Event Notification Report for June 18, 1999
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
06/17/1999 - 06/18/1999
General Information or Other
Event Number: 35844
Rep Org: ROSEMOUNT NUCLEAR INSTRUMENTS, INC.
Licensee: ROSEMOUNT NUCLEAR INSTRUMENTS, INC
Region: 3
City: EDEN PRAIRIE State: MN
County:
License #:
Agreement: N
Docket:
NRC Notified By: GERRY HANSON
HQ OPS Officer: BOB STRANSKY
Licensee: ROSEMOUNT NUCLEAR INSTRUMENTS, INC
Region: 3
City: EDEN PRAIRIE State: MN
County:
License #:
Agreement: N
Docket:
NRC Notified By: GERRY HANSON
HQ OPS Officer: BOB STRANSKY
Notification Date: 06/18/1999
Notification Time: 12:28 [ET]
Event Date: 06/18/1999
Event Time: 10:00 [CDT]
Last Update Date: 06/18/1999
Notification Time: 12:28 [ET]
Event Date: 06/18/1999
Event Time: 10:00 [CDT]
Last Update Date: 06/18/1999
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
VERN HODGE (via fax) (NRR)
VERN HODGE (via fax) (NRR)
10 CFR PART 21 REPORT REGARDING POTENTIAL NEED FOR CAPACITOR REPLACEMENT IN SPECIFIC TRIP/CALIBRATION SYSTEMS
The following is text of a letter submitted to the NRC Operations Center via facsimile:
"The purpose of this letter is to inform you that a specific capacitor C25 used in Rosemount Trip/Calibration Systems may require replacement. This affects all Rosemount Model 510DU Trip Units (obsolete since 1987) and certain Rosemount Model 710DU Trip Units.
"Prior to 1987, there were 10 confirmed random failures of C25 capacitors used in slave and master trip unit circuit boards. In 1987, a number of tests were run on sample capacitors in an attempt to duplicate the failure modes. It was found that unless the capacitors were exposed to temperatures and voltages in excess of normal operating conditions, no failures were induced. However, introduction of a contaminant such as moisture could possibly cause this type of failure. As a precautionary measure, a component change was instituted on the C25 capacitor beginning with manufacturer date code 8630. This change featured an enhanced moisture seal. No confirmed failures of the post 8630 date code C25 capacitor have been identified. However, Rosemount Nuclear Instruments. Inc is sending this letter to affected utilities to remind users that the model 710 and 510 trip units with pre-8630 date code C25 capacitors could be susceptible to this condition. Failure of the C25 capacitor can cause a low resistance condition between the power supply input and chassis ground. This low resistance can allow over-current conditions to cause failure of the affected unit.
"A letter [enclosed in facsimile] was issued by Rosemount Nuclear Instruments Inc to the Scram Frequency Reduction Committee of the BWR Owners' Group in response to a BWROG request for information [enclosed in facsimile], documenting the change which had been made in 1987. Since that time, there have been five additional confirmed C25 failures and fourteen additional failures possibly related to the C25 capacitor. Again, all of these failures occurred in units with C25 capacitors that had date codes of pre-8630. The total number of model 510 and 710 trip units shipped with the pre-8630 date code capacitors was 18,164. We believe that recent failures may warrant further examination of your 510DU/710DU systems to determine whether replacement of the pre-8630 date code capacitors is warranted."
The following is text of a letter submitted to the NRC Operations Center via facsimile:
"The purpose of this letter is to inform you that a specific capacitor C25 used in Rosemount Trip/Calibration Systems may require replacement. This affects all Rosemount Model 510DU Trip Units (obsolete since 1987) and certain Rosemount Model 710DU Trip Units.
"Prior to 1987, there were 10 confirmed random failures of C25 capacitors used in slave and master trip unit circuit boards. In 1987, a number of tests were run on sample capacitors in an attempt to duplicate the failure modes. It was found that unless the capacitors were exposed to temperatures and voltages in excess of normal operating conditions, no failures were induced. However, introduction of a contaminant such as moisture could possibly cause this type of failure. As a precautionary measure, a component change was instituted on the C25 capacitor beginning with manufacturer date code 8630. This change featured an enhanced moisture seal. No confirmed failures of the post 8630 date code C25 capacitor have been identified. However, Rosemount Nuclear Instruments. Inc is sending this letter to affected utilities to remind users that the model 710 and 510 trip units with pre-8630 date code C25 capacitors could be susceptible to this condition. Failure of the C25 capacitor can cause a low resistance condition between the power supply input and chassis ground. This low resistance can allow over-current conditions to cause failure of the affected unit.
