| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
10/08/2004 - 10/12/2004 ** EVENT NUMBERS ** |
 | General Information or Other | Event Number: 40962 | Rep Org: GENERAL ELECTRIC NUCLEAR ENERGY
Licensee: GENERAL ELECTRIC NUCLEAR ENERGY
Region: 1
City: WILMINGTON State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JASON S. POST
HQ OPS Officer: BILL GOTT | Notification Date: 08/16/2004
Notification Time: 16:04 [ET]
Event Date: 08/16/2004
Event Time: 16:00 [EDT]
Last Update Date: 10/11/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
TODD JACKSON (R1)
CAUDLE JULIAN (R2)
BRENT CLAYTON (R3)
WILLIAM JONES (R4)
VERN HODGE (NRR)
| Event Text PART 21 60-DAY INTERIM NOTIFICATION: NARROW RANGE WATER LEVEL INSTRUMENT LEVEL 3 TRIP
This letter provides a 10 CFR21(a)(2) 60 - Day Interim Report notification regarding a potential issue with the Level 3 trip from the narrow range water level instruments that initiate reactor scram. A conservative evaluation by GE Nuclear Energy (GENE) has determined that water level instruments may indicate high by as much as 8 inches, should the reactor water level drop below the dryer seal skirt. At issue is whether with the actual water level as much as 8 inches lower than indicated, the top of active fuel (TAF) will be uncovered for the limiting loss of feedwater event due to
1. Actual water level being lower than indicated when the Level 3 trip occurs, or
2. Failure of the Level 3 trip to occur if water level drops below the narrow range instrument variable leg tap prior to reaching the Level 3 trip setpoint.
Because TAF is a Technical Specification Safety Limit, TAF uncovery for a loss of feedwater event would be a Reportable Condition. However, it would not lead to a significant safety hazard due to multiple automatic and passive protection features of a BWR.
GENE has completed analysis for BWR 2/3 plants and determined that this is not a reportable condition (i.e., the TAF safety limit is protected). GENE has not completed the analyses for BWR /4-/6 plants. For these plants, GENE has determined that actual water level being lower than indicated by up to 8 inches when the Level 3 trip occurs does not lead to TAF uncovery. However, for these plants GENE has not determined if water level could drop below the narrow range instrument variable leg tap prior to reaching the Level 3 trip setpoint.
Therefore, this letter is issued as a 60 - Day Interim Notification under 10 CFR21.21(a)(2) to the BWR/4-/6 plants listed in Attachment 1. The potentially affected plants are being concurrently notified of this situation by a GENE Safety Communication letter. GENE is committed to complete the evaluation by October 11, 2004.
Attachment 1 Plants List Below: (Affected Plants - Evaluation incomplete)
Clinton, Brunswick Units 1 & 2, Nine Mile Point Unit 2, Fermi Unit 2, Columbia, Fitzpatrick, Grand Gulf, River Bend, La Salle Units 1&2, Limerick Units 1 & 2, Peach Bottom Units 1 & 2, Perry, Cooper, Duane Arnold, Susquehanna Units 1 & 2, Hope Creek, Hatch Units 1 & 2, Browns Ferry Units 1 (plant is in an extended shutdown),2 & 3.
* * * UPDATE TO HUFFMAN FROM B. BAHN (FOR J POST) AT 16:16 EDT ON 10/11/04 * * *
"GENE has now completed the evaluation and has determined that all BWR/2/3/6 and ABWR plants, and most BWR/4-5 plants are not susceptible to the condition that could result in a failure of the Level 3 trip to occur. These plants are identified as Not Reportable in Attachment 1 [of GENE Part 21 final report to the NRC dated October 11, 2004]. However, for same BWR/4-5 plants, GENE does not have sufficient information to determine if the condition exists that could result in a failure of the Level 3 trip to occur. These plants are notified, as a Transfer of Information under 10CFR21.21(b), as identified in Attachment 1 [of GENE Part 21 final report to the NRC dated October 11, 2004]."
The plants where insufficient information for a final determination have been identified by GENE as Columbia, Susquehanna 1 and 2, Browns Ferry 2 and 3.
