United States Nuclear Regulatory Commission - Protecting People and the Environment
Home > NRC Library > Document Collections > Reports Associated with Events > Event Notification Reports > 2018 > October 19

Event Notification Report for October 19, 2018

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
10/18/2018 - 10/19/2018

** EVENT NUMBERS **


53643 53647 53658 53659 53673 53674 53675 53676

To top of page
Power Reactor Event Number: 53643
Facility: VOGTLE
Region: 2     State: GA
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: KEVIN LOWE
HQ OPS Officer: VINCE KLCO
Notification Date: 10/04/2018
Notification Time: 07:57 [ET]
Event Date: 10/04/2018
Event Time: 00:00 [EDT]
Last Update Date: 10/19/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
OMAR LOPEZ (R2DO)
Unit SCRAM Code RX Crit Initial PWR Initial RX Mode Current PWR Current RX Mode
1 A/R Y 0 Startup 0 Hot Standby

Event Text

MANUAL REACTOR TRIP DURING LOW POWER PHYSICS TESTING

"At 0544 EDT on October 4, 2018, with Unit 1 in Mode 2 with reactor power in the intermediate range performing low power physics testing, the reactor was manually tripped due to a rod control urgent failure alarm. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam system.

"Unit 2 was not affected.

"There was no impact to the health and safety of the public or plant personnel.

"The NRC Resident Inspectors have been notified.

"All control rods inserted as expected. The cause of the rod control urgent failure is being investigated."


* * * UPDATE FROM KEVIN LOWE TO DONALD NORWOOD AT 1408 EDT ON 10/19/2018 * * *

"This Event Notification is being updated to clarify that the reactor was not critical when this event occurred. Therefore, the reporting requirement is changed from 10 CFR 50.72(b)(2)(iv)(B) to 10 CFR 50.72 (b)(3)(iv)(A).

"During Dynamic Rod Worth Measurement testing, Control Bank Charlie was inserted approximately 153 steps when the urgent failure occurred (CBC positioned at 75 steps out). Following the scram, additional analysis concluded that the reactor was subcritical when the Reactor Protection System was actuated."

The licensee notified the NRC Resident Inspector.

Notified the R2DO (McCoy).

To top of page
!!!!! THIS EVENT HAS BEEN RETRACTED.THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 53647
Facility: WOLF CREEK
Region: 4     State: KS
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: MARCY BLOW
HQ OPS Officer: ANDREW WAUGH
Notification Date: 10/05/2018
Notification Time: 09:59 [ET]
Event Date: 10/05/2018
Event Time: 00:00 [CDT]
Last Update Date: 10/18/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
THOMAS FARNHOLTZ (R4DO)
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

UNPLANNED LOSS OF THE ASSESSMENT CAPABILITY DUE TO TECHNICAL SUPPORT DIESEL

"At 05:52 CDT on 10/5/2018, the [Wolf Creek Nuclear Operating Corporation] (WCNOC) Technical Support Center (TSC) Diesel fuel oil transfer pump would not run. There was ongoing modification to the facility Halon system at the time. The modification process had included a jumper to the fuel oil transfer pump to allow it to continue to be available. This issue was discovered during testing as the modification was progressing. The pump was verified to function on 10/4/2018 by normal operations rounds.

"If an emergency is declared requiring the TSC activation during the time the TSC diesel is non-functional, the TSC will be staffed and activated using existing emergency planning procedure. If offsite power is lost, the TSC will relocate to the Alternate TSC using existing emergency planning procedures.

"There was no impact to the health and safety of the public or plant personnel.

"The NRC Resident Inspector has been notified."

* * * RETRACTION ON 10/18/2018 AT 1306 EDT FROM MARCY BLOW TO ANDREW WAUGH * * *

"Event Notification (EN) 53647, made on October 8, 2018, is being retracted because during the time that the TSC Diesel fuel oil transfer pump was not available, normal power was continuously available. The Alternative TSC was also available.

