| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
05/04/2011 - 05/05/2011 ** EVENT NUMBERS ** |
 | Power Reactor | Event Number: 46813 | Facility: DUANE ARNOLD
Region: 3 State: IA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: KEN CARLSON
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 05/03/2011
Notification Time: 15:39 [ET]
Event Date: 05/03/2011
Event Time: 13:44 [CDT]
Last Update Date: 05/04/2011 | Emergency Class: ALERT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
50.72(b)(1) - DEVIATION FROM T SPEC
| Person (Organization):
MARK RING (R3DO)
ANNE BOLAND (R3 R)
BRUCE BOGER (NRR)
STEVEN WEST (R3)
WILLIAM GOTT (IRD)
R. BRADSHAW (DHS)
T. KUZIA (FEMA)
S. MORRONE (DOE)
B. EMERSON (HHS)
R. JONES (USDA)
|
| 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 ALERT DECLARED DUE TO FIRE AFFECTING ACCESS TO SAFETY RELATED EQUIPMENT
The licensee has declared an Alert due to a hydrogen fire on the hydrogen pad outside the protected area that resulted in evacuating outbuildings (pump house and air compressor buildings) that contain safety related equipment. The Alert is classified under Emergency Action Level HA-3.2 based on report or detection of gases in concentrations higher than lower flammability limit or contiguous to a safe shutdown or vital area. The licensee has also suspended certain security measures under 50.54(x) and 50.54(y). Security measures were suspended due to risks from the fire with no compensatory measures in-place at this time.
The fire involves hydrogen cylinders delivered to the site to a trailer pad outside the protected area. The fire developed while exchanging a newly delivered trailer of hydrogen cylinders with an expended hydrogen cylinder trailer.
The fire brigade has responded and offsite fire departments have responded to the site. Fire water is currently being sprayed on the hydrogen cylinders. There is no indication of any damage to any plant equipment other than the hydrogen trailer area equipment. The driver of the hydrogen supply truck was reported to have sustained some injuries. There are no other reported injuries to plant personnel.
The hydrogen water chemistry system has been isolated and makeup hydrogen to the main generator is isolated. The hydrogen isolation has no impact on the licensee at this time. The plant continues to operate at full power and there have been no immediate impacts from the fire to plant operation.
The licensee has notified State, and local authorities and the NRC Resident Inspector.
* * * UPDATE FROM SCHWERTFEGER TO TEAL AT 1550 EDT ON 5/3/11 * * *
The licensee made a one-hour notification of deviation from technical specifications under 10 CFR 50.54(x) and (y) due to the ongoing hydrogen fire. This was due to safeguards system degradation related to monitoring and detection of the OCA. Contact the Headquarters Operations Officer for details.
R3 IRC was notified of this report.
* * * UPDATE FROM SCHWERTFEGER TO HUFFMAN AT 1925 EDT ON 5/3/11 * * *
The licensee reported all safeguards systems have been restored and they have exited 10 CFR 50.54(x) and (y).
R3 IRC was notified of this report.
* * * UPDATE FROM CARLSON TO HUFFMAN AT 2023 EDT ON 5/3/11 * * *
At 1855 CDT, Duane Arnold has re-classified the hydrogen trailer fire event to an Unusual Event level. The event is classified as an UE under EAL HU-5.1 based on other conditions that exist which, in the judgment of the emergency director, indicate events are in progress or have occurred which indicate a potential for degradation for the level of safety of the plant.
The licensee has allowed access to all areas that were previously restricted due to potential safety concerns from the ongoing fire. The hydrogen trailer cylinders are still being sprayed with water and several cylinders are still showing thermal images above ambient. The licensee will evaluate event termination when the fire can be conclusively demonstrated to be extinguished. The licensee has notified state and local authorities and the NRC Resident Inspector.
NRC Region 3 continues to monitor the event. Notified R3 (West), NRR EO (Holian), FEMA (Via), and DHS (Knox).
