Event Notification Report for March 31, 2014
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/30/2014 - 03/31/2014
Part 21
Event Number: 49974
Rep Org: GE-HITACHI NUCLEAR ENERGY
Licensee: GE-HITACHI NUCLEAR ENERGY
Region: 1
City: WILMINGTON State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: DALE PORTER
HQ OPS Officer: JOHN SHOEMAKER
Licensee: GE-HITACHI NUCLEAR ENERGY
Region: 1
City: WILMINGTON State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: DALE PORTER
HQ OPS Officer: JOHN SHOEMAKER
Notification Date: 03/31/2014
Notification Time: 09:07 [ET]
Event Date: 03/31/2014
Event Time: 00:00 [EDT]
Last Update Date: 08/29/2014
Notification Time: 09:07 [ET]
Event Date: 03/31/2014
Event Time: 00:00 [EDT]
Last Update Date: 08/29/2014
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
Person (Organization):
MARVIN SYKES (R2DO)
PART 21 REACTOR GROU (EMAI)
MARVIN SYKES (R2DO)
PART 21 REACTOR GROU (EMAI)
INTERIM PART 21 REPORT - CONTAINMENT LOADS POTENTIALLY EXCEED LIMITS WITH HIGH SUPPRESSION POOL WATER LEVEL IN THE ABWR DESIGN
The following summary was excerpted from GE Hitachi Interim Part 21 Report received via email:
"A potential analysis error has been identified that is associated with the ABWR (Advanced Boiling Water Reactor) hydrodynamic loads determined by using the Technical Specification Suppression Pool High Water Level (HWL) as an analysis input condition. Vessel coolant inventory is transferred into the containment Suppression Pool during a postulated LOCA blowdown, thereby increasing the Suppression Pool water level. The correction in the analysis may lead to a Suppression Pool water level greater than what is currently assumed in structural analyses which apply the containment hydrodynamic loads generated during a postulated LOCA event."
Facility Identification: South Texas Project Units 3 and 4, Clinton ESP Site, Grand Gulf ESP Site, North Anna ESP Site, and includes the ESP application for the PSEG Site and Victoria County Station ESP application.
If you have any questions, then contact: Dale E. Porter, GE-Hitachi Nuclear Energy Americas LLC, Ph. #(910) 819-4491.
* * * UPDATE AT 1343 EDT ON 06/26/14 FROM JIM HARRISON TO S. SANDIN VIA EMAIL * * *
"June 26, 2014
"MFN 14-013 R1
"This letter provides supplemental information concerning an evaluation being performed by GE Hitachi Nuclear Energy (GEH) regarding the potential increase in hydrodynamic loads that may be experienced by containment structures during a postulated Loss of Coolant Accident (LOCA) associated with Reference 1, and requests additional time to complete the evaluation for the determination of reportability of this condition.
"A potential analysis error has been identified that is associated with the ABWR hydrodynamic loads determined by using the Technical Specification Suppression Pool High Water Level (HWL) as an analysis input condition. Vessel coolant inventory is transferred into the containment Suppression Pool during a postulated LOCA blowdown, thereby increasing the Suppression Pool water level. The correction in the analysis may lead to a Suppression Pool water level greater than what is currently assumed in structural analyses which apply the containment hydrodynamic loads generated during a postulated LOCA event. For example, a postulated Feedwater Line Break (FWLB) may transfer a large quantity of FW liquid into the Suppression Pool with a notable increase in pool water level, even assuming a portion of the discharged fluid spills over into the lower drywell region of the ABWR containment. A higher Suppression Pool water level may result in increased hydrodynamic loads acting on the submerged walls and structures in the containment. The higher Suppression Pool water level can extend the wetted regions of the Suppression Pool walls and the ABWR access tunnel, as well as result in wetted submerged structure segments that were not previously considered wetted. This potential analysis error affects the LOCA containment hydrodynamic loads including condensation oscillation (CO) and chugging, as well as Safety Relief Valve (SRV) actuation loads.
