05000327/LER-2010-003, Regarding Manual Reactor Trip as a Result of a Fire in the Main Generator Neutral Bushing Bus Duct Housing
| ML110550159 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/17/2011 |
| From: | Skaggs M Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 10-003-00 | |
| Download: ML110550159 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3272010003R00 - NRC Website | |
text
Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy, Tennessee 37384-2000 February 17, 2011 10 CFR 50.73 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Unit 1 Facility Operating License Nos. DPR-77 NRC Docket Nos. 50-327
Subject:
License Event Report 327/2010-003, "Manual Reactor Trip as a Result of a Fire in the Main Generator Neutral Bus Duct Housing" The enclosed licensee event report provides details concerning a manual reactor trip and automatic engineered safety feature actuation of auxiliary feedwater as the result of a fire located on the main electrical generator.
This report is being submitted in accordance with 10 CFR 50.73 (a)(2)(iv)(A), as an event that resulted in a valid actuation of the reactor protection system.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact G. M. Cook, SQN Site Licensing Manager, at (423) 843-7170.
Respectfully, Michael D.Skaggs Site Vice President Sequoyah Nuclear Plant
Enclosure:
Licensee Event Report 327/2010-Enclosure cc: See page 2 C-9 -_
U.S. Nuclear Regulatory Commission Page 2 February 17, 2011 Enclosure cc (Enclosure):
Regional Administrator - Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/20 (10-2010)
, the NRC may sfor each block) not conduct or sponsor, and a person is not required to respond to, the digits/characters finformation collection.
.l 3. PAGE Sequoyah Nuclear Plant (SQN), Unit 1 05000327 1 OF 5
- 4. TITLE: Manual Reactor Trip as a Result of a Fire in the Main Generator Neutral Bushing Bus Duct Housing
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
RE MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 12 20 2010 2010 -
003 -
00 02 17 2011
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 EL 20.2201(b)
LI 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
LI 50.73(a)(2)(vii)
LI 20.2201(d)
LI 20.2203(a)(3)(ii)
LI 50.73(a)(2)(ii)(A)
LI 50.73(a)(2)(viii)(A)
,I 20.2203(a)(1)
[1 20.2203(a)(4)
LI 50.73(a)(2)(ii)(B)
[I 50.73(a)(2)(viii)(B)
E:1 20.2203(a)(2)(i)
[1 50.36(c)(1)(i)(A)
Ej 50.73(a)(2)(iii)
E] 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL LI 20.2203(a)(2)(ii)
LI 50.36(c)(1)(ii)(A) 50.73(a)(2)(iv)(A)
LI 50.73(a)(2)(x) 100 LI 20.2203(a)(2)(iii)
LI 50.36(c)(2)
LI 50.73(a)(2)(v)(A)
L] 73.71(a)(4)
[j 20.2203(a)(2)(iv)
LI 50.46(a)(3)(ii)
Li 50.73(a)(2)(v)(B)
LI 73.71(a)(5)
[l 20.2203(a)(2)(v)
LI 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
Ej OTHER 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify in Abstract below or in
D. Other Systems or Secondary Functions Affected
Following the reactor trip, the condensate and feedwater system [EIIS Code SN]
experienced a feedwater heater drain tank pump [EIIS Code P] trip and feedwater heater [EIIS Code HX] string isolation. The Number 7 Heater Drain Tank Pump 1B Motor tripped because of an apparent low level in the Number 7 Heater Drain Tank.
Subsequently, the low pressure "C" feedwater heater string (i.e., stages 5, 6, and 7) isolated. Maintenance was performed on this equipment to determine and correct identified issues. No equipment issues were noted. It is suspected based on operating experience of this equipment that the drain tank did experience a low level condition resulting in the pump trip with subsequent heater string isolation.
E.
Method of Discovery
Site personnel reported smoke in and around the SQN, Unit 1 main generator of the turbine building. Operations personnel located the source of the smoke and initiated an emergency alarm.
F.
Operator Actions
Based on the valid information of a fire located in the neutral bus duct housing of the main generator, the SQN, Unit 1 Unit Supervisor in accordance with an operating procedure directed a manual reactor trip.
