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The query [[Category:ENS Notification]] [[Site::Columbia]] [[Reporting criterion::10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation]] was answered by the SMWSQLStore3 in 0.1095 seconds.


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 Start dateReporting criterionTitleEvent descriptionSystemLER
ENS 5341018 May 2018 13:51:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationAutomatic Reactor Scram Caused by Main Transformer TripAt 0651 (PDT) on May 18th, 2018, Columbia Generating station experienced a Main Transformer trip, that caused a Reactor Scram. Reactor Power, Pressure and Level were maintained as expected for this condition. MS-RV-1A (Safety Relief Valve) and MS-RV-1B (Safety Relief Valve) opened on reactor high pressure during the initial transient. MS-RV-1B appeared to remain open after pressure lowered below the reset point. The operating crew removed power supply fuses for MS-RV-1B and it currently indicates intermediate position. SRV (Safety Relief Valve) tail pipe temperatures indicate all valves are closed. Suppression pool level and temperature have remained steady within normal operating levels. All control rods inserted and reactor power is being maintained subcritical. RPV (Reactor Pressure Vessel) water level is being maintained with condensate and feed system with startup flow control valves in automatic. Reactor Pressure is being maintained with the Turbine Bypass valves controlling in automatic. The main condenser is the heat sink. No ECCS (Emergency Core Cooling Systems) systems actuated or injected; the EOC-RPT (End of Cycle-Recirculation Pump Trip) and RPS (Reactor Protection System) systems actuated causing a trip of the RRC pumps and a reactor scram. Core recirculation is being maintained with RRC-P-1A (Reactor Recirculation Pump) running. No release has occurred. At this time there will be no notifications to state, local or other public agencies. The NRC Senior Resident has been notified. The cause of the event is currently under investigation. Plant conditions are stable. The plant is in its normal electrical alignment and offsite power is available to the site.Reactor Pressure Vessel
Reactor Protection System
Main Transformer
ENS 5291820 August 2017 23:05:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Reactor Scram Due to a Rise in Main Condenser Back Pressure

On August 20, 2017 at 1605 PDT, Columbia Generating Station was manually scrammed from 100 percent power due to a rise of Main Condenser back pressure. Manual scram of the unit is procedurally required upon a loss of Main Condenser back pressure. Preliminary investigations indicate that the Main Condenser air removal suction valve (AR-V-1) closed, resulting in the Condenser back pressure rising to within 1.0 inch Hg of the setpoint with reactor power greater than 25 percent. Further investigations continue. All control rods fully inserted. In addition to the closure of the air removal suction valve, one of two Reactor Feedwater startup flow control valves did not adequately operate to control Reactor vessel level and resulted in a high-level (Level-8) actuation tripping the Reactor Feedwater System. All other systems operated as expected. Reactor water level is currently being controlled manually with the start-up level control isolation valve. AR-V-1 has been manually opened with a jumper and temporary air supply. Reactor decay heat is being removed via bypass valves to the Main Condenser. This event is being reported under the following: 10 CFR 50.72(b)(2)(iv)(B), which requires a four-hour notification for any event or condition that results in actuation of the Reactor Protection System when the reactor is critical. The licensee notified the NRC Resident Inspector. The licensee plans to issue a press release.

  • * * UPDATE ON 8/24/17 AT 1937 EDT FROM MATT HUMMER TO DONG PARK * * *

The licensee is updating the notification to include an 8 hour notification under 10 CFR 50.72(b)(3)(iv)(A) for a specified system actuation due to a Level 3 isolation signal which occurred approximately 20 minutes after the scram. The licensee is currently in cold shutdown to repair the Reactor Feedwater startup flow control valve. The licensee has notified the NRC Resident Inspector. Notified R4DO (Farnholtz).

