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Event Notification Report for June 2, 2015

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
06/01/2015 - 06/02/2015

** EVENT NUMBERS **


50979 51088 51091 51106 51107 51108 51109 51112

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!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 50979
Facility: FITZPATRICK
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: JOHN WALKOWIAK
HQ OPS Officer: JEFF HERRERA
Notification Date: 04/12/2015
Notification Time: 20:51 [ET]
Event Date: 04/12/2015
Event Time: 18:00 [EDT]
Last Update Date: 06/01/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
MARC FERDAS (R1DO)

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

TEMPORARY LOSS OF CONTROL ROOM ENVELOPE BOUNDARY

"The purpose of this report is to provide a telephone notification under 10 CFR 50.72(b)(3)(v)(D) to notify the NRC of a temporary loss of the Control Room Envelope (CRE) boundary. The safety function of the CRE boundary is to ensure the in-leakage of unfiltered air into the CRE will not exceed the in-leakage assumed in the licensing basis analysis of design basis accident (DBA) consequences to CRE occupants. In addition to an intact CRE boundary maintaining CRE occupant dose from a large radioactive release below the calculated dose in the licensing basis consequence analysis for DBAs, it also ensures the occupants are protected from hazardous chemicals and smoke.

"The loss of the CRE boundary was due to a failed latching mechanism for a CRE boundary door used for normal passage of personnel into and out of the CRE. The failure of the door to latch as designed is considered a condition that could have prevented the fulfillment of a safety function at the time of discovery, and is therefore reportable as required by paragraph 50.72(b)(3), 'Eight-hour reports.'

"Procedural controls have restored the safety function of the CRE boundary by mechanically locking the subject door in the closed position through the use of a specifically designed mechanical strong-back until a permanent repair is made.

"The NRC Resident Inspector has been notified."

* * * RETRACTION FROM MARK HAWES TO JOHN SHOEMAKER AT 1642 EDT ON 6/1/15 * * *

"The main control room corridor fire door (76FDR-A-300-10) was found to not be able to latch. The latch was stuck in the latch mechanism because the latch bolt was bent. The latch was replaced on 4/15/2015.

"The Control Room Emergency Ventilation Air Supply System (CREVAS) provides a protected environment from which occupants can control the plant following an uncontrolled release of radioactivity, hazardous chemicals, or smoke. The Control Room Envelope (CRE) is the physical boundary around the CREVAS environment. The Operability of the CRE boundary depends on its ability to minimize in-leakage of unfiltered air such that after a design bases accident a habitable environment can be maintained for 31 days without exceeding 5 rem whole body dose or its equivalent to any part of the body.

"The control room is normally pressurized greater than the 0.125 inches water gauge. This causes air to leak out rather than allowing infiltration of air from surrounding areas into the CRE boundary. The pressurized control room pushes this door (76FDR-A-300-10) outward, toward the open direction; however, even though the latch to the door did not work the door was still able to close. The closed door minimized in-leakage and a positive differential pressure was maintained in the control room during this event. These doors are kept closed against the door seals primarily by the closure mechanism. The latch is a secondary means of ensuring that the doors remain closed as well as a means to control personnel access to the control room.

"The Control Room Envelope (CRE) remained Operable with this deficiency and there was no loss of safety function per 10 CFR 50.72(b)(3)(v)(D). The original notification may be retracted."

The licensee has notified the NRC Resident Inspector.

Notified the R1DO (Powell).

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Agreement State Event Number: 51088
Rep Org: VIRGINIA RAD MATERIALS PROGRAM
Licensee: ELEKTA, INC.
Region: 1
City: HAMPTON State: VA
County:
License #: GA-1153-2
Agreement: Y
Docket:
NRC Notified By: CHARLES COLEMAN
HQ OPS Officer: JEFF HERRERA
Notification Date: 05/22/2015
Notification Time: 17:08 [ET]
Event Date: 05/04/2015
Event Time: [EDT]
Last Update Date: 05/22/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
JAMES DWYER (R1DO)
NMSS_EVENTS_NOTIFIC (EMAI)

Event Text

AGREEMENT STATE REPORT - TRANSPORTABLE HIGH DOSE RATE UNIT DAMAGED WHILE BEING UNLOADED

The following report was provided by the Virginia Department of Health via facsimile:

