(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
On April 28, 2003 at 1648 hours0.0191 days <br />0.458 hours <br />0.00272 weeks <br />6.27064e-4 months <br /> Indian Point Unit 2 experienced an automatic reactor trip with all control rods fully inserting. N The trip was initiated by main turbine trip on auto stop oil.
The auto stop oil turbine trip was caused by a trip of the over frequency relays actuated by a disturbance associated with the 345kV North Ring Bus at the Buchanan Substation and the Consolidated Edison 138kV system. Following the grid disturbance breakers 11 and 7 opened followed by a Phase A ground fault on breaker F7. When Consolidated Edison attempted to re-close breaker 11, a relay [EIIS:FK:68) malfunction occurred causing Unit 2 output breaker 9 to open resulting in a turbine trip.
The resultant trip placed the plant in natural circulation with 480-volt buses 2A and 3A de-energized as per design. N All three Emergency Diesel Generators ( EDGs) started and buses 2A and 3A were manually energized by 22 EDG, this was an expected response. 480-volt buses 5A and 6A remained energized from off- site sources during this event. N No steam generator or pressurizer safety valves lifted and actuation of the Safety Injection System was not required. N No radioactive release to the environment occurred as a result of this transient. |
FACILITY NAME (1)
DOCKET
l) I SEOUENT REVISI IAL � ON � NI IKARFP � 'hoe us aor LER NUMBER (6
PLANT AND SYSTEM IDENTIFICATION
Westinghouse 4-Loop Pressurized Water Reactor
EVENT IDENTIFICATION
Automatic Reactor trip initiated by a main turbine trip on auto stop oil.
EVENT DATE
April 28, 2003
REFERENCE
Condition Reporting System Number: 200302511
PAST SIMILAR EVENTS
Licensee Event Report Number: 2001-007-00 Licensee Event Report Number: 1997-018-00
EVENT DESCRIPTION
On April 28, 2003 at 1648 hours0.0191 days <br />0.458 hours <br />0.00272 weeks <br />6.27064e-4 months <br /> Indian Point Unit 2 experienced an automatic reactor trip with all control rods fully inserting. The trip was initiated by a main turbine trip on auto stop oil.
The auto stop oil turbine trip was caused by a trip of the over frequency relays actuated by a disturbance associated with the 345kV North Ring Bus at the Buchanan Substation and the Consolidated Edison 138kV system. This disturbance was caused by a Phase A ground fault on 345kV transmission line Y94. Following the grid disturbance breakers 11 and 7 opened at the Buchanan Substation followed by a Phase A ground fault on 138kV breaker F7. When Consolidated Edison attempted to re-energize line Y94 by closing breaker 11 a CEYB (General Electric type) relay malfunction occurred causing output breaker 9 to open resulting in a turbine trip.
The resultant trip placed the plant in natural circulation with 480-volt buses 2A and 3A de-energized as per design. All three Emergency Diesel Generators (EDGs) started and buses 2A and 3A were manually energized by 22 EDG, this was an expected response. 480-volt buses 5A and 6A remained energized from off-site sources during this event. No steam generator or pressurizer safety valves lifted and actuation of the Safety Injection System was not required. No radioactive release to the environment occurred as a result of this transient.
DOCKET
SEQUENT
IAL
MHPARPP
REOSI
ON
MIORAft
EVENT ANALYSIS
This event is reportable in accordance with 10CFR50.73(a)(2)(iv)(A) which requires a Licensee Event Report (LER) for any event that resulted in manual or automatic actuation of the Reactor Protection System (RPS) including: reactor scram or reactor trip.
EVENT SAFETY SIGNIFICANCE
This event was initiated as a result of a grid disturbance on the North 345kV ring bus at the Buchanan switchyard. This is an expected plant response due to the actuation of the over-frequency protection circuit:
These relays were added as part of a plant modification after a similar event in July 1997 resulted in a 100% load reject. Since this event is bounded by section 14.1.12 (Loss of all power to the Station Auxiliaries) of the Updated Final Safety Analysis Report (UFSAR) the safety significance was determined to be minimal.
FACILITY NAME (1)
DOCKET
SEQUENT
IAL
NI URPP
ON
CORRECTIVE ACTION
The root cause of this event was the malfunction of the CEYB relay.
This relay detected the fault in the wrong direction. The CEYB relay detected a fault on line W93 when the fault was actually on line Y94.