* * * UPDATE FROM RUSHWORTH TO KLCO AT 0302 EDT ON 5/4/11* * *
At 0135 CDT on 5/4/11, Duane Arnold terminated from the Unusual Event. The fire was verified to be out, with no hot spots. The hydrogen tanks have been isolated and the licensee verified no hydrogen is present in the area of the hydrogen pad.
The licensee has notified State, local authorities and the NRC Resident Inspector.
R3 IRC was notified of this report. Notified R3 (West), NRR (Boger), FEMA (Via), DHS (Hill), HHS (Collins), USDA (Timmons) and DOE (Yates). Will notify R3DO (Ring), NRR EO (Holian), IRD (Marshall). |
| Power Reactor | Event Number: 46816 | Facility: BEAVER VALLEY
Region: 1 State: PA
Unit: [ ] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: DAVID HASER
HQ OPS Officer: HOWIE CROUCH | Notification Date: 05/04/2011
Notification Time: 10:25 [ET]
Event Date: 03/07/2011
Event Time: 11:18 [EDT]
Last Update Date: 05/04/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
| Person (Organization):
HAROLD GRAY (R1DO)
|
| 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 INVALID ACTUATION OF THE AUXILIARY FEEDWATER SYSTEM
"On March 7, 2011 at 1118 hours with Beaver Valley Power Station Unit 2 in Mode 5 during a scheduled refueling outage an invalid actuation of the Auxiliary Feedwater (AFW) System occurred. Instrumentation & Control (I&C) technicians were using a procedure to place simulated signals in two out of three channels of the narrow range level indication on all three steam generators (SG). During this activity, I&C technicians repositioned the input test relays to the test position on both channel 1 and channel 2 in two out of three SGs before inserting the simulated signals. This action resulted in a zero level input to the SG level circuits in two out of three SGs. An Engineered Safety Feature Actuation System (ESFAS) actuation signal for the Auxiliary Feedwater (AFW) System was generated based on having a two out of three low-low level signal in at least one SG (starts turbine driven AFW pump) and subsequently having a two out of three steam generators low-low level signal in at least two SGs (starts motor driven AFW pumps). The AFW initiation signal resulted in a successful automatic start of both the Train A and Train B motor driven AFW pumps. The AFW pump discharge flow control valves were closed per procedure prior to this event, so the motor driven AFW pumps operated on 100% recirculation flow following their automatic start. The turbine driven AFW pump did not start since steam pressure was not present in the main steam lines due to the plant being in Mode 5. A Reactor Protection Signal (RPS) signal was also generated due to having a two out of three low-low level signal in at least one SG. The RPS actuation signal did not result in opening of the reactor trip breakers since the breakers were previously opened as part of the plant shutdown procedures. The reactor control rods had been fully inserted into the reactor core prior to this event. Plant operators took appropriate actions to secure AFW flow after determining that the actuation was invalid and the issue was entered into the corrective action program for evaluation.
"The initiation of an ESFAS and RPS actuation signal, due to a SG low-low level signal, was not in response to any valid system or plant condition. There was no event, transient or condition that required any type of mitigation in Mode 5. Plant equipment responded as expected based on the conditions prior to the event.
'The invalid actuation of the AFW system is reportable pursuant to 10 CFR 50.73(a)(2)(iv)(A) since it involved an actuation of a PWR auxiliary feedwater system train per 10 CFR 50.73(a)(2)(iv)(B)(6). Per 10 CFR 50.73(a)(2)(iv)(A)(2)(ii), the actuation of the RPS due to a SG low-low signal is not reportable since the actuation was invalid and occurred after the safety function had already been completed (i.e. the RPS actuation occurred after the control rods had already been inserted in the core).
"Pursuant to 10 CFR 50.73(a)(1), this event is being reported via this telephone notification, in lieu of a written Licensee Event Report, since the automatic actuation of both trains of AFW was invalid."
