05000458/LER-2011-003, Regarding Automatic Reactor Scram Due to a Main Turbine Trip
| ML12059A096 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 02/21/2012 |
| From: | Olson E Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RBG-47210 LER 11-003-00 | |
| Download: ML12059A096 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(iv)(A), System Actuation 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 |
| 4582011003R00 - NRC Website | |
text
SEntergy Entergy Operations, Inc.
River Bend Station 5485 U. S. Highway 61 N St. Francisville, LA 70775 Tel 225 381 4374 Fax 225 381 4872 eolson~entergy.com Eric W. Olson Site Vice President RBG-4721 0 February 21, 2012 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
Licensee Event Report 50-458 / 2011-003-00 River Bend Station - Unit 1 Docket No. 50-458 License No. NPF-47 RBF1-12-0021
Dear Sir or Madam:
In accordance with 10 CFR 50.73, enclosed is the subject Licensee Event Report.
This document contains no commitments. If you have any questions, please contact Mr. Joseph Clark at 225-381-4177.
Sincerely, EWO/dhw Enclosure
Licensee Event Report 50-458 / 2011-003-00 February 21, 2012 RBG-4721 0 RBF1-12-0021 Page 2 of 2 cc:
U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Blvd.
Arlington, TX 76011-4511 NRC Sr. Resident Inspector P. O. Box 1050 St. Francisville, LA 70775 INPO Records Center E-Mail (MS Word format)
Ms. Tracie Lowery Public Utility Commission of Texas 1701 N. Congress Ave.
Austin, TX 78711-3326 Department of Environmental Quality Office of Environmental Compliance Radiological Emergency Planning and Response Section JiYoung Wiley P.O. Box 4312 Baton Rouge, LA 70821-4312
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/3,1/2013 (10-2010) digits/characters for each block) the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE River Bend Station - Unit 1 05000-458 1 OF 5
- 4. TITLE Automatic Reactor Scram Due to a Main Turbine Trip
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY
ýYEAR YEAR SEQUENTIAL REV MONTH DAY YEAR NUMBER NO.
05000 FACILITY NAME DOCKET NUMBER 12 23 2011 2011-003-00 02 21 2012 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
E] 20.2203(a)(3)(i)
E] 50.73(a)(2)(i)(C)
[] 50.73(a)(2)(vii) 1
[
20.2201(d)
[
20.2203(a)(3)(ii)
E] 50.73(a)(2)(ii)(A)
EZ 50.73(a)(2)(viii)(A)
[
20.2203(a)(1)
LI 20.2203(a)(4)
[] 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
Li 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
LI 50.73(a)(2)(iii)
F1 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
[
20.2203(a)(2)(ii)
[: 50.36(c)(1)(ii)(A)
[
50.73(a)(2)(iv)(A)
LI 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
[] 50.36(c)(2)
LI 50.73(a)(2)(v)(A)
LI 73.71(a)(4) 100 El 20.2203(a)(2)(iv)
[L 50.46(a)(3)(ii)
EL 50.73(a)(2)(v)(B)
LI 73.71 (a)(5)
LI 20.2203(a)(2)(v)
LI 50.73(a)(2)(i)(A)
LI 50.73(a)(2)(v)(C)
El 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code)
Joseph A. Clark, Manager - Licensing 225-381-4177CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX n/a
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION
[: YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
E NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On December 23, 2011, at approximately 6:10 a.m. CST, the main turbine tripped unexpectedly, resulting in a reactor scram. The plant was stable at 100 percent power at the time of the event, and no safety-related systems were out of service. Operators implemented the appropriate response procedures, and began to stabilize reactor vessel pressure and water level. The closure of the turbine control valves resulted in the actuation of at least fifteen of sixteen main steam safety relief valves. A subsequent high reactor water level caused a trip of all three reactor feedwater pumps. As reactor water level lowered back through the normal operating range, operators attempted to restart a feedwater pump, but component malfunctions were encountered on "B" and "C" pumps. The reactor core isolation cooling (RCIC) system was manually actuated approximately nine minutes after the scram and injected water into the reactor for approximately two minutes.
