United States Nuclear Regulatory Commission - Protecting People and the Environment

Event Notification Report for April 16, 2012

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
04/13/2012 - 04/16/2012

** EVENT NUMBERS **


47810 47820 47828 47833 47834 47836 47837 47838

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Power Reactor Event Number: 47810
Facility: OCONEE
Region: 2 State: SC
Unit: [ ] [2] [ ]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: STEPHEN NEWMAN
HQ OPS Officer: PETE SNYDER
Notification Date: 04/06/2012
Notification Time: 03:57 [ET]
Event Date: 04/05/2012
Event Time: 22:38 [EDT]
Last Update Date: 04/13/2012
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
Person (Organization):
GERALD MCCOY (R2DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
2 N Y 100 Power Operation 85 Power Operation

Event Text

STANDBY SHUTDOWN FACILITY (SSF) NOT ANALYZED FOR ALL OPERATING CONDITIONS

"On 3/29/2012 Duke Energy identified that unanalyzed conditions exist for SSF mitigated events since associated thermal and hydraulic analyses do not consider all initial operating conditions, especially lower operating modes and lower decay heat. Specifically there are four (4) conditions where the SSF is not currently analyzed:

"1. SSF operating at less than 525 degrees F and less than normal operating pressure (approximately 2155 psig),
2. SSF operation before four (4) Effective Full Power Days (EFPDs),
3. SSF reactor coolant make up at low Reactor Coolant System (RCS) pressure.
4. A reactor trip from less than 85 percent power and less than 579 degrees F.

"On 4/4/2012, an immediate determination of operability concluded that for the first three (3) conditions the SSF was operable but degraded/nonconforming (OBDN). For the 4th issue, there was reasonable assurance that 1% delta k/k shutdown margin would be maintained if T average. remained above 500 degrees F. Based on a lack of analysis and an increased likelihood of reducing T average. below 525 degrees F during a 72 hour event, the SSF was declared OBDN with a separate operability determination required to validate the Unit 3 power coastdown and end of life T average. reduction analysis. Until additional analysis is performed, the SSF is inoperable on any unit where the power level is reduced below 85 percent.

"A second operability determination for Unit 3 concluded that the SSF will maintain greater than or equal to 525 degrees F with an initial power level of 70 percent and a 570 degree F T average. The SSF will be declared inoperable on Unit 3 if power is reduced to less than 70 percent. Seventy percent was chosen as a conservative value to ensure the unit stayed inside the bounds of existing analyses. Unit 3 is currently at approximately 85 percent power and reducing power at approximately 1 percent per day in preparation for the Unit 2 end of core 26 refueling outage. For Unit 3, the SSF is OBDN based on preliminary calculation results.

"On 4/5/2012, due to a worsening component cooling water system leak on Unit 2, it was necessary to bring the unit down to Mode 3 to implement repairs. Upon down power, when Unit 2 transitioned below 85 percent power, the ability of the SSF to perform its design function, in consideration of the information above, could not be confirmed and the SSF was declared inoperable for Unit 2.

"Currently, there is no conclusive information that would support SSF operability while Unit 2 is below 85 percent power. As such, this event is being conservatively reported under 50.72(b)(3)(ii)(B), 'The nuclear plant being in an unanalyzed condition that significantly degrades plant safety.' Due to their current power levels, this condition does not affect Units 1 and 3.

"Initial Safety Significance: Until confirmed by analysis, the lack of decay heat may result in an initial over cooling of the RCS and potentially an interruption of natural circulation or inadequate shutdown margin. Consequently, the SSF was declared inoperable.

"Corrective Action(s): Additional analyses are being completed to reestablish SSF operability to bound the unanalyzed entry conditions.

"The NRC Resident Inspector has been informed."

* * * UPDATE FROM DEAN PORTER TO JOHN KNOKE AT 1403 EDT ON 04/13/12 * * *

"Update to ENS Notification Number 47810: ENS Notification number 47810 identified an unanalyzed condition for Oconee Unit 2. This update includes Oconee Units 1, 2 and 3 because the condition of the SSF affects all three units during certain plant conditions.

