| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
08/13/2015 - 08/14/2015 ** EVENT NUMBERS ** |
 | Power Reactor | Event Number: 51120 | Facility: HATCH
Region: 2 State: GA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: PAUL UNDERWOOD
HQ OPS Officer: DONG HWA PARK | Notification Date: 06/04/2015
Notification Time: 12:56 [ET]
Event Date: 06/04/2015
Event Time: 10:03 [EDT]
Last Update Date: 08/13/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
STEVE ROSE (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 UNANALYZED CONDITION FOR A POSTULATED FIRE
"In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10CFR50.48(c) (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Unit 1 and Unit 2 Reactor Buildings. This updated analysis has identified circuit configurations in four Fire Areas where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. These are Category 1 barrier impairments.
"In the Unit 1 Safe Shutdown Analysis, RCIC (1E51C001) (Path 1) components are impacted by a fire in Fire Area 1203. The postulated failure described above impacts HPCI (1E41C001) (Path 2) operation. Therefore, in the updated analysis there is no safe shutdown method for high pressure injection that remains free of fire damage for an Appendix R postulated fire in Fire Area 1203. While this represents an unanalyzed condition for Appendix R, the described scenario is only possible given a fire has occurred in Fire Area 1203.
"In the Unit 1 Safe Shutdown Analysis, Path 2 components are impacted by a fire in Fire Area 1205. The postulated failure described above impacts the 1E 4160 Kv (1R22S005) emergency bus power to Path 1 components. Therefore, in the updated analysis there is no safe shutdown method that remains available for an Appendix R postulated fire in Fire Area 1205. While this represents an unanalyzed condition for Appendix R, the described scenario is only possible given a fire has occurred in Fire Area 1205.
"In the Unit 2 Safe Shutdown Analysis, Path 2 components are impacted by a fire in Fire Area 2205. The postulated failure described above impacts the 2E 4160 Kv (2R22S005) emergency bus power to Path 1 components. Therefore, in the updated analysis there is no safe shutdown method that remains available for an Appendix R postulated fire in Fire Area 2205. While this represents an unanalyzed condition for Appendix R, the described scenario is only possible given a fire has occurred in Fire Area 2205.
"In the updated post-fire safe shutdown model, both safe shutdown paths include the same three options for Torus Water Temperature indication (1T48R072, 1T47R611 or 1T47R612). Only one of these three components is required to succeed, however, all are impacted by the postulated fire. Thus, there is no Unit 1 Torus Water Temperature Indication available for a fire in Fire Area 1205. While this represents an unanalyzed condition for Appendix R, the described scenario is only possible given a fire has occurred in Fire Area 1205.
"Based on the updated Plant Hatch Appendix R Safe Shutdown analysis recommendations and the plant's Fire Hazard Analysis (FHA), compensatory measures have been taken and will remain in place until the conditions are resolved.
"The presence of the compensatory measures, in addition to portable fire protection equipment and installed fire protection and detection equipment, ensures the safe shutdown paths are preserved until the conditions are resolved. CR 10079009, 10079019, 10079022, 10079025"
The licensee has notified the NRC Resident Inspector.
* * * UPDATE FROM STANLEY STONE TO DONALD NORWOOD AT 1634 EDT ON 6/17/2015 * * *
"In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10CFR50.48(c) (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Unit 1 and Unit 2 Turbine Building. This updated analysis has identified circuit configurations in two Fire Areas where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. These are Category 1 barrier impairments.
"1) In the Unit 1 Safe Shutdown Analysis, Path 1 RCIC components are impacted by a fire in Fire Area 1105. The postulated failure would impact Path 2 (HPCI) operation. Therefore, in the current analysis for the credited safe shutdown method for high pressure injection may be affected for an Appendix R postulated fire in Fire Area 1105. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1105.
"2) In the updated post-fire safe shutdown model, both safe shutdown paths include the same two options for Torus Water Level Indication: 2T48-R622A and 2T48-R622B. Only one of these two components is required to succeed, however both would be impacted by a postulated fire in Fire Area 2104. Consequently, both credited paths of Unit 2 Torus Water Level Indication could potentially be affected due to a fire in Fire Area 2104. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 2104.
"Based on the updated Plant Hatch Appendix R Safe Shutdown analysis recommendations and the plant's Fire Hazard Analysis (FHA), compensatory measures have been taken and will remain in place until the conditions are resolved.