Notified R1DO (Barkley), R2DO(Ogle), R3DO (Clayton) and R4DO (Bywater) and NRR (Hodge). |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 40983 | Facility: CLINTON
Region: 3 State: IL
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: STEVEN MENG
HQ OPS Officer: JOHN MacKINNON | Notification Date: 08/24/2004
Notification Time: 17:36 [ET]
Event Date: 08/24/2004
Event Time: 09:17 [CDT]
Last Update Date: 10/08/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
ERIC DUNCAN (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 95 | Power Operation | 95 | Power Operation | Event Text HIGH PRESSURE CORE SPRAY (HPCS) DECLARED INOPERABLE
At 1128 hours on 8/23/04, the Division 3 Essential Switchgear Heat Removal System (VX) was removed from service and declared inoperable for performance of system flow verification and balance. The test includes an as found flow check on the Division 3 Essential Switchgear Heat Removal System Condensing Unit, rendering the Division 3 VX safety-related chiller 1VX06CC INOPERABLE. The non-safety VX subsystem remained OPERABLE during the test.
"At 0917 hours on 8/24/04, the non-safety Division 3 VX Heat Removal Supply Fan 1VX04CC, tripped due to the breaker for the safety-related fan being removed for replacement. The Main Control Room received alarm 5042-6A, Auto Trip Pump/Fan. Since both the safety and non-safety subsystems of VX were unavailable Operators declared the High Pressure Core Spray (HPCS) System inoperable per Technical Specification 3.7.2, Action A.1.
"At 1153 hours, the breaker replacement was complete, 1VX04CC was restored to service, and the HPCS System was declared OPERABLE.
"The VX System maintains safety-related switchgear, battery and inverter room, and cable spread areas within the design temperature limits of the equipment. The VX system is support system for the HPCS System. With both subsystems of the VX System out of service, the HPCS System may not have been capable of performing its safety function to provide Emergency Core Cooling, aid in depressurization and maintain reactor vessel water level following a loss of coolant accident.
"An engineering evaluation is currently in progress to determine if the HPCS System would have been capable of performing its safety function with both safety and non-safety subsystems of VX out of service.
"This issue is being reported in accordance with 10CFR50.72(b)(3)(v)(D), as an event or condition that at the time of discovery could have prevented the fulfillment of the safety function needed to mitigate the consequences of an accident."
The NRC Resident Inspector was notified of this event by the licensee.
* * * RETRACTION FROM BILL CARSKY TO BILL HUFFMAN AT 17:47 EDT ON 10/08/04 * * *
"Upon further review of this event, additional analysis has been performed which bounds the design bases heatup of the associated rooms cooled by the Division III Essential Switchgear Heat Removal System (VX). This analysis concludes that the areas cooled by the Division 3 VX subsystem would not have exceeded design temperatures while the cooling was secured, prior to cooling recovery, and that the supported systems remained operable.
"Based upon this additional analysis, it can be reasonably concluded that the safety function of High Pressure Core Spray, as a single train safety system, was fulfilled. Therefore this event is not reportable and Event #40983 is being retracted."
The NRC Resident Inspector was notified of this retraction by the licensee. R3DO (Clayton) has been notified. |
| General Information or Other | Event Number: 41096 | Rep Org: NEW MEXICO RAD CONTROL PROGRAM
Licensee: RIVERSIDE TECHNOLOGIES
Region: 4
City: FAIRVIEW State: NM
County: RIO ARRIBA
License #: DM345
Agreement: Y
Docket:
NRC Notified By: WALTER MEDINA
HQ OPS Officer: BILL GOTT | Notification Date: 10/06/2004
Notification Time: 10:28 [ET]
Event Date: 10/01/2004
Event Time: [MDT]
Last Update Date: 10/06/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
RUSSELL BYWATER (R4)
SANDRA WASTLER (NMSS)
| Event Text AGREEMENT STATE REPORT
"Riverside Technologies notified the Radiation Control Bureau with the New Mexico Environment Department that one of their density moisture gauges was stolen sometime between Friday night, October 1, and Monday morning, October 4,2004. The locked storage room was broken into and the gauge was removed from its transport container and taken from the company office in Espanola, New Mexico. The gauge was a Campbell Pacific Nuclear, Model MC-3, serial number M30069661, and containing Cesium-137 and Americium-241/Be radioactive sealed sources. The gauge was locked in the safe position." |
| Power Reactor | Event Number: 41103 | Facility: HATCH
Region: 2 State: GA
Unit: [ ] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: GREG JOHNSON
HQ OPS Officer: HOWIE CROUCH | Notification Date: 10/08/2004
Notification Time: 09:13 [ET]
Event Date: 10/08/2004
Event Time: 01:50 [EDT]
Last Update Date: 10/08/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
| Person (Organization):
CHARLES R. OGLE (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 99 | Power Operation | 99 | Power Operation | Event Text HIGH PRESSURE COOLANT INJECTION (HPCI) SYSTEM DECLARED INOPERABLE
"On 10/08/04 on Unit Two, the HPCI Valve Operability was being performed. During the course of this evolution the suction path was transferred from the Condensate Storage Tank (CST) to the Suppression Pool. When the HPCI System was aligned to the Suppression Pool the Suction Pressure decreased from 25.5 psig to 1.5 psig. With HPCI aligned to the suppression pool and with suction pressure less than 14 psig the HPCI System was declared INOPERABLE.
