05000390/LER-2003-001, Re Automatic Reactor Trip Due to Moisture Intrusion Into an Electrical Connector
| ML031330052 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 05/09/2003 |
| From: | Lagergren W Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 03-001-00 | |
| Download: ML031330052 (9) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3902003001R00 - NRC Website | |
text
Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 William R. Lagergren, Jr.
Site Vice President, Watts Bar Nuclear Plant 10 CFR 50.73 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D. C. 20555 Gentlemen:
In the Matter of Tennessee Valley Authority Docket No.50-390
)
WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 -
DOCKET NO. 50-390 -
FACILITY OPERATING LICENSE NPF LICENSEE EVENT REPORT (LER) 50-390/2003-001 This submittal provides Licensee Event Report 390/2003-001.
This LER addresses an event that occurred on March 10, 2003, which resulted in an automatic actuation of engineered safety features which included the reactor protection system and auxiliary feedwater system.
This event is being reported under 10 CFR 50.73(a)(2) (iv) (A).
There are no regulatory commitments in this submittal.
If you have any questions about this change, please contact P. L.
Pace at (423) 365-1824.
Sincerely, W
R Lager en Enclosure cc:
See page 2 P
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U.S. Nuclear Regulatory Commission Page 2 MAY 0 9 2003 cc (Enclosure):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. K. N. Jabbour, Senior Project Manager U.S. Nuclear Regulatory Commission MS 08G9 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303
Abstract
On March 10, 2003, at 0012 Eastern Standard Time, with Watts Bar 1 at 100 percent power, a Generator Backup Relay unexpectedly actuated causing an automatic turbine/reactor trip. The relay actuated due to a ground fault caused by a broken o-ring in the C phase main transformer's high side bushing capacitance tap connector.
Plant safety equipment performed as designed which included the auto-start of the auxiliary feedwater system.
The root cause of this event was inadequate preventive maintenance procedure. Corrective actions include repairing the connector, revising preventative maintenance (PM) procedures, and a design change to the affected single point vulnerability relay scheme.
NRC FORM 366 (7.2001) i I
'.U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE (3) 05000 YEAR SEQUENTIAL REVISION NUMBER 2
of 7
Watts Bar Nuclear Plant 05000 - 390 2003 001 00 TEXT /If more space is required, use additional copies of NRC Form 366A) 17)
I. PLANT CONDITION(S)
On March 10, 2003, at approximately 0012 Eastem Standard Time, Unit 1 was in Mode 1, steady state operation at 100 percent power. The Reactor Coolant System (RCS) (Energy Industry Identification System (EIIS) Code AB) pressure was 2235 psig and RCS Tavg was 588 degrees F.
II. DESCRIPTION OF EVENT
A. Event At 0012 on 3/10/03, the Main Generator Breaker (EIIS Code BKR) opened unexpectedly, which initiated an automatic turbine/reactor (EIIS Code RCT/TRB) trip. Plant safety equipment functioned as designed in response to the trip including auto-start of the auxiliary feedwater system (EIIS Code BA). Control room indication showed the event was initiated by generator backup relay 121 GB (EIIS Code RLY), an impedance type relay that uses both current and voltage potential to detect faults. This relay is connected to a current transformer (CT) (EIIS Code XCT) and a potential device (PD) (EIIS Code FD) on the high side bushing of each 500 kV main transformer (EIIS Code XFMR)
(3 - one per phase), and provides backup protection for the generator (EIIS Code GEN) and the transformers.
Meggar testing showed C phase of the PD was shorted to ground. Further troubleshooting found evidence of arcing on the C phase main transformer's high side bushing capacitance tap connector, damage to its o-ring seal, cracks in the external RTV sealant, and blackening of the petrolatum insulating grease inside the bushing well, all of which indicated this was the faulted component. This was confirmed following an evaluation of the operational characteristics of the 121 GB relay and other generator protective relays. The A phase potential connection was also found to have a damaged o-ring. An analysis is being performed on the o-rings to substantiate the failure mechanism.
B. Inoperable Structures, Components, or Systems that Contributed to the Event:
There were no inoperable systems that contributed to this event.
i I U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER 16)
PAGE (3) 05000 YEAR SEQUENTIAL REVISION NUMBER 3 of 7
Watts Bar Nuclear Plant 05000 - 390 2003 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
C. Dates and Approximate Times of Major Occurrences
Time Event 00:12:45.516 00:12:45.570 00:12:45.738 00:12:45.738 00:12:45.758 00:12:45.759 00:12:45.793 00:12:45.798 00:12:45.809 00:12:45.840 00:12:45:939 00:12:47.917 00:12 00:13:00.603 00:13:00.750 00:40 01:14 01:20 01:22 01:25 01:40 02:00 Generator Breaker open.
