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Event Notification Report for March 7, 2016

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
03/04/2016 - 03/07/2016

** EVENT NUMBERS **


51759 51769 51770 51771

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Non-Agreement State Event Number: 51759
Rep Org: SAINT ALPHONSUS HEALTH SYSTEM
Licensee: SAINT ALPHONSUS HEALTH SYSTEM
Region: 4
City: BOISE State: ID
County:
License #: 11-27396-01
Agreement: N
Docket:
NRC Notified By: ERIC COLIANNI
HQ OPS Officer: JOHN SHOEMAKER
Notification Date: 02/26/2016
Notification Time: 18:36 [ET]
Event Date: 02/26/2016
Event Time: 11:20 [MST]
Last Update Date: 02/26/2016
Emergency Class: NON EMERGENCY
10 CFR Section:
35.3045(a)(1) - DOSE <> PRESCRIBED DOSAGE
Person (Organization):
JACK WHITTEN (R4DO)
NMSS_EVENTS_NOTIFIC (EMAI)

Event Text

MEDICAL EVENT - UNDERDOSE TO PATIENT

Notification from the licensee's Radiation Safety Officer, of a medical event that occurred on February 26, 2016 at 1120 MST, in which the Y-90 SIR-Sphere dose delivered to the patient's liver was less than the prescribed dose. The intended dose to the patient was 86 mCi, however, most of the dose remained in the catheter and it is estimated the patient received 13 mCi. The patient and attending physician have been informed and there was no harm to the patient.

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.

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Power Reactor Event Number: 51769
Facility: BRUNSWICK
Region: 2 State: NC
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: KEN KALEY
HQ OPS Officer: DONALD NORWOOD
Notification Date: 03/04/2016
Notification Time: 20:02 [ET]
Event Date: 03/04/2016
Event Time: 12:35 [EST]
Last Update Date: 03/04/2016
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
MARVIN SYKES (R2DO)

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

Event Text

EMERGENCY DIESEL GENERATOR DECLARED INOPERABLE

"On March 3, 2016, during restoration of power to a Unit 1 electrical bus following planned work, an error in the restoration sequence resulted in an invalid auto-start signal to Emergency Diesel Generators (EDGs) 1, 2, 3 and 4. EDG 1 was out-of-service under clearance to support Unit 1 refueling outage modifications and maintenance and, as such, did not start. EDGs 2 and 4 auto-started as designed. However, EDG 3 failed to auto-start. At 1235 EST on March 4, 2016, EDG 3 was declared inoperable when troubleshooting identified a broken fuse block connection in the EDG 3 auto-start circuitry, which would have prevented a Technical Specification (TS) required auto-start of EDG 3. This condition concurrent with EDG 1 out-of-service would have precluded emergency power supply to emergency busses needed to mitigate the consequences of an accident.

"Technical Specification 3.8.1, Required Action D.3, requires declaring the required features supported by the inoperable EDG 3 inoperable when the redundant required features are inoperable. As a result, both required Conventional Service Water (CSW) pumps were declared inoperable at 1635 EST on March 4, 2016. This also required declaring both Control Building Instrument Air Compressors inoperable. As a result, both Control Room Emergency Ventilation (CREV) subsystems and all three Control Building Air Conditioners were declared inoperable at 1635 EST on March 4, 2016.

"The above conditions are reportable under 10 CFR 50.72(b)(3)(v)(D), as an event or condition that could have prevented the fulfillment of a safety function needed to mitigate the consequences of an accident.

"This event did not result in any adverse impact to the health and safety of the public.

"The risk significance of this event is considered to be low. Both EDG 2 and EDG 4 were available and protected, along with the supplemental diesel generator and offsite electrical sources. Except for the periods of time for repair activities and post-repair testing, EDG 3 remained available via manual start. Actions were taken to protect other redundant safety systems and additional defense-in-depth was provided.

"EDG 3 was restored to Operable status March 4, 2016 at 1834 EST and this has restored the safety functions of the above mentioned systems."

The licensee notified the NRC Resident Inspector.

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Power Reactor Event Number: 51770
Facility: OCONEE
Region: 2 State: SC
Unit: [1] [2] [3]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: ARCHIE NEWBERRY
HQ OPS Officer: HOWIE CROUCH
Notification Date: 03/06/2016
Notification Time: 16:09 [ET]
Event Date: 03/06/2016
Event Time: 15:20 [EST]
Last Update Date: 03/06/2016
Emergency Class: ALERT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
CATHY HANEY (R2RA)
BILL DEAN (NRR)
BILL GOTT (IRD)
MARVIN SYKES (R2DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 A/R Y 100 Power Operation 0 Hot Standby
2 N Y 100 Power Operation 100 Power Operation
3 N Y 100 Power Operation 100 Power Operation

Event Text

EMERGENCY DECLARATION DUE TO FIRE/EXPLOSION IN THE MAIN TRANSFORMER

At 1512 EST, a fire/explosion occurred in the Unit 1 Main Transformer which resulted in a reactor trip. At 1520 EST, the licensee declared a Notification of Unusual Event. Offsite assistance was requested. At 1633 EST, smoke and flame were no longer visible. Fire brigade personnel were applying additional foam to prevent a re-flash. No personnel injuries occurred.