The "A" feedwater pump was restored to service approximately one minute after RCIC was initiated. The cause of the turbine trip was a spurious backup over-speed trip resulting from an electrical discharge from the turbine shaft in the vicinity of the EHC turbine speed pickup probe. The cause of the electrical discharge was due to a failure of the shaft grounding system. The plant responded as designed, and no emergency core cooling system actuation setpoints were exceeded. This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as a condition that resulted in the automatic actuation of the reactor protection system (RPS).
ILICENSEE EVENT REPORT (LER)
U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR I SEQUENTIAL REV.
INUMBER NO.
River Bend Station - Unit 1 05000 -458 2 OF 5 2011 -- 003 -- 00 REPORTED CONDITION On December 23, 2011, at approximately 6:10 a.m. CST, the main turbine (**TRB**) tripped unexpectedly, resulting in a reactor scram. The plant was stable at 100 percent power at the time of the event, and no safety-related systems were out of service. Operators implemented the appropriate response procedures, and began to stabilize reactor vessel pressure and water level.
The rapid closure of the turbine control valves caused a rise in reactor pressure that actuated at least fifteen of sixteen main steam safety relief valves (SRVs). The initial shrink in reactor water level accompanying the reactor scram caused a Level 3 alarm, with water level reaching a low point of -0.1 inches approximately 15 seconds after the scram (Level 3 is 9.7 inches). The main feedwater control system responded, and the subsequent increase in reactor water level caused a Level 8 trip of all three reactor feedwater pumps (**PMP**)
approximately three minutes into the event.
As reactor water level lowered back through the normal operating range, operators attempted to restart a feedwater pump, but component malfunctions were encountered on "B" and "C" pumps. The reactor core isolation cooling (RCIC) (BN) system was manually actuated approximately nine minutes after the scram and injected water into the reactor for approximately two minutes. The "A" feedwater pump was restored to service approximately one minute after RCIC was initiated.
This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as a condition that resulted in the automatic actuation of the reactor protection system (RPS) (JC).
In the immediate notification performed in accordance with 10 CFR 50.72, it was reported that the reactor vessel Level 3 condition caused the actuation of primary containment isolation valves in the suppression pool cooling system. Those valves were already closed at the time of this event due to the system being out of service.
INVESTIGATION and CAUSAL ANALYSIS
- 1. Main turbine trip The trip signal to the main turbine originated in the electro-hydraulic control (EHC) system, and caused a fast closure of all turbine control valves. Analysis of transient data found that both the primary and back-up turbine speed signals became erratic and indicated an overspeed condition. Within approximately one second, all four turbine control valves were commanded to go fully closed. The RPS system responded to the fast closure signals from the turbine control valves, initiating a reactor scram.
ILICENSEE EVENT REPORT (LER)
U.s. NUCLEAR REGULATORY COMMISSION (10-2010)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REV.
NUMBER NO.
River Bend Station - Unit 1 05000 -458 3 OF 5 2011 -- 003 -- 00 A root cause team was formed following the turbine trip. This investigation was performed using internal and external technical experts in EHC controls, magnetization, and electrostatic discharge.
The cause of the turbine trip was a spurious backup over-speed trip. An electrical discharge from the turbine shaft to the vicinity of the EHC turbine speed pickup probes generated electromagnetic pulses that influenced the speed probes, which act on magnetic flux associated with a toothed wheel coupled to the turbine shaft. Over-speed signals in at least 2 out of 3 channels caused a turbine trip signal. The cause of the electrical discharge was due to a failure of the shaft grounding system.
The turbine shaft grounding system was modified in 2004 to add a new brush at the mid-standard location (between the high-pressure turbine inboard bearing and the thrust bearing). There are four brushes assigned to the turbine shaft, three of which provide ground protection. The mid-standard brush was removed as part of troubleshooting the turbine trip and it was found to have very little wear for the time in service. An inspection conducted by both internal and external technicians concluded that the brush was not providing protection, given the level of wear observed. The brush is designed to pivot with bristle wear to maintain shaft contact. Since less wear was observed than expected, it is concluded the brush wore until the maximum range of the brush pivot was achieved, after which the brush lost contact with the shaft due to making hard contact within the indicator housing.
