| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
12/05/2006 - 12/06/2006 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 42906 | Facility: BRUNSWICK
Region: 2 State: NC
Unit: [ ] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: MICHAEL POTTER
HQ OPS Officer: STEVE SANDIN | Notification Date: 10/13/2006
Notification Time: 16:55 [ET]
Event Date: 10/13/2006
Event Time: 13:57 [EDT]
Last Update Date: 12/05/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
JAY HENSON (R2)
|
| 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 HIGH PRESSURE COOLANT INJECTION (HPCI) SYSTEM DECLARED INOPERABLE DUE TO POTENTIAL TURBINE EXHAUST DIAPHRAGM FAILURE
"On October 13, 2006 at 1357 (EST) Unit 2 HPCI system was declared inoperable due to indications of a leaking rupture diaphragm in the turbine steam exhaust line.
"This determination was made when water was discovered in the E41-PSH-N012A and E41-PSH-N012C instrument lines during the performance of 0MST- HPCI23Q (HPCI Turbine Exhaust Diaphragm High Pressure Instrument Channel Calibration).
"TS LCO 3.5.1. 'ECCS - Operating,' Condition D was entered for the HPCI system being declared inoperable. The Reactor Core Isolation Cooling (RCIC) system was verified operable per Required Action D.1. Required Action D.2 of TS LCO 3.5.1 requires the HPCI system to be returned to operable status within 14 days.
"The initial safety significance of this condition is considered to be minimal. The RCIC system and all other required ECCS, are operable at this time.
"HPCI has been isolated and will be placed under clearance to allow the turbine exhaust diaphragm to be inspected and replaced if necessary
" The [NRC] Resident Inspector has been notified."
* * * RETRACTION FROM M. TURKAL TO P. SNYDER ON 12/05/06 * * *
"Upon further review, it has been determined that the HPCI system was not rendered inoperable as a result of the condition identified on October 13, 2006.
"The HPCI system is equipped with two rupture diaphragms on the steam exhaust line installed in series. Between the rupture discs are four instrument lines leading to pressure switches (i.e., E41-PSH-N012A, N012B, N012C, and N012D) and a header vent line that vents to the HPCI room atmosphere. The HPCI Turbine Exhaust Diaphragm Pressure - High signals are initiated from these pressure switches; which are required to be operable per Technical Specification 3.3.6.1, 'Primary Containment Isolation Instrumentation,' to isolate the HPCI exhaust line in the event of a degraded inner rupture disc, before the redundant outer disc is significantly challenged. This isolation provides equipment protection and [is] not assumed in any transient or accident analysis.
"The suspected leaking inner rupture diaphragm was confirmed to be fully intact and, as such, not a source of the water (i.e., approximately 1.125 quarts) in the E41-PSH-N012A and E41-PSH-N012C instrument lines. The most likely source of this water is residual water remaining from a rupture disc failure that occurred in November 2003. Engineering has evaluated the potential impact of the residual water and determined that both the HPCI system and the HPCI Turbine Exhaust Diaphragm Pressure - High isolation function remained operable. There was not sufficient water in the lines to affect the function of the HPCI rupture diaphragms if required. In addition, the quantity of water discovered would not have prevented a HPCI initiation, if required, as evidenced by successful operation of the Unit 2 HPCI system on at least 10 occasions since November 2003. The Technical Specification required function of the pressure switches was not impacted by the presence of residual water. Investigation of this condition is documented in the corrective action program is Nuclear Condition Report (NCR) 209265.
"On this basis, the HPCI system was capable of performing its function to mitigate the consequences of an accident and the issue is not reportable under 10 CFR 50.72(b)(3)(v)(D).
"The NRC Resident Inspector was notified of this retraction."
