05000390/LER-2010-001, Regarding Reactor/Turbine Trip
| ML102090141 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 07/20/2010 |
| From: | Grissette D Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 10-001-00 | |
| Download: ML102090141 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), 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)(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 |
| 3902010001R00 - NRC Website | |
text
Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381 July 20, 2010 10 CFR 50.73 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 1 Facility Operating License No. NPF-90 NRC Docket No. 50-390
Subject:
Licensee Event Report 390/2010-001, Reactor/Turbine Trip This submittal provides Licensee Event Report (LER) 390/2010-001. This LER documents an instance where the reactor tripped automatically on a turbine trip signal above 50 percent rated thermal power. The condition is reported as an LER in accordance with 10 CFR50.73(a)(2)(iv).
There are no regulatory commitments in this letter. Please direct any questions concerning this matter to Robert Clark, WBN Site Senior Licensing Engineer, at (423) 365-1818.
Re pe t l,
D. E. Grissette Site Vice President Watts Bar Nuclear Plant Enclosure cc: See Page 2
U.S. Nuclear Regulatory Commission Page 2 Enclosure cc (Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
(See reverse for required number of digits/characters for each block)
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010
, the NRC may not conduct or sponsor, and a person is not required to respond to, he informa ion collection.
- 1. FACILITY NAME Watts Bar Nuclear Plant
- 2. DOCKET NUMBER 05000390
- 3. PAGE 1 OF 5
- 4. TITLE Reactor / Turbine Trip
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.
MONTH DAY YEAR FAC LITY NAME N/A DOCKET NUMBER N/A 05 21 2010 2010 - 001 -
0 07 20 2010 FAC LITY NAME N/A DOCKET NUMBER N/A
- 9. OPERATING MODE 1
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii) 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 20.2203(a)(2)(ii) 50.36(c)(1)(ii)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71(a)(4) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(B) 73.71(a)(5) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)
OTHER 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 10. POWER LEVEL 100%
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME Robert Clark, Senior Licensing Engineer TELEPHONE NUMBER (Include Area Code)
(423) 365-1818 CAUSE SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX
CAUSE
SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX D
IT EC W
Y D
BA CNV F
Y
- 14. SUPPLEMENTAL REPORT EXPECTED YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
NO
- 15. EXPECTED SUBMISSION DATE MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On May 21, 2010 at 1936 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.36648e-4 months <br /> with Watts Bar Nuclear Plant, Unit 1 at 100 percent rated thermal power, the reactor tripped automatically in response to a turbine trip signal generated by closure of the Main Turbine Throttle/Stop Valves. Following reactor trip, main feedwater isolation occurred due to reactor trip coincident with low Reactor Coolant System TAVG, which then resulted in automatic startup of the Auxiliary Feedwater System.
The most probable cause was determined to be an intermittent failure of a circuit card in the Westinghouse Mark IV Analog Electro-Hydraulic (AEH) turbine control system that caused the servo control valves for all four Main Turbine Throttle Valves to close their respective valves. TVA replaced the suspect circuit cards in the AEH control system.
All safety systems performed their required functions in response to this event.
This event is reportable as an LER in accordance with 10 CFR 50.73(a)(2)(iv).
NRC FORM 366 (9-2007) (9-2007)
LICENSEE EVENT REPORT (LER)
U.S. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE Watts Bar Nuclear Plant 05000390 YEAR SEQUENTIAL NUMBER REV No.
2 OF 5 2010 001 0
I.
PLANT CONDITIONS
Plant in Mode 1 at 100% RTP.
II. DESCRIPTION OF EVENT
A. Event On May 21, 2010 at 1936 EDT hours with Watts Bar Nuclear Plant (WBN), Unit 1 at 100% rated thermal power the reactor tripped automatically in response to a turbine trip signal generated by closure of the Main Turbine Throttle/Stop Valves [EIIS1 Code TA]. According to the Sequence of Events record from the Integrated Computer System (ICS) [EIIS Code ID], the Solid State Protection System (SSPS) [EIIS Code JC]
generated an alarm at 19:36:47 indicating that Main Turbine Throttle/Stop Valves 1 and 3 had closed. Two seconds later at 19:36:49 the SSPS generated a second alarm indicating that Main Turbine Throttle/Stop Valves 2 and 4 had closed. With the reactor operating above interlock setpoint P9 and all Main Turbine Throttle/Stop Valves closed, a Reactor Trip signal was generated by the SSPS. Immediately following reactor trip, the Steam Dump Control System [EIIS Code J] automatically opened the condenser bypass valves [EIIS Code SB] to remove stored energy and residual heat from the Reactor Coolant System (RCS) [EIIS Code AB]. During plant cooldown to no-load equilibrium conditions, a main feedwater isolation signal was generated by the SSPS due to Reactor Trip coincident with low RCS TAVG. The purpose of this signal is to prevent a potential RCS overcooling transient following a turbine trip. The closure of the main feedwater isolation valves [EIIS Code SJ] generated a Main Feed Pump Turbine (MFPT) [EIIS Code JK] trip signal which then resulted in the automatic startup of the Auxiliary Feedwater (AFW) System [EIIS Code BA]. All systems performed their required safety functions in response to this event.
This event was documented in TVAs Corrective Action Program as Problem Evaluation Report (PER) 231080.
B. Inoperable Structures, Components, or Systems that Contributed to the Event.
The most probable cause was determined to be an intermittent failure of a circuit card in the Westinghouse Mark IV Analog Electro-Hydraulic (AEH) turbine control system [EIIS Code IT] that caused the servo control valves for all four Main Turbine Throttle/Stop Valves to close their respective valves.