"A letter [enclosed in facsimile] was issued by Rosemount Nuclear Instruments Inc to the Scram Frequency Reduction Committee of the BWR Owners' Group in response to a BWROG request for information [enclosed in facsimile], documenting the change which had been made in 1987. Since that time, there have been five additional confirmed C25 failures and fourteen additional failures possibly related to the C25 capacitor. Again, all of these failures occurred in units with C25 capacitors that had date codes of pre-8630. The total number of model 510 and 710 trip units shipped with the pre-8630 date code capacitors was 18,164. We believe that recent failures may warrant further examination of your 510DU/710DU systems to determine whether replacement of the pre-8630 date code capacitors is warranted."
Power Reactor
Event Number: 35845
Facility: PALO VERDE
Region: 4 State: AZ
Unit: [] [2] []
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: DAN MARKS
HQ OPS Officer: DICK JOLLIFFE
Region: 4 State: AZ
Unit: [] [2] []
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: DAN MARKS
HQ OPS Officer: DICK JOLLIFFE
Notification Date: 06/19/1999
Notification Time: 04:16 [ET]
Event Date: 06/18/1999
Event Time: 22:29 [MST]
Last Update Date: 06/19/1999
Notification Time: 04:16 [ET]
Event Date: 06/18/1999
Event Time: 22:29 [MST]
Last Update Date: 06/19/1999
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
Person (Organization):
JOHN PELLET (R4)
JOHN PELLET (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | A/R | Y | 100 | Power Operation | 0 | Hot Standby |
- Auto reactor trip from 100% power due to low DNBR trips due to a failed control element assembly calculator -
On 06/18/99, at approximately 2229 MST, Palo Verde Unit 2 experienced a reactor trip (RPS actuation) from 100% rated thermal power due to low departure from nucleate boiling ratio (DNBR) trips on all four channels of the core protection calculators (CPCs). Unit 2 was at normal temperature and pressure prior to the trip.
Prior to the reactor trip, at approximately 2204 MST, a CPC channel 'C' sensor fail alarm and control element assembly (CEA) withdrawal prohibit (CWP) alarm were received and cleared about one second later. While
investigating the alarms, at 2229 MST, all four CPC channel sensor fail alarms and the CWP alarm were received. The reactor automatically tripped six seconds later. A CEA calculator (CEAC) fail alarm was received on CEAC #2. The apparent cause is presently suspected to be the failure of CEAC #2. An investigation has commenced to determine the root cause of the reactor trip.
All control rods fully inserted into the core. Four of eight steam bypass control valves quick opened, per design, directing steam flow to the main condenser. No main steam or primary relief valves lifted and none were required. There was no loss of heat removal capability or loss of safety functions associated with this event. Electrical buses transferred to offsite power, as designed. The Shift Manager determined that this event was an uncomplicated reactor trip. No significant LCOs have been entered as a result of this event. No major equipment was inoperable prior to the event nor contributed to the event, other than the CPCs discussed above.
Unit 2 is stable in Mode 3 (hot standby) at normal operating temperature and pressure. No ESF actuations occurred and none were required. The event did not result in any challenges to the fission product barrier or result in any releases of radioactive materials. There were no adverse safety consequences or implications as a result of this event. The event did not adversely affect the safe operation of the plant or health and safety of the public.
The licensee notified the NRC Resident Inspector.
On 06/18/99, at approximately 2229 MST, Palo Verde Unit 2 experienced a reactor trip (RPS actuation) from 100% rated thermal power due to low departure from nucleate boiling ratio (DNBR) trips on all four channels of the core protection calculators (CPCs). Unit 2 was at normal temperature and pressure prior to the trip.
Prior to the reactor trip, at approximately 2204 MST, a CPC channel 'C' sensor fail alarm and control element assembly (CEA) withdrawal prohibit (CWP) alarm were received and cleared about one second later. While
investigating the alarms, at 2229 MST, all four CPC channel sensor fail alarms and the CWP alarm were received. The reactor automatically tripped six seconds later. A CEA calculator (CEAC) fail alarm was received on CEAC #2. The apparent cause is presently suspected to be the failure of CEAC #2. An investigation has commenced to determine the root cause of the reactor trip.
All control rods fully inserted into the core. Four of eight steam bypass control valves quick opened, per design, directing steam flow to the main condenser. No main steam or primary relief valves lifted and none were required. There was no loss of heat removal capability or loss of safety functions associated with this event. Electrical buses transferred to offsite power, as designed. The Shift Manager determined that this event was an uncomplicated reactor trip. No significant LCOs have been entered as a result of this event. No major equipment was inoperable prior to the event nor contributed to the event, other than the CPCs discussed above.
Unit 2 is stable in Mode 3 (hot standby) at normal operating temperature and pressure. No ESF actuations occurred and none were required. The event did not result in any challenges to the fission product barrier or result in any releases of radioactive materials. There were no adverse safety consequences or implications as a result of this event. The event did not adversely affect the safe operation of the plant or health and safety of the public.
The licensee notified the NRC Resident Inspector.