"The presence of the compensatory measures, in addition to portable fire protection equipment and installed fire protection and detection equipment, ensures the safe shutdown paths are preserved until the conditions are resolved. The analysis associated with the transition of the Plant Hatch Fire Protection Licensing Basis from Appendix R to NFPA 805 is continuing, and this and any subsequent similar conditions that meet reporting requirements will be in included in an ENS Update Report. CR 10084753, CR 10084757."
The licensee notified the NRC Resident Inspector.
Notified R2DO (HAAG).
* * * UPDATE FROM SCOTT BRITT TO VINCE KLCO ON 6/24/15 AT 2114 EDT * * *
"In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10CFR50.48(c) (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Diesel Generator Building. This updated analysis has identified circuit configurations in five Fire Areas where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. These are Category 1 barrier impairments.
"1) An Appendix R postulated fire in Fire Area 1404 is assessed to impact a cable required for RHR Inboard Injection Valve A, 1E11-F015A, to open. This cable was not identified in the current Safe Shutdown Analysis Report (SSAR) for this component. This valve is normally closed and is required to open to support the operation of RHR Loop A in LPCI mode, which is the credited Low Pressure Injection system for Unit 1 in support of Inventory Control to the RPV for a fire in Fire Area 1404. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1404. RHR Loop B is not available in this fire area due to fire impacts.
2) An Appendix R postulated fire in Fire Area 1408 is assessed to impact cables required for RHR Inboard Injection Valve B, 1E11-F015B, to open. These cables were not identified in the current Safe Shutdown Analysis Report (SSAR) for this component. This valve is normally closed and is required to open to support the operation of RHR Loop B in LPCI mode, which is the credited Low Pressure Injection system for Unit 1 in support of Inventory Control to the RPV for a fire in Fire Area 1408. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1408. RHR Loop A is not available in this fire area due to fire impacts.
3) An Appendix R postulated fire in Fire Area 1412 is assessed to impact a cable required for RHR Inboard Injection Valve B, 1E11-F015B, to open. This cable was not identified in the current Safe Shutdown Analysis Report (SSAR) for this component. This valve is normally closed and is required to open to support the operation of RHR Loop B in LPCI mode, which is the credited Low Pressure Injection system for Unit 1 in support of Inventory Control to the RPV for a fire in Fire Area 1412. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 1412. RHR Loop A is not available in this fire area due to fire impacts.
4) An Appendix R postulated fire in Fire Area 2404 is assessed to impact a cable required for RHR Inboard Injection Valve B, 2E11-F015B, to open. This cable was not identified in the current Safe Shutdown Analysis Report (SSAR) for this component. This valve is normally closed and is required to open to support the operation of RHR Loop B in LPCI mode, which is the credited Low Pressure Injection system for Unit 2 in support of Inventory Control to the RPV for a fire in Fire Area 2404. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 2404. RHR Loop A is not available in this fire area due to fire impacts.
5) An Appendix R postulated fire in Fire Area 2408 is assessed to impact cables required for RHR Inboard Injection Valve B, 2E11-F015B, to open. These cables were not identified in the current Safe Shutdown Analysis Report (SSAR) for this component. This valve is normally closed and is required to open to support the operation of RHR Loop B in LPCI mode, which is the credited Low Pressure Injection system for Unit 2 in support of Inventory Control to the RPV for a fire in Fire Area 2408. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 2408. RHR Loop A is not available in this fire area due to fire impacts.
"Based on the updated Plant Hatch Appendix R Safe Shutdown analysis recommendations and the plant's Fire Hazard Analysis (FHA), compensatory measures have been taken and will remain in place until the conditions are resolved.
"The presence of the compensatory measures, in addition to portable fire protection equipment and installed fire protection and detection equipment, ensures the safe shutdown paths are preserved until the conditions are resolved. The analysis associated with the transition of the Plant Hatch Fire Protection Licensing Basis from Appendix R to NFPA 805 is continuing, and this and any subsequent similar conditions that meet reporting requirements will be in included in an ENS Update Report.
"CR 10088142"
The licensee will notify the NRC Resident Inspector.
Notified the R2DO (O'Donohue).
* * * UPDATE AT 1739 EDT ON 08/13/15 FROM PAUL UNDERWOOD TO JEFF HERRERA * * *
"In preparation for transitioning the Plant Hatch Fire Protection Licensing Basis from 10 CFR 50.48(b) (Appendix R) to 10CFR50.48(c) (NFPA 805), an update to the Plant Hatch Appendix R Safe Shutdown Analysis has been performed for the Control Building. This updated analysis has identified circuit configurations in a Fire Area where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. This is a Category 1 barrier impairment.