"Assessing the overall impact of increased hydrodynamic loads calculated with higher Suppression Pool water level requires an evaluation of the containment structural components' design bases. GEH is in the process of examining revised containment loads, and determining available margin in the ABWR containment component design specifications to accommodate potentially increased load source forcing functions. ABWR plants may then compare affected plant-specific containment structural design bases to these specifications for relative margin. An extended time period is needed in order to complete the revised containment load determination and evaluate the impact on containment structures.
"GEH is requesting additional time to complete the analysis previously noted in Reference 1. The information required for this GEH 60-Day Interim Report Notification per 21.21(a)(2) is provided in Attachment 1. The commitment for follow-on actions is provided in Attachment 1, item (vii).
"If you have any questions, please call me at (910) 819-4491.
"Sincerely,
"Dale E. Porter
Safety Evaluation Program Manager
GE-Hitachi Nuclear Energy Americas LLC"
Notified R2DO (Rich) and Part 21 Reactor Group via email.
* * * UPDATE FROM LISA SCHICHLEIN TO JOHN SHOEMAKER AT 0745 ON 8/29/14 VIA EMAIL * * *
"August 29, 2014, MFN 14-013 R2, Specification 000N7289-R2
"When considered with the realistic assumptions and the lowered scale factors, the condition reported in Reference 1 [of the final report] is determined as non-reportable, and there is no Substantial Safety Hazard nor will it lead to exceeding a Technical Specification Safety Limit for the affected plants and plant designs. The GEH evaluation within 10 CFR Part 21 is now closed."
If you have any questions, please call;
Dale E. Porter
Safety Evaluation Program Manager
GE-Hitachi Nuclear Energy Americas LLC
Ph. (910) 819-4491.
Notified R2DO (Rose) and Part 21 Reactor Group via email.
The following summary was excerpted from GE Hitachi Interim Part 21 Report received via email:
"A potential analysis error has been identified that is associated with the ABWR (Advanced Boiling Water Reactor) hydrodynamic loads determined by using the Technical Specification Suppression Pool High Water Level (HWL) as an analysis input condition. Vessel coolant inventory is transferred into the containment Suppression Pool during a postulated LOCA blowdown, thereby increasing the Suppression Pool water level. The correction in the analysis may lead to a Suppression Pool water level greater than what is currently assumed in structural analyses which apply the containment hydrodynamic loads generated during a postulated LOCA event."
Facility Identification: South Texas Project Units 3 and 4, Clinton ESP Site, Grand Gulf ESP Site, North Anna ESP Site, and includes the ESP application for the PSEG Site and Victoria County Station ESP application.
If you have any questions, then contact: Dale E. Porter, GE-Hitachi Nuclear Energy Americas LLC, Ph. #(910) 819-4491.
* * * UPDATE AT 1343 EDT ON 06/26/14 FROM JIM HARRISON TO S. SANDIN VIA EMAIL * * *
"June 26, 2014
"MFN 14-013 R1
"This letter provides supplemental information concerning an evaluation being performed by GE Hitachi Nuclear Energy (GEH) regarding the potential increase in hydrodynamic loads that may be experienced by containment structures during a postulated Loss of Coolant Accident (LOCA) associated with Reference 1, and requests additional time to complete the evaluation for the determination of reportability of this condition.
"A potential analysis error has been identified that is associated with the ABWR hydrodynamic loads determined by using the Technical Specification Suppression Pool High Water Level (HWL) as an analysis input condition. Vessel coolant inventory is transferred into the containment Suppression Pool during a postulated LOCA blowdown, thereby increasing the Suppression Pool water level. The correction in the analysis may lead to a Suppression Pool water level greater than what is currently assumed in structural analyses which apply the containment hydrodynamic loads generated during a postulated LOCA event. For example, a postulated Feedwater Line Break (FWLB) may transfer a large quantity of FW liquid into the Suppression Pool with a notable increase in pool water level, even assuming a portion of the discharged fluid spills over into the lower drywell region of the ABWR containment. A higher Suppression Pool water level may result in increased hydrodynamic loads acting on the submerged walls and structures in the containment. The higher Suppression Pool water level can extend the wetted regions of the Suppression Pool walls and the ABWR access tunnel, as well as result in wetted submerged structure segments that were not previously considered wetted. This potential analysis error affects the LOCA containment hydrodynamic loads including condensation oscillation (CO) and chugging, as well as Safety Relief Valve (SRV) actuation loads.
