| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
10/04/2013 - 10/07/2013 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 49335 | Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] GE-3
NRC Notified By: RONALD BLENKER
HQ OPS Officer: PETE SNYDER | Notification Date: 09/09/2013
Notification Time: 22:39 [ET]
Event Date: 09/09/2013
Event Time: 14:00 [CDT]
Last Update Date: 10/04/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
CHRISTINE LIPA (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 99 | Power Operation | 99 | Power Operation | Event Text HIGH ENERGY LINE BREAK BARRIER IMPROPERLY OBSTRUCTED
"At 1400 [CDT] on September 9, 2013, plant personnel found HELB barrier HATCH-1/TB blocked. The hatch is diamond plate steel located on the turbine floor. Pallets, a fan and a gantry were positioned on top of the hatch possibly preventing pressure relief during a HELB event. This issue is being reported as an unanalyzed condition per 10CFR50.72(b)(3)(ii)(B). All items were removed from the hatch by 1800 on September 9, 2013."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION FROM JEREMY TANNER TO JOHN SHOEMAKER AT 1118 EDT ON 10/04/13 * * *
"The licensee reviewed the design basis calculations and analyses for HELB events in the area where Hatch-1/TB is located. The review determined that it is acceptable to block or hold Hatch-1/TB down and that the painted markings on the hatch are overly restrictive.
"In looking at HELB calculation of record, feed water break at the feed water pumps, there is no flow path modeled between the HELB volumes. Therefore, if Hatch-1/TB is blocked, the physical flow path is in accordance with the Gothic model of the HELB volume. Further, the analyses also indicated that no credit is taken for the hatch to relieve as no HELBs are postulated in the room under the hatch. Finally, Hatch 1/TB structural integrity was verified assuming a HELB occurred with the hatch blocked as described in notification 49335 (pallets, fan and gantry) the barrier would function as designed.
"Licensee initiated a Work Request to remove any markings on Hatch-l/TB that indicate do not block or hold down."
The licensee will notify the NRC Resident Inspector.
The Region 3 Duty Officer (Valos) was notified. |
| Power Reactor | Event Number: 49407 | Facility: QUAD CITIES
Region: 3 State: IL
Unit: [ ] [2] [ ]
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: MARK BRIDGES
HQ OPS Officer: PETE SNYDER | Notification Date: 10/04/2013
Notification Time: 02:12 [ET]
Event Date: 10/03/2013
Event Time: 20:45 [CDT]
Last Update Date: 10/04/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
NICK VALOS (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text UNIT 2 HIGH PRESSURE COOLANT INJECTION (HPCI) INOPERABLE DUE TO DRAIN LINE LEAK
"On October 3, 2013, at 2045 [CDT] hours, a defect (pinhole through-wall leak) was identified on the drain line for the LS 2-2365, HPCI TURBINE INLET DRAIN POT LEVEL SWITCH. The defect was identified during investigation of leakage near LS 2-2365. The LS 2-2365, HPCI TURBINE INLET DRAIN POT LEVEL SWITCH, is provided to detect a failure of the HPCI steam trap during standby line-up. The location of the defect, is in Class 2 Safety related piping. HPCI is a single train safety system and this notification is being made in accordance with 10CFR50.72(b)(3)(v)(D). The instrument isolations for LS 2-2365 have been close and the leak has been isolated."
There is no increase to plant risk and RCIC (Reactor Core Isolation Cooling) is available.
The licensee will inform the NRC Resident Inspector. |
| Research Reactor | Event Number: 49409 | Facility: UNIV OF WISCONSIN
RX Type: 1000 KW TRIGA (CONVERSION)
Comments:
Region: 3
City: MADISON State: WI
County: DANE
License #: R-74
Agreement: Y
Docket: 05000156
NRC Notified By: ROBERT AGASIE
HQ OPS Officer: DANIEL MILLS | Notification Date: 10/04/2013
Notification Time: 16:34 [ET]
Event Date: 10/04/2013
Event Time: 10:01 [CDT]
Last Update Date: 10/04/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
NON-POWER REACTOR EVENT
| Person (Organization):
GEOFFREY WERTZ (NRR)
NICK VALOS (R3DO)
| Event Text TECHNICAL SPECIFICATION VIOLATION DUE TO OPERATOR LEAVING CONTROL ROOM
"Technical Specification 6.7.2(1)(c) Special Report Violation of Technical Specification 6.1.3(1)(a)
"The following special report is being made in accordance with the requirements of Technical Specification (TS) 6.7.2(1)(c) which states in part: There shall be a report not later than the following day by telephone or similar conveyance to the NRC Headquarters Operation Center of any reportable occurrence as defined in TS 1.3. TS 1.3 defines an observed inadequacy in the implementation of administrative or procedural controls, such that the inadequacy caused the existence of a condition which results in a violation of technical specifications as a reportable occurrence.