"Consequently, the condition did not meet the criteria for a 8-hour notification per 10 CFR 50.72(b)(3)(xiii) for any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).

"The NRC Resident Inspector has been notified of the Event Notification retraction."

Notified R4DO (Pick).

To top of page
Non-Agreement State Event Number: 53658
Rep Org: TEAM INDUSTRIAL SERVICES
Licensee: TEAM INDUSTRIAL SERVICES
Region: 4
City: ALVIN   State: TX
County:
License #: 42-32219-01
Agreement: Y
Docket:
NRC Notified By: DAVID TEBO
HQ OPS Officer: BETHANY CECERE
Notification Date: 10/10/2018
Notification Time: 17:40 [ET]
Event Date: 10/10/2018
Event Time: 00:00 [CDT]
Last Update Date: 10/10/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
30.50(b)(2) - SAFETY EQUIPMENT FAILURE
Person (Organization):
AARON McCRAW (R3DO)
HEATHER GEPFORD (R4DO)
NMSS_EVENTS_NOTIFICATION (EMAIL)

Event Text

RADIOGRAPHY SOURCE FAILED TO RETRACT

"During retraction of the source following an exposure inside a storage tank [at the BP Refinery, Whiting, IN], the radiographer noted the source did not retract with the drive cable as survey readings indicated the source was still exposed. An immediate call was made to the local RSO [Radiation Safety Officer], who is source retrieval trained, to report the issue. Barricades were continuously monitored while waiting for his arrival. Upon arrival and evaluation of the situation, the local RSO determined a source disconnect had occurred due to a possible broken drive cable. Source retrieval procedures were then enacted with the source safely returned to the shielded position in the exposure device at approximately 1400 CDT. The exact cause of the event is not clear at this time and the drive cable and connector are being sent to the manufacturer for evaluation.

"Equipment involved: QSA Ir192 model A424-9 source assembly with approximate activity 84 Curies; QSA model 880 Delta exposure device and associated equipment (drive assembly and guide tube).

"Approximately 4-5 individuals were involved in the retrieval process with the RSO (Authorized Source Retriever) indicating he received the highest exposure, as recorded on pocket dosimeters, of 25 mrem."

To top of page
Agreement State Event Number: 53659
Rep Org: TEXAS DEPT OF STATE HEALTH SERVICES
Licensee: BRASKEM AMERICA, INC.
Region: 4
City: FREEPORT   State: TX
County:
License #: LL06443
Agreement: Y
Docket:
NRC Notified By: ROBERT FREE
HQ OPS Officer: KAREN COTTON
Notification Date: 10/11/2018
Notification Time: 08:24 [ET]
Event Date: 10/10/2018
Event Time: 00:00 [CDT]
Last Update Date: 10/11/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
HEATHER GEPFORD (R4DO)
NMSS_EVENTS_NOTIFICATION (EMAIL)

Event Text

AGREEMENT STATE REPORT- FIXED GAUGE SHUTTER BROKE

The following information was received from the State of Texas by email:

"On October 10, 2018, the Agency [Texas Department of State Health Services] was notified by the licensee that while performing a leak test of a Vega SH-F1 nuclear gauge the operating handle broke preventing manipulation of the shutter. The gauge contains a 100 milliCurie (original activity) cesium - 137 source. The gauge is located 160 ft. above ground level and does not present an exposure threat to workers. The manufacturer and a service contractor have been contacted to perform the repair of the gauge. Additional information will be provided as it is received in accordance with SA-300."