"Assessing the overall impact of increased hydrodynamic loads calculated with higher Suppression Pool water level requires an evaluation of the containment structural components' design bases. GEH is in the process of examining revised containment loads, and determining available margin in the ABWR containment component design specifications to accommodate potentially increased load source forcing functions. ABWR plants may then compare affected plant-specific containment structural design bases to these specifications for relative margin. An extended time period is needed in order to complete the revised containment load determination and evaluate the impact on containment structures.
"GEH is requesting additional time to complete the analysis previously noted in Reference 1. The information required for this GEH 60-Day Interim Report Notification per 21.21(a)(2) is provided in Attachment 1. The commitment for follow-on actions is provided in Attachment 1, item (vii).
"If you have any questions, please call me at (910) 819-4491.
"Sincerely,
"Dale E. Porter
Safety Evaluation Program Manager
GE-Hitachi Nuclear Energy Americas LLC"
Notified R2DO (Rich) and Part 21 Reactor Group via email.
* * * UPDATE FROM LISA SCHICHLEIN TO JOHN SHOEMAKER AT 0745 ON 8/29/14 VIA EMAIL * * *
"August 29, 2014, MFN 14-013 R2, Specification 000N7289-R2
"When considered with the realistic assumptions and the lowered scale factors, the condition reported in Reference 1 [of the final report] is determined as non-reportable, and there is no Substantial Safety Hazard nor will it lead to exceeding a Technical Specification Safety Limit for the affected plants and plant designs. The GEH evaluation within 10 CFR Part 21 is now closed."
If you have any questions, please call;
Dale E. Porter
Safety Evaluation Program Manager
GE-Hitachi Nuclear Energy Americas LLC
Ph. (910) 819-4491.
Notified R2DO (Rose) and Part 21 Reactor Group via email.
Power Reactor
Event Number: 49977
Facility: QUAD CITIES
Region: 3 State: IL
Unit: [] [2] []
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: WILLIAM STRICKLAND
HQ OPS Officer: JEFF ROTTON
Region: 3 State: IL
Unit: [] [2] []
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: WILLIAM STRICKLAND
HQ OPS Officer: JEFF ROTTON
Notification Date: 03/31/2014
Notification Time: 18:57 [ET]
Event Date: 03/31/2014
Event Time: 13:02 [CDT]
Last Update Date: 03/31/2014
Notification Time: 18:57 [ET]
Event Date: 03/31/2014
Event Time: 13:02 [CDT]
Last Update Date: 03/31/2014
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS 50.72(b)(3)(ii)(A) - DEGRADED CONDITION
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS 50.72(b)(3)(ii)(A) - DEGRADED CONDITION
Person (Organization):
DAVE PASSEHL (R3DO)
DAVE PASSEHL (R3DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 87 | Power Operation | 54 | Power Operation |
UNIT 2 REACTOR PRESSURE BOUNDARY LEAKAGE AND UNIT SHUTDOWN
"On March 31, 2014, at 1302 [CDT], with Unit 2 operating in Mode 1 (coastdown), leakage was identified from the body of a one inch isolation valve associated with a Control Rod Drive Hydraulic Control Unit (CRD HCU 18-27 insert valve 101). The leakage is several drops per minute; attempts to isolate the leak have been unsuccessful. The associated piping communicates with the reactor coolant system (through the CRD mechanism).
"Technical Specification 3.4.4 (RCS Operational Leakage), Condition C, was entered at 1302 hours. Condition C requires the unit to be placed in Mode 3 (Hot Shutdown) in 12 hours. At 1616 hours, a Technical Specification required shutdown was initiated on Unit 2.
"The cause and resolution of the leakage are being pursued.
"This condition is being reported under 50.72(b)(3)(ii)(A) given the defect is associated with the primary coolant system pressure boundary. This notification is also being made in accordance with 10 CPR 50.72(b)(2)(i) given the initiation of a plant shutdown required by the plant Technical Specifications."
The licensee has notified the NRC Resident Inspector.
