Event Notification Report for October 06, 2002
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
10/05/2002 - 10/06/2002
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 39250
Facility: VERMONT YANKEE
Region: 1 State: VT
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: MITCH McCLUSKIE
HQ OPS Officer: FANGIE JONES
Region: 1 State: VT
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: MITCH McCLUSKIE
HQ OPS Officer: FANGIE JONES
Notification Date: 10/06/2002
Notification Time: 18:00 [ET]
Event Date: 10/06/2002
Event Time: 11:30 [EDT]
Last Update Date: 10/11/2002
Notification Time: 18:00 [ET]
Event Date: 10/06/2002
Event Time: 11:30 [EDT]
Last Update Date: 10/11/2002
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
PETE ESELGROTH (R1)
PETE ESELGROTH (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
RCIC SYSTEM DECLARED INOPERABLE
Following reactor core isolation cooling system injection check valve surveillance, the check valve apparently did not fully close. This resulted in high pump suction pressure trip which would have prevented further system operation. The licensee conservatively declared the RCIC system inoperable and entered Tech Spec LCO 3.5.G.2, a 14 day action statement. Shortly afterwards the plant pressure was dropped to less than 150 psig as part of normal plant shutdown to refueling, this takes the plant out of the Tech Spec action statement as it no longer applies. The licensee is conducting an investigation of the problem with RCIC injection check valve. Repairs will be completed prior to plant restart.
The licensee notified the NRC Resident Inspector and the State of Vermont.
* * * RETRACTION FROM Andy Wisniewski on 10/11/02 at 1428 EDT to Gerry Waig * * *
"BASIS FOR RETRACTION:
NRC Event Notification 39250 was made with the concern that this condition may have rendered the Reactor Core Isolation Cooling (RCIC) system inoperable due to a 'high pump suction pressure trip'. The RCIC pump does not have this aforementioned trip device. Therefore, the predicament that Operations has been confronted with for declaring the RCIC system inoperable may be more aptly stated as; 'Could the leakage that occurred past the discharge check valve, after the RCIC system run, have over-pressurized the RCIC system suction piping and caused the system to be unavailable for service if called upon for a 'Loss of All AC Event'. A team of engineers was assembled to determine the extent of this condition that resulted in the completion of an Operability Determination on 10/08/02 by the RCIC System Engineer, concluding that the RCIC system remained operable based upon the following two facts:
1. The RCIC suction line relief valve lifts at 150 psi.
2. The outer pump discharge valve, combined with a pressure switch, provides an additional barrier of defense. The pressure switch actuates at 74 psi and causes the RCIC PUMP SUCT PRESS HI alarm in the control room. The alarm response procedure directs operators to close the isolation valve that is downstream of the discharge check valve.
"Additionally, the leakage has been determined to be 'slight' and well within the capacity of the relief valve, based upon system performance and observation of the event as it occurred. The RCIC system would have performed it's specified safety function during this condition if required.
"Therefore, ENS Event Number 39250, made on 10/06/02, is being retracted."
The licensee notified the NRC Resident Inspector of this retraction. The R1DO (John White) was notified by the NRC Headquarters Operations Officer.
Following reactor core isolation cooling system injection check valve surveillance, the check valve apparently did not fully close. This resulted in high pump suction pressure trip which would have prevented further system operation. The licensee conservatively declared the RCIC system inoperable and entered Tech Spec LCO 3.5.G.2, a 14 day action statement. Shortly afterwards the plant pressure was dropped to less than 150 psig as part of normal plant shutdown to refueling, this takes the plant out of the Tech Spec action statement as it no longer applies. The licensee is conducting an investigation of the problem with RCIC injection check valve. Repairs will be completed prior to plant restart.
The licensee notified the NRC Resident Inspector and the State of Vermont.
* * * RETRACTION FROM Andy Wisniewski on 10/11/02 at 1428 EDT to Gerry Waig * * *
"BASIS FOR RETRACTION:
NRC Event Notification 39250 was made with the concern that this condition may have rendered the Reactor Core Isolation Cooling (RCIC) system inoperable due to a 'high pump suction pressure trip'. The RCIC pump does not have this aforementioned trip device. Therefore, the predicament that Operations has been confronted with for declaring the RCIC system inoperable may be more aptly stated as; 'Could the leakage that occurred past the discharge check valve, after the RCIC system run, have over-pressurized the RCIC system suction piping and caused the system to be unavailable for service if called upon for a 'Loss of All AC Event'. A team of engineers was assembled to determine the extent of this condition that resulted in the completion of an Operability Determination on 10/08/02 by the RCIC System Engineer, concluding that the RCIC system remained operable based upon the following two facts:
1. The RCIC suction line relief valve lifts at 150 psi.
2. The outer pump discharge valve, combined with a pressure switch, provides an additional barrier of defense. The pressure switch actuates at 74 psi and causes the RCIC PUMP SUCT PRESS HI alarm in the control room. The alarm response procedure directs operators to close the isolation valve that is downstream of the discharge check valve.