Turbine auto stop oil pressure low Turbine trip electrical trouble Generator 1 backup and transformer 1 feeder differential relay operation 500 kV PCB 5088 operation 500 kV PCB 5044 operation Turbine Trip-Auto Stop oil pressure low Generator 1 exciter power supply abnormal Reactor trip/Turbine trip Generator 1 voltage regulator trip Turbine Throttle Valves Closed Rods at Bottom Control room crew entered E-0 ( Reactor Trip / SI)
Lo Tavg with Reactor trip Main Feedwater (MFW) Isolation Turbine Trip-MFPT A and B tripped Transitioned from ES-0.1 to GO-5.
Started Standby MFW pump Established normal feedwater flow to Steam Generator Removed AFW from service Completed balance of Plant (BOP) realignment Main Turbine did not go on turning gear Restored Raw Cooling Water (RCW) to SBMFP
D. Other Systems or Secondary Functions Affected
There were no other systems affected other than equipment required for plant shutdown.
E. Method of Discovery
The operators were first alerted of the event by the annunciation of Window 75-C, Electrical Trouble.U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT TEXT CONTINUATION FACILITY NAME 11 DOCKET LER NUMBER (6l PAGE 31 05000 YEAR SEQUENTIAL REVISION NUMBER 4 of 7
Watts Bar Nuclear Plant 05000 - 390 2003 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) 171
F. Operator Actions
Operations crew performance for the Reactor/Turbine Trip on March 10, 2003, at 0012 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> was satisfactory.
However, some procedural challenges were encountered, but these issues did not adversely impact the crew's response to the transient or the plant's stabilization in Mode 3. These procedural issues have been entered into the corrective action program.
G. Safety System Responses:
Plant safety systems operated as designed which included the auto start of the auxiliary feedwater system.
Ill. CAUSE OF THE EVENT A Immediate Cause:
The initial indication was that the turbine trip was initiated by the generator backup relay 121 GB (impedance type relay that uses both current and voltage potential to detect faults). This relay is connected to a current transformer (CT) and a potential device (PD) on the high side of the 500KV main transformers and provides backup protection for the generator and the transformer. The operation of the 121 GB relay will cause the operation of auxiliary relay 186GB (auxiliary tripping device) after a time delay of 10 cycles. Operation of the 186GB relay results in opening of the main generator breakers, and closure of the steam supply to the turbine.
B. Root Cause:
The root cause of this event was inadequate preventive maintenance (PM) because of the failure to incorporate vendor recommendations into the applicable PM procedure.
The PM procedure was inadequate in that it contained improper guidance for filling the connector with the insulating compound and for tightening the connector cap. Over filling these connectors may produce high internal pressures, increasing the potential for the petrolatum to leak out.The vendor recommended insulating compound is heavy insulating oil. Following this event, it was learned that petrolatum becomes fluid above 54 degrees C (129.2 degrees F) and can absorb moisture. The use of RTV sealant, although not required per the PM, was an additional barrier incorporated from a 1997 event. (Refer to Section VI.B, Previous LERs on Similar Events.")U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER 16)
PAGE 3) 05000 YEAR I SEQUENTIAL I REVISION NUMBER 5
5 of 7
Watts Bar Nuclear Plant 05000 - 390 2003 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) 117)
IV. ANALYSIS OF THE EVENT
The plant responded as designed to the initiating condition. This event is compared to the LOSS OF EXTERNAL ELECTRICAL LOAD AND/OR TURBINE TRIP as described in Final Safety Analysis Report (FSAR) Section 15.2.7.
The complete loss of load/turbine trip from full power is examined to show the adequacy of the pressure relieving devices and also to demonstrate protection and the departure from nucleate boiling (DNBR). The plant trip on March 10, 2003 was less challenging than and bounded by the event described in the FSAR. The following plant conditions were bounded by the event described in the FSAR:
- 1. Reactor power was at 100% and less than the FSAR value of 102%.
- 2. The anticipatory reactor trip occurred on turbine trip versus the reactor protection system trip setpoints.
- 3. Reactor control was in automatic versus manual assumed in the FSAR.
- 4. Steam dumps operated as designed. The FSAR design basis does not credit the operation of the steam dump system or steam generator power operated relief valves (SG-PORVs).
The reactor trip occurred as designed from a turbine trip and since station power was not lost during the event, the plant response remained within the FSAR boundary analysis. The main condenser steam dump valves opened per design and as a result it was not necessary for the SG-PORVs to operate. Pressurizer level and pressure did not increase to challenge the pressurizer PORVs and safeties to limit RCS pressure. RCS pressure and loop average temperatures decreased during the transient rather than increasing as predicted by the conservative FSAR assumptions and the DNBR was not challenged. The differences between the FSAR and the plant event are associated with the conservatism assumed in the FSAR analysis and the benign nature of the actual plant event which was quickly brought to a stable condition.
