| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
05/29/2009 - 06/01/2009 ** EVENT NUMBERS ** |
 | Power Reactor | Event Number: 45074 | Facility: HOPE CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: MICHAEL REED
HQ OPS Officer: HOWIE CROUCH | Notification Date: 05/17/2009
Notification Time: 05:24 [ET]
Event Date: 05/17/2009
Event Time: 03:35 [EDT]
Last Update Date: 05/29/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
DANIEL HOLODY (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | M/R | Y | 100 | Power Operation | 0 | Hot Shutdown | Event Text MANUAL REACTOR SCRAM DUE TO LOSS OF SCRAM AIR HEADER
"At 0335, Hope Creek was manually scrammed due to indications of multiple control rods drifting. All rods indicate fully inserted. Reactor level is being controlled in the normal band with Start-up level control in automatic. Reactor pressure is being controlled by bypass valves to the main condenser. Recirc pumps are in service and no ECCS system actuations were reached. The failure is a solder joint on the air supply to HCU 22-15. A manual scram was reinserted at 0445 to mitigate the air leak."
The licensee reset the scram to re-pressurize the scram air header. Once the leak was located, a second manual scram signal was initiated to secure the leak.
No safety relief valves lifted during the transient. The electrical grid is stable and the plant is in a normal shutdown electrical lineup.
The licensee will be notifying the Lower Alloways Creek Township and has notified the NRC Resident Inspector.
* * * UPDATE ON 5/17/2009 AT 0552 FROM MICHAEL REED TO MARK ABRAMOVITZ * * *
The failure was on HCU 22-11 not 22-15.
The licensee notified the NRC Resident Inspector. Notified the R1DO (Holody) via e-mail.
* * * UPDATE ON 5/29/2009 AT 1133 FROM JIM PRIEST TO VINCE KLCO * * *
" On 5/17/09, at 0335, Hope Creek automatically scrammed due to low Reactor Pressure Vessel water level approximately two seconds prior to locking the Reactor Mode Switch in Shutdown due to indications of multiple control rods drifting. All rods indicate fully inserted. Reactor level is being controlled in the normal band with Start-up level control in automatic. Reactor pressure is being controlled by bypass valves to the main condenser. Recirc. Pumps are in service and no ECCS system actuations were reached. The failure is a solder joint on the air supply to HCU 22-11. A manual scram was reinserted at 0445 to mitigate the air leak."
The licensee notified the NRC Resident Inspector. Notified the R1DO(Dentel). |
| General Information or Other | Event Number: 45095 | Rep Org: WA DIVISION OF RADIATION PROTECTION
Licensee: ANVIL CORPORATION
Region: 4
City: BELLINGHAM State: WA
County:
License #: WN-IR031-1
Agreement: Y
Docket:
NRC Notified By: ARDEN SCROGGS
HQ OPS Officer: DONALD NORWOOD | Notification Date: 05/26/2009
Notification Time: 14:59 [ET]
Event Date: 06/06/2006
Event Time: [PDT]
Last Update Date: 05/26/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
MICHAEL SHANNON (R4DO)
MARK THAGGARD (FSME)
| Event Text AGREEMENT STATE REPORT - DROPPED RADIOGRAPHY DEVICE
The following was received from the State of Washington via e-mail regarding a previously unreported event:
"This is notification of an event in Washington State as reported to or investigated by the WA Department of Health, Office of Radiation Protection.
"A radiography crew was attempting to hoist a Co-60 radiography device to an elevated platform to perform work at an oil refinery. During the process of hoisting the device it slipped out of its rigging and fell an estimated 23 feet to the pavement below. No associated equipment was connected to the device and the plugs were still inserted. The metal flange on the device was dented but there was no other obvious damage. The device is an AEA model 741B, serial number B100. It contained a 5.5 Curie Co-60 Model 424-18 AEA source, serial number 2418.
"A radiation survey was performed by the radiography companies' crew and later by the RSO. The readings were normal for that device. A control cable was attached to the source inside the device and the source was rotated. It was determined to move freely and not bound in the shielded position within the device. Operations with this device were terminated and the device was sent back to its licensed storage location in Burlington, Washington. The radiography company sent the device back to QSA Global for evaluation and maintenance.
"The licensee's RSO stated that repairs were made to a bracket and the handle was replaced. The maintenance was performed. The device was resourced and relabeled and returned to the licensee.
"This incident, given [Washington state] report number WA-06-043 is closed. There was no release of activity. Personnel exposure was kept low so that no exposure limits were exceeded." |
| General Information or Other | Event Number: 45096 | Rep Org: MA RADIATION CONTROL PROGRAM
Licensee: ALLEGHENY RODNEY
Region: 1
City: NEW BEDFORD State: MA
County:
License #: G-0112
Agreement: Y
Docket:
NRC Notified By: BRUCE PACKARD
HQ OPS Officer: JOHN KNOKE | Notification Date: 05/26/2009
Notification Time: 17:17 [ET]
Event Date: 05/26/2009
Event Time: [EDT]
Last Update Date: 05/26/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
GLENN DENTEL (R1DO)
MARK THAGGARD (FSME)
| Event Text AGREEMENT STATE REPORT - GAUGE SHUTTER STUCK OPEN
The following was received from the State of Massachusetts via fax:
"A gauge on the Z-24 press is sometimes stuck open. The equipment is out of use. A licensed contractor was contacted to fix the gauge. Gauge s/n Z3239, model number SS-3a. Radionuclide is 1000 mCi of Am-241." |
| Power Reactor | Event Number: 45103 | Facility: COOPER
Region: 4 State: NE
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: STEVE WHEELER
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 05/29/2009
Notification Time: 03:37 [ET]
Event Date: 05/28/2009
Event Time: 22:06 [CDT]
Last Update Date: 05/29/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
MICHAEL SHANNON (R4DO)
|
| 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 CONTROL ROOM EMERGENCY FILTRATION SYSTEM EXCESSIVE LEAKAGE
"On 28 May 2009 the Control Room Emergency Filtration System (CREFS) was declared inoperable due to a degraded Control Room Envelope (CRE). Two CRE boundary doors were found with excessive leakage. Based on the identified leakage, reasonable assurance that CREFS would fulfill its safety function could not be established.
