Event Notification Report for October 17, 2006
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
10/16/2006 - 10/17/2006
EVENT NUMBERS
4291342914429154291642917
Power Reactor
Event Number: 42913
Facility: POINT BEACH
Region: 3 State: WI
Unit: [] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: RYAN RODE
HQ OPS Officer: JOHN KNOKE
Region: 3 State: WI
Unit: [] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: RYAN RODE
HQ OPS Officer: JOHN KNOKE
Notification Date: 10/17/2006
Notification Time: 10:22 [ET]
Event Date: 10/17/2006
Event Time: 08:32 [CDT]
Last Update Date: 10/17/2006
Notification Time: 10:22 [ET]
Event Date: 10/17/2006
Event Time: 08:32 [CDT]
Last Update Date: 10/17/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
Person (Organization):
RICHARD SKOKOWSKI (R3)
RICHARD SKOKOWSKI (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown |
OFFSITE NOTIFICATION - PRESS RELEASE
"On 10/17/06 at 0832 Nuclear Management Company, LLC issued press release concerning the Unusual Event at Point Beach Nuclear Plant that was declared on 10/16/06 at 21:26 CDT. The Unusual Event was previously reported via EN# 42912 and occurred as a result of a momentary loss of power to the power-operated relief valve (PORV) instrumentation which caused the valve to open. The valve opening resulted in reactor coolant leakage into the pressurizer relief tank. The tank is located in the Unit 2 containment building and is designed to collect water flow from the PORV's. At no time was reactor coolant released outside of the piping systems. The NRC Resident Inspector has been informed of the press release and this subsequent notification."
"On 10/17/06 at 0832 Nuclear Management Company, LLC issued press release concerning the Unusual Event at Point Beach Nuclear Plant that was declared on 10/16/06 at 21:26 CDT. The Unusual Event was previously reported via EN# 42912 and occurred as a result of a momentary loss of power to the power-operated relief valve (PORV) instrumentation which caused the valve to open. The valve opening resulted in reactor coolant leakage into the pressurizer relief tank. The tank is located in the Unit 2 containment building and is designed to collect water flow from the PORV's. At no time was reactor coolant released outside of the piping systems. The NRC Resident Inspector has been informed of the press release and this subsequent notification."
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 42914
Facility: HATCH
Region: 2 State: GA
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: BARRY D. COLEMAN
HQ OPS Officer: JOE O'HARA
Region: 2 State: GA
Unit: [1] [] []
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: BARRY D. COLEMAN
HQ OPS Officer: JOE O'HARA
Notification Date: 10/17/2006
Notification Time: 17:02 [ET]
Event Date: 10/17/2006
Event Time: 10:21 [EST]
Last Update Date: 11/16/2006
Notification Time: 17:02 [ET]
Event Date: 10/17/2006
Event Time: 10:21 [EST]
Last Update Date: 11/16/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
JAY HENSON (R2)
JAY HENSON (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
ACCIDENT MITIGATION - HPCI SYSTEM INOPERABLE
"Unit HPCI system declared inoperable. During performance of the quarterly surveillance, HPCI pump operability, the HPCI system was secured when the "HPCI Turbine Oil Pressure Low" alarm was received and confirmed. Turbine governor end bearing oil pressure was 2 PSIG. The alarm setpoint is 6 PSIG and procedure limit is 10-12 PSIG. Pressure adjust valve was throttled open to raise pressure to 11 PSIG. The HPCI system was not immediately declared inoperable since an evaluation was being performed to determine if 2 PSIG turbine bearing oil pressure was adequate. Evaluation by the vendor will not be complete until 10/18/06. At 1655 hours, HPCI declared operable after a successful run with adequate oil pressure. HPCI is a single train system."
The licensee notified the NRC Resident Inspector.
* * * UPDATE PROVIDED BY NRC ON 11/06/06 AT 1106 EST DUE TO EVENT ENTRY ERROR * * *
Original report was entered in error on 10/17/06 with Unit 2 versus Unit 1. Changed EN #42914 to accurately reflect the affected unit (Unit 1).