G.
Safety System Responses:
The safety systems performed as designed for the reactor trip. Auxiliary feedwater (AFW) system [EIIS Code BA] automatically initiated following the reactor trip. Operators reduced the flow of AFW following the reactor trip to mitigate a decrease in reactor coolant system (RCS) [EIIS Code AB] average temperature. All safety systems remained within Technical Specifications (TS) and Updated Final Safety Analysis Report (UFSAR) limits.
Ill.
CAUSE OF THE EVENT
A. Immediate Cause:
The immediate cause of the unit trip was the initiation of a manual reactor trip signal.
B. Root Cause:
The root cause was identified as a cyclic vibration induced failure of the 'A' phase neutral bushing porcelain, that allowed a hydrogen leak that ignited.
C.
Contributing Factor:
The SQN, Unit 1 generator ran from 2004 to the fall of 2010 with a rotor that showed evidence of a shorted turn in the rotor. The shorted turn in the rotor increased vibration, which could have been a contributor to the degradation of the bushing porcelain. This rotor with suspected shorted turn was replaced during the fall 2010 refueling outage, which significantly reduced the vibration levels.
IV.
ANALYSIS OF THE EVENT
This event is most similar to and bounded by the analyzed event of loss of external electrical load and/or turbine trip, with the plant safety systems operating as designed during and following the manual reactor trip. Prior to the event, SQN, Unit 1 was operating in Mode 1 at approximately 100 percent power. The reactor coolant system (RCS) pressure was approximately 2235 pounds per square inch gauge (psig) with an average temperature near the program value of 578 degrees Fahrenheit (F) prior to the plant transient. Following the reactor trip, the loss of nuclear heat generation and the introduction of cold AFW resulted in a decrease in RCS average temperature to approximately 545 degrees F and RCS pressure declined to 2030 psig. Operators at approximately 5 minutes following the reactor trip, reduced the AFW flow to approximately 200 gallons per minute (gpm) to each steam generator (SG) [EIIS Code SG] to mitigate the RCS temperature and pressure decrease.
Steam pressure was nearly 867 psig before the reactor trip and increased to approximately 1023 psig when the turbine stop valves closed. Steam dumps to the condenser [EIIS Code COND] operated as expected and remained available. The introduction of AFW decreased steam pressure to approximately 990 psig within 30 minutes and once AFW flow was matched to decay heat generation, steam pressure returned to the no-load value of approximately 1005 psig. RCS temperature was restored to its no-load value of 547 degrees F as the secondary side pressure recovered. Emergency boration was not required based on the shutdown margin requirements. Additionally, AFW system actuated as designed on the SG low-low level signal, recovering the SG water levels following the reactor trip. The plant responded as expected for the conditions of the trip, with exception to the secondary plant equipment described above. No TS limits were exceeded and the UFSAR analysis of this event remained bounding.
V.
ASSESSMENT OF SAFETY CONSEQUENCES
Based on the above "Analysis of The Event," this event did not adversely affect the health and safety of plant personnel or the general public.
VI.
CORRECTIVE ACTIONS
A.
Immediate Corrective Actions
The immediate corrective actions included replacement of the failed neutral bushing.
,B. Corrective Actions to Prevent Recurrence:
A preventive maintenance instruction is being developed to define a service life for refurbishment/replacement of SQN, Unit 1 and Unit 2 main and neutral bushings.
Additionally, the SQN, Unit 2 bushings are being visually inspected during the next scheduled refueling outage.
VII.
ADDITIONAL INFORMATION
A.
Failed Components:
The failed component was a main generator bushing from Westinghouse Electrical Corporation, style number 5621 D28G01, insulation class 25kV.
B. Previous LERs on Similar Events:
A review of previous reportable events for the past three years did not identify any
previous similar events
C.
Additional Information
None.
D. Safety System Functional Failure:
This event did not result in a safety system functional failure in accordance with 10 CFR 50.73(a)(2)(v).
E.
Unplanned Scram with Complications:
This event did not result in an unplanned scram with complications.
VIII.
COMMITMENTS
None.