RWCU (Reactor Water Clean Up)), Group 6 (ventilation), and group 8 (TIP (Transverse Incore Probe)) were received along with the auto start of CREV (Control Room Emergency Ventilation) and the 3 SGT (Standby Gas Treatment) trains. Four MSRV's (Main Steam Safety Relief Valves) lifted momentarily to stabilize reactor pressure. This event is reportable as a 4-hour and 8-hour Non-Emergency Notification along with a 60-day written report in accordance with 10CFR50.72(b)(2)(iv)(B), 10CFR50.72(b)(3)(iv)(A) and 10CFR50.73(a)(2)(iv)(A) as 'Any event or condition that results in a valid actuation of RPS (Reactor Protection System) and PCIS .. The plant was performing restoration from switchyard maintenance at the time of the scram. All safety relief valves that lifted properly re-seated. Decay heat is being removed via the steam bypass valves to condenser. The electrical grid is stable. The licensee has notified the NRC Resident Inspector. |
| ENS 41348 | 21 January 2005 08:19:00 | At 23:53 CST Unit 2 HPCI pump suction automatically transferred from the Condensate Storage Tank to the Suppression Pool. This caused HPCI to be declared inoperable due to not meeting Technical Specification Surveillance Requirement 3.5.1.1. HPCI keep fill could not be verified so HPCI was declared inoperable. This event is being reported pursuant to 10CFR 50.72(b)(3)(v) as 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (D) Mitigate the consequences of an accident.' HPCI remains inoperable pending venting of the system. The licensee is conducting an investigation to determine the root cause for the automatic transfer. The licensee informed the NRC Resident Inspector. |