"Investigation continues as to the cause of the low suction pressure. Preliminarily it is suspected that the Suppression Pool suction path was not adequately filled and vented following a recent tag out of that suction path for maintenance inspection activities. Investigation continues."
The Licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 41104 | Facility: NORTH ANNA
Region: 2 State: VA
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: DAVE NUNBERG
HQ OPS Officer: BILL HUFFMAN | Notification Date: 10/08/2004
Notification Time: 14:14 [ET]
Event Date: 10/08/2004
Event Time: 13:55 [EDT]
Last Update Date: 10/08/2004 | Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
| Person (Organization):
CHARLES R. OGLE (R2)
TERRY REISS (EO)
PETER WILSON (IRD)
JANIE EVERETTE (DHS)
JIM DUNKER (FEMA)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 74 | Power Operation | 74 | Power Operation | Event Text UNUSUAL EVENT DUE TO ACTUATION OF CARBON DIOXIDE FIRE SUPPRESSION SYSTEM IN TURBINE BUILDING
The licensee experienced an apparently spurious actuation of the CO2 (carbon dioxide) suppression system in the North Anna Unit 1 turbine building (specifically the main turbine, low pressure turbine, and exciter area). The licensee responded and determined there was no fire and secured the CO2 release within two minutes. The event was classified as an unusual event per item K.13 of the emergency action levels due to onsite release of a toxic gas.
The licensee has pulled the actuation fuses on the system, isolated the CO2 storage tank, and is in the process of ventilating the area. A fire watch has been posted and the licensee is checking CO2 levels throughout the area.
There were no injuries as a result of the discharge and no significant operational impairment. The licensee has notified appropriate state and local authorities as required due to the declaration of an unusual event. The licensee will also notify the NRC resident inspector.
* * * UPDATE 14:53 EDT ON 10/08/04 FROM DAVE NUNBERG TO BILL HUFFMAN * * *
The licensee terminated the unusual event at 14:45 EDT following inspection of the turbine building and confirmation that CO2 levels did not represent a personnel hazard. R2DO (Ogle); NRR EO (Reis); NSIR IRD (Wilson); FEMA (Dunker); and the DHS senior watch officer have been notified. |
| Power Reactor | Event Number: 41105 | Facility: GINNA
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP
NRC Notified By: ERIC MATZ
HQ OPS Officer: BILL HUFFMAN | Notification Date: 10/08/2004
Notification Time: 15:38 [ET]
Event Date: 10/08/2004
Event Time: 14:10 [EDT]
Last Update Date: 10/08/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
RICHARD BARKLEY (R1)
TERRT REIS (EO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text OFFISTE NOTIFICATION TO FIRE DEPARTMENT DUE TO NATURAL GAS LINE BREAK
A natural gas pipeline was broken by a backhoe while digging at a location between 300 to 400 feet outside the protected area fence southwest of the plant. The licensee made offsite notifications to local fire departments and Rochester Gas and Electric which have responded to the scene and are in the process of isolating the line break.
The broken gas line is 2.5 inches in diameter and the licensee does not consider the break a threat to the plant and will continue normal operation. The licensee is monitoring the situation and is monitoring for excessive natural gas concentrations.
