| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
03/25/2011 - 03/28/2011 ** EVENT NUMBERS ** |
 | Agreement State | Event Number: 46691 | Rep Org: UTAH DIVISION OF RADIATION CONTROL
Licensee: APPLIED GEOTECHNICAL ENGINEERING CONSULTANTS INC
Region: 4
City: LINDON State: UT
County:
License #: UT1800298
Agreement: Y
Docket:
NRC Notified By: GWYN GALLOWAY
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 03/22/2011
Notification Time: 20:45 [ET]
Event Date: 03/22/2011
Event Time: 17:35 [MDT]
Last Update Date: 03/22/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
GREG PICK (R4DO)
JIM WHITNEY (ILTA)
ANGELA MCINTOSH (FSME)
|
| This material event contains a "Less than Cat 3" level of radioactive material. | Event Text MOISTURE DENSITY GAUGE STOLEN AND RECOVERED
"A Troxler Electronic Laboratories, Inc. Model 3430, portable gauging device [serial number 22936, containing approximately 8.0 millicuries of cesium-137, and approximately 40 millicuries of americium-241/beryllium] was stolen from the licensee's vehicle while parked at the Home Depot in Lindon, Utah. The Cs-137 source was in the safe shielded position when it was stolen and the transportation case was also secured. The device had been secured by two independent physical barriers, but both barriers were breached. The device was recovered at approximately 5:55 p.m. MST by licensee personnel. The transportation case had been opened, but the source rod was still secured in the shielded position.
"The licensee's vehicle was an open bed pickup truck with a mechanism to secure the device as required."
Utah Report: UT110001
THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL
Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf |
| Power Reactor | Event Number: 46697 | Facility: NORTH ANNA
Region: 2 State: VA
Unit: [1] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: DON TAYLOR
HQ OPS Officer: CHARLES TEAL | Notification Date: 03/25/2011
Notification Time: 11:14 [ET]
Event Date: 03/24/2011
Event Time: 14:00 [EDT]
Last Update Date: 03/25/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
OTHER UNSPEC REQMNT
| Person (Organization):
DANIEL RICH (R2DO)
MARISSA BAILEY (NMSS)
|
| 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 | Event Text NON-COMPLIANCE WITH NUHOMS TECHNICAL SPECIFICATION
"This 24-hour report is being issued in accordance with the requirements of NRC Certificate of Compliance 1030, Amendment 0, for the NUHOMS Storage System, Technical Specification (TS) 2.2, Functional Operating Limit Violations.
"During a review of historical North Anna NUHOMS dry storage canister (DSC) loading certification documents, a discrepancy was identified. The NUHOMS Certificate of Compliance 1030 Amendment 0 Technical Specifications include a Figure 2, "Heat Load Zones" which specifies the maximum decay heat load for each of the 32 assembly locations in a DSC. The figure includes limits for two zone '1b' locations and two zone '1a' locations in the four center locations of the DSC. The zone '1b' decay heat limit of 0.8 kw is specified for the two 'upper compartments' and zone '1a' decay heat limit of 1.05 kw is specified for the two 'lower compartments' on the figure. Contrary to this, the loading certifications for 7 of 10 DSCs already loaded at NAPS (North Anna Power Station) were not developed to maintain this orientation when loaded in the horizontal storage module (HSM). As a result, the DSC zone '1b' heat load limits were exceeded in some cases for these 7 DSCs.
"The heat load limit for all other zones in the DSCs are symmetric, and those assemblies were verified to the correct limit and are unaffected by this error. In addition the total heat load limit for the sum of the center assemblies was met for all DSCs. The maximum heat load of any zone '1b' assembly at the time of loading was 0.859 kw, which is slightly higher than the 0.8 kw limit. The lower heat load of assemblies in the other compartments offset the slightly higher heat load effects, and it is expected that the thermal analysis acceptance criteria would still have been met at the time of loading.
"The decay heat of the assemblies has continued to decrease since their initial loading and it was confirmed that 12 of the 13 assemblies that initially exceeded the 0.8 kw limit currently meet the zone '1b' heat load limits. The current decay heat of the remaining assembly is slightly above the 0.8 kw limit. Based on the offsetting margins identified above all of the affected DSCs are currently in a safe condition as loaded in the HSMs."