"The Standby Shutdown Facility (SSF) serves as a backup for existing safety systems. Currently the events that rely upon the SSF for mitigation are unanalyzed on Units 1 and 2 when the power level is reduced below 85 percent and on Unit 3 when the power level is reduced below 70 percent, or for any unit operating with less than 4 Effective Full Power Days (EFPDs) at 100% full power since its most recent shutdown. Since the SSF events are unanalyzed for these conditions and until further analysis and evaluation can be completed, Duke Energy is conservatively calling the SSF inoperable when in these conditions.

"For these conditions, this event is being reported for Oconee Units I, 2, and 3 under 50. 72(b )(3)(ii)(B), that is, the nuclear plant being in an unanalyzed condition that significantly degrades plant safety. Based on current analyses, the SSF will be declared inoperable, and action statements entered on Units 1 or 2 if power is reduced below 85% power and on Unit 3 if power is reduced below 70% power, or for any unit operating with less than 4 EFPDs since its most recent shutdown.

"Initial Safety Significance: Until confirmed by analysis or evaluation, the lack of decay heat may result in an initial over cooling of the RCS and potentially fail the various acceptance criteria of the events required to be augmented by the SSF. Consequently, the SSF will be declared inoperable for the conditions stated above. Unit 3 is being shut down for a routine refueling outage. Units 1 and 2 continue to operate at 100% power with no problems.

"Corrective Action(s): Additional evaluations are being completed to establish whether the existing analyses are applicable to the conditions outside of which they were performed and, if not, there is a reasonable assurance that successful mitigation can be accomplished with the existing procedure. If not, further licensing action may be required. These evaluations were initiated on April 4, 2012, and are ongoing."

The licensee has notified the NRC Resident Inspector. Notified the R2DO (Kathleen O'Donohue).

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Power Reactor Event Number: 47820
Facility: SUSQUEHANNA
Region: 1 State: PA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: ALEX MCLELLAN
HQ OPS Officer: JOHN KNOKE
Notification Date: 04/11/2012
Notification Time: 01:18 [ET]
Event Date: 04/10/2012
Event Time: 21:48 [EDT]
Last Update Date: 04/14/2012
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
BLAKE WELLING (R1DO)

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

SPDS AND ERDS REMOVED FROM SERVICE DUE TO PLANNED MAINTENANCE

"At 2148 EDT, on April 10, 2012 the Unit 1 and Unit 2 Safety Parameter Display System (SPDS) and Emergency Response Data System (ERDS) were removed from service to support a planned maintenance outage on the 1A Engineered Safeguards System (ESS) Bus as part of the U1 17th Refueling and Inspection outage. The Bus Outage is expected to have a duration greater than 8 hours, but less than 72 hours. During this time, required control room hardwire indications will be available from the unaffected ESS buses. An update will be provided when SPDS/ERDS becomes available.

"Since the Unit 1 and Unit 2 SPDS/ERDS computer system will be unavailable for greater than 8 hours, this is considered a Loss of Emergency Assessment Capability and reportable under 10 CFR50.72(b)(3)(xiii)."

The licensee has notified the NRC Resident Inspector.

* * * UPDATE FROM RON FRY TO JOHN KNOKE AT 2313 EDT ON 04/14/12 * * *

"As of 1800 hours EDT on 04/13/12, Unit 1 and Unit 2 ERDS and SPDS were restored to normal operation. Operation of these systems has been monitored since that time and operation has been determined to be reliable."

The licensee has not notified the NRC Resident Inspector. Notified the R1DO (Blake Welling)

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Power Reactor Event Number: 47828
Facility: ROBINSON
Region: 2 State: SC
Unit: [2] [ ] [ ]
RX Type: [2] W-3-LP
NRC Notified By: GEORGE CURTIS
HQ OPS Officer: HOWIE CROUCH
Notification Date: 04/12/2012
Notification Time: 09:42 [ET]
Event Date: 04/12/2012
Event Time: 09:30 [EDT]
Last Update Date: 04/13/2012
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
KATHLEEN O'DONOHUE (R2DO)

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

TSC AND EOF OUT OF SERVICE DUE TO MAINTENANCE

"At approximately 0930 hours EDT on Thursday, April 12, 2012, the H. B. Robinson Steam Electric Plant, Unit No. 2, Technical Support Center (TSC)/Emergency Response Facility (EOF) air conditioning and charcoal filtration systems will be removed from service to facilitate the replacement of the charcoal filtration media. The duration of work is expected to be approximately 11 hours. Since the unavailability will last greater than 8 hours, this is considered a Loss of Emergency Assessment Capability, and reportable under 10 CFR 50.72(b)(3)(xiii).