The licensee has notified the NRC resident inspector. |
| Power Reactor | Event Number: 41106 | Facility: QUAD CITIES
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: BRIAN WAHLHEIM
HQ OPS Officer: BILL HUFFMAN | Notification Date: 10/08/2004
Notification Time: 22:58 [ET]
Event Date: 10/08/2004
Event Time: 21:20 [CDT]
Last Update Date: 10/08/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
BRENT CLAYTON (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 85 | Power Operation | 85 | Power Operation | | 2 | N | Y | 85 | Power Operation | 85 | Power Operation | Event Text CONTOL ROOM EMERGENCY VENTILATION SYSTEM DECLARED INOPERABLE
"At 2120 [hrs. CDT] on October 8, 2004, while performing the 'Control Room Emergency Ventilation System Test,' which verifies the integrity of the control room envelope, it was determined that the positive pressure requirement of greater than or equal to 0.125 inches water gauge for the control room envelope in Technical Specification Surveillance Requirement 3.7.4.4 could not be met for all specified test points. As a result, Control Room Emergency Ventilation System was declared inoperable and Technical Specification 3.7.4, Condition A was entered.
"Recently completed surveillance testing has demonstrated that a positive pressure ranging from 0.056 to 0.301 inches water gauge is being maintained in the control room envelope; therefore it is expected that the safety function is being met.
"However, this notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(D) because the Control Room Emergency Ventilation System is a single train safety system and the Technical Specification requirement is not met. The affect of the failure to meet the Technical Specification requirements on the ability to perform the safety function is continuing to be evaluated."
The licensee notified the NRC resident inspector. |
| Power Reactor | Event Number: 41107 | Facility: PEACH BOTTOM
Region: 1 State: PA
Unit: [ ] [3] [ ]
RX Type: [2] GE-4,[3] GE-4
NRC Notified By: WILLIAM DALTON
HQ OPS Officer: HOWIE CROUCH | Notification Date: 10/09/2004
Notification Time: 05:13 [ET]
Event Date: 10/09/2004
Event Time: 01:15 [EDT]
Last Update Date: 10/09/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
RICHARD BARKLEY (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 3 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text HIGH PRESSURE COOLANT INJECTION SYSTEM CONDENSATE STORAGE TANK SUCTION VALVE FAILED TO AUTO-CLOSE DURING SURVEILLANCE
The following information was obtained from the licensee via facsimile:
"During the performance of scheduled instrument surveillance testing it was discovered that the Unit 3 High Pressure Coolant Injection (HPCI) pump suction valve from the Condensate Storage Tank (CST) would not automatically close. This valve is designed to automatically close when both Torus suction valves are fully open. The CST suction valve is required to close to support operability of the HPCI System. The CST suction valve was subsequently closed with the Torus suction valves open to allow performance of the HPCI instrumentation testing. After the completion of scheduled instrumentation testing, the HPCI system will be restored to an operable status with pump suction remaining from the Torus and isolated from the CST. The failure of the CST suction valve to auto close is being investigated."
The licensee has informed the NRC Resident Inspector. |
| Power Reactor | Event Number: 41108 | Facility: OCONEE
Region: 2 State: SC
Unit: [ ] [ ] [3]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: KEVIN MOSES
HQ OPS Officer: HOWIE CROUCH | Notification Date: 10/10/2004
Notification Time: 00:51 [ET]
Event Date: 10/09/2004
Event Time: 19:45 [EDT]
Last Update Date: 10/10/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
CHARLES R. OGLE (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 3 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown | Event Text REACTOR BUILDING PRESSURE DECREASED BELOW SPECIFIED LIMITS
The following information was obtained from the licensee via facsimile:
"On October 9, 2004 at 1945 hrs. EST, Operators discovered that Oconee Nuclear Station Unit 3 Reactor Building pressure decreased to less than the limit specified in Oconee Selected Licensee Commitment 16.6.13, 'Additional Requirements to Support Low Pressure Injection (LPI) Operability'. This initial Reactor Building Pressure is used in the NPSH Analysis for the LPI Pumps in the Sump Recirculation phase of post-LOCA operation. In addition to effecting LPl, this condition also affects the Reactor Building Spray system. Engineering Evaluation performed on May 25, 2004 identified that guidance contained within the Selected Licensee Commitment may be inadequate and as a result, the Reactor Building Spray and Low Pressure Injection systems were determined to be Operable But Degraded/Nonconforming. For the interim, until appropriate changes are made to the Selected Licensee Commitment, Engineering recommended that Oconee Operations enter Technical Specification 3.0.3 any time that the limits of Selected Licensee Commitment 16.6.13 are exceeded.