"Specifically, on Friday October 4, 2013, while performing a training session, an operator left the control room while the console key remained in the console key switch and the switch was in the ON position. This is in direct violation of the staffing requirements defined in TS 6.1.3(1)(a) which states a licensed reactor operator must be in the control room when the reactor is not secured. The reactor was not secured by the fact that the console key switch was not in the OFF position and the key remained in the console key switch. However, all control elements were fully inserted, a normal reactor SCRAM condition was present, the reactor was shut down, no work was in progress involving core fuel, core structure, control elements or drives and no experiments were being moved. This condition remained for a period of 67 seconds until the reactor operator returned to the control room. The reactor operator subsequently secured the reactor and notified the Reactor Supervisor. A review of security records confirms the control room remained locked and unoccupied during the entire duration.
"A follow-up written report to the NRC describing the circumstance of the event including an analysis of the cause of the occurrence, corrective actions and measures to prevent or reduce the probability of the recurrence shall be made no later than 14 days from today, October 4, 2013." |
| Power Reactor | Event Number: 49411 | Facility: PALO VERDE
Region: 4 State: AZ
Unit: [1] [2] [3]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: N. AARONSCOOKE
HQ OPS Officer: BILL HUFFMAN | Notification Date: 10/04/2013
Notification Time: 18:20 [ET]
Event Date: 10/04/2013
Event Time: 09:46 [MST]
Last Update Date: 10/04/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
HEATHER GEPFORD (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 | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | | 3 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text POSTULATED FIRE EVENT COULD RESULT IN A HOT SHORT THAT COULD ADVERSELY IMPACT SAFE SHUTDOWN EQUIPMENT
"A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits in the control room has determined the described condition to be applicable to Palo Verde Nuclear Generating Station resulting in an unanalyzed condition with respect to 10 CFR 50 Appendix R analysis requirements. The original plant wiring design and associated analysis for the Class 1E Train B and D batteries and chargers (including the BD Swing charger) control room ampere indications do not include overcurrent protection features to limit the fault current.
"In the postulated event, a fire in the control room could cause one of the ammeter wires to hot short to the ground plane; simultaneously, the fire causes another DC wire from the opposite polarity on the same battery to also hot short to the ground plane. This would cause a ground loop through the unprotected ammeter wiring. This event could result in excessive current flow (heating) in the ammeter wiring to the point of causing a secondary fire in the raceway system. The secondary fire could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10CFR50 Appendix R.
"The train B and D DC circuits were reviewed first since they are part of alternate safe shutdown capability for the control room fire event. An extent of condition review is ongoing for the A and C train DC circuits and other similar circuit designs that could potentially cause a secondary fire.
"This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Compensatory measures (fire watches) have been implemented for affected areas of the plant.
"The NRC Resident Inspector has been notified." |
| Power Reactor | Event Number: 49413 | Facility: BEAVER VALLEY
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: PATRICK HARTIG
HQ OPS Officer: PETE SNYDER | Notification Date: 10/05/2013
Notification Time: 01:19 [ET]
Event Date: 10/04/2013
Event Time: 23:24 [EDT]
Last Update Date: 10/05/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
JUDY JOUSTRA (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 UNEXPECTED CONTAINMENT LINER PENETRATION FOUND
"On October 4, 2013, during the Beaver Valley Unit 1 refueling outage, a planned visual examination of the interior containment liner and coatings was being performed. The containment design consists of an interior steel liner that is surrounded by reinforced concrete. An area approximately 0.4 inches by 0.28 inches was discovered that penetrated through the containment steel liner plate.
"With the plant currently shut down and in Mode 6 the containment as specified in Technical Specification 3.6.1 is not required to be operable. The cause of this discrepancy is currently being evaluated.
"This is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A).
"The Site Resident Inspector has been notified." |
| Power Reactor | Event Number: 49414 | Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-2
NRC Notified By: CHARLES SPAGNUOLO
HQ OPS Officer: BILL HUFFMAN | Notification Date: 10/06/2013
Notification Time: 14:26 [ET]
Event Date: 10/06/2013
Event Time: 11:30 [EDT]
Last Update Date: 10/06/2013 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
JUDY JOUSTRA (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | M/R | Y | 20 | Power Operation | 0 | Hot Shutdown | Event Text MANUAL REACTOR SCRAM DUE TO DEGRADED CONDENSER VACUUM
"Today at approximately 1040 [EDT], during a planned reactor power ascension with reactor power at approximately 20% of rated thermal power, main condenser vacuum began to lower. In accordance with the abnormal operating procedure for degrading vacuum, Operators inserted a manual scram of the reactor at 1130 [EDT]. The cause of the degraded vacuum is currently under investigation.
"All rods inserted into the core and all systems functioned as expected during the scram. No electromatic (EMRVs) or safety relief valves lifted during the transient. The plant is currently shutdown and parameters are stable. The plant is in its normal shutdown electrical lineup and decay heat is being removed via steam bypass valves to the main condenser.
"This event is reportable within 4 hours per 10CFR50.72(b)(2)(iv)(B) - any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation."
The licensee has notified the NRC Resident Inspector and will be notifying state authorities. | |