* * * RETRACTION PROVIDED BY E. BURKETT TO KOZAL ON 11/16/06 AT 1039 * * *
"EN #42914 was submitted by Southern Nuclear Operating Company based upon a conservative decision to declare the HPCI system inoperable pending further evaluation to support its operability. Southern Nuclear Operating Company retracts EN #42914 based on the following discussion.
"During a subsequent review of the parameters by the HPCI Turbine Vendor, Dresser-Rand, and site engineering it was concluded that the HPCI system would have been capable of performing its intended safety function with the lower turbine governor end bearing oil pressure.
"During the operation of the system, the visual local indication was approximately 2.5 PSIG oil pressure at the governor end bearing. A review of the data showed that with a governor end oil pressure of the procedural minimal of 10 PSIG, the predicted oil flow rate would be 1.08 gpm with a minimum film thickness of 0.48 mils and a maximum bearing temperature of 228 deg F. With a degraded oil pressure of 2.5 PSIG, the predicted oil flow rate would be 0.54 gpm with a minimum film thickness of 0.46 mils and a maximum bearing temperature of 233 deg F.
"Based on the calculated data, the turbine governor end bearing would have performed satisfactorily for at least 8 hours at an oil pressure of 2.5 PSIG. Using the design basis success criteria, HPCI operation is successful if the system can inject water through the core Feedwater line for a total of 4 hours over a 24 hour period. The 4 hour mission time for HPCI is based on the design basis - if HPCI fails, it is backed up by the Automatic Depressurization System (ADS) in combination with Core Spray and Low Pressure Coolant Injection. The HPCI system is not credited for long term injection or late injection.
"Although the oil flow rate was reduced by 50% and the minimum film thickness reduced by 4%, the bearing temperatures were predicted to only increase a maximum of 5 deg F. Supporting this conclusion is the fact that the bearing was not damaged during the operation with low oil pressure when the turbine was run for 9 minutes at 2.5 PSIG governor end bearing oil pressure."
The licensee will notify the NRC Resident Inspector.
Notified R2DO (Haag)
"Unit HPCI system declared inoperable. During performance of the quarterly surveillance, HPCI pump operability, the HPCI system was secured when the "HPCI Turbine Oil Pressure Low" alarm was received and confirmed. Turbine governor end bearing oil pressure was 2 PSIG. The alarm setpoint is 6 PSIG and procedure limit is 10-12 PSIG. Pressure adjust valve was throttled open to raise pressure to 11 PSIG. The HPCI system was not immediately declared inoperable since an evaluation was being performed to determine if 2 PSIG turbine bearing oil pressure was adequate. Evaluation by the vendor will not be complete until 10/18/06. At 1655 hours, HPCI declared operable after a successful run with adequate oil pressure. HPCI is a single train system."
The licensee notified the NRC Resident Inspector.
* * * UPDATE PROVIDED BY NRC ON 11/06/06 AT 1106 EST DUE TO EVENT ENTRY ERROR * * *
Original report was entered in error on 10/17/06 with Unit 2 versus Unit 1. Changed EN #42914 to accurately reflect the affected unit (Unit 1).
* * * RETRACTION PROVIDED BY E. BURKETT TO KOZAL ON 11/16/06 AT 1039 * * *
"EN #42914 was submitted by Southern Nuclear Operating Company based upon a conservative decision to declare the HPCI system inoperable pending further evaluation to support its operability. Southern Nuclear Operating Company retracts EN #42914 based on the following discussion.
"During a subsequent review of the parameters by the HPCI Turbine Vendor, Dresser-Rand, and site engineering it was concluded that the HPCI system would have been capable of performing its intended safety function with the lower turbine governor end bearing oil pressure.
"During the operation of the system, the visual local indication was approximately 2.5 PSIG oil pressure at the governor end bearing. A review of the data showed that with a governor end oil pressure of the procedural minimal of 10 PSIG, the predicted oil flow rate would be 1.08 gpm with a minimum film thickness of 0.48 mils and a maximum bearing temperature of 228 deg F. With a degraded oil pressure of 2.5 PSIG, the predicted oil flow rate would be 0.54 gpm with a minimum film thickness of 0.46 mils and a maximum bearing temperature of 233 deg F.