The NRC Resident Inspector has been informed. |
| Power Reactor | Event Number: 46698 | Facility: SURRY
Region: 2 State: VA
Unit: [1] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: RETT GARNER
HQ OPS Officer: CHARLES TEAL | Notification Date: 03/25/2011
Notification Time: 11:40 [ET]
Event Date: 03/25/2011
Event Time: [EDT]
Last Update Date: 03/25/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
OTHER UNSPEC REQMNT
| Person (Organization):
DANIEL RICH (R2DO)
MARISSA BAILEY (NMSS)
|
| 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 | 98 | Power Operation | 98 | Power Operation | Event Text NON-COMPLIANCE WITH NUHOMS TECHNICAL SPECIFICATION
"This 24-hour report is being issued in accordance with the requirements of NRC Certificate of Compliance 1030, Amendment 0, for the NUHOMS Storage System, Technical Specification (TS) 2.2, Functional Operating Limit Violations.
"During a review of historical Surry NUHOMS Dry Storage Canister (DSC) loading certification documents, a discrepancy was identified. The NUHOMS Certificate of Compliance 1030 Amendment 0 Technical Specifications include a Figure 2, 'Heat Load Zones' which specifies the maximum decay heat load for each of the 32 assembly locations in a DSC. The figure includes limits for two zone '1b' locations and two zone '1a' locations in the four center locations of the DSC. The zone '1b' decay heat limit of 0.800 kw is specified for the two 'upper compartments' and zone '1a' decay heat limit of 1.05 kw is specified for the two 'lower compartments' on the figure. Contrary to this, the loading certifications for 6 of 12 DSCs already loaded at Surry were not developed to maintain this orientation when loaded in the horizontal storage module (HSM). As a result, the DSC zone '1b' heat load limits were exceeded in some cases for these 4 DSCs.
"The heat load limit for all other zones in the DSCs are symmetric, and those assemblies were verified to the correct limit and are unaffected by this error. In addition the total heat load limit for the sum of the center assemblies was met for all DSCs. The maximum heat load of any zone '1b' assembly at the time of loading was 0.806 kw, which is slightly higher than the 0.800 kw limit. The lower heat load of assemblies in the other compartments offset the slightly higher heat load effects, and it is expected that the thermal analysis acceptance criteria would still have been met at the time of loading.
"The decay heat of the assemblies has continued to decrease since their initial loading and all assemblies currently meet the upper central compartment limits. The affected fuel assemblies are in a safe condition as required by NUHOMS TS 2.2.1.
"The NRC Resident Inspector has been notified of this event." |
| Power Reactor | Event Number: 46699 | Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: TERRY BACON
HQ OPS Officer: BILL HUFFMAN | Notification Date: 03/25/2011
Notification Time: 17:27 [ET]
Event Date: 03/25/2011
Event Time: 10:53 [CDT]
Last Update Date: 03/25/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
JAMNES CAMERON (R3DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 99 | Power Operation | 98 | Power Operation | Event Text POTENTIAL UNANALYZED CONDITION DUE TO POWER LEVEL GREATER THAN LIMIT
"During performance of maintenance to troubleshoot the B feed water regulating bypass valve, the Thermal Power Monitor (TPM) indication exceeded the maximum thermal power assumed in the Safety Analysis Report. Operators were maintaining 12 Steam Generator (SG) water level in a band from 40 to 48 percent by controlling the B Feed Regulating Valve (FRV) in manual from the Control Room. Operators noted a power increase; adjustments were made via the FRV to reduce SG water level, however the valve response was sluggish and thermal power exceeded 100%. Immediate steps were taken to reduce power to below 100% by reducing 1st stage turbine pressure and inserting Bank D control rods 7 steps.
"The TPM indication was above the maximum thermal power limit of 100.36% for 1.68 minutes. The TPM indication peak was 100.39%. No concurrent increase in power was observed by the nuclear indication system.