Feedwater
Reactor Protection System
05000397/LER-2017-004
ENS 5244218 December 2016 19:24:0010 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge
10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Scram Due to Load Reject from SubstationOn December 18, 2016 at time 1124 PST the plant experienced a full reactor scram. Preliminary investigations indicate that the scram was caused by a load reject from the Bonneville Power Administration (BPA) Ashe substation. Further investigations continue. The following conditions have occurred: Turbine Governor valve closure Reactor high pressure trip +13 inches reactor water level activations E-TR-B (backup transformer) supplying E-SM-7/SM-8 (vital power electrical busses) Complete loss of Reactor Closed Cooling (RCC) E-TR-S (Startup transformer) supplying SM-1/2/3 (non-vital power electrical busses) E-DG-1/2/3 (emergency diesel generators) auto start Low Pressure Core Spray (LPCS) and Residual Heat Removal (RHR) A/B/C initiation signals Main Steam Isolation Valves (MSIV) are closed Reactor Core Isolation Cooling (RCIC) RCIC and High Pressure Core Spray (HPCS) were manually activated and utilized to inject and maintain reactor water level. Pressure control is with Safety Relief Valves (SRV) in, manual. Level control is with RCIC and Control Rod Drive (CRD). RCIC has experienced an over speed trip that was reset so that level control could be maintained by RCIC. This event is being reported under the following: 10 CFR 50.72(b)(2)(iv)(A) which requires a 4 hour notification for Emergency Core Cooling System (ECCS) discharge into the reactor coolant system. 10 CFR 50.72(b)(2)(iv)(B) which requires a 4 hour notification for any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical. 10 CFR 50.72(b)(3)(iv)(A) which requires an 8 hours notification for actuation of ECCS systems. All control rods fully inserted. The NRC Resident Inspector has been informed. The licensee indicated that no increase in radiation levels were detected.Emergency Diesel Generator
Core Spray
Residual Heat Removal
Main Steam Isolation Valve
Reactor Core Isolation Cooling
High Pressure Core Spray
Reactor Coolant System
Reactor Protection System
05000397/LER-2016-005
ENS 5182628 March 2016 20:22:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Scram Following Loss of Reactor Closed CoolingAt 1322 PDT on Monday, March 28, 2016, Columbia Generating Station was manually scrammed from 100% thermal power due to the loss of Reactor Closed Cooling (RCC). Manual scram of the unit is procedurally required upon loss of RCC. The cause of the loss of RCC is being investigated. Regulation 10 CFR 50.72(b)(2)(iv)(B) requires reporting within 4 hours of any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical. All control rods were fully inserted. Valve RWCU-V-4 automatically closed upon high water temperature due to loss of RCC flow. No other safety system actuations were reported. All systems operated as expected. Reactor decay heat is being removed via bypass valves to the Main Condenser. The station is in normal shutdown electrical lineup. The NRC Resident Inspector has been informed. No safety/relief valves lifted and no emergency core cooling systems injected following the reactor scram.Reactor Protection System05000397/LER-2016-001
ENS 454847 November 2009 15:25:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Reactor Scram Due to a Main Turbine Digital Electro-Hydraulic Control System LeakAt 0725 hours, a manual scram was inserted due to a Main Turbine Digital Electro-Hydraulic (DEH) control system leak and anticipated turbine trip on loss of DEH tank level. Initial investigation identified a DEH leak at the area of the Quad-Voter hydraulic trip subsystem on the high pressure turbine. All rods fully inserted, main steam isolation valves (MSIVs) remained open, no safety/relief valves (SRVs) opened, and all other safety systems operated as designed. Reactor water level was restored and maintained with the Reactor Feedwater and Condensate systems during the post-scram transient within normal operating bands. Reactor pressure was controlled using main steam drain lines. A normal cooldown to the condenser is in progress. Offsite power is available. All three emergency diesel generators are operable and available. The plant was operating at 52% for planned maintenance. All safety systems remain available. The licensee has notified the NRC Resident Inspector.Main Turbine
Main Steam Isolation Valve
Feedwater
Emergency Diesel Generator
05000397/LER-2009-005
ENS 452455 August 2009 14:50:0010 CFR 50.72(a)(1)(i), Emergency Class Declaration
10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Unusual Event Declared Due to Toxic Gas from a Switchgear Fire in the Turbine Building