"Based on the calculated data, the turbine governor end bearing would have performed satisfactorily for at least 8 hours at an oil pressure of 2.5 PSIG. Using the design basis success criteria, HPCI operation is successful if the system can inject water through the core Feedwater line for a total of 4 hours over a 24 hour period. The 4 hour mission time for HPCI is based on the design basis - if HPCI fails, it is backed up by the Automatic Depressurization System (ADS) in combination with Core Spray and Low Pressure Coolant Injection. The HPCI system is not credited for long term injection or late injection.
"Although the oil flow rate was reduced by 50% and the minimum film thickness reduced by 4%, the bearing temperatures were predicted to only increase a maximum of 5 deg F. Supporting this conclusion is the fact that the bearing was not damaged during the operation with low oil pressure when the turbine was run for 9 minutes at 2.5 PSIG governor end bearing oil pressure."
The licensee will notify the NRC Resident Inspector.
Notified R2DO (Haag)
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 42915
Facility: SUMMER
Region: 2 State: SC
Unit: [1] [] []
RX Type: [1] W-3-LP
NRC Notified By: ROBERT SWEET
HQ OPS Officer: PETE SNYDER
Region: 2 State: SC
Unit: [1] [] []
RX Type: [1] W-3-LP
NRC Notified By: ROBERT SWEET
HQ OPS Officer: PETE SNYDER
Notification Date: 10/17/2006
Notification Time: 17:50 [ET]
Event Date: 10/17/2006
Event Time: 14:15 [EDT]
Last Update Date: 11/04/2006
Notification Time: 17:50 [ET]
Event Date: 10/17/2006
Event Time: 14:15 [EDT]
Last Update Date: 11/04/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
Person (Organization):
JAY HENSON (R2)
JAY HENSON (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown |
PRESSURE BOUNDARY LEAKAGE IDENTIFIED DURING AN OUTAGE
"On October 14th V.C. Summer Nuclear Station shut down for a routine refueling outage. At 1415 on October 17, 2006 during an inspection of the reactor vessel upper head, personnel identified a small leak in a weld in the Reactor Vessel Level Instrumentation System (RVLIS). This weld is located in a 3/4 inch stainless steel instrumentation line. The line is attached to, and is approximately three feet from the reactor vessel head. This leak resulted in a small amount of boron accumulation in an isolated area. There is no indication that any safety systems were affected."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION PROVIDED BY ARNIE CRIBB TO JEFF ROTTON AT 1521 ON 11/04/06 * * *
"EN #42915 was submitted by South Carolina Electric & Gas Company based on the initial indication of a pressure boundary leak in the Reactor Vessel Level Instrumentation System (RVLIS). South Carolina Electric & Gas Company retracts EN #42915 based on the following discussion.
"During subsequent cleaning and inspection of the suspect weld, which was preserved in its as-found condition for root cause analysis, no flaw in the analyzed RVLIS piping or pipe coupling was found. Further investigation was conducted of the area under the Control Rod Drive Mechanism (CRDM) shroud near the RVLIS line. The source was determined to be leakage through the lower canopy seal weld of CRDM No. 25, which resulted in accumulation of boric acid on the suspect RVLIS line.
"Since the RVLIS line was not leaking and the CRDM canopy connection is a mechanical (screwed) fitting with a seal weld, there has been no pressure boundary leakage as defined by Technical Specifications. The canopy seal weld is neither a structural weld nor a pressure retaining weld comprising the reactor coolant system pressure boundary. A permanent plant design was developed and the canopy seal was repaired using a standard canopy seal clamping process.
"A 100 % bare metal visual inspection of the reactor vessel head was completed during this refueling outage. The results confirmed that there has been no degradation."
The licensee will notify the NRC Resident Inspector.