Texas Incident: 9620

To top of page
Part 21 Event Number: 53673
Rep Org: ITT ENIDINE
Licensee: ITT ENIDINE
Region: 1
City: WESTMINSTER   State: SC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JEFFREY B GOTTHELF
HQ OPS Officer: OSSY FONT
Notification Date: 10/19/2018
Notification Time: 08:40 [ET]
Event Date: 08/23/2018
Event Time: 00:00 [EDT]
Last Update Date: 10/19/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
Person (Organization):
MATT YOUNG (R1DO)
GERALD MCCOY (R2DO)
DAVID HILLS (R3DO)
GREG PICK (R4DO)
PART 21 MATERIALS (EMAIL)

Event Text

PART 21 REPORT - ENVIRONMENTAL QUALIFICATION TEST DISCREPANCY

The following report synopsis was received via fax from ITT Enidine:

On August 23, 2018, ITT Enidine discovered a discrepancy between the published specifications in sales literature, internal test acceptance procedures and the environmental qualification (EQ) test report for the ITT Conoflow model GT25 series. The current to pressure transducer EQ report 3419 lists the temperature range as -40 to 150 degrees Fahrenheit, and the linearity as 0.75 percent of span. The current temperature range specification is 0 to 150 degrees Fahrenheit, and the linearity is 1.5 percent of span. Units manufactured after October 2010 may not conform to the specifications of the EQ test report for linearity and temperature range.

Customers who procured safety related GT25 series I/P transducers and the specific part numbers since 2010 will be notified of this potential product defect by October 23, 2018.

ITT GT25 Series customer list (2010 to present):
- ALMARAZ: GT25CD1826
- ARIZONA PUBLIC SERVICE - GT25CA1826
- CALVERT CLIFFS NUCLEAR - GT25CD1826
- DUKE ENERGY (CATAWBA NUCLEAR STATION) - GT25CA1826; GT25CD1826
- CONTROL COMPONENTS INC - GT25CD1826
- ENERTECH - GT25CD1826
- ENRICO FERMI POWER - GT25CD1826
- FLOWSERVE - GT25CA1826
- TRACTABEL (VIA GEFRAN BENELUX) - GT25CA1826; GT25CD1826
- KHNP - GT25CA1826
- KOREA HYDRO - GT25CA1826
- KRSKO NUCLEAR PLANT - GT25CA1826
- NEBRASKA PUBLIC POWER - GT25CA1826
- ONTARIO POWER GENERATION - GT25CA1826
- SPX FLOW TECHNOLOGY (SPX MCKEAN) - GT25CA1826
- WEIR VALVES - GT25CA1826
- ERGYTECH - GT25CD1826
- OPPG FORT CALHOUN STATION - GT65FA1826
- BECHTEL POWER CORPORATION (TVA) - GT25CD1826; GT65CD1826

To top of page
Power Reactor Event Number: 53674
Facility: FERMI
Region: 3     State: MI
Unit: [2] [] []
RX Type: [2] GE-4
NRC Notified By: JUSTIN D. WILHELM
HQ OPS Officer: OSSY FONT
Notification Date: 10/19/2018
Notification Time: 10:00 [ET]
Event Date: 10/19/2018
Event Time: 00:00 [EDT]
Last Update Date: 10/19/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
DAVID HILLS (R3DO)
Unit SCRAM Code RX Crit Initial PWR Initial RX Mode Current PWR Current RX Mode
2 N N 0 Cold Shutdown 0 Cold Shutdown

Event Text

LOW PRESSURE CORE INJECTION PUMPS INCAPABLE OF AUTOMATIC STARTUP

"On 10/19/2018, at approximately 0400 EDT, during an investigation into a failed surveillance test for a Loss of Offsite Power (LOP) coincident with a Loss of Coolant Accident (LOCA), it was identified that the Engineered Safety System Bus degraded voltage relay scheme contained a time delay setting that could inhibit all Low Pressure Core Injection (LPCI) pumps from automatically starting and operating during a LOP/LOCA, thus making LPCI incapable of meeting its functional requirement of automatic startup and operation regardless of the availability of offsite power supply (UFSAR Section 6.3.1.4 and Tech. Spec. Surveillance Requirement 3.8.1.17).

"The condition was identified during the first-time performance of a revised surveillance procedure for a LOP coincident with a LOCA signal. Fermi is currently in Mode 4 (Cold Shutdown) and LPCI auto start on a LOP/LOCA signal is not required. However, the initial investigation identified the condition likely existed in the past during modes of operation where LPCI auto start on LOP/LOCA was required. Investigation into the cause and corrective actions is ongoing.