"On May 4, 2015, a transportable HDR [high dose rate] unit (Elekta microSelectron Model 106.900, serial number 14514) licensed for use by a Virginia licensee was damaged while being unloaded from its transport trailer. The source activity at the time was approximately 8 curies of IR-192. The damage appeared to be limited to the unit's covers. The licensee contacted Elekta, Inc., (which performs work in Virginia under reciprocal recognition of its Georgia license) and a field service engineer was sent to investigate. The service engineer found the head covers and collar cover were broken and other damages, but tests indicated the unit functioned properly. New covers were ordered. During the following week the source was uploaded into an emergency container while the covers were replaced. After the source was returned to the HDR it was found to be stuck in the safe. A kink was found in the cable and a new source was ordered. A source exchange was scheduled on May 19th, but the source could not be manually unloaded as before. Instead, it had to be removed from the back of the HDR. The frayed cable was cut and the source was placed in the emergency container by the service engineer. The source fell to the bottom of the emergency container and the service engineer could not retrieve it. The container was placed in storage at the Virginia licensee's facility after additional shielding was placed around it to reduce the exposure rate to 200 microR/hour. The dose received by the service engineer as a result of the event was estimated by Elekta, using a worst case scenario, as 327 mrem whole body. The service engineer's dosimeter was sent to the dosimetry supplier for an emergency evaluation. Elekta has contacted the source manufacturer (Alpha-Omega Services) [AOS] to assist in the retrieval of the source from the emergency container and to send it to AOS for further investigation. Elekta will provide additional information as it investigates the event.

"Virginia Event Report ID No.: VA-15-06"

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Power Reactor Event Number: 51091
Facility: HATCH
Region: 2 State: GA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: SCOTT BRITT
HQ OPS Officer: JEFF HERRERA
Notification Date: 05/26/2015
Notification Time: 17:45 [ET]
Event Date: 08/07/2014
Event Time: 17:07 [EDT]
Last Update Date: 06/01/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
KATHLEEN O'DONOHUE (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

THIS IS A CONTINUATION OF EN #50351 AND EN #50998

* * * UPDATE ON 05/26/15 AT 1745 EDT FROM SCOTT BRITT TO JEFF HERRERA * * *

"During an expanded scope inspection, deficiencies in the Control Building 147[foot] elevation were observed that caused the affected barrier to be considered nonfunctional:
- Gaps were identified around cables in the foam cable tray penetration seal for penetration 1Z43H006F in the floor of the Cable Spreading Room (separating Fire Areas 0024A and 1104).

"The nonconforming condition observed for the affected fire barrier was identified as affecting both safe shutdown paths for Units 1 and 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity.

"CR 10074859"

The licensee will be notifying the NRC Resident Inspector.

* * * UPDATE FROM STANLEY STONE TO JOHN SHOEMAKER AT 2017 EDT ON 6/1/15 * * *

"During an expanded scope inspection for penetration seals, using more intrusive tools and methods, fire barriers in the Control Building El. 112 [foot] were found not to meet acceptance criteria. The fire protection engineering staff has examined the situations and recommends that these conditions be considered NON-FUNCTIONAL.

- An issue was identified with the wall separating the el. 112 [foot] Control Building Working Floor, Fire Area (FA) 0001 from the Station Battery Room 1B, FA 1005.
- An issue was identified with the wall separating the Station Battery Room 2A, Fire Area (FA) 2004 from the Station Battery Room 2B, FA 2005, on el. 112 [foot].

"The nonconforming conditions observed for the affected fire barriers were identified as affecting both safe shutdown paths for Units 1 and 2. Compensatory measures were already in place in accordance with the plant's Fire Hazard Analysis (FHA) as a result of previous conditions involving degraded barriers in the same fire areas and will remain in place until the fire barriers are repaired. The presence of the compensatory measures in addition to portable fire protection equipment located in adjacent areas ensures the safe shutdown paths are preserved until the degraded conditions are repaired. The expanded scope inspection activity is continuing, and this and any subsequent similar condition(s) that meets the reporting requirements will be included in an ENS Update Report as required and will be documented in a revised LER at the end of the inspection activity. CR 10077573 & 10077574."

The licensee has notified the NRC Resident Inspector.

Notified the R2DO(O'Donohue).

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Power Reactor Event Number: 51106
Facility: COOK
Region: 3 State: MI
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: RICHARD HARRIS
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 05/31/2015
Notification Time: 16:32 [ET]
Event Date: 05/31/2015
Event Time: 16:00 [EDT]
Last Update Date: 06/01/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
Person (Organization):
CHRISTINE LIPA (R3DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 100 Power Operation 99 Power Operation

Event Text

TECHNICAL SPECIFICATION REQUIRED SHUTDOWN DUE TO FAILURE OF AN EMERGENCY DIESEL GENERATOR

"At 1600 [EDT] on May 31, 2015, [DC Cook] operations commenced a shutdown of DC Cook Unit 1 to comply with LCO 3.8.1 Condition G, when the 14 day limit to complete Condition B Required Action could not be met.