"At the time of discovery, Oconee Unit 3 was cooling down for refueling outage. Unit 3 was in Mode 4 at approximately 235 degrees, 275 psig with one train of Reactor Building Spray deactivated per the Controlling Procedure for Unit Shutdown. When it was identified that the limits of Selected Licensee Commitment 16.6.13 were exceeded, Operations began increasing Oconee Unit 3 Reactor Building pressure. Oconee Unit 3 Reactor Building pressure was restored to within the limits of Selected Licensee Commitment 16.6.13 on October 9, 2004 at 2023 hrs. EST.
"Initial Safety Significance:
"The NPSH analysis for the Low Pressure Injection pumps in the sump recirculation phase of post-LOCA operation credit reactor building overpressure of 2.2 psig as permitted by a license amendment granted July 19, 1999 and supplemented August 19, 1999. Operation with Reactor Building pressure less than the limits specified in Selected Licensee Commitment 16.6.13 cannot ensure that 2.2 psig overpressure will always be available.
"Corrective Action(s):
"Actions were taken by Operations to restore Unit 3 Reactor Building pressure to within the limits of Selected Licensee Commitment 16.6.13. These actions were successful in restoring Reactor Building Pressure to within limits of Selected Licensee Commitment 16.6.13 approximately 38 minutes from time of discovery."
The licensee has informed the NRC Resident Inspector. |
| Power Reactor | Event Number: 41109 | Facility: HOPE CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: BILL KOPCHICK
HQ OPS Officer: MIKE RIPLEY | Notification Date: 10/10/2004
Notification Time: 21:48 [ET]
Event Date: 10/10/2004
Event Time: 18:14 [EDT]
Last Update Date: 10/11/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(2)(iv)(A) - ECCS INJECTION
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
RICHARD BARKLEY (R1)
CHRISTOPHER GRIMES (NRR)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | M/R | Y | 69 | Power Operation | 0 | Hot Shutdown | Event Text MANUAL REACTOR SCRAM DUE TO A STEAM LEAK IN THE TURBINE BUILDING
"At 1814 [hrs. EDT] on October 10, 2004, Hope Creek Generating Station was manually scrammed due to a steam leak in the Turbine Building. All Control Rods inserted fully. Subsequent to the manual actuation of the Reactor Protection System, reactor pressure was reduced to minimize the effects of the steam leak. Degrading Main Condenser Vacuum following the scram resulted in trips of all operating Reactor Feed Pump Turbines at 10 [inches] HgA. The High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems were manually initiated for reactor level control and the Main Steam Isolation Valves (MSIV's) were closed to isolate the leak - MSIV closure was completed prior to reaching the Main Condenser Vacuum isolation setpoint of 21.5 [inches] HgA. During plant stabilization, Reactor Water Level lowered below the RPS actuation setpoint of 12.5 inches four separate times. First, following the initial scram. Second, immediately following initiation of the HPCI and RCIC systems, when the 'A' and 'B' Reactor Water Level channels lowered to -38 inches (Level 2). Level 2 is the HPCI and RCIC actuation setpoint and Primary Containment Isolation actuation setpoint for Groups 2, 7, 8, 9, 12, 13, 14, 17, 18, 19, and 20 valves. Because only two of the four Level 2 instrument channels actuated, the isolation of these systems was channel dependent and occurred as required by the respective isolation logic. Third, following manual closure of the MSIVs. Finally, Reactor Water Level lowered below 12.5 inches following reset of the original manual scram signal which resulted in an automatic scram signal. RCIC was re-initiated manually to restore Reactor Water Level.
"No personnel were injured during this event. The plant is currently stable in OPCON 3 with reactor pressure at 615 psig. Pressure control (decay heat removal) was transitioned to HPCI in pressure control mode during plant stabilization. Reactor Water Level is being maintained with the Secondary Condensate Pumps. Two loops of RHR in Suppression Pool Cooling mode are in service with Suppression Pool Temperature at 110 degrees F in compliance with Technical Specification 3.6.2.1 Action b.2. Actions to determine the cause of the steam leak and effect repairs are in progress."
The licensee will inform Lower Alloway Creek Township and has informed the NRC resident inspector.
* * * UPDATE ON 10/11/04 @ 0049 HRS EDT BY BAUER TO GOULD * * *
On steam leak investigation, a walk down of the turbine building condenser bay determined the source of the leak to be a failure of an 8 inch moisture separator dump line. The line break is located approximately one foot from the condenser shell penetration. An additional investigation into the root cause of the failure has commenced.
