| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
06/15/2007 - 06/18/2007 ** EVENT NUMBERS ** |
 | Power Reactor | Event Number: 43414 | Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: TIM BOLAND
HQ OPS Officer: JOE O'HARA | Notification Date: 06/09/2007
Notification Time: 15:53 [ET]
Event Date: 06/09/2007
Event Time: 11:00 [CDT]
Last Update Date: 06/15/2007 | 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):
PAUL FREDRICKSON (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | A/R | Y | 79 | Power Operation | 0 | Hot Shutdown | Event Text AUTOMATIC REACTOR SCRAM DUE TURBINE TRIP AS A RESULT OF HIGH MOISTURE SEPARATOR REHEATER TANK LEVEL
"On 06/09/2007 at 1100 CDT, Browns Ferry Unit 1 received an automatic SCRAM due to a Turbine Trip Signal caused by a Moisture Separator Drain Tank Level High. All control rods inserted and all systems responded as required to the automatic SCRAM signal. Two Main Steam Relief Valves (MSRVs) momentarily lifted in response to the pressure transient. No Emergency Core Cooling System (ECCS) initiations occurred as a result of the SCRAM. Reactor water level lowered below Level 3 (+2") as a result of the SCRAM and was recovered to the normal level band by the reactor feed pumps (RFPs). The expected Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolations were received due to reactor water level lowering below Level 3 (+2") with all systems isolating as required. Additionally, the expected initiation of the RPT breakers from the turbine trip was received which resulted in the trip of both reactor recirculation pumps. Reactor pressure is being controlled using Main Steam Bypass Valves. Reactor Level is being maintained in band using RFPs. Cooldown is in progress to Mode 4. Investigation into the event is proceeding.
"The NRC Senior Resident Inspector has been informed of this event.
"This event is reportable under 10CFR50.72(b)(2)(iv)(B) 'any event or condition that results in a valid actuation of the Reactor Protection System'; 10 CFR50.72(b)(3)(iv)(A), ' Any event that results in an actuation of the specified systems'. This event also requires a 60 day written report in accordance with 10CFR50.73(a)(2)(iv)(A)."
* * * UPDATE FROM BOLAND TO HUFFMAN AT 1850 EDT ON 6/9/07 * * *
Follow-up review of the reported reactor scram revealed that the Group 8 isolation did not function as required. Specifically, the licensee provided the update below concerning the function of one of the reactor's five Traversing Incore Probes (TIP) that had been used the previous day for flux mapping and were in the drywell (not fully retracted) to permit decay when the scram occurred.
"The Group 8 isolation signal received during Unit 1 Rx SCRAM on 06/09/2007 @ 1100 did not automatically go to completion as designed. The 'D' TIP failed to automatically withdraw as required. When the TIP was manually withdrawn, the TIP Ball valve closed as required. The local resident was notified. A work order and PER was written to correct the deficiency."
The other four TIPs did retract and the corresponding ball valves shut as expected. R2DO (Fredrickson) notified.
* * * UPDATE FROM TIM GOLDEN TO JOE O'HARA AT 1804 ON 6/15/07 * * *
"Review of available data indicates that no Main Steam safety relief valves (MSRVs) opened in response to the Unit 1 reactor scram on 06-09-2007. There were no indications of an open MSRV on any discharge tailpipe thermocouple or acoustic monitor. Initial indications of the discharge tailpipe thermocouples for MSRVs 1-PCV-1-5, 1-PCV-1-30, and 1-PCV-1-31 did indicate slight increases in temperature (5 to 18 degrees F) as reactor pressure decreased, which resulted in the initial assumption of two SRVs opening. However, this behavior is a classical indication of slight main seat leakage. This equipment condition is under review via an open PER. The multiple reactor pressure instrumentation responses were reviewed. The peak reactor pressure was indicated at approximately 1093 psig which is 42 psi below the lowest nominal MSRV setpoint. Based upon the observed peak reactor pressure and no indication of an MSRV opening, it would appear that the MSRVs performed as required to during the reactor pressure transient event.
The licensee will notify the NRC Resident Inspector.
Notified R2DO(Shaeffer). |
| Power Reactor | Event Number: 43424 | Facility: POINT BEACH
Region: 3 State: WI
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: RICK ROBBINS
HQ OPS Officer: JOE O'HARA | Notification Date: 06/14/2007
Notification Time: 20:21 [ET]
Event Date: 06/14/2007
Event Time: 18:19 [CDT]
Last Update Date: 06/15/2007 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
| Person (Organization):
MONTE PHILLIPS (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 98 | Power Operation | Event Text TECH SPEC. REQUIRED SHUTDOWN DUE TO 72 HOUR COMPLETION TIME NOT MET
"At 1819 hours, an orderly shutdown on Unit 1 commenced as the result of Technical Specification Action Condition (TSAC) 3.7.5.B.1 completion time of 72 hours not being met for the 1P-29 turbine-driven auxiliary feedwater pump. The pump was declared inoperable on June 12, 2007, at 0131 hours as a result of high pump bearing temperatures. Repairs and testing performed to date have not satisfactorily resolved the problem.
"This non-emergency notification is being made in accordance with 10CFR50.72(b)(2)(i). The PBNP resident inspector has been notified."
The licensee plans to conduct an extended duration test of the 1P-29 AFW pump. This test is scheduled to commence shortly and will last approximately 4 to 6 hours. The licensee will have two different hold points in the test to take pump bearing temperature data. The licensee has one emergency diesel generator (EDG) out of service (OOS), but an alternate EDG is aligned to ensure all four vital buses have vital power in the event of an loss of offsite power. No other safety related systems are OOS at this time.
The licensee notified the NRC Resident Inspector.
* * * UPDATE FROM RICK ROBBINS TO JOE O'HARA AT 2006 ON 06/15/07 * * *
"This is an update to EN#43424: Regarding Unit 1 turbine driven auxiliary feedwater pump inoperability and Technical Specification required shutdown.
"PBNP Unit 1 entered MODE 3 on 6/15/07 at 0407.
"PBNP Unit 1 entered MODE 4 on 6/15/07 at 1712.
"Technical Specification Action Condition 3.7.5.D.1 required Unit 1 to be in MODE 3 by 0731 and MODE 4 by 1931 on 6/15/07. All Technical Specification Required Actions for the AFW pump OOS have been met within the required times."
The licensee informed the NRC Resident Inspector. Notified R3DO(Stone). | |