Notified R2DO (Landis). |
| General Information or Other | Event Number: 43019 | Rep Org: KANSAS DEPT OF HEALTH & ENVIRONMENT
Licensee: BOEING COMMERCIAL AIRPLANE GROUP
Region: 4
City: WITCHITA State: KS
County:
License #: 29-C064-01
Agreement: Y
Docket:
NRC Notified By: TOM CONLEY
HQ OPS Officer: JASON KOZAL | Notification Date: 12/01/2006
Notification Time: 16:55 [ET]
Event Date: 12/01/2006
Event Time: [CST]
Last Update Date: 12/01/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
RUSSELL BYWATER (R4)
SCOTT MOORE (NMSS)
|
| This material event contains a "Less than Cat 3" level of radioactive material. | Event Text AGREEMENT STATE REPORT - LOST SOURCE
"The licensee reported a lost static air gun containing 10 millicuries of Po-210. The source is approximately 1 year old and the licensee believes the source was disposed of while attached to an air hose which was thrown away."
The State is continuing its investigation into this event and will provide updated information when it is available.
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. |
| Power Reactor | Event Number: 43030 | Facility: BEAVER VALLEY
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: RONALD T. GREEN
HQ OPS Officer: PETE SNYDER | Notification Date: 12/05/2006
Notification Time: 11:04 [ET]
Event Date: 12/03/2006
Event Time: 18:04 [EST]
Last Update Date: 12/05/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
MARVIN SYKES (R1)
|
| 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 NON-FUNCTIONAL SAFETY PARAMETER DISPLAY SYSTEM
"At 1804 on 12/03/2006, both Safety Parameter Display System (SPDS) displays in the Beaver Valley Power Station (BVPS) Unit 1 control room were identified as not functioning. Initial evaluation indicated that the system would be returned in less than a day. However, subsequent evaluation determined that vendor assistance would be needed to return the system to operation. The vendor is currently performing maintenance, with the system anticipated to be returned to service today (12/05/2006). Thus, this is being reported as a loss of emergency assessment capability pursuant to 10 CFR 50.72(b)(3)(xiii).
"Other BVPS Unit 1 equipment that remain available to adequately address potential emergent events include the Inadequate Core Cooling Monitor (ICCM), the BVPS Unit 1 Plant Computer and the Control Room alarm system."
The licensee notified the NRC Resident Inspector.
* * * UPDATE FROM RONALD T. GREEN TO JOE O'HARA AT 1449 ON 12/5/06 * * *
" At 1401 on 12/05/2006, the Safety Parameter Display System (SPDS) was repaired and returned to service at Beaver Valley Power Station (BVPS) Unit 1."
The licensee notified the NRC Resident Inspector.
Notified R1DO (M. Sykes) |
| Power Reactor | Event Number: 43031 | Facility: SUMMER
Region: 2 State: SC
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP
NRC Notified By: CHARLES MCKINNEY
HQ OPS Officer: JOE O'HARA | Notification Date: 12/05/2006
Notification Time: 13:52 [ET]
Event Date: 12/04/2006
Event Time: 13:25 [EST]
Last Update Date: 12/05/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
MARVIN SYKES (R1)
KERRY LANDIS (R2)
CHRISTINE LIPA (R3)
LINDA SMITH (R4)
VERN HODGE (EMAIL) ()
|
| 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 POTENTIAL DEFECTIVE WOODWARD GOVERNOR FOR FAIRBANKS MORSE EDG
"South Carolina Electric & Gas Company makes the following notification under 10 CFR 21.21(d)(3)(i) of a defect found during pre-installation testing of a Fairbanks Morse Woodward 2301 A/DRU governor. This governor was not installed on the Virgil C. Summer Nuclear Station Emergency Diesel Generators (EDG).
"The Fairbanks failure mode effect analysis (FMEA) states that when 125 VDC power is lost to the governor Digital Reference Unit (DRU), the speed reference signal goes to 0 VDC and that the Control Module will respond to drive the EDG to rated speed of 514 rpm. When this failure mode was tested, removal of the 125 VDC power resulted in a speed reference signal of -4 VDC. This would result in the governor Control Module driving the EDG to the low speed set point of 300 rpm. The failure to achieve rated speed would prevent the EDG from performing its safety function.