At 0750 PDT the licensee experienced a turbine trip and reactor scram. At approximately the same time, a fire was detected in the non-safety related 6.9 kV feed bus to switchgear SH-5 and SH-6 in the turbine building. The fire and associated fault to the switchgear feed bus caused a loss of power to both reactor recirculation pumps and the automatic reactor scram. The fire produced smoke and potentially toxic gases. Due to the presence of potentially toxic gas in the power plant, an Unusual Event was declared at 0812 PDT based on EAL 9.3.U.3. At the present time, the fire is out and the smoke is being cleared from the plant. All rods fully inserted upon the reactor scram. Systems functioned as expected except for problems with the EHC system which resulted in the bypass valves remaining open which caused the reactor to depressurize to approximately 390 psi. MSIVs were manually shut to halt the reactor depressurization and cooldown. Decay heat is being removed via relief valves to the suppression pool with reactor pressure being maintained between 500 and 600 psi. Suppression pool cooling is via RHR. Makeup water to the reactor is via normal feed. Licensee is on natural circulation at the time of this report. Normal shutdown electrical alignment is established with the exception of Division II emergency switchgear which is aligned to the backup transformer. A request for assistance was made to the Hanford fire department, however, the fire was out prior to their arrival on site. The licensee has notified State, local agencies and the NRC Resident Inspector.

  • * * UPDATE FROM JOHN SLACK TO DONALD NORWOOD AT 1332 ON 08/05/09 * * *

The Unusual Event has been terminated as of 1006 PDT. The licensee is taking the unit to cold shutdown. The licensee will notify the NRC Resident Inspector. Notified NRR EO (Ross-Lee), R4DO (Walker), DHS (Enzer), and FEMA (Eaches).

  • * * UPDATE FROM NICK RULLMAN TO PETE SNYDER AT 1827 ON 08/05/09 * * *

Following the reactor scram, with the turbine bypass valves failed fully open, the inboard main steam isolation valves had to be manually closed to prevent excessive cool down. The outboard main steam isolation valves automatically closed from the inability to maintain condenser vacuum. The inability to maintain condenser vacuum was due to limited equipment access in the Turbine Building from heavy smoke. This is reportable under Part 50.72(b)(3)(iv)(A) as a valid actuation of one of the systems listed in Part 50.72(b)(3)(iv)(B)(2). The licensee notified the NRC Resident Inspector. Notified R4DO (Walker).

Main Steam Isolation Valve
ENS 4516927 June 2009 03:05:0010 CFR 50.72(a)(1)(i), Emergency Class Declaration
10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
Unusual Event Declared Based on Fire Lasting Greater than 15 Minutes

At 1949 PDT, a small fire was observed between the #1 and #2 bearings on the main turbine involving some lube oil leakage and lagging. The fire brigade was dispatched and at 1953 PDT, the reactor was manually scrammed. At 2005 PDT, the licensee declared an Unusual Event (EAL 9.2.U.1) based on a fire lasting greater than 15 minutes. At 2006 PDT, the fire was reported out. The manual scram was uncomplicated and all systems functioned as required. The reactor is being cooled by normal feedwater and discharging decay heat to the condenser. The licensee is cooling down the reactor to Mode 4. Currently reactor pressure is 495 psi. The licensee has stationed a re-flash watch at the fire location and is assessing any damage that may have occurred. The only damage currently reported involves lagging at the fire location. The NRC Resident Inspector, State and local authorities have been notified.

  • * * UPDATE FROM BILL HART TO HOWIE CROUCH @ 0128 EDT ON 6/27/09 * * *

The licensee terminated the NOUE at 2159 PDT. The termination criteria was the fire is out, a re-flash watch stationed and the plant is stable and transitioning to Mode 4. The licensee has made State and local notifications and has notified the NRC Resident Inspector. Notified R4DO (Powers), IRD (Grant), NRR ET (Lubinski), FEMA (Casto) and DHS (Vestal).