"Additionally, the leakage has been determined to be 'slight' and well within the capacity of the relief valve, based upon system performance and observation of the event as it occurred. The RCIC system would have performed it's specified safety function during this condition if required.
"Therefore, ENS Event Number 39250, made on 10/06/02, is being retracted."
The licensee notified the NRC Resident Inspector of this retraction. The R1DO (John White) was notified by the NRC Headquarters Operations Officer.
Power Reactor
Event Number: 39251
Facility: COMANCHE PEAK
Region: 4 State: TX
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MARK VERBECK
HQ OPS Officer: STEVE SANDIN
Region: 4 State: TX
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MARK VERBECK
HQ OPS Officer: STEVE SANDIN
Notification Date: 10/06/2002
Notification Time: 18:45 [ET]
Event Date: 10/06/2002
Event Time: 16:50 [CDT]
Last Update Date: 10/06/2002
Notification Time: 18:45 [ET]
Event Date: 10/06/2002
Event Time: 16:50 [CDT]
Last Update Date: 10/06/2002
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
Person (Organization):
WILLIAM JONES (R4)
WILLIAM JONES (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
4-HOUR REPORT PURSUANT TO 10 CFR 50.72(b)(2) PER LICENSEE ADMINISTRATIVE REQUIREMENTS
"Results of the steam generator eddy current testing of the hot leg top of the tube sheet rotating plus point coil program indicated that greater than 1% of the tubes are defective in Steam Generator #2.
"Technical Specification category C-3. 5.6.10."
The licensee informed the NRC resident inspector.
* * * UPDATE 2326EDT ON 10/6/02 FROM DENNIS REDINGER TO S.SANDIN * * *
The licensee is updating this report to include the following additional SG inspection findings for Steam Generators #3 and #4:
"Results of the steam generator eddy current testing of the hot leg top of the tube sheet rotating pancake plus point coil indicates that greater than 1% of the tubes are defective in CPSES Unit 1 Steam Generators # 3 and Steam Generators # 4.
"CPSES Technical Specification states that, '[R]esults of steam generator tube inspections which fall into Category C-3 shall be reported to the Commission pursuant to 10 CFR 50.72(b)(2) within four hours of initial discovery, and in a report within 30 days and prior to resumption of plant operation.'
"Note:
"We understand that CPSES Technical Specification is conservative with respect to timing of reporting this condition; and that the current 10 CFR 50.72 requirements are eight hour notification. However, until the license amendment (which is in process) is issued, CPSES will continue to report this condition within four hours."
The licensee informed the NRC resident inspector. Notified R4DO(Jones).
"Results of the steam generator eddy current testing of the hot leg top of the tube sheet rotating plus point coil program indicated that greater than 1% of the tubes are defective in Steam Generator #2.
"Technical Specification category C-3. 5.6.10."
The licensee informed the NRC resident inspector.
* * * UPDATE 2326EDT ON 10/6/02 FROM DENNIS REDINGER TO S.SANDIN * * *
The licensee is updating this report to include the following additional SG inspection findings for Steam Generators #3 and #4:
"Results of the steam generator eddy current testing of the hot leg top of the tube sheet rotating pancake plus point coil indicates that greater than 1% of the tubes are defective in CPSES Unit 1 Steam Generators # 3 and Steam Generators # 4.
"CPSES Technical Specification states that, '[R]esults of steam generator tube inspections which fall into Category C-3 shall be reported to the Commission pursuant to 10 CFR 50.72(b)(2) within four hours of initial discovery, and in a report within 30 days and prior to resumption of plant operation.'
"Note:
"We understand that CPSES Technical Specification is conservative with respect to timing of reporting this condition; and that the current 10 CFR 50.72 requirements are eight hour notification. However, until the license amendment (which is in process) is issued, CPSES will continue to report this condition within four hours."
The licensee informed the NRC resident inspector. Notified R4DO(Jones).