"Since LPCI auto start is not required at the time of discovery (Mode 4), this event is being reported pursuant to 50.72(b)(3)(ii)(b).

"The NRC Resident Inspector has been notified."

To top of page
Power Reactor Event Number: 53675
Facility: PERRY
Region: 3     State: OH
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: EDWARD CONDO
HQ OPS Officer: ANDREW WAUGH
Notification Date: 10/19/2018
Notification Time: 12:04 [ET]
Event Date: 10/19/2018
Event Time: 00:00 [EDT]
Last Update Date: 10/19/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
DAVID HILLS (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

Event Text

POSTULATED HOT-SHORT FIRE EVENT THAT COULD ADVERSELY IMPACT SAFE SHUTDOWN EQUIPMENT

"During extent of condition review of a previously identified fire induced hot-short (Ref. EN#53644) an unfused circuit associated with the 0M23C0002A, Miscellaneous Switchgear Recirculation Fan was discovered. This condition is not bounded by existing design and licensing documents.

"This results in an unanalyzed condition due to the possibility for a postulated fire induced hot-short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in the Appendix R Evaluation, Safe Shutdown Capabilities Report, due to an unfused circuit.

"Without overcurrent protection for this circuit, the potential exists that an initial fire event affecting this circuit could cause a short circuit that would cause excessive current through the circuit beyond the capacity rating of the conductors. This could lead to a secondary fire in another plant area where this circuit is routed, challenging the ability to achieve and maintain safe shutdown.

"The postulated event would affect multiple fire zones in the control complex.

"This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant.

"The licensee has notified the NRC Resident Inspector."

* * * UPDATE ON 10/19/2018 AT 1454 EDT FROM EDWARD CONDO TO ANDREW WAUGH * * *

"Further extent of condition reviews have discovered another unfused circuit. The circuitry is related to 0M24C001A, Battery Room Exhaust Fan. This condition is not bounded by existing design and licensing documents.

"This results in an unanalyzed condition due to the possibility for a postulated fire induced hot-short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in the Appendix R Evaluation, Safe Shutdown Capabilities Report, due to an unfused circuit.

"This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant.

"The licensee has notified the NRC Senior Resident Inspector."

Notified R3DO (Hills).

To top of page
Power Reactor Event Number: 53676
Facility: DUANE ARNOLD
Region: 3     State: IA
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: TERRY BRANDT
HQ OPS Officer: ANDREW WAUGH
Notification Date: 10/19/2018
Notification Time: 21:44 [ET]
Event Date: 10/19/2018
Event Time: 00:00 [CDT]
Last Update Date: 10/19/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(2)(iv)(A) - ECCS INJECTION
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
DAVID HILLS (R3DO)
Unit SCRAM Code RX Crit Initial PWR Initial RX Mode Current PWR Current RX Mode
1 A/R Y 100 Power Operation 0 Hot Shutdown

Event Text

AUTOMATIC REACTOR TRIP DUE TO FEEDWATER REGULATING VALVE FAILING CLOSED

"At 1725 CDT, a Feedwater Regulating valve failed closed, resulting in a reactor level transient, which initiated a reactor trip, Primary Containment Isolation System signals to valves in Groups 2, 3, and 4 and initiation of High Pressure Coolant Injection and Reactor Core Isolation Cooling. All control rods inserted and level has been restored to normal. The cause of the feedwater valve failure is under investigation. All other systems responded as expected.

"This report is being made under 10 CFR 50.72 (b)(2)(iv)(B), (b)(3)(iv)(A) and (b)(2)(iv)(A).

"The Senior Resident Inspector has been informed."

Decay heat is being removed via the main condenser and reactor vessel water level is being maintained by the condensate and feedwater systems.


Page Last Reviewed/Updated Friday, May 03, 2019
Friday, May 03, 2019