05000461/LER-2004-001

LER-2004-001, Clinton Power Station 05000461 1 OF 4

Docket Number
Event date:
Report date:
4612004001R01 - NRC Website

A. PLANT OPERATING CONDITIONS PRIOR TO THE EVENT

Event Date: 3/22/2004 Event Time: 1931 Central Standard Time Mode: 1 (POWER OPERATION)�Reactor Power: 93 percent Unit: 1�

B. DESCRIPTION OF THE EVENT

On March 22, 2004, at about r931 hours, an automatic reactor scram occurred with the plant at 93 percent power. Operators in the Main Control Room received a trouble alarm for the Hydrogen and Stator Cooling Cabinet [TIC] [TJ] [CAB] followed by a Main Generator [GEN] [TB] neutral over-voltage trip and Generator Trip System 2 Lockout.

The Generator trip caused a Main Turbine [TRB] [TA] trip and Turbine Control Valve [V] fast closure, resulting in an automatic reactor scram. All control rods fully inserted.

Following the scram, as expected, reactor pressure vessel water level dropped below the Low Level 3 trip setpoint to 0.0 inches Narrow Range, initiating the Reactor Protection System [JC]. (Low Level 3 is 8.9 inches Narrow Range indication.) At 1932 hours0.0224 days <br />0.537 hours <br />0.00319 weeks <br />7.35126e-4 months <br /> operators entered the actions of procedure CPS 4001.01, "Reactor Scram Off-Normal," in response to the reactor scram and the lowering RPV water level. At 1939 hours0.0224 days <br />0.539 hours <br />0.00321 weeks <br />7.377895e-4 months <br />, operators completed the immediate actions of the reactor scram off­ normal procedure and entered Emergency Operating Procedure (EOP) 1, "RPV Control," due to the low reactor water level transient.

At 2006 hours0.0232 days <br />0.557 hours <br />0.00332 weeks <br />7.63283e-4 months <br />, the reactor scram signal was reset.

At 2215 hours0.0256 days <br />0.615 hours <br />0.00366 weeks <br />8.428075e-4 months <br />, operators exited EOP 1 and transitioned into Procedure CPS 3006.01, "Unit Shutdown," as reactor pressure was stable at 858 psig and reactor water level was stable at 34 inches Narrow Range. At 0130 hours0.0015 days <br />0.0361 hours <br />2.149471e-4 weeks <br />4.9465e-5 months <br /> on March 23, operators exited the reactor scram off-normal procedure.

As expected, the Low Level 3 RPV water level trip caused Primary Containment Isolation Valves [ISV] in Group 2 (Residual Heat Removal (RHR)[130]), Group 3 (RHR), and Group 20 (miscellaneous systems) to receive signals to shut; these valves were already shut prior to the event in accordance with the normal plant lineup.

The reactor remained in Mode 3 (HOT SHUTDOWN) with reactor coolant pressure being controlled between 800 and 1065 psig using the Turbine Bypass Valves and steam drains [DRN], and reactor coolant level being maintained between Low Level 3 and High Level 8 using the Motor-Driven Reactor Feed Pump [MO] [P] [SJ].

No Main Steam Isolation Valves closed and no Safety Relief Valves lifted during this event.

Condition Report 210033 was initiated to track the investigation and resolution of this event.

B. DESCRIPTION OF THE EVENT (continued) :

No automatic or manually initiated safety system actuations were necessary to place the plant in a safe and stable condition. No other inoperable equipment or components directly affected this event.

C. CAUSE OF THE EVENT

A complex troubleshooting plan was initiated to investigate this event.

Troubleshooting identified two components within the Isolated Phase Bus Duct Cooling System [BDUC] [EL] failed. A Partial Discharge Analysis (PDA) System cable [CBL] was severed and a conductor expansion joint (EXJ] was torn during the initial inspection of the 'B' Isolated Phase Bus Duct.

The root cause investigation determined that the failed PDA cable and/or a piece of aluminum laminate from a degraded bus conductor expansion joint contacted the grounded bus duct and caused a ground fault and subsequent generator neutral over-voltage trip. The failures of the PDA cable and the aluminum laminate were both due to mechanical fatigue (vibration fatigue) caused by an increase in design air flow rate (from 21,700 scfm to 36,700 scfm) in the bus duct implemented during the Spring 2004 refueling outage. The increased air flow was implemented in order to accommodate increased cooling requirements for Extended Power Uprate.

D. SAFETY ANALYSIS

There were no actual safety consequences associated with his event. The event was reviewed for analyzed transients discussed in Chapter 15 of the Clinton Power Station Updated Safety Analysis Report. The analysis determined that this event was within the design basis of the plant.

No safety system functional failures occurred during this event.

E. CORRECTIVE ACTIONS

All three Isolated Phase Bus Ducts were inspected for degradation, the PDA System cables have been removed from all three Isolated Phase Bus Ducts, and an expansion link on the conductor in the 'B' Isolated Phase Bus Duct has been repaired. A design change will be implemented for the bus duct expansion joint that is not susceptible to fatigue failure.

F. PREVIOUS OCCURRENCES

Based on a review of industry operating experience, no previous similar events are known.

G. COMPONENT FAILURE DATA:

Manufacturer�Nomenclature�Model�Mfg. Part Number 22 kV bus duct General Electric 357A2528P0003� H.K. Porter n/a n/a Partial Discharge n/a Assembly Kit Coaxial Connector Note: The H.K. Porter isolated phase bus duct product line was purchased by Delta- Unibus.

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