"At 0010 [EDT] on May 18, 2015, Unit 1 AB Emergency Diesel Generator was removed from service for planned maintenance. LCO 3.8.1 Condition B was entered which allows 14 days to restore diesel to operable.

"At 1049 [EDT] on May 21, 2015, Unit 1 AB Emergency Diesel Generator tripped during post maintenance testing due to high bearing temperatures. Subsequent actions to repair and restore the diesel to operable status have been unsuccessful.

"This event is reportable under 10 CFR 50.72(b)(2)(i), the initiation of any nuclear plant shutdown required by the plant's Technical Specifications, as a four (4) hour report. The DC Cook Sr. Resident NRC Inspector has been notified."

Unit 1 is expected to be in Mode 5 by 2030 EDT on June 1, 2015. There is no impact on Unit 2.

* * * UPDATE FROM CHRIS PEAK TO JOHN SHOEMAKER ON 6/1/15 AT 1704 EDT * * *

"This update is to correct the information contained in the block titled 'Power/Mode After'. The power and mode after the event requiring notification (TECHNICAL SPECIFICATION REQUIRED SHUTDOWN DUE TO INABILITY TO RESTORE UNIT 1 AB EDG WITHIN THE COMPLETION TIME PRESCRIBED IN LCO 3.8.1 CONDITION B) was 99% power and mode 1.

"The licensee has notified the NRC Resident Inspector."

D.C. Cook Unit 1 is currently in Mode 3 and conducting a normal cooldown to Mode 4.

Notified R3DO (Passehl).

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Power Reactor Event Number: 51107
Facility: PRAIRIE ISLAND
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: JEFFREY HUMAN
HQ OPS Officer: DANIEL MILLS
Notification Date: 06/01/2015
Notification Time: 02:25 [ET]
Event Date: 05/31/2015
Event Time: 22:20 [CDT]
Last Update Date: 06/01/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
CHRISTINE LIPA (R3DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 M/R Y 100 Power Operation 0 Hot Standby

Event Text

MANUAL REACTOR TRIP DUE TO TRIP OF CONDENSATE AND MAIN FEEDWATER PUMP

"On May 31, 2015 at 2220 CDT, the Unit 1 reactor was manually tripped while operating at 100 percent power due to a lockout trip of 11 Condensate Pump followed by a lockout trip of 11 Main Feedwater Pump. Manual Reactor Trip is directed by the annunciator response procedure for the lockout alarm, C47010-0101, 11 Feedwater Pump Locked Out. This also resulted in a turbine trip. The crew entered the reactor trip emergency operating procedures and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. The manual trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the auxiliary feedwater pumps as designed on low narrow range steam generator level and provided makeup flow to the steam generators. The auxiliary feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam generator levels have been returned to normal. Following the reactor trip, 15A Feedwater Heater relief lifted and failed to reseat. 12 Main Feedwater Pump was subsequently secured resulting in 15A Feedwater Heater relief valve reseating successfully. Steam generators are being supplied by 12 Motor Drive Auxiliary Feedwater Pump and decay heat is being removed by the condenser steam dump system. The cause of 11 Condensate Pump trip remains under investigation. There was no effect on Unit 2 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event. The NRC Senior Resident Inspector has been notified."

Unit 2 is unaffected and remains at 100 percent power.

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Power Reactor Event Number: 51108
Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [ ] [ ]
RX Type: [1] GE-2
NRC Notified By: BRYAN EAGAN
HQ OPS Officer: DANIEL MILLS
Notification Date: 06/01/2015
Notification Time: 06:57 [ET]
Event Date: 06/01/2015
Event Time: 07:00 [EDT]
Last Update Date: 06/01/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
FRED BOWER (R1DO)

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

TECHNICAL SUPPORT CENTER OUT OF SERVICE FOR PLANNED MAINTENANCE

"A planned maintenance evolution at the Oyster Creek Generating Station has removed the Technical Support Center (TSC) ventilation system from service. The TSC ventilation system will be rendered non-functional during the course of the work activities. The TSC ventilation is expected to be out of service for approximately sixteen hours and will return to service at approximately 2200 [EDT] June 1, 2015.

"If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature, radiological, or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff to an alternate location in accordance with applicable site procedures.

"This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to potential loss of the TSC. An update will be provided once the TSC ventilation has been restored to normal operation."

The NRC Resident Inspector has been notified.

