| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
06/26/2014 - 06/27/2014 ** EVENT NUMBERS ** |
 | 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 | Notification Date: 03/31/2014
Notification Time: 09:07 [ET]
Event Date: 03/31/2014
Event Time: [EDT]
Last Update Date: 06/26/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
| Person (Organization):
MARVIN SYKES (R2DO)
PART 21 REACTOR GROU (EMAI)
| Event Text 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. |
| Fuel Cycle Facility | Event Number: 50208 | Facility: NUCLEAR FUEL SERVICES INC.
RX Type: URANIUM FUEL FABRICATION
Comments: HEU CONVERSION & SCRAP RECOVERY
NAVAL REACTOR FUEL CYCLE
LEU SCRAP RECOVERY
Region: 2
City: ERWIN State: TN
County: UNICOI
License #: SNM-124
Agreement: Y
Docket: 07000143
NRC Notified By: RANDY SHACKELFORD
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 06/18/2014
Notification Time: 09:58 [ET]
Event Date: 06/17/2014
Event Time: 19:00 [EDT]
Last Update Date: 06/18/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
PART 70 APP A (a)(4) - ALL SAFETY ITEMS UNAVAILABLE
| Person (Organization):
ROBERT HAAG (R2DO)
TIM MCCARTIN (NMSS)
JEFFERY GRANT (IRD)
FUELS GROUP (OUO) (EMAI)
| Event Text ITEMS RELIED ON FOR SAFETY (IROFS) BYPASSED
"At approximately 1900 hours (EDT) on June 17, 2014, an employee was observed improperly operating two (2) spring return valves identified as Items Relied On For Safety (IROFS) and Safety Related Equipment (SRE). The spring return valves were observed to be 'propped' open. These spring return valves were intended to be manually operated to prevent [a chemical solution] from overfilling a column, spilling to the floor, and causing an acute chemical exposure. No actual overflow occurred. Although the operator was observing and monitoring the filling of the column, the operation of the spring return valves was improper. Operations in the area have been placed in a safe condition and an investigation is underway. This condition was determined to be reportable at 0910 (Eastern Time) on June 18, 2014.
"There were no actual or potential safety consequences to the public or the environment. There were no actual safety consequences to the workers. The potential safety consequences to the workers include exposure to [a hazardous chemical] solution.
"The licensee notified the NRC Resident Inspector and NRC Region 2."
NFS Event (PIRCS) No. - P44298 |
| Power Reactor | Event Number: 50230 | Facility: MCGUIRE
Region: 2 State: NC
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: TERRY KING
HQ OPS Officer: VINCE KLCO | Notification Date: 06/26/2014
Notification Time: 15:29 [ET]
Event Date: 06/26/2014
Event Time: 09:18 [EDT]
Last Update Date: 06/26/2014 | Emergency Class: NON EMERGENCY
10 CFR Section:
26.719 - FITNESS FOR DUTY
| Person (Organization):
DANIEL RICH (R2DO)
|
| 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 FITNESS FOR DUTY - NON-LICENSED SUPERVISOR IN VIOLATION OF DUKE ENERGY FITNESS FOR DUTY POLICY
A non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness for duty test. The employee's access to both the McGuire and Oconee plants has been restricted.
The NRC Resident Inspectors at both the McGuire and Oconee sites have been notified. | |