The licensee has notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 46817 | Facility: BEAVER VALLEY
Region: 1 State: PA
Unit: [ ] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: DAVID HASER
HQ OPS Officer: HOWIE CROUCH | Notification Date: 05/04/2011
Notification Time: 10:35 [ET]
Event Date: 03/15/2011
Event Time: 17:16 [EDT]
Last Update Date: 05/04/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
| Person (Organization):
HAROLD GRAY (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Refueling | 0 | Refueling | Event Text INVALID ACTUATION OF THE STANDBY SERVICE WATER PUMP
"On March 15, 2011 with Beaver Valley Power Station Unit 2 in Mode 6 during a scheduled refueling outage, the 'A' Standby Service Water (SWE) pump 2SWE-P21A automatically started at 1716 hours when electrical power was removed from the Secondary Process Rack 'A' . The removal of power from the process rack caused pressure transmitter 2SWS-PT113A (Service Water Pump Discharge Pressure Transmitter) to read low resulting in an invalid automatic start of pump 2SWE-P21A. As a result of the automatic start of the SWE pump, the SWE pump discharge isolation valve (2SWE-MOV116A) also opened to provide a flow path from the SWE header to the main service water supply header. The SWE pump is designed to automatically start on a low service water header pressure (i.e. < 34 psig). At the time of the actuation, the service water header pressure was stable at 93 psig. This invalid actuation resulted in a successful start of the Train 'A' SWE system which functioned properly. Plant operators secured the SWE pump after determining that the pump start was invalid and entered the issue into the corrective action program for evaluation.
"The SWE system is required by the Beaver Valley Unit 2 Licensing Requirements Manual requirement number 3.7.5 to be Functional in plant operating Modes 1, 2, 3, and 4. The SWE system is not normally in operation and serves as a backup cooling water supply on a loss of the main intake structure. The non-safety related automatic start of the SWE pump on low service water pressure is provided to prevent an inadvertent plant trip on a loss of a running service water pump and is not required for the design basis event. The plant was in Mode 6 when the invalid actuation occurred. The automatic start of the SWE pump was not initiated in response to any plant event.
"This event is reportable 10 CFR 50.73(a)(2)(iv)(A) since it involved an actuation of an emergency service water system as per 10 CFR 50.73(a)(2)(iv)(B)(9). Pursuant to 10 CFR 50.73(a)(1), this event is being reported via this telephone notification, in lieu of a written Licensee Event Report, since the automatic actuation of the Train 'A' SWE was invalid."
The licensee has notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 46820 | Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-2
NRC Notified By: JOHN CLARK
HQ OPS Officer: PETE SNYDER | Notification Date: 05/04/2011
Notification Time: 17:57 [ET]
Event Date: 05/04/2011
Event Time: 16:13 [EDT]
Last Update Date: 05/04/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
HAROLD GRAY (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text CHANGES IN FUEL VENDOR CALCULATION METHODOLOGY TO COMPLY WITH THE ECCS PERFORMANCE REQUIREMENTS OF 10 CFR 50.46(b).
"Oyster Creek has been informed of a change in its vendor's calculation of Peak Cladding Temperature (PCT) and the Maximum Local Oxidation (MLO) that is based on corrections to errors in the previous calculation of record the vendor has identified. Based on 10 CFR 50.46 Appendix K inputs and assumptions, the correction of errors resulted in an increase of 115 degrees F in the PCT for GE11 fuel and 145 degrees F for the GNF2 fuel. The impact on the MLO due to these errors resulted in an increase of 16.0% for GE11 fuel and 33.0% for the GNF2 fuel.
"To ensure compliance with the 10 CFR 50.46 requirements, administrative Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit adjustments have been applied to bring the PCT and MLO limits below the 10 CFR 50.46 criteria of 2200 degrees F and 17% acceptance criteria. The current LOCA analysis of record remains applicable with the applied MAPLHGR limit adjustments.