"NRC Resident had been informed." |
| Power Reactor | Event Number: 46700 | Facility: BEAVER VALLEY
Region: 1 State: PA
Unit: [ ] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: DANIEL SCHWER
HQ OPS Officer: BILL HUFFMAN | Notification Date: 03/25/2011
Notification Time: 21:35 [ET]
Event Date: 03/25/2011
Event Time: 14:00 [EDT]
Last Update Date: 03/27/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
JOHN ROGGE (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 BOTH TRAINS OF EMERGENCY DIESEL GENERATORS ARE INOPERABLE
"On March 25, 2011, the Train B emergency diesel generator (2EGS-EG2-2) was inoperable and unavailable due to being out of service for scheduled maintenance. At 1400 hours, the Train A emergency diesel generator (2EGS-EG2-1) was declared inoperable, but available, after questions were raised about the adequacy of the assembly method for fuel injection line compression fittings by the manufacturer. Without assurance that the fittings meet full qualification requirements, the Train A emergency diesel generator was declared inoperable.
"The Unit is currently in Mode 6 with fuel loaded and the upper internals installed in the reactor vessel. The reactor vessel head is removed with 23 feet of water in the cavity, two Operable Residual Heat Removal trains one of which is in operation. With both emergency diesels inoperable, the safety functions needed for accident mitigation could be impaired in the event of a loss of off-site power. Actions are currently in progress to restore one emergency diesel generator to an Operable status.
"This event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(D) due to both emergency diesel generators being inoperable. This event will be evaluated for 10 CFR Part 21 applicability."
The licensee is in Technical Specification 3.8.2. With both diesels declared inoperable, they have to suspend all core alterations and possible reactivity additions and return an EDG to service. The licensee plans to return the B train EDG an available status (but not Operable per Technical Specification) by tomorrow. The licensee will then replace the discrepant fuel injector line compression fittings on the A train and return it to Technical Specification Operable status.
The licensee has notified the NRC Resident Inspector.
* * * UPDATE AT 0226 ON 3/27/11 FROM DANIEL SCHWER TO MARK ABRAMOVITZ * * *
"Following replacement of the questionable fuel injection line compression fittings and successful surveillance and post maintenance testing, the Train A emergency Diesel Generator (2EGS-EG2-1) was declared OPERABLE at 0058 hours on 3/27/2011. One Diesel Generator was maintained available at all times while the issue was being addressed."
The licensee notified the NRC Resident Inspector.
Notified the R1DO (Rogge). |
| Power Reactor | Event Number: 46701 | Facility: DIABLO CANYON
Region: 4 State: CA
Unit: [ ] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DOUGLAS DYE
HQ OPS Officer: HOWIE CROUCH | Notification Date: 03/26/2011
Notification Time: 21:10 [ET]
Event Date: 03/26/2011
Event Time: 14:49 [PDT]
Last Update Date: 03/26/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
GREG PICK (R4DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | M/R | Y | 88 | Power Operation | 0 | Hot Standby | Event Text MANUAL REACTOR TRIP WHEN A LOSS OF THE 2-1 MAIN FEEDWATER PUMP OCCURRED
"This notification provides the 4-hour non-emergency event report for the manual reactor trip of Diablo Canyon Power Plant Unit 2 in accordance with 10 CFR 50.72(b)(2)(iv)(B) 'RPS Actuation (scram)'. Additionally, this notification provides the 8-hour non-emergency event report of the automatic actuation of the auxiliary feedwater system as a result of the reactor trip in accordance with 10 CFR 50.72(b)(3)(iv)(A) 'Specified System Actuation'.
"On March 26, 2011 at 1449 PDT operators at Diablo Canyon Power Plant Unit 2 manually initiated a reactor trip in response to loss of main feedwater pump 2-1. The main feedwater pump is believed to have tripped due to non-radioactive water spray on its control console. The water spray was caused by leakage from the flange of the relief valve on the feedwater heater 2-1A condenser dump valve line. Emergency plan activation was not required.
The unit is stable in mode 3 (Hot Standby) with offsite power being supplied to all buses via the 230 kV startup circuit. Diesel generators 2-1 and 2-2 remain OPERABLE in standby. Diesel generator 2-3 remains unavailable due to pre-planned maintenance. All rods fully inserted on the reactor trip. The reactor is being cooled by the auxiliary feedwater system with the condenser in service. All systems performed as designed with no unexpected pressure or level transients. ECCS actuation was not required. Automatic main feedwater isolation, auxiliary feedwater actuation, and steam generator blowdown isolation occurred as expected.
"Unit 1 was unaffected by this event and remains at 100% power."
The licensee has notified the NRC Resident Inspector, San Luis Obispo County, and the State of California. | |