"1) An Appendix R postulated fire in Fire Area 0014 is assessed to impact a cable that is required for Diesel Building MCC 1C, 1R24-S027, to remain energized. Further analysis has shown that an inter-cable hot short between two conductors could cause the feeder breaker to this MCC to trip. This MCC is required to support the operation of Diesel Generator 1C, which is a credited power source in the Safe Shutdown analysis for both Unit 1 and Unit 2 in the event of a fire in this area. While this represents an unanalyzed condition for Appendix R, the described scenario presumes a fire has occurred in Fire Area 0014.
"Based on the updated Plant Hatch Appendix R Safe Shutdown analysis recommendations and the plant's Fire Hazard Analysis (FHA), compensatory measures have been taken and will remain in place until the conditions are resolved.
"The presence of the compensatory measures, in addition to portable fire protection equipment and installed fire protection and detection equipment, ensures the safe shutdown paths are preserved until the conditions are resolved.
"CR 10108999."
The licensee notified the NRC Resident Inspector.
Notified the R2DO (Nease). |
| Non-Agreement State | Event Number: 51288 | Rep Org: ADVANCE TESTING COMPANY, INC
Licensee: ADVANCE TESTING COMPANY INC
Region: 1
City: CAMPBELL HALL State: NY
County:
License #: 31-31284-01
Agreement: Y
Docket:
NRC Notified By: MARK CLARK
HQ OPS Officer: JOHN SHOEMAKER | Notification Date: 08/05/2015
Notification Time: 10:02 [ET]
Event Date: 08/05/2015
Event Time: [EDT]
Last Update Date: 08/07/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
30.50(b)(2) - SAFETY EQUIPMENT FAILURE
| Person (Organization):
ART BURRITT (R1DO)
NMSS_EVENTS_NOTIFIC (EMAI)
| Event Text DAMAGED PORTABLE MOISTURE DENSITY GAUGE
The licensee reported that a portable moisture density gauge was damaged, by a dump truck, during a road paving project in Glastonbury, CT. The gauge case and keypad were damaged however, the sources are in the stored/safe position. There is no indication of leakage and no radiation exposures have occurred. The licensee has stored the gauge in their secure location pending further evaluation. The gauge is an Instrotek Model 3500 gauge containing 11mCi Cs-137 and 44 mCi Am/Be sources.
* * * UPDATE PROVIDED BY MARK CLARK TO JEFF ROTTON AT 1258 EDT ON 08/07/2015 * * *
Licensee called to correct initial report. The damaged gauge was actually a Troxler 3440, Serial number 18428 and not an Instrotek model 3500. Reported activity of sources remains the same.
Notified R1DO (Burritt) and NMSS Events Notification Group via email. |
| Agreement State | Event Number: 51292 | Rep Org: NC DIV OF RADIATION PROTECTION
Licensee: VIDANT MEDICAL CENTER
Region: 1
City: GREENVILLE State: NC
County:
License #: 074-1457-1
Agreement: Y
Docket:
NRC Notified By: DAVID CROWLEY
HQ OPS Officer: JEFF ROTTON | Notification Date: 08/06/2015
Notification Time: 12:22 [ET]
Event Date: 08/05/2015
Event Time: [EDT]
Last Update Date: 08/06/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
ART BURRITT (R1DO)
NMSS_EVENTS_NOTIFIC (EMAI)
| Event Text AGREEMENT STATE REPORT - MEDICAL EVENT DUE TO ADMINISTRATION OF HIGHER DOSE THAN PRESCRIBED
The following information was provided by the State of North Carolina via email:
"Pitt County Memorial Hospital, Inc. dba Vidant Medical Center (License No. 074-1457-1) had a medical event occur yesterday afternoon [08/05/2015]. In brief, a patient with a low GFR [Glomerular Filtration Rate] was being treated for thyroid carcinoma. The original plan was to give the patient 50 mCi of I-131, which was received, assayed and ready for the Radiologist's approval.
"The Radiologist on site was not the original Radiologist who planned the treatment for this patient. The physician onsite felt that with the low GFR (score to indicate kidney function) a lower dose, 35 mCi, would be prudent and a second order was placed with the radiopharmacy.
"Around 1200 EDT, the dose was received, assayed and ready for administration. The Radiation Safety Representative identified the patient as required and discussed the home-going instructions with the patient prior to the administration. After the patient acknowledged the instructions, the Radiation Safety Representative went to the hot lab, confirmed the written directive, identified an assayed dose with the patient's name on it (of which there were two), failed to confirm the activity on the pig and slip, and administered the dose.