Notified the R2DO (Evans)
"On October 14th V.C. Summer Nuclear Station shut down for a routine refueling outage. At 1415 on October 17, 2006 during an inspection of the reactor vessel upper head, personnel identified a small leak in a weld in the Reactor Vessel Level Instrumentation System (RVLIS). This weld is located in a 3/4 inch stainless steel instrumentation line. The line is attached to, and is approximately three feet from the reactor vessel head. This leak resulted in a small amount of boron accumulation in an isolated area. There is no indication that any safety systems were affected."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION PROVIDED BY ARNIE CRIBB TO JEFF ROTTON AT 1521 ON 11/04/06 * * *
"EN #42915 was submitted by South Carolina Electric & Gas Company based on the initial indication of a pressure boundary leak in the Reactor Vessel Level Instrumentation System (RVLIS). South Carolina Electric & Gas Company retracts EN #42915 based on the following discussion.
"During subsequent cleaning and inspection of the suspect weld, which was preserved in its as-found condition for root cause analysis, no flaw in the analyzed RVLIS piping or pipe coupling was found. Further investigation was conducted of the area under the Control Rod Drive Mechanism (CRDM) shroud near the RVLIS line. The source was determined to be leakage through the lower canopy seal weld of CRDM No. 25, which resulted in accumulation of boric acid on the suspect RVLIS line.
"Since the RVLIS line was not leaking and the CRDM canopy connection is a mechanical (screwed) fitting with a seal weld, there has been no pressure boundary leakage as defined by Technical Specifications. The canopy seal weld is neither a structural weld nor a pressure retaining weld comprising the reactor coolant system pressure boundary. A permanent plant design was developed and the canopy seal was repaired using a standard canopy seal clamping process.
"A 100 % bare metal visual inspection of the reactor vessel head was completed during this refueling outage. The results confirmed that there has been no degradation."
The licensee will notify the NRC Resident Inspector.
Notified the R2DO (Evans)
Power Reactor
Event Number: 42916
Facility: VOGTLE
Region: 2 State: GA
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: ROBERT SWEET
HQ OPS Officer: PETE SNYDER
Region: 2 State: GA
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: ROBERT SWEET
HQ OPS Officer: PETE SNYDER
Notification Date: 10/17/2006
Notification Time: 18:07 [ET]
Event Date: 10/17/2006
Event Time: 14:30 [CDT]
Last Update Date: 10/17/2006
Notification Time: 18:07 [ET]
Event Date: 10/17/2006
Event Time: 14:30 [CDT]
Last Update Date: 10/17/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
JAY HENSON (R2)
JAY HENSON (R2)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
DEGRADED TECHNICAL SUPPORT CENTER VENTILATION SUPPLY FAN
"At 1430 CDST on 10/17/2006 the Control Room was notified by the Site HVAC engineer that a condition existing in the Technical Support Center (TSC) HVAC called into question the habitability of the TSC under accident conditions. Specifically, the TSC HVAC Supply fan was noted on the previous shift to be putting out less air than normal. Engineering was asked to look into it and discovered that the fan belts were slipping and that the fan was currently running at approximately half normal speed. While this was sufficient to maintain the required positive pressure in the TSC at the present time, they were not confident of it's ability to maintain it long-term under accident conditions. Receipt of this information and consultation with Southern Nuclear Company Emergency Preparedness employees has led the Control Room staff to the decision to make this notification under 10CFR50.72(b)(3)(xiii), for a major loss of emergency assessment capability due to the uncertainty of the long-term habitability of the TSC. Preparations are underway for the repair of the affected fan, which is expected to be completed tonight."
The licensee notified the NRC Resident Inspector.
"At 1430 CDST on 10/17/2006 the Control Room was notified by the Site HVAC engineer that a condition existing in the Technical Support Center (TSC) HVAC called into question the habitability of the TSC under accident conditions. Specifically, the TSC HVAC Supply fan was noted on the previous shift to be putting out less air than normal. Engineering was asked to look into it and discovered that the fan belts were slipping and that the fan was currently running at approximately half normal speed. While this was sufficient to maintain the required positive pressure in the TSC at the present time, they were not confident of it's ability to maintain it long-term under accident conditions. Receipt of this information and consultation with Southern Nuclear Company Emergency Preparedness employees has led the Control Room staff to the decision to make this notification under 10CFR50.72(b)(3)(xiii), for a major loss of emergency assessment capability due to the uncertainty of the long-term habitability of the TSC. Preparations are underway for the repair of the affected fan, which is expected to be completed tonight."