Main Turbine
Feedwater
ENS 450518 May 2009 17:45:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Reactor Scram After Loss of Generator Seal OilAt 1045 hours, a manual scram was inserted due to a loss of main generator seal oil pressure and subsequent loss of hydrogen pressure in the Main Generator. Initial investigation revealed that a filter in the seal oil system became clogged during a system test. Even though the test was stopped and the system lineup restored, seal oil system pressure did not return to normal until the main turbine generator was secured. No indications of turbine or generator damage exist. All rods fully inserted, MSIVs remained open, no SRVs opened, and all other safety systems operated as designed. Reactor Water Level was restored and maintained with the Reactor Feedwater and Condensate systems during the post scram transient within normal operating bands. Reactor Pressure was controlled using the Main Turbine Bypass Valves in automatic control. Normal cool down to the Main Condenser is in progress. Offsite power is available. EDG-2 was inoperable but available due to surveillance testing, restoration is in progress. All other EDGs are operable and in standby status. This shutdown occurred 13 hours before a planned shutdown for the refueling outage, so the plant will not be restarted until after the outage. The licensee notified the NRC Resident Inspector.Main Turbine
Feedwater
05000397/LER-2009-002
ENS 448398 February 2009 19:25:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationAutomatic Reactor Scram Due to a Governor Valve Fast Closure SignalFollowing a downpower to 75% for replacement of DEH-SV-TRIP/B (solenoid for quadvoter valve) a reactor SCRAM occurred during the performance of the DEH (Digital Electric Hydraulic) quadvoter valve testing. The cause of the reactor SCRAM was Governor Valve Fast Closure signal due to DEH trip header pressure fluctuation. The cause of the DEH trip header pressure fluctuation is unknown and being investigated. All rods fully inserted. MSIVs remained open. No SRVs opened. RPV level is being controlled in the normal band using the feedwater and condensate systems. RPV pressure is being controlled in the normal band using the Bypass valves and Main Steam Line drains. All other safety systems operated as designed. Off-site power is available. All three emergency diesel generators are operable and available. The licensee will be making a press release. The licensee notified the NRC Resident Inspector.Feedwater
Emergency Diesel Generator
05000397/LER-2009-001
ENS 4443221 August 2008 23:06:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationAutomatic Reactor Scram Due to a Main Turbine TripFollowing a downpower to 65% for maintenance on one of the reactor feedwater turbines, a main turbine digital electro-hydraulic (DEH) quadvoter valve, that had failed a previous surveillance, was replaced. During the operability testing of the new DEH quadvoter valve, a leak developed. DEH reservoir level dropped 8 inches (approx. 52 gal) and tripped the main turbine on low DEH pressure. An automatic reactor trip occurred due the main turbine trip. All rods fully inserted. RPV pressure was controlled using the bypass valves and main steam line drains. No SRV's opened. RPV level was maintained using the feedwater and condensate system in the normal RPV level band. All other safety systems operated as designed. Off-site power is available. Three emergency diesel generators are operable and available. The licensee has notified the NRC Resident Inspector.Main Turbine
Feedwater
Emergency Diesel Generator
05000397/LER-2008-001
ENS 4345729 June 2007 00:17:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationReactor Trip Due to Condensate Pump TripReactor trip at 1717 hrs PDT due to (condensate pump) COND-P-2B trip. Reactor power was at 70% with (condensate pump) COND-P-2A secured. Reactor vessel level reached -50 inches and was restored with High Pressure Core Spray and Reactor Core Isolation Cooling. Main Steam Isolation Valves closed as expected due to the reaching -50 inches. All systems operated as expected. Further investigation into COND-P-2B trip is underway. Plant is stable in mode three, heat removal is being maintained by RHR-P-2B and Safety Relief Valves. All rods fully inserted on the automatic reactor scram. All safety systems were available at the time of the trip. The trip was considered uncomplicated. The reactor pressure is currently being maintained between 500 to 600 psi and water level between 60 to 80 inches. The licensee was at 70% power at the time of the trip due to maintenance of condensate pump P-2A. The licensee will notify the NRC Resident Inspector.High Pressure Core Spray
Reactor Core Isolation Cooling
Main Steam Isolation Valve
05000397/LER-2007-004
ENS 4295031 October 2006 12:45:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationReactor Trip from Turbine Trip on Low Auto Stop Oil Pressure

Reactor trip at 0445 following a turbine trip. Initial indication of cause is turbine trip due to low auto stop oil pressure. There were no complications in plant response. Plant is stabilized in mode three, heat removal is being maintained by turbine bypass valves. All control rods fully inserted on the trip, no safety or relief valves lifted during the transient, reactor water level 3 isolations did isolate, and the minimum level attained during the transient was -6 inches. Vessel water level is being maintained with normal feedwater flow and the electrical lineup is the normal shut-down electrical lineup. The cause of the low auto stop oil pressure is under investigation. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM LICENSEE (M. HUMMER) TO M. RIPLEY AT 1439 EST ON 10/31/06 * * *