"The CRE boundary doors support CREFS at CNS. CREFS is a single train system and per 10CFR50.72(b)(3)(v)(D) an 8 hour report is required due to the fact that at the time of discovery this condition could have prevented the fulfillment of the safety function of an SSC [system, structure or component] that is required to mitigate the consequences of an accident.
"Actions to implement mitigating actions have been initiated in accordance with plant Technical Specifications.
"The NRC Senior Resident Inspector has been notified of the condition." |
| Power Reactor | Event Number: 45104 | Facility: BRAIDWOOD
Region: 3 State: IL
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JAMES SMIT
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 05/29/2009
Notification Time: 04:49 [ET]
Event Date: 05/28/2009
Event Time: 21:40 [CDT]
Last Update Date: 05/29/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP
| Person (Organization):
JULIO LARA (R3DO)
|
| 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 CONTROL POWER TO ECCS VALVES
"At 2140 on 5/28/09 Braidwood Station identified a loss of control power for a Safety Related MCC (Motor Control Center) which provided power to SVAG (Single Valve Actuation Group) valves associated with both trains of the ECCS system. The MCC is normally de-energized to maintain the valve power removed in accordance with Tech Specs for ECCS. Loss of the control power for the associated MCC would prevent operation of these valves, which would prevent realignment of components required for transfer to cold leg recirculation and hot leg recirculation for long term core cooling.
"Entry was made into LCO 3.5.2, ECCS Operating, and LCO 3.0.3 due to inoperability of both trains of ECCS based on the inability to realign portions of both trains of the ECCS system from injection to cold leg recirculation and subsequent hot leg recirculation. At 2230 on 5/28/09 preparations had been completed for ramp off line per LCO 3.0.3.
"Troubleshooting was performed and a blown control power fuse was identified and replaced at 2319 on 5/28/09. No Unit ramp was initiated.
"The NRC Resident Inspector was notified." |
| General Information or Other | Event Number: 45105 | Rep Org: ASCO VALVE INCORPORATED
Licensee: ASCO VALVE INCORPORATED
Region: 1
City: FLORHAM PARK State: NJ
County:
License #:
Agreement: N
Docket:
NRC Notified By: KAREN BOLIO
HQ OPS Officer: HOWIE CROUCH | Notification Date: 05/29/2009
Notification Time: 11:54 [ET]
Event Date: 04/29/2009
Event Time: [EDT]
Last Update Date: 05/29/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
GEORGE HOPPER (R2DO)
TOM HERRITY (email) (NRR)
OMID TABATAI (email) (NRO)
| Event Text PART 21 FOR MISSING PARTS ON ASCO SOLENOID VALVES
The following information was obtained from ASCO Valve QA Engineering via facsimile:
On 4/24/09 ASCO Engineering was made aware of a non-conformance in the valve assembly area. During the assembly of a solenoid valve, part number NPEFKX8300141EG 10688 125/DC, the valve assembler noticed that the valve was missing an insulating barrier and a wiring label used with screw terminal coils. Neither item was included on the valve Bill of Materials.
Purchase orders were reviewed and it was determined that North Anna Nuclear Power Station had purchased two of the subject valves. ASCO notified North Anna on April 27, 2009.
ASCO has taken measures to prevent future occurrences. |
| Power Reactor | Event Number: 45107 | Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [ ] [2] [ ]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: WILLIAM BAKER
HQ OPS Officer: HOWIE CROUCH | Notification Date: 05/31/2009
Notification Time: 17:39 [ET]
Event Date: 05/31/2009
Event Time: 13:30 [CDT]
Last Update Date: 05/31/2009 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
GEORGE HOPPER (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown | Event Text DEGRADED CONDITION DUE TO DISCOVERY OF PRESSURE BOUNDARY LEAKAGE
"During the performance of 2-SI-3.3.1.A ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping (ASME Section III, Class 1 and 2), a pressure boundary leak was identified on the RHR Shutdown Cooling Root Valve (2-SHV-074-0049) at a weld between the valve body and a 3/4 inch schedule 160 pipe nipple. The pipe nipple is a 6 inch long capped valve leak off line. This valve and line are ASME Code Class 1 equivalent components located between the A recirculation piping and the inboard Shutdown Cooling isolation valve (2-FCV-074-0048). This condition 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 with Shutdown Cooling in service.
"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 [Browns Ferry Nuclear] Corrective Action program as PER# 172551.
"Unit 1 and 3 remain at 100% power and are not affected by this event." | |