Offsite assistance was requested with three local fire departments responding.

All rods inserted on the trip. Steam generator feed is by the normal path. The plant is in its normal shutdown electrical lineup.

The licensee has notified state and local authorities and the NRC Resident Inspector.

Notified DHS SWO, FEMA, and DHS NICC. Notified FEMA NWS and Nuclear SSA via email.

* * * UPDATE FROM DAVID HALE TO HOWIE CROUCH AT 1711 EST ON 03/06/16 * * *

At 1658 EST, the licensee declared an Alert based on EAL Alert A.1. The cause of entry was that the fire damaged an overhead power line that supplies emergency power to all three units at Oconee. Offsite power is still available to all units.

At 1708 EST, the fire is declared out.

The licensee has notified the NRC Resident Inspector.

Notified DHS SWO, FEMA, DHS NICC, USDA, HHS, DOE, and EPA. Notified FEMA NWS, FDA and Nuclear SSA via email.

* * * UPDATE FROM DAVID HALE TO HOWIE CROUCH AT 1805 EST ON 03/06/16 * * *

The licensee made notification of the RPS actuation as a result of the transformer fault.

The licensee has notified the NRC Resident Inspector.

Notified R2DO (Sykes).

* * * UPDATE FROM DAVID HALE TO HOWIE CROUCH AT 2026 EST ON 03/06/16 * * *

At 2016 EST, the licensee terminated all emergency declarations.

The licensee has notified the NRC Resident Inspector.

Notified R2DO (Sykes), IRD (Gott), NRR EO (Morris), DHS SWO, FEMA, DHS NICC, USDA, HHS, DOE, and EPA. Notified FEMA NWS, FDA and Nuclear SSA via email.

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Power Reactor Event Number: 51771
Facility: BRAIDWOOD
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JAMES PETTY
HQ OPS Officer: DANIEL MILLS
Notification Date: 03/07/2016
Notification Time: 02:00 [ET]
Event Date: 03/06/2016
Event Time: 20:00 [CST]
Last Update Date: 03/07/2016
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
NICK VALOS (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 Y 100 Power Operation 100 Power Operation

Event Text

UNANALYZED CONDITION INVOLVING DIESEL DRIVEN AUXILIARY FEEDWATER PUMP AIR INTAKES

"The Auxiliary Feedwater (AF) system at Braidwood automatically supplies feedwater to the Steam Generators (SG) to remove decay heat from the Reactor Coolant System following a loss of normal feedwater supply. The AF System consists of a motor driven pump (A) and a diesel driven pump (B) configured into two trains for each unit. Each pump provides 100% of the required AF capacity to the SGs as assumed in the accident analysis. One pump at full flow is sufficient to remove decay heat and cool the unit to Residual Heat Removal (RHR) entry conditions. The diesel driven AF pump is powered from an independent diesel whose combustion air intake is located in the Seismic Category II (non-seismically qualified) Turbine Building but the diesel and pump are located in the Seismic Category I (seismically qualified) Auxiliary Building.

"During the ongoing NRC Component Design Basis Inspection at Braidwood Station, inspectors asked about the acceptability of the diesel combustion air intake being located in the non-seismic Turbine Building. During the review of available documentation related to the AF diesel engine combustion air intake, it was identified that the documentation did not support operation of the diesel with High Energy Line Break (HELB) environmental conditions in the Turbine Building. Specifically, prior evaluations did not account for air displacement by steam release during the event. After running different models for the Turbine Building HELB, diesel driven AF pump operability was supported for all but the Main Feedwater (FW) HELB. For the FW HELB, the best air density obtained failed to remain above the required levels deemed acceptable for engine operation and remained suppressed for extended periods of time. Additional efforts to qualify the FW piping in the Turbine Building to eliminate this piping from HELB considerations were not successful. This condition applies to both Units 1 & 2 but does not affect the motor driven AF pumps. There were periods within the last 3 years when the motor driven AF pump on each unit was also inoperable for surveillance testing, but remained available; therefore, this does not constitute a loss of safety function.

"This event is reportable per 10 CFR 50.72(b)(3)(ii)(B) for 'any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.'

"The licensee has notified the NRC Resident Inspector."

The licensee entered a 72-hour Action Statement and is preparing to address the issue with a configuration change.

Page Last Reviewed/Updated Monday, March 07, 2016
Monday, March 07, 2016