In response to questions raised during this event notification, a verbal response was provided to the NRC that an 8 hour notification in accordance with 50.72(b)(3)(iv)(A) (Specified System Actuation) was required. Upon further review it has been determined by Energy Northwest that this notification was not required because all required reporting was satisfied by 50.72(b)(2)(iv)(B). The licensee will notify the NRC Resident Inspector. Notified R4 DO (D. POWERS)

Feedwater05000397/LER-2006-001
ENS 4179023 June 2005 20:46:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Unit Experienced an Automatic Reactor Scram Due to a Loss of Feedwater FlowOn June 23, 2005 at 1348 PDT Columbia Generating Station experienced a Reactor Protection System (RPS) actuation and closure of all Main Steam Isolation Valves (MSIV) while operating at approximately 25% power. Currently, Reactor Pressure Vessel (RPV) level is being controlled using the Reactor Core Isolation Cooling (RCIC) system and RPV pressure is stable at approximately 800 psi. There was no automatic ECCS system injection. At this time, the plant is stable and indications show that the RPS actuation and closure of MSIVs was initiated in response to a valid low RPV water level signals that occurred because of a loss of feedwater flow. Determination of the cause of the loss of feedwater flow is ongoing at this time. All control rods fully inserted. The licensee informed the NRC Resident Inspector.Feedwater
Reactor Protection System
Main Steam Isolation Valve
Reactor Pressure Vessel
Reactor Core Isolation Cooling
05000397/LER-2005-004
ENS 4177915 June 2005 21:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor Scram Due to Turbine Throttle Valve ClosureOn June 15, 2005 at 1400 PDT Columbia Generating Station experienced a Reactor Protection System (RPS) actuation while operating at full power. Currently, reactor level is being controlled with the feedwater system and reactor pressure is being controlled via the turbine bypass valves to the condenser. All safety systems functioned as expected and there was no ECCS system injection. The initial attempt to trip the main turbine from the control room was unsuccessful and it was subsequently tripped locally. At this time the plant is stable and indications show that the RPS actuation originated from closure of turbine throttle valves. Determination of the cause of the RPS actuation is ongoing at this time. All rods fully inserted and no relief valves lifted. The NRC Resident Inspector was notified.Reactor Protection System
Feedwater
Main Turbine
05000397/LER-2005-003
ENS 4096417 August 2004 12:28:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Reactor Scram Due to Trip of Reactor Feedwater Pump "a" on Low Suction Pressure.With a reactor startup in progress at 0528 PDT, operators at Columbia Generating Station inserted a manual reactor scram when the operating Reactor Feed Water (RFW) pump RFW-P-1A tripped. Reactor power was approximately 20% at the time of RFW pump trip. The cause of the RFW pump trip was due to low suction pressure; the cause of the low suction pressure is currently under investigation. The Reactor Core Isolation Cooling (RCIC) system (was manually started and) was used to maintain reactor vessel water level until reactor pressure was reduced to within the capacity of the condensate booster pumps (500 to 600 psi). The RCIC system has been returned to a standby lineup. The reactor is in Mode 3 (Hot Shutdown) with both reactor recirculation pumps running at minimum speed (15 Hertz). Decay heat is being rejected to the main condenser via auxiliary steam loads. All ECCS systems are operable. All emergency diesel generators are operable. No Safety Relief valves lifted during the scram. The NRC Resident Inspector was notified of this event by the licensee. See similar event number 40959 that occurred on 08/15/04.Feedwater
Reactor Core Isolation Cooling
Emergency Diesel Generator
ENS 4095915 August 2004 20:03:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Reactor Scram Due to Reactor Feedwater Pump TripWith a reactor startup in progress at 1303 PDT, operators at Columbia Generating Station inserted a manual reactor scram when the running Reactor Feedwater (RFW-P-1A) pump tripped. Reactor power was approximately 18% at the time of RFW-P-1A trip. The cause of the RFW-P-1A trip was a high RFW Turbine Drain Tank (MD-TK-1) level; the cause of the MD-TK-1 high level is under investigation. The Reactor Core Isolation Cooling (RCIC) was used (manually started) to maintain reactor vessel water level until reactor pressure was reduced to within the capacity of the condensate booster pumps; the RCIC system has been returned to a standby lineup. The reactor is in mode 3 with both reactor recirculation pumps running at minimum speed (15 Hertz). Decay heat is being rejected to the main condenser via auxiliary steam loads. One control rod that indicated full-in immediately after the scram lost full-in indication eleven seconds after the scram. This control rod indicated full-in again 84 seconds after the scram. This ENS notification is made pursuant to 10 CFR 50.72(b)(2)(iv)(B). The licensee has notified the NRC Resident Inspector.Feedwater
Reactor Core Isolation Cooling
ENS 4091030 July 2004 17:00:0010 CFR 50.72(a)(1)(i), Emergency Class Declaration
10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
Alert Declared at Columbia Generating Station