"Due to the inability of the TSC/EOF ventilation system to maintain a habitable atmosphere, as a compensatory measure, Emergency Responders assigned to these facilities have been informed to report to the alternate facilities until such time that the TSC/EOF ventilation system has been returned to service.

"TSC/EOF ventilation system maintenance and post maintenance testing is scheduled to be completed by 2030 hours EDT on Thursday April 12, 2012.

"The NRC Resident Inspector has been informed."

* * * UPDATE FROM WARREN WONKA TO HOWIE CROUCH AT 0435 EDT ON 4/13/12 * * *

At 1814 EDT on 4/12/12, maintenance on the TSC/EOF ventilation system was completed and the TSC/EOF was returned to service.

Notified R2DO (O'Donohue).

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Part 21 Event Number: 47833
Rep Org: MITSUBISHI NUCLEAR ENERGY SYSTEMS
Licensee: MITSUBISHI HEAVY INDUSTRIES, LTD
Region: 1
City: ARLINGTON State: VA
County:
License #:
Agreement: Y
Docket:
NRC Notified By: EI KADOKAMI
HQ OPS Officer: JOHN KNOKE
Notification Date: 04/13/2012
Notification Time: 15:58 [ET]
Event Date: 04/13/2012
Event Time: [EDT]
Last Update Date: 04/16/2012
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
Person (Organization):
BLAKE WELLING (R1DO)
KATHLEEN O'DONOHUE (R2DO)
DAVID HILLS (R3DO)
VINCENT GADDY (R4DO)
PART 21 GROUP (EMAI)

Event Text

PART 21 INTERIM REPORT - STEAM GENERATOR TUBE WEAR

This interim Part 21 is in regard to San Onofre Nuclear Generating Station, Unit 2, Steam Generator replacement.

"During the first refueling outage following steam generator replacement, eddy current testing identified ten total tubes with depths of 90 to 28 percent of the tube wall thickness. Some of the affected tubes were located adjacent to retainer bars. The retainer bars are part of the floating anti-vibration bar (AVB) structure that stabilizes the u-bend region of the tubes.

"Other tubes in the two steam generators had detectable wear associated with support points elsewhere in the AVB structure. Each steam generator has 9727 tubes with an 8 percent (778 tubes) design margin for tube plugging.

"Discovery Date: February 13, 2012

"Evaluation completion schedule date: May 31, 2012"

"Those Mitsubishi Heavy Industries customers potentially affected by this issue have been notified and will receive a copy of this interim report."

Reference Document: UET-20120089
Interim Report No: U21-018-IR (0)

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Power Reactor Event Number: 47834
Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [1] [2] [3]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: JOHN RIDINGER
HQ OPS Officer: JOHN KNOKE
Notification Date: 04/13/2012
Notification Time: 19:31 [ET]
Event Date: 04/13/2012
Event Time: 15:25 [CDT]
Last Update Date: 04/13/2012
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
3 N N 0 Refueling 0 Refueling

Event Text

UNANALYZED CONDITION IMPACTING EMERGENCY DIESEL GENERATOR LOADING

"On March 14, 2012, it was determined that in the event of an Appendix R fire, fire damage to cables in certain fire areas could cause a Residual Heat Removal Service Water System (RHRSW) pump to spuriously start, overload EDG A and B, and render them inoperable during certain Appendix R fires. This was reported as an unanalyzed condition (Ref. EN #47764).

"An extent of condition analysis was completed on April 13, 2012. From this analysis it was determined that EDG A, D, 3EC, and 3ED could exceed the maximum rated loading due to the potential for an automatic or spurious start of RHRSW Pumps B3 and D3 that supply Emergency Equipment Cooling Water (EECW) to essential safety equipment.

"The following are the Fire Areas (FA) affected:

EDG A in FA 21
EDG D for FA 2-3 and 9
EDG 3EC in FA 1-1, 1-3, and 20, and
EDG 3ED in FA 1-1, 1-3, 1-4, and 20.

"This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B). This is also reportable as a 60 day written report IAW 10 CFR 50.73(a)(2)(ii)(B). This event was entered into the licensee's Corrective Action Program as PER 536176.

"The NRC Resident Inspector has been notified of this event."