"On March 31, 2014, at 1302 [CDT], with Unit 2 operating in Mode 1 (coastdown), leakage was identified from the body of a one inch isolation valve associated with a Control Rod Drive Hydraulic Control Unit (CRD HCU 18-27 insert valve 101). The leakage is several drops per minute; attempts to isolate the leak have been unsuccessful. The associated piping communicates with the reactor coolant system (through the CRD mechanism).
"Technical Specification 3.4.4 (RCS Operational Leakage), Condition C, was entered at 1302 hours. Condition C requires the unit to be placed in Mode 3 (Hot Shutdown) in 12 hours. At 1616 hours, a Technical Specification required shutdown was initiated on Unit 2.
"The cause and resolution of the leakage are being pursued.
"This condition is being reported under 50.72(b)(3)(ii)(A) given the defect is associated with the primary coolant system pressure boundary. This notification is also being made in accordance with 10 CPR 50.72(b)(2)(i) given the initiation of a plant shutdown required by the plant Technical Specifications."
The licensee has notified the NRC Resident Inspector.
Power Reactor
Event Number: 49978
Facility: WATERFORD
Region: 4 State: LA
Unit: [3] [] []
RX Type: [3] CE
NRC Notified By: PATRICK TUMY
HQ OPS Officer: STEVE SANDIN
Region: 4 State: LA
Unit: [3] [] []
RX Type: [3] CE
NRC Notified By: PATRICK TUMY
HQ OPS Officer: STEVE SANDIN
Notification Date: 03/31/2014
Notification Time: 20:30 [ET]
Event Date: 03/31/2014
Event Time: 15:58 [CDT]
Last Update Date: 03/31/2014
Notification Time: 20:30 [ET]
Event Date: 03/31/2014
Event Time: 15:58 [CDT]
Last Update Date: 03/31/2014
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):
HEATHER GEPFORD (R4DO)
HEATHER GEPFORD (R4DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | N | Y | 100 | Power Operation | 100 | Power Operation |
LOSS OF SIX EMERGENCY SIRENS
"At approximately 15:58 hours (CDT) on March 31, 2014, Waterford 3 was informed that five emergency sirens were inoperable during the performance of a monthly siren inspection. Subsequent review during preparation of this notification has identified an additional one inoperable siren, which brings the total number of inoperable sirens to six. There are a total of seventy-three sirens distributed among two parishes (counties). The loss of these six sirens for more than one hour is considered a major loss of offsite response capability and is reported pursuant to 10 CFR 50.72(b)(3)(xiii).
"The affected parish Emergency Operations Center was notified of the condition and it was confirmed that they will use the preplanned alternative method of Route Alerting for the affected areas until notified that repairs to the sirens have been completed.
"Waterford 3 has initiated preparations to repair the sirens, with actual repairs expected to commence tomorrow. The performance of the monthly siren inspection will continue during daylight hours today and resume tomorrow. There is no effect on the plant. This issue has been entered into the Waterford 3 Corrective Action Program and appropriate corrective actions will be developed.
"The NRC Resident Inspector and local agencies were notified."
"At approximately 15:58 hours (CDT) on March 31, 2014, Waterford 3 was informed that five emergency sirens were inoperable during the performance of a monthly siren inspection. Subsequent review during preparation of this notification has identified an additional one inoperable siren, which brings the total number of inoperable sirens to six. There are a total of seventy-three sirens distributed among two parishes (counties). The loss of these six sirens for more than one hour is considered a major loss of offsite response capability and is reported pursuant to 10 CFR 50.72(b)(3)(xiii).
"The affected parish Emergency Operations Center was notified of the condition and it was confirmed that they will use the preplanned alternative method of Route Alerting for the affected areas until notified that repairs to the sirens have been completed.
"Waterford 3 has initiated preparations to repair the sirens, with actual repairs expected to commence tomorrow. The performance of the monthly siren inspection will continue during daylight hours today and resume tomorrow. There is no effect on the plant. This issue has been entered into the Waterford 3 Corrective Action Program and appropriate corrective actions will be developed.
"The NRC Resident Inspector and local agencies were notified."