The NRC Resident Inspector was notified and Lower Alloway Creek Township will be notified.
The Reg 1 RDO (Richard Barkley) and EO (Chris Grimes) were informed.
HOO Note: See Event # 41110 |
| Power Reactor | Event Number: 41110 | Facility: HOPE CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: CLYDE BAUER
HQ OPS Officer: CHAUNCEY GOULD | Notification Date: 10/11/2004
Notification Time: 00:49 [ET]
Event Date: 10/10/2004
Event Time: 21:53 [EDT]
Last Update Date: 10/11/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
RICHARD BARKLEY (R1)
CHRISTOPHER GRIMES (NRR)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown | Event Text A RPS ACTUATION SIGNAL OCCURRED DUE TO LOW REACTOR WATER LEVEL WHILE IN MODE 3
At 2153 [hrs. EDT] on October 10, 2004, the Hope Creek Generating Station experienced an automatic reactor scram signal on low reactor level +12.5 inches (Level 3) while cooling down following a manual scram. As previously reported under Event Notification 41109, the Main Steam Isolation Valves (MSIV's) were closed as the result of a steam leak in the Turbine Building. The +12.5 inch (Level 3) scram occurred from the manual closure of a Safety Relief Valve (SRV) while it was being manually operated to reduce reactor pressure. The SRV was closed when reactor level was +24 inches, resulting in a reactor level shrink. Reactor level lowered to +8 inches, and stabilized. The secondary condensate pumps immediately restored reactor level to its normal band following the scram signal. SRV's were being utilized to assist the plant cool down because the High Pressure Coolant Injection (HPCI) system had been manually taken out of service. The HPCI vacuum tank vacuum pump tripped on an overload/power failure condition, and use was not desired. The Reactor Core Isolation Cooling (RCIC) system was out of service because of a high reactor level condition, due to plant cool down. Also, the Reactor Water Cleanup (RWCU) system was out of service due to the initial manual scram that occurred at 1814 hours which prevented normal reactor level blow down.
The NRC Resident Inspector was notified and Lower Alloway Creek Township will be notified.
HOO note: See Event # 41109 |
| Power Reactor | Event Number: 41111 | Facility: COOK
Region: 3 State: MI
Unit: [ ] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: TOD KASPAR
HQ OPS Officer: HOWIE CROUCH | Notification Date: 10/11/2004
Notification Time: 07:16 [ET]
Event Date: 10/11/2004
Event Time: 00:20 [EDT]
Last Update Date: 10/11/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
BRENT CLAYTON (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Refueling | 0 | Refueling | Event Text CONTAINMENT VENTILATION ISOLATION SYSTEM INOPERABLE DURING FUEL MOVEMENT
The following information was obtained from the licensee via facsimile:
"In accordance with 10 CFR 50.72(b)(3)(v)(D) ['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 mitigate the consequences of an accident'], D.C. Cook Unit 2 is making an 8-hour non-emergency report.
"At 00:20 [EDT] on 10-11-2004, it was discovered that neither train of the Containment Ventilation Isolation System would have automatically isolated containment purge on a high radiation signal within the containment. Also, a manual Phase A containment isolation actuation would not have isolated containment purge.
"At all times, containment purge could have been isolated using individual control switches from the Unit 2 Control Room.
"D.C. Cook Unit 2 Technical Specification 3.9.9 requires during core alterations or movement of irradiated fuel within the Containment that the Containment Purge and Exhaust Isolation System be operable. Technical Specification 3.9.4.c requires during core alterations or movement of irradiated fuel within the containment that each containment penetration providing direct access from the containment atmosphere to the outside atmosphere be closed by an isolation valve, blind flange, manual valve, or equivalent, OR be capable of being closed by an operable automatic Containment Purge and Exhaust isolation valve.
"Technical Specification 3.0.4 requires when a Limited Condition of Operation is not met, entry into an operational mode or other specified condition in the applicability shall be made only in specified conditions. Core alterations and movement of irradiated fuel assemblies began at 14:41 on 10-09-2004, approximately 71 minutes after the Containment Ventilation Isolation System had been made inoperable.