The licensee notified the NRC Resident Inspector.
Power Reactor
Event Number: 42917
Facility: BEAVER VALLEY
Region: 1 State: PA
Unit: [] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: DAVE HELD
HQ OPS Officer: JASON KOZAL
Region: 1 State: PA
Unit: [] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: DAVE HELD
HQ OPS Officer: JASON KOZAL
Notification Date: 10/18/2006
Notification Time: 00:19 [ET]
Event Date: 10/17/2006
Event Time: 21:00 [EDT]
Last Update Date: 10/18/2006
Notification Time: 00:19 [ET]
Event Date: 10/17/2006
Event Time: 21:00 [EDT]
Last Update Date: 10/18/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
RICHARD BARKLEY (R1)
RICHARD BARKLEY (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | N | 0 | Refueling | 0 | Refueling |
REACTOR VESSEL HEAD NOZZLE WELD AREA FLAW
"During the performance of NDE examinations of the Beaver Valley Unit 2 (BV-2) reactor pressure vessel (RPV) closure head Alloy 600 Control Rod Drive Mechanism (CRDM) penetrations required by NRC Order (EA-03-009), indications were identified in penetrations #16, #56, and #61. The ultrasonic examination (UT) signal for the three (3) penetrations were in the J-groove weld area and extend into the CRDM tube base material.
"Confirmatory eddy current testing (ECT) was performed on penetrations #16, #56 and #61. Liquid Penetrant (PT) examinations were performed on penetrations #56 and #61 and displayed indications that required further investigation. Following minor excavation of the weld surface in the area of the indications on both penetrations (#56 and #61), additional PT examinations were performed. The result of these examinations on 10/17/06 at 2100 confirmed the presence of flaws in the weld that do not meet acceptance criteria. Penetration #16 is in the state of discovery at this time through PT examination techniques.
"Based upon the preliminary analysis, it is expected that these indications would not be acceptable under ASME standards. Although this condition does not exceed the specific 10 CFR 50.72(b)(3)(ii)(A) criteria since there was no leakage as addressed in ASME Table IWB-3410-1 acceptance criteria and the component structures defined in the ASME IWB-3600 acceptance guidance do not address the J-groove weld configuration, reporting under voluntary considerations is appropriate.
"BV-2 is planning to repair penetrations #56 and #61. Penetration #16 will also be repaired if warranted through the discovery process. All repairs will be completed prior to returning the RPV head to service."
The licensee notified the NRC Resident Inspector.
"During the performance of NDE examinations of the Beaver Valley Unit 2 (BV-2) reactor pressure vessel (RPV) closure head Alloy 600 Control Rod Drive Mechanism (CRDM) penetrations required by NRC Order (EA-03-009), indications were identified in penetrations #16, #56, and #61. The ultrasonic examination (UT) signal for the three (3) penetrations were in the J-groove weld area and extend into the CRDM tube base material.
"Confirmatory eddy current testing (ECT) was performed on penetrations #16, #56 and #61. Liquid Penetrant (PT) examinations were performed on penetrations #56 and #61 and displayed indications that required further investigation. Following minor excavation of the weld surface in the area of the indications on both penetrations (#56 and #61), additional PT examinations were performed. The result of these examinations on 10/17/06 at 2100 confirmed the presence of flaws in the weld that do not meet acceptance criteria. Penetration #16 is in the state of discovery at this time through PT examination techniques.
"Based upon the preliminary analysis, it is expected that these indications would not be acceptable under ASME standards. Although this condition does not exceed the specific 10 CFR 50.72(b)(3)(ii)(A) criteria since there was no leakage as addressed in ASME Table IWB-3410-1 acceptance criteria and the component structures defined in the ASME IWB-3600 acceptance guidance do not address the J-groove weld configuration, reporting under voluntary considerations is appropriate.
"BV-2 is planning to repair penetrations #56 and #61. Penetration #16 will also be repaired if warranted through the discovery process. All repairs will be completed prior to returning the RPV head to service."
The licensee notified the NRC Resident Inspector.