"The error was not identified until the hot lab nuclear medicine technologist noted that the 35 mCi dose was still in the hot lab. The Radiologist and Radiation Safety Office was notified immediately. As of 1420 EDT, the referring physician was notified and patient was to be notified by the end of the day. At this time, it is not probable that there will be any health impact from the discrepancy.
"A NC Health Physicist will be doing a reactive inspection before the end of this week. The radiation safety team is conducting an investigation and will be filing a formal report (15-day report) by August 20, 2015."
NC NMED Report Identification number: NC150023
A Medical Event may indicate potential problems in a medical facility's use of radioactive materials. It does not necessarily result in harm to the patient. |
| Power Reactor | Event Number: 51313 | Facility: FERMI
Region: 3 State: MI
Unit: [2] [ ] [ ]
RX Type: [2] GE-4
NRC Notified By: BRETT JEBBIA
HQ OPS Officer: STEVEN VITTO | Notification Date: 08/12/2015
Notification Time: 13:31 [ET]
Event Date: 08/12/2015
Event Time: 10:07 [EDT]
Last Update Date: 08/13/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
| Person (Organization):
ROBERT ORLIKOWSKI (R3DO)
|
| 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 SECONDARY CONTAINMENT TECHNICAL SPECIFICATIONS NOT MET
"At 1007 [EDT] on August 12, 2015, while restoring Reactor Building (RB) HVAC (RBHVAC) after surveillance testing, an equipment malfunction resulted in improper damper alignment resulting in Secondary Containment Technical Specifications (TS) to not be met.
"The plant TS require Secondary Containment pressure be maintained greater than or equal to -0.125 inches of vacuum water gauge (TS SR 3.6.4.1.1). This specification was not maintained for five seconds and the highest pressure observed was -0.095 inches of vacuum water gauge. This value was observed on only one of two installed recorders, of the Secondary Containment pressure recorders. The highest observed pressure on the other recorder was -0.14 inches of vacuum water gauge. Secondary Containment was restored by the Standby Gas Treatment System (SGTS) already in operation and shutting down the affected train of RBHVAC.
"The technical specification requirement is to maintain secondary containment at -0.125 inches of vacuum water gauge for secondary containment operability. Declaring secondary containment inoperable is reportable under 10 CFR50.72(b)(3)(v)(c) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material."
The licensee notified the NRC Resident Inspector.
* * * UPDATE AT 1159 EDT ON 08/13/15 FROM BRETT JEBBIA TO S. SANDIN * * *
The licensee is updating this report to delete the minus sign for all references to inches of vacuum water gauge.
"FOLLOW UP - CORRECTED INFORMATION: At 1007 [EDT] on August 12, 2015, while restoring Reactor Building (RB) HVAC (RBHVAC) after surveillance testing, an equipment malfunction resulted in improper damper alignment resulting in Secondary Containment Technical Specifications (TS) to not be met.
"The plant TS require Secondary Containment pressure be maintained greater than or equal to .125 inches of vacuum water gauge (TS SR 3.6.4.1.1). This specification was not maintained for five seconds and the highest pressure observed was .095 inches of vacuum water gauge. This value was observed on only one, of two installed recorders, of the Secondary Containment pressure recorders. The highest observed pressure on the other recorder was .14 inches of vacuum water gauge. Secondary Containment was restored by the Standby Gas Treatment System (SGTS) already in operation and shutting down the affected train of RBHV AC.
"The technical specification requirement is to maintain secondary containment greater than or equal to .125 inches of vacuum water gauge for secondary containment operability. Declaring secondary containment inoperable is reportable under 10CFR50.72(b)(3)(v)c as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material."
the licensee informed the NRC Resident Inspector. Notified R1DO (Powell). |
| Power Reactor | Event Number: 51318 | Facility: LASALLE
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] GE-5,[2] GE-5
NRC Notified By: JOHN VAN FLEET
HQ OPS Officer: JEFF HERRERA | Notification Date: 08/13/2015
Notification Time: 16:52 [ET]
Event Date: 08/13/2015
Event Time: 09:38 [CDT]
Last Update Date: 08/13/2015 | Emergency Class: NON EMERGENCY
10 CFR Section:
26.719 - FITNESS FOR DUTY
| Person (Organization):
ROBERT ORLIKOWSKI (R3DO)
|
| 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 | N | 0 | Startup | 0 | Startup | Event Text FITNESS-FOR-DUTY REPORT INVOLVING A LICENSED EMPLOYEE
"A licensed, non-supervisory employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated."
The licensee notified the NRC Resident Inspector as well as the NRC Region 3 Office Safeguards Inspector. | |