"At approximately 13:30 on 10-09-2004, breaker 2-CRID-1-7 (Reactor Protection and Safeguards Actuation Cabinet RPS-A Input Channel I & All RPS A Output) and 2-GRID-4-8 (Reactor Protection & Safeguards Actuation Cabinet RPS-B Input Channel IV & ALL RPS B) were opened and tagged as part of a clearance. This prevented the automatic actuation of both trains of the Containment Ventilation Isolation System and prevented a manual phase A containment isolation from isolating containment purge. Thereafter, at 14:41 on 10-09-04, fuel movement commenced from the reactor vessel to the Spent Fuel Pit.
"At 23:25 on 10-10-2004, breakers 2-CRID-1-7 and 2-CRID-4-8 were re-closed.
"At 0020 on 10-11-2004, it was discovered that during the time from approximately 13:30 on 10-09-2004 to 23:25 on 10-10-2004, the Containment Ventilation isolation system had been inoperable.
"D.C. Cook Unit 2 is currently stable in Mode 6 during the Unit 2 Cycle 15 refueling outage with core offload to the Spent Fuel Pit in progress."
The licensee has notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 41113 | Facility: WOLF CREEK
Region: 4 State: KS
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP
NRC Notified By: MARK JENKINS
HQ OPS Officer: STEVE SANDIN | Notification Date: 10/11/2004
Notification Time: 13:35 [ET]
Event Date: 10/11/2004
Event Time: 07:25 [CDT]
Last Update Date: 10/11/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
RUSSELL BYWATER (R4)
CHRISTOPHER GRIMES (NRR)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 98 | Power Operation | Event Text LOSS OF SWITCHYARD WEST BUS
"On 10/11/2004 at 0725 CDT, Wolf Creek Generating Station experienced a loss of the west bus in the switchyard causing a loss of power to the Startup Transformer and the 'B' Train 4.16 Kv ESFAS bus NB02. The 'B' Emergency Diesel Generator started and loaded, as expected, to supply power to the NB02 bus. The shutdown sequencer started the required components. Turbine load was reduced by the control room staff following the expected start of the steam driven auxiliary feedwater pump to maintain reactor thermal power below license limits.
"Following the shutdown sequencer start of the 'B' Essential Service Water (ESW) pump it was noted the 'B' Control Room Air Conditioning unit condenser inlet end bell gasket had started leaking. The Control Room Air Conditioning unit was secured and ESW was isolated to and from the unit. All other equipment operated as required.
"System Operations and Site personnel are investigating the cause of the power loss to the west switchyard bus, no cause has been identified at this time. Turbine load is being reduced to 950 MWe net per System Operations request to ensure grid stability is maintained.
"The Senior Resident has been contacted concerning this issue."
The licensee intends to reduce reactor power to 78-80% while investigating the cause for the loss of the west bus. There is no indication of any malevolent intent involved. The "B" EDG will maintain loads on the NB02 bus. |
| Power Reactor | Event Number: 41114 | Facility: SUMMER
Region: 2 State: SC
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP
NRC Notified By: ROBERT SWEET
HQ OPS Officer: STEVE SANDIN | Notification Date: 10/11/2004
Notification Time: 13:51 [ET]
Event Date: 10/11/2004
Event Time: 10:57 [EDT]
Last Update Date: 10/11/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
CHARLES R. OGLE (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text LOSS OF EMERGENCY SIREN CAPABILITY FOLLOWING PLANNED SIREN MAINTENANCE
"On October 11, 2004, at 1057 [hrs. EDT] , while performing modification work to support security upgrades, the radio system controller that activates the plant offsite warning system was removed from service during a pre-planned activity. At 1148, the system was rebooted but failed to perform correctly. Approximately 70 % of the sirens were communicating with the system, while plant procedures direct that less than 75 % of the offsite warning system meets the reporting threshold. At 1226, the offsite warning system capability was restored to greater than 75%.
"Since the removal of the offsite notification network from service was a planned evolution, the county, state and the NRC resident inspector were notified prior to the removal from service. Appropriate compensatory measures were taken by the state and local agencies.
"The cause of the event was the new Federal Signal activation frequency not reloading after the outage since it was not hard-coded into the software for the radio system controller. This caused all 27 of the Federal Signal sirens from being activated. Three other sirens had pre-existing conditions which made 30 sirens inoperable out of a total of 106.
"The correct information was loaded into the software via modem and a silent siren test confirmed 93% capability at 1226. The frequency information was hard coded into the software to prevent future occurrences."
The licensee notified state/local agencies and the NRC resident inspector. | |