* * * UPDATE FROM JOSHUA MCGUIRE TO JOHN SHOEMAKER AT 2324 EDT ON 6/1/15 * * *

At 2240 EDT on 6/1/15, the TSC ventilation system was restored.

The licensee will notify the NRC Resident Inspector.

Notified the R1DO (Powell).

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Part 21 Event Number: 51109
Rep Org: ABB MEDIUM VOLTAGE SERVICE
Licensee: ASEA BROWN BOVERI
Region: 1
City: Florence State: SC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JAY LAVRINC
HQ OPS Officer: DANIEL MILLS
Notification Date: 06/01/2015
Notification Time: 10:30 [ET]
Event Date: 06/01/2015
Event Time: [EDT]
Last Update Date: 06/01/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE
Person (Organization):
PART 21/50.55 REACT (EMAI)
RAY POWELL (R1DO)
KATHLEEN O'DONOHUE (R2DO)
DAVE PASSEHL (R3DO)
JACK WHITTEN (R4DO)

Event Text

PART 21 - DEFECTIVE CIRCUIT BREAKER SECONDARY CLOSE LATCH

The following is an excerpt of communication received via email:

"This letter provides notification of a failure to comply with specifications associated with a secondary close latch, part number 716610K01, used in K-Line 225/800 and 1600/2000 amp electrically operated Model 7 circuit breakers.

"Nature of the deviation: During installation of a primary close latch and subsequent bench testing at a nuclear utility, mechanical binding was observed between the primary and secondary close latches. This binding prevented the breaker from operating. Inspection of both latches by ABB showed that the issue lies with the secondary close latch. It was determined that the secondary close latch dimension from the centerline of the hub that the latch rotates about, to the edge of the secondary latch surface with the half pin on the primary latch, was oversized. The failure to comply was identified when replacing latches and verified with a calculated dimension.

"The additional length from the center of the hub in the secondary close latch to the corner of the latch surface caused an interference between the pin cam interface and the half pin on the primary close latch, where it rolls off of the latch surface on the secondary close latch.

"It is recommended that affected licensees with this latch in inventory, ensure that bench testing is performed prior to installation to verify that the primary and secondary close latches work together without any evidence of binding.

"ABB currently cycles K-Line breakers that are refurbished approximately 55 close/open operations before they ship from the Florence facility. New breakers get at least the same number of close/open operations before shipment. This level of operational testing validates that there is no binding between the primary and secondary close latches.

"Questions concerning this notification should be directed to the Quality Manager at the Medium Voltage Service Center in Florence, SC at (843) 413-4727 or Fax (843) 413-4853."

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Power Reactor Event Number: 51112
Facility: RIVER BEND
Region: 4 State: LA
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: JACK MCCOY
HQ OPS Officer: DANIEL MILLS
Notification Date: 06/02/2015
Notification Time: 02:00 [ET]
Event Date: 06/01/2015
Event Time: 21:11 [CDT]
Last Update Date: 06/02/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
JACK WHITTEN (R4DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 A/R Y 90 Power Operation 0 Hot Shutdown

Event Text

REACTOR SCRAM DUE TO LOW REACTOR WATER LEVEL

"At 2111 [CDT] River Bend Nuclear Station sustained an Automatic Reactor Scram due to low Reactor Water Level (Level 3). The plant is currently stable, with level being maintained in a normal band of 10 - 51 inches with Condensate and Feedwater. Reactor Pressure is in the prescribed band of 500-1090 psig. The plant is in Mode 3, Hot Shutdown, and will remain in Mode 3 until investigation of the scram is complete. The transient began with a trip of Reactor Feed Pump 'A', followed by a Reactor Scram and a trip of Reactor Feed Pump 'C'. Reactor water level was recovered with Reactor Feed Pump 'B' to a normal post scram level band. There was a problem noted with the Reactor Feedwater Master Level Controller; this was mitigated by the Operator placing the controller to manual. There was no subsequent Level transient. Reactor Pressure was stabilized in normal pressure band with Turbine bypass valves. During the transient, a Reactor Recirculating Flow Control Valve Runback was not received as expected. Reactor Recirculating Pump 'A' responded as expected to transient [switching to low pump speed], Reactor Recirculating Pump 'B' tripped during transient. A Level 3 isolation signal was received, all expected isolations occurred.

"The cause of the transient is currently under investigation."

The reactor is stable in Mode 3 with decay heat being removed via turbine bypass valves, and a normal electrical line up.

The NRC Resident Inspector has been notified.

Page Last Reviewed/Updated Tuesday, June 02, 2015
Tuesday, June 02, 2015