This malfunction caused breaker 9 to open resulting in Unit 2 tripping off line. Consolidated Edison removed the CEYB relay from the system to determine cause of malfunction. There are redundant relays in place .
monitoring the distribution system. Entergy has assigned Corrective Actions to follow up with Consolidated Edison and obtain their root cause report for the faults on the Y94 feeder and breaker F7 and the failure of the CEYB relay. The root cause report from Consolidated Edison is expected by the end of the third quarter 2003. This report will also contain Consolidated Edison's actions to prevent re- occurrence. Entergy Management has increased their involvement in the substation activities since the event. Senior levels of Entergy management have been meeting with their Consolidated Edison counterparts to ensure that appropriate actions are being taken to increase the reliability of the electrical system. Entergy has named a switchyard coordinator who is responsible for the interface between the plant and Consolidated Edison.
PREVIOUS OCCURRENCES
Similar events occurred December 26, 2001 and July 26, 1997 and are documented in LER 2001-007-00 and LER 1997-018-00. The root cause of the December 2001 event was the failure of a blocking relay on Consolidated Edison's 345kV line Y94. The root cause of the July 1997 event was malfunction of a directional relay device associated with transformer TA5. As a result of the July 1997 event, over-frequency relays were added to the overall unit protection scheme. The over- frequency relays actuated as per design for the December 26, 2001 event.
|
|---|
|
|
| | | Reporting criterion |
|---|
| 05000397/LER-2003-010 | | | | 05000528/LER-2003-001 | Pressurizer Safety Valve As-Found Lift Pressure Outside of Technical Specification Limits | 10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000277/LER-2003-001 | | | | 05000282/LER-2003-001 | | | | 05000301/LER-2003-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000261/LER-2003-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000251/LER-2003-001 | Channel Failure of Qualified Safety Parameter Display System | | | 05000316/LER-2003-001 | Unit 2 Shutdown In Accordance With Technical Specification 3.8.1.1, A.C. Sources, Action b | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | | 05000324/LER-2003-001 | Main Steam Line Drain Isolation Valve Local Leak Rate Test Failures | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | | 05000352/LER-2003-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000353/LER-2003-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000397/LER-2003-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000364/LER-2003-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000529/LER-2003-001 | Reactor Trip with Loss of Forced Circulation Due to Failed Pressurizer Main Spray Valve | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000278/LER-2003-001 | | | | 05000305/LER-2003-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000331/LER-2003-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000313/LER-2003-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000352/LER-2003-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000301/LER-2003-002 | | 10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | | 05000305/LER-2003-002 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | | 05000316/LER-2003-002 | Supplemental LER for Unit 2 Reactor Trip due to Instrument Rack 24 Volt Power Supply Failure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(iv), System Actuation | | 05000458/LER-2003-002 | | 10 CFR 50.73(a)(2)(v)(c) | | 05000348/LER-2003-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000341/LER-2003-002 | Automatic Reactor Shutdown Due to Electric Grid Disturbance and Loss of Offsite Power | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000261/LER-2003-002 | | | | 05000285/LER-2003-002 | 4 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000397/LER-2003-002 | | | | 05000499/LER-2003-002 | Safety Injection Actuation | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000440/LER-2003-002 | Reactor Scram as a Result of a Loss of Off-site Power | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
10 CFR 50.73(a)(2)(iii) | | 05000400/LER-2003-002 | 1 O OF 3 3 | | | 05000266/LER-2003-002 | | 10 CFR 50.73(a)(2)(iv), System Actuation | | 05000250/LER-2003-003 | Unescorted Access Inappropriately Approved Due to Falsified Pre-Access Information | | | 05000261/LER-2003-003 | | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor | | 05000219/LER-2003-003 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000247/LER-2003-003 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000530/LER-2003-003 | | 10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000331/LER-2003-003 | | | | 05000529/LER-2003-003 | SOURCE RANGE MONITOR INOPERABLE DURING CORE RELOAD | 10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000440/LER-2003-003 | Unrecognized Diesel Generator Inoperability During Mode Changes | | | 05000348/LER-2003-003 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000482/LER-2003-003 | REACTOR PROTECTION SYSTEM ACTUATION AND REACTOR TRIP DUE TO FEEDWATER ISOLATION VALVE CLOSURE | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(iv), System Actuation | | 05000301/LER-2003-003 | | | | 05000302/LER-2003-003 | Reactor Coolant System Pressure Boundary Leakage Limit Exceeded Due To Pressurizer Instrument Tap Nozzle Cracks | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | | 05000382/LER-2003-003 | RCS Pressure Boundary Leakage Due to Primary Water Stress Corrosion Cracking (PWSCC) | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | | 05000397/LER-2003-003 | | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(iv), System Actuation | | 05000458/LER-2003-003 | | 10 CFR 50.73(a)(2)(v)(c) | | 05000454/LER-2003-003 | Licensed Maximum Power Level Exceeded Due to Inaccuracies in Feedwater Ultrasonic Flow Measurements | | | 05000282/LER-2003-003 | | | | 05000346/LER-2003-014 | Steam Feedwater Rupture Controls System Re-Energizes in a Blocked Condition | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
|