The UFSAR discussion on Turbine Trip has been recently impacted by Westinghouse letter WAT-D-1 1111 issued February 12, 2003. This letter transmits Nuclear Safety Advisory Letter (NSAL)-03-1 on "Safety Analysis Modeling Loss of LoadlTurbine Trip." The NSAL determined that the assumption of a lower reactor vessel coolant average temperature at full power could result in a 3 psig higher peak RCS pressure for this analysis with a peak pressure of 2655 psia. During this trip event, the peak RCS pressure remained at approximately 2235 psig and the RCS Tavg was approximately 588 degrees F prior to the trip. Therefore, the NSAL issue was not an issue for the event.
V. ASSESSMENT OF SAFETY CONSEQUENCES
Based on the discussion in Section IV above, there was no safety significance to this event.U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE 13) 05000 YEAR SEQUENTIAL REVISION NUMBER l
6 of 7
Watts Bar Nuclear Plant 05000 - 390 2003 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
VI. CORRECTIVE ACTIONS
A Immediate Corrective Actions:
The immediate corrective actions included repairing the connector (cleaning; o-ring replacement; filling with the proper insulating compound), replacing cables, and changing the position of an existing transfer switch to make the 121 GB relay voltage potential source a switchyard bus section versus the main transformer high side bushings. The latter action removed these potential device connectors from any protective relaying function.
B. Corrective Actions to Prevent Recurrence:
The following actions are tracked under TVA's Corrective action program and therefore not consider to be regulatory
commitments
WBN supervisory personnel are to be briefed on the importance of identifying the cause of an equipment failure versus just treating the symptom(s) to prevent an event from recurring.
Applicable PM procedures are being revised to incorporate vendor recommendations for maintaining a one-quarter inch expansion space when filling the tap well with either GE A13A1 B heavy oil or GE A13A3A1 transformer insulating oil, and to tighten the cap only snug tight.
GE Instruction (GEl) 79087D is being added to the applicable vendor manual.
A design change will be made to improve the reliability of Generator Backup relay 121GB by installing redundant trip logic (e.g., 2/2 or 2/3), or a loss of relay function to detect a blown fuse or loss of potential to block the possible incorrect tripping of the protective relays.
VIl. ADDITIONAL INFORMATION A Failed Components:
As previously discussed, the root cause of this event was inadequate preventive maintenance because of the failure to incorporate vendor recommendations into the applicable PM procedure, which was also not identified as the cause of the similar 1997 event. (Refer to Section VI.B, Previous LERs on Similar Events.') This resulted in a short to ground of C phase of the PD (GE Type KA-108). Troubleshooting found evidence of arcing on the C phase main transformer's high side bushing capacitance tap connector, damage to its o-ring seal, cracks in the external RTV sealant, and blackening of the petrolatum insulating grease inside the bushing well, all of which indicated this was the faulted component. This was confirmed following an evaluation of the operational characteristics of the 121GB relay and other generator protective relays. The A phase potential connection was also found to have a damaged o-ring.
An analysis is being performed on the o-rings to substantiate the failure mechanism.:1 i.U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT TEXT CONTINUATION FACILITY NAME (11)
DOCKET LER NUMBER 16)
PAGE (3) 05000 YEAR SEQUENTIAL REVISION NUMBER 7 of 7
Watts Bar Nuclear Plant 05000 - 390 2003 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) 1171 B. Previous LERs on Similar Events:
At 0125 Eastern Daylight Time (EDT) on April 20, 1997, the WBN Unit 1 main generator PCB opened initiating a turbine trip and a subsequent automatic reactor trip. Investigations found that the turbine trip was initiated by operation of the generator backup relay, relay 121GB. Operation of this relay results in the opening of the main generator PCB and closure of the steam supply to the turbine. Meggar testing was performed to establish a possible cause for the operation of the protective relay. For the A phase main transformer high side potential device (PD) circuit, a low meggar reading was found. Additional troubleshooting found evidence that the phase A capacitance tap connector from the 500KV bushing to the PD had experienced arcing. Moisture was also found in the bushing well and the lubricant around the bushing well was blackened indicating a fault on the A phase capacitance tap connector.
Once this problem was found, the capacitance tap bushing area was cleaned, and new lubricant was applied. In addition, the exposed surfaces of the connector were sealed with an RTV sealant. Resistance tests were subsequently performed with the readings found to be normal. Due to the problems found on the A phase, the B and C phase capacitance tap connectors were inspected even though the PD circuits meggar tested satisfactorily. No problems were found on the B phase. C phase was found to have lost a portion of its lubricant, and the remaining lubricant was blackened but no significant burn odor was noted. The same type of repairs performed on the A phase were made to both the B and the C phases. This event was documented as LER 390/97-010.
C. Additional Information
None.
D. Safety System Functional Failure Consideration:
This event is not considered a safety system functional failure in accordance with NEI 99-02 in that the principal plant safety systems operated as designed. Therefore, the functional capability of the overall system was not jeopardized.
E. Loss Of Normal Heat Removal Consideration:
This event is not considered a scram with loss of normal heat removal.
VIII. COMMITMENTS
None.