"This notification is being made as a result of the 10 CFR 50.46(a)(3)(ii) requirement to report this issue in accordance with 10 CFR50.72 and 10 CFR50.73.
"Based on current core thermal power level and existing margin to limits on power operation there is sufficient margin for analyzed accident scenarios requiring ECCS operation including appropriate MAPLHGR compensation to restore the PCT and the MLO within 10 CFR 50.46 acceptance criteria and therefore there is no impact on safe operation. As a result of the MAPLHGR limit adjustment the current LOCA analysis of record remains applicable and therefore, the offsite dose is still bounded by our current safety analysis. Therefore this event is not significant with respect to the health and safety of the public.
"Corrective Action(s):
"(1) Administrative adjustments to allowable MAPLHGR limits have been imposed to restore the applicability of the current LOCA analysis of record.
"(2) Revised MAPLHGR limits have been provided by the fuel vendor and will be implemented into the plant monitoring system, followed by removal of the administrative adjustments to the MAPLGHR.
"(3) A 10 CFR 50.46 report will be submitted within 30-days."
The licensee will notify the NRC Resident Inspector. |
| Part 21 | Event Number: 46821 | Rep Org: SOR INC PROCESS INSTRUMENTATION
Licensee: SOR INC PROCESS INSTRUMENTATION
Region: 4
City: LENEXA State: KS
County:
License #:
Agreement: Y
Docket:
NRC Notified By: COLBERT TURNEY
HQ OPS Officer: BILL HUFFMAN | Notification Date: 05/04/2011
Notification Time: 12:14 [ET]
Event Date: 05/04/2011
Event Time: [CDT]
Last Update Date: 05/04/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
DAVID PROULX (R4DO)
PART 21 GROUP ()
MARK RING (R3DO)
| Event Text PRESSURE SWITCH WITH POSSIBLE SUB-STANDARD ELECTRICAL LEAD TERMINATIONS
SOR Process Instruments has determined that certain model pressure switches it has manufactured may have leads attached to the terminal blocks that are not in conformance with good manufacturing practice. The pressure switches in question may have lead wire attached to the terminal block in which the connection between the stripped, stranded wire and a crimped terminal may not be flush with the terminal end. SOR states that there have been no known field problems associated with this issue.
This condition applies to SOR model switches with terminal block "X" options 9013-737, 9013-747, 9013-7674. SOR has provided this information to those customers that purchased these switches. The U.S. plants that received these switches are Braidwood and Duane Arnold. |
| Power Reactor | Event Number: 46822 | Facility: FITZPATRICK
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: DAVID RICHARDSON
HQ OPS Officer: JOE O'HARA | Notification Date: 05/05/2011
Notification Time: 02:06 [ET]
Event Date: 05/05/2011
Event Time: 04:00 [EDT]
Last Update Date: 05/05/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
HAROLD GRAY (R1DO)
|
| 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 TECHNICAL SUPPORT SYSTEM VENTILATION OUT OF SERVICE FOR PLANNED MAINTENANCE
"A planned maintenance evolution at the James A. FitzPatrick (JAF) Nuclear Power Plant will remove the Technical Support Center (TSC) ventilation system from service. The TSC ventilation system will be rendered non-functional during the course of the work activity. The TSC ventilation is expected to be out of service for approximately 9 hours from 0400 to 1300 today.
"If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature, radiological or other conditions. If relocation of the TSC becomes necessary, the station Emergency Plant Manager will relocate the TSC staff to an alternate location in accordance with applicable site procedures giving first consideration to the Control Room. TSC facility leads have been made aware of this contingency.
"This notification is being made in accordance with 10CFR50.72 (b)(3)(xiii) due to the potential loss of an emergency response facility (ERF). An update will be provided one the TSC ventilation has been restored to normal operation."
The NRC Resident Inspector has been notified. | |