The following information was obtained by the licensee via facsimile: Reactor SCRAM received at 0924 (hrs) PDT. Initial indications are the scram signal was caused by RPS (Reactor Protection System) High Pressure. Following the scram, 2 (two) control rods did not immediately indicate (fully inserted). Control room staff entered (procedure) PPM5.1.2 and took the required actions. All control rods subsequently indicated (fully inserted). An ALERT Classification was declared at 1000 (hrs) based on PPM 13.1.1 Criteria 2.2.A.1, 'RPS Setpoint exceeded and automatic actions failed to result in a rod pattern which alone assures reactor shutdown'. Manual actions resulted in all rods (fully inserted) and reactor power (less than or equal to) 5 percent. All other plant systems responded as expected with the exception of a wetwell-to-drywell vacuum breaker which indicates open. Investigation into the cause of the scram and actual control rod position is ongoing. Further details will be provided when available. The licensee also reports that no relief valves lifted during the transient. Decay heat removal is via the main condenser. Reactor level is steady at 36 inches. Reactor temperature and pressure are 507 degrees and 710 psig respectively. Offsite power is available. All emergency systems are available in standby. The licensee has notified the NRC Resident Inspector of the incidents. The NRC entered Monitoring mode at 1327 hrs EDT with Region IV leading. The NRC exited Monitoring at 1530 hrs. EDT and returned to Normal mode. Notified HHS (Ayles) as well as others noted in notification block.

  • * * UPDATE AT 2130 HRS EDT ON 7/30/04 FROM COLEMAN TO CROUCH * * *

At 1358 (hrs.) EDT on 7/30/04, NRC was notified of an Alert at Columbia Generating Station (EN #40910). This is a follow-up to inform NRC that the event was terminated at 1457 (hrs) (EDT) (1157 PDT). All control rods are inserted. Reactor is shutdown and water level is normal. All required emergency systems are operable. All offsite and onsite power sources are operable. Reactor pressure is normal. The licensee has notified the NRC Resident Inspector, State of Washington and local authorities of the termination. The NRC Operations Center notified R4DO(Bywater), DHS, FEMA, DOE(NRC), USDA, EPA and CDC(HHS).

  • * * UPDATE AT 1930 EDT ON 08/03/04 FROM M. HEDGES TO A. COSTA * * *

At 1358 EDT on July 30, 2004, NRC was notified of an Alert at Columbia Generating Station (EN #40910). A subsequent notification was made to inform the NRC that the event was terminated at 1157 PDT (1457 EDT) on July 30, 2004. This is a follow-up to inform the NRC that Columbia Generating Station is retracting its Alert Emergency Declaration due to the following reason. Following the RPS actuation, control rod position indication for two control rods was indeterminate for approximately two minutes to the control room staff. A subsequent review of control rod position indication from the Plant Data Information System (PDIS), Rod Worth Minimizer (RWM) logs, and Auto Scram Timer (AST) data by Columbia Generating Station personnel shows that all rods were successfully inserted to the 'Full-in' position following the initial RPS actuation, assuring that the reactor was shutdown under all conditions. The Emergency Action Level for this Alert classification requires that the following three conditions be met: Any RPS set point (including manual) has been exceeded per T.S. 3.3.1.1 AND RPS actuation failed to result in a control rod pattern which alone always assures reactor shutdown under all conditions AND Manual actions (mode switch in shutdown, manual push buttons, and ARI) result in reactor power LE 5%. Since all rods were successfully inserted without the assistance of any manual actions and within the Technical Specification required time, Columbia Generating Station staff now believes that no emergency classification should have been made, and we are retracting the Alert emergency classification from this event notification. The problem with control rod position indication following the scram is being addressed through our corrective action program. The licensee notified the NRC Resident Inspector and will notify local, State, and other Government Agencies of this update. Notified R4 DO (Runyan).

05000397/LER-2004-004