This investigation also found that the mounting bracket for that brush was improperly fabricated, such that the angle between the brush head and the shaft was not correct.
Following this forced outage, the as-left reading on the brush wear indicator is about halfway between "replace" and "new." This allows adequate brush movement and shaft contact.
Actions will be taken in the next refueling outage to correct the angle on this bracket to make it read accurately.
There were significant contributing factors in this event:
The grounding brush at turbine bearing no. 2 was installed as part of a modification to add a new grounding point. At that time, the preventative maintenance (PM) task for measurement of shaft voltage should have been revised to include shaft voltage measurements from either the new grounding brush or the shaft voltage monitoring.
The PM was not revised. Increased shaft voltage would indicate that the shaft grounding brush was not working properly.LICENSEE EVENT REPORT (LER)
U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REV.
NUMBER NO.
River Bend Station - Unit 1 05000 -458 4 OF 5 2011 -- 003-- 00 The post-modification testing following installation of the new brush in 2004 was not performed properly. An improperly fabricated mounting bracket built for that modification apparently does not allow the wear indicator to accurately show the "new" indication for a new brush head.
Maintenance and Engineering personnel recognized that the wear indicator did not accurately measure actual brush wear, but did not document the deficiency in the corrective action program.
- 2. Reactor feedwater system malfunctions The malfunctions that impeded the recovery of the reactor feedwater system are described as follows.
The operator first attempted to start the "C" feedwater pump. Part of the start sequence is the opening of the minimum flow valve, initiated by depressing the pump "start" button.
When the pump start sequence was initiated, the operator observed the indication for the minimum flow valve start to travel from closed position to an intermediate position, instead of going fully open. The pump start circuitry is electrically interlocked with that valve position indication, so this failure prevented the pump from starting. Troubleshooting found that the valve was actually opening as commanded, but the limit switches were out of adjustment, preventing the fully open indication from being applied to the start circuitry.
The operator then attempted to start the "B" feedwater pump. One of the actions involved in starting a feedwater pump is to verify that the auxiliary lubricating oil pumps on the pump and the gearbox are operating correctly. Upon initiation of the pump start, the operator observed that the gearbox auxiliary oil pump was cycling "on" then "off," which is abnormal.
Troubleshooting found that a pressure regulator in that lube oil system was out of adjustment. This issue was corrected prior to plant restart.
PREVIOUS OCCURRENCE EVALUATION River Bend Station submitted Licensee Event Report 050-458/08-002-00 on May 2, 2008, following an automatic reactor scram that resulted from a malfunction in the main turbine EHC system. The root cause of that event was determined to be a loose electrical termination plug in the turbine speed sensing circuit. That failure mode has been eliminated in the investigation of the December 2011 event.LICENSEE EVENT REPORT (LER)
U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REV.
NUMBER NO.
River Bend Station - Unit 1 05000 -458 5 OF 5 2011 -- 003-00 CORRECTIVE ACTION TO PREVENT RECURRENCE The following actions are planned, and will be documented in the corrective action program:
the wear indicator on the replaced brush will be corrected a monitoring plan has been established to monitor and trend speed signals, shaft voltages and power supply voltages outside technical expertise will be applied to independently review possible causes of shaft voltage buildup and develop corrective actions for the causes. In addition, further connection verifications of casing and turbine components to building grounds will be conducted
SAFETY SIGNIFICANCE
The reactor protection system and main steam safety relief valves responded as designed to the main turbine trip. Reactor pressure was controlled automatically by the brief SRV actuation and the main turbine bypass system throughout the event, until a controlled cooldown was established. Reactor water level reached a low point of -0.1 inches (the Level 3 alarm setpoint is 9.7 inches). The "A" feedwater pump was restored to service approximately nine minutes after the scram.
This event was of minimal safety significance to the health and safety of the public. The RPS actuation was successful, and no emergency core cooling systems actuation setpoints were exceeded.
(NOTE: Energy Industry Component Identification codes are annotated as (**XX**).)