"Following a loss of 125 VDC power, the DRU must be manually reset in order for the Control Module to drive the EDG to rated speed. Users of this governor, that are not aware of this failure mode, may be unaware that the in-service EDG would be unable to meet the Technical Specification requirements unless surveillance tests were performed following the loss of power."
The licensee will notify the NRC Resident Inspector. |
| General Information or Other | Event Number: 43033 | Rep Org: ROSEMOUNT NUCLEAR INSTRUMENTS INC.
Licensee: ROSEMOUNT NUCLEAR INSTRUMENTS INC.
Region: 3
City: CHANHASSEN State: MN
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JEFFREY W. SCHMITT
HQ OPS Officer: JOE O'HARA | Notification Date: 12/05/2006
Notification Time: 17:56 [ET]
Event Date: 09/18/2006
Event Time: [CST]
Last Update Date: 12/05/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
CHRISTINE LIPA (R3)
LINDA SMITH (R4)
VERN HODGE EMAIL ()
| Event Text POTENTIALLY DEFECTIVE PRESSURE TRANSMITTERS SHIPPED TO DIABLO CANYON
"Notification under 10 CFR Part 21 for Model 1154 Series H Pressure Transmitters pursuant to 10 CFR Part 21, section 21.21(a), Rosemount Nuclear Instruments, Inc. (RNII) is writing to inform you that: (a) two Model 1154 Series H pressure transmitters shipped to Pacific Gas & Electric on September 18, 2006, may not conform to RNII's published specification's under accident conditions.
"1.0 Name and address of the Individual providing the information: Mr. Jeffrey W. Schmitt, Vice President & General Manager Rosemount Nuclear Instruments, Inc. 8200 Market Blvd Chanhassen, MN 55317
"2.0 Identification of Items supplied: Two Model 1154 Series H pressure transmitters are affected, as shown below:
Purchase Order - 130124, Sales Order 1732406, Model Number - 1154DH5RC, Serial Number(s) - 0533268, 0533275 Ship date - 18SEP06, Site - Diablo Canyon
"3.0 Identification of firms supplying the item: Rosemount Nuclear Instruments, Inc. 8200 Market Blvd Chanhassen, MN 55317
"4.0 Nature of the failure and potential safety hazard: RNII Model 1154 and Model 1154 Series H transmitters are similar in construction except that the Model 1154 Series H transmitters have a metal shroud, filled with an insulating material, surrounding the sensor module. This feature assures that the Model 1154 Series H transmitter can perform to specifications under certain accident conditions.
"When an order for a pressure transmitter is released to production, a travel card is generated by RNII's business operating system. In the production process a specific list of manufacturing and inspection procedures is joined to the travel card based on the transmitter model number pre-printed on the travel card. These procedures are completed in the order listed to produce the specified pressure transmitter.
"For the two transmitters in question, a list of procedures pertaining to the Model 1154 was attached to the travel card, rather than the procedures for the Model 1154 Series H transmitter. This error was not detected during the production process, and, as a result, the transmitters were assembled and inspected as Model 1154 transmitters.
"5.0 The corrective action which is taken, the name of the individual or organization responsible for that action, and the length of time taken to complete that action: (a) As a precautionary measure, RNII held shipments of all Model 1154 Series H transmitters while production records were reviewed for discrepancies. No additional suspect transmitters were identified. (b) All RNII transmitter travel cards for the prior three years of production were examined for similar discrepancies. No additional suspect transmitters were identified. (c) RNII's field returns history was reviewed for similar nonconformances. None were found. (d) An internal corrective action request was initiated. Corrective actions have been implemented to eliminate the potential for attaching the incorrect list of production procedures. (Completed: Dec 4, 2006)
"6.0 Any advice related to the potential failure of the item: The end user is advised to determine the impact of this defect upon its plant's operation and safety, and take action as deemed necessary. RNII recommends that the affected transmitters be returned to RNII for rework. (Arrangements have been made with PG&E to return the affected transmitters.)" | |