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Power Reactor Event Number: 47836
Facility: VOGTLE
Region: 2 State: GA
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JEFF TODD
HQ OPS Officer: JOHN KNOKE
Notification Date: 04/14/2012
Notification Time: 16:12 [ET]
Event Date: 04/14/2012
Event Time: 13:46 [EDT]
Last Update Date: 04/14/2012
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):
KATHLEEN O'DONOHUE (R2DO)

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 LOW MAIN FEEDWATER FLOW

"At 1346 EDT, Vogtle Unit 1 reactor was manually tripped from 100% power due to Main Feedwater Pump 'B' discharge flow lowering unexpectedly. All control rods fully inserted. AFW system automatically actuated as expected. System responses allowed for an uncomplicated reactor trip response. Plant is stable in Mode 3 during cause investigation."

The electrical lineup remained normal. No safety valves lifted due to the trip. Decay heat is being removed via the steam dumps to the main condenser.

The licensee has notified the NRC Resident Inspectors.

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Power Reactor Event Number: 47837
Facility: PALO VERDE
Region: 4 State: AZ
Unit: [ ] [ ] [3]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: MICHAEL KOHRT
HQ OPS Officer: JOHN KNOKE
Notification Date: 04/15/2012
Notification Time: 17:13 [ET]
Event Date: 04/15/2012
Event Time: 12:20 [MST]
Last Update Date: 04/15/2012
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
VINCENT GADDY (R4DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
3 M/R Y 0 Startup 0 Hot Standby

Event Text

MANUAL REACTOR TRIP DUE TO CONTROL ROD DEVIATION DURING STARTUP

"The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.

"On April 15, 2012 at approximately 1220 Mountain Standard Time (MST), Palo Verde Unit 3 was manually tripped during low power physics testing.

"While conducting low power physics testing following a refueling outage, Regulating Group 1 rods were being inserted while simultaneously diluting to maintain a constant power level below the Point of Adding Heat. While inserting rods one rod deviated from its subgroup when it stopped moving. The Reactor Operator immediately ceased rod motion and the dilution was stopped. The residual positive reactivity in the core caused a corresponding reactor power increase that approached procedural power limits set forth in the low power physics testing procedure. Based on these indications, operators initiated a manual reactor trip.

"Following the reactor trip, all CEAs inserted fully into the core. All systems operated as expected and this event was diagnosed as an uncomplicated reactor trip. No emergency plan classification was required per the Emergency Plan. Safety related busses remained powered during the event from offsite power and the offsite power grid is stable. No major equipment was inoperable prior to the event that contributed to the event or complicated operator response. Unit 3 is stable and in Mode 3 feeding Steam Generators with Auxiliary Feedwater Pump 'N'. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public.

"The NRC Resident Inspector was informed of the Unit 3 reactor trip."

The electrical lineup remained normal. Decay heat is being removed via the steam bypass to the main condenser.

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Power Reactor Event Number: 47838
Facility: HARRIS
Region: 2 State: NC
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP
NRC Notified By: CURTIS BULLOCK
HQ OPS Officer: VINCE KLCO
Notification Date: 04/16/2012
Notification Time: 07:38 [ET]
Event Date: 04/16/2012
Event Time: 08:00 [EDT]
Last Update Date: 04/16/2012
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
KATHLEEN O'DONOHUE (R2DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 92 Power Operation 92 Power Operation

Event Text

PLANNED MAINTENANCE ON THE TECHNICAL SUPPORT CENTER NORMAL POWER SUPPLY

"At approximately 0800 [EDT] on April 16, 2012, the Harris Nuclear Plant (HNP) Technical Support Center (TSC) normal power feed will be removed from service for scheduled maintenance.

"The maintenance will consist of first switching the TSC to the TSC backup power supply. The normal supply will be disconnected and replaced with another offsite power source which is independent of the Harris switchyard. This power arrangement will remain in place while maintenance is performed on the TSC normal power supply and is expected to last approximately two months. A backup diesel generator is stationed near the TSC which can be connected if necessary during an emergency.

"An update will be provided when the TSC normal power supply has been returned to its normal alignment

"This event is reportable per 10CFR50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 2, since this work activity affects an emergency response facility for the duration of the maintenance.

"The [NRC] Senior Resident Inspector has been informed."

Page Last Reviewed/Updated Monday, April 16, 2012
Monday, April 16, 2012