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 35842
Facility: LASALLE
Region: 3 State: IL
Unit: [] [2] []
RX Type: [1] GE-5,[2] GE-5
NRC Notified By: JIM SPIELER
HQ OPS Officer: DOUG WEAVER
Region: 3 State: IL
Unit: [] [2] []
RX Type: [1] GE-5,[2] GE-5
NRC Notified By: JIM SPIELER
HQ OPS Officer: DOUG WEAVER
Notification Date: 06/18/1999
Notification Time: 05:48 [ET]
Event Date: 06/18/1999
Event Time: 02:50 [CDT]
Last Update Date: 06/22/1999
Notification Time: 05:48 [ET]
Event Date: 06/18/1999
Event Time: 02:50 [CDT]
Last Update Date: 06/22/1999
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION
Person (Organization):
JOHN MADERA (R3)
JOHN MADERA (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
ESF ACTUATION - VACUUM BREAKER
At 0250 CDT, #2PC001D, the 2D Primary Containment Vacuum Breaker went from the (normal) closed position to an indicated dual position. Locally, it was determined that the valve was just cracked open. The valve is designed to open when suppression chamber pressure is approximately 0.5 psig greater than drywell pressure. When the vacuum breaker opened the aforementioned differential pressure was approximately 0.2 psid. The vacuum breaker was closed by an equipment attendant and then cycled per LOS-PC-M2 "Unit 2 Drywell-Suppression Pool Vacuum Breaker Operability Test for Condition 1, 2 and 3" to verify proper operation. The valve cycled satisfactorily.
The licensee notified the NRC resident inspector.
NOTE: THIS EVENT IS SIMILAR TO EVENT #35840.
*** UPDATE ON 6/22/99 @ 1611 BY MARIK TO GOULD *** RETRACTION
BASED ON A REVIEW BY STATION MANAGEMENT AND CONCURRENCE FROM CORPORATE REGULATORY ASSURANCE, THE LICENSEE IS RETRACTING THIS EVENT. FOLLOWING ADDITIONAL REVIEWS OF 10CFR50.72 AND NUREG-1022 REV 1, IT HAS BEEN DETERMINED THAT THE ACTUATION OF THE "D" CONTAINMENT VACUUM BREAKER IS AN ACTUATION OF A SINGLE COMPONENT OF THE CONTAINMENT VACUUM RELIEF SYSTEM (CVARS). THIS ACTUATION BY ITSELF WOULD NOT BE CAPABLE OF PROVIDING THE CVARS ESF DESIGN FUNCTION OF PROTECTING THE DRYWELL FLOOR AGAINST ADVERSE NEGATIVE PRESSURE DIFFERENTIALS. THEREFORE, THIS EVENT IS NOT REPORTABLE AS AN ESF AND ACCORDINGLY IS BEING RETRACTED.
THE RESIDENT INSPECTOR WAS NOTIFIED. REG 3 RDO (JORDAN) WAS NOTIFIED.
At 0250 CDT, #2PC001D, the 2D Primary Containment Vacuum Breaker went from the (normal) closed position to an indicated dual position. Locally, it was determined that the valve was just cracked open. The valve is designed to open when suppression chamber pressure is approximately 0.5 psig greater than drywell pressure. When the vacuum breaker opened the aforementioned differential pressure was approximately 0.2 psid. The vacuum breaker was closed by an equipment attendant and then cycled per LOS-PC-M2 "Unit 2 Drywell-Suppression Pool Vacuum Breaker Operability Test for Condition 1, 2 and 3" to verify proper operation. The valve cycled satisfactorily.
The licensee notified the NRC resident inspector.
NOTE: THIS EVENT IS SIMILAR TO EVENT #35840.
*** UPDATE ON 6/22/99 @ 1611 BY MARIK TO GOULD *** RETRACTION
BASED ON A REVIEW BY STATION MANAGEMENT AND CONCURRENCE FROM CORPORATE REGULATORY ASSURANCE, THE LICENSEE IS RETRACTING THIS EVENT. FOLLOWING ADDITIONAL REVIEWS OF 10CFR50.72 AND NUREG-1022 REV 1, IT HAS BEEN DETERMINED THAT THE ACTUATION OF THE "D" CONTAINMENT VACUUM BREAKER IS AN ACTUATION OF A SINGLE COMPONENT OF THE CONTAINMENT VACUUM RELIEF SYSTEM (CVARS). THIS ACTUATION BY ITSELF WOULD NOT BE CAPABLE OF PROVIDING THE CVARS ESF DESIGN FUNCTION OF PROTECTING THE DRYWELL FLOOR AGAINST ADVERSE NEGATIVE PRESSURE DIFFERENTIALS. THEREFORE, THIS EVENT IS NOT REPORTABLE AS AN ESF AND ACCORDINGLY IS BEING RETRACTED.
THE RESIDENT INSPECTOR WAS NOTIFIED. REG 3 RDO (JORDAN) WAS NOTIFIED.