C. Dates and Approximate Times of Major Occurrences
Date Time (EDT)
Event May 21, 2010 19:36:47 SSPS Alarm - Main Turbine Throttle/Stop Valves 1 & 3 closed May 21, 2010 19:36:49 SSPS Alarm - Main Turbine Throttle/Stop Valves 2 & 4 closed May 21, 2010 19:36:49 SSPS Alarm - Reactor Trip From Turbine Trip Above P9 May 21, 2010 19:37:07 SSPS Alarm - Main Feedwater Isolation May 21, 2010 19:37:07 AFW Auto Start 1 Energy Industry Identification System (9-2007)
LICENSEE EVENT REPORT (LER)
U.S. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE Watts Bar Nuclear Plant 05000390 YEAR SEQUENTIAL NUMBER REV No.
3 OF 5 2010 001 0
II. DESCRIPTION OF EVENT (continued):
D. Other Systems or Secondary Functions Affected
Discharge Valve 1-PCV-3-132 located downstream of Motor Driven Pump (MDP) 1B-B in the AFW system failed to open following pump start-up. The PCV is a safety-grade air-operated valve designed to create sufficient backpressure to prevent pump run-out whenever Steam Generator (SG) pressure is low (i.e.,
during plant cooldown or in the event of a faulted SG). PER 231297 and Work Order 08-812356-01 were issued to investigate this problem. The most likely cause was determined to be an intermittent failure of the I/P converter in the control circuit for this valve. Corrective action was to replace and calibrate the new I/P converter (1-PM-3-0132) associated with 1-PCV-3-132. The I/P converter is designed to throttle the PCV closed until the pump differential pressure corresponds to an acceptable operating point on the pump curve.
E. Method of Discovery
Self-revealing. The reactor tripped automatically in response to a turbine trip signal generated by closure of the Main Turbine Throttle/Stop Valves.
F. Operator Actions
Following Reactor/Turbine trip, the operators entered the following procedures to stabilize the plant at Hot Standby Conditions:
E Reactor Trip or Safety Injection ES 0.1 - Reactor Trip Response GO Unit Shutdown from 30% Power to Hot Standby AOI Turbine Trip G. Safety System Responses The Reactor Protection System actuated a reactor trip in response to a turbine trip signal generated by closure of the Main Turbine Throttle/Stop Valves. Main Feedwater Isolation was initiated in response to a Reactor Trip coincident with low Reactor Coolant System TAVG. The AFW System automatically started on trip of all MFPTs due to Main Feedwater isolation. No safety injection signals were initiated. All safety systems responded as required, although failure of the discharge valve for the 1B-B AFW pump reduced margin.
III. CAUSE OF EVENT
The most probable cause was determined to be an intermittent failure of a circuit card in the Westinghouse Mark IV Analog Electro-Hydraulic (AEH) turbine control system that caused the servo control valves for all four Main Turbine Throttle Valves to close their respective valves. This conclusion is based on a thorough Root Cause investigation and a series of tests (i.e., Kepner-Tregoe Analysis).
Manual and automatic turbine trips were considered and discarded as a potential cause because there was no evidence in the sequence of event recordings or the operator logs/interviews that either had occurred.
Examination of the Sequence of Events logs identified that Throttle valves 2 and 4 closed two seconds after Throttle valves 1 and 3, which is atypical. A normal trip results in the four Throttle valves all closing within a fraction of a second of each other. This indicated that this was not a normal Turbine Trip; but that the servo control valves (also called Moog valves) could have been inadvertently commanded by the turbine control system, a Westinghouse Mark IV Analog Electro-Hydraulic (AEH) system, to close the Throttle valves.
Turbine test trips confirmed that a Manual or Automatic Turbine Trip did not exhibit the two second time difference. Therefore, a malfunction of the AEH main turbine control system was considered. The test trips did (9-2007)
LICENSEE EVENT REPORT (LER)
U.S. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE Watts Bar Nuclear Plant 05000390 YEAR SEQUENTIAL NUMBER REV No.
4 OF 5 2010 001 0 not identify any failed components, which led to the conclusion that the malfunction was intermittent in nature.
External factors that could cause intermittent malfunction, such as welding by Unit 2 construction and solar activity, were evaluated, and were determined not to be likely causes.
Based on its evaluation, TVA concluded that the most probable cause was that one of the circuit cards involved in the generation of the signal to the Throttle valve Moog valves experienced an intermittent failure. Those circuit cards are the Up/Down Counters, Input Expanders, HTL Gates, Mixing Amplifiers, and Analog Switches. The corrective action for that most probable cause was to replace each of those cards that were in the signal path that was in use at the time of the trip.
Consistent operation during post-maintenance testing supports the identified cause.
IV. ANALYSIS OF THE EVENT
The reactor protection system functioned properly in response to closure of the Main Turbine Throttle/Stop Valves. The reactor trip system initiated a turbine trip signal in response to the reactor trip, as designed. The steam dump and the pressurizer control systems functioned properly thereby minimizing RCS and SG [EIIS Code SB] temperature and pressure transients. No pressurizer [EIIS Code AB] or SG Power Operated Relief Valves (PORVs) or Safety Valves lifted during the transient. Main Feedwater isolation occurred as designed following plant cool down to prevent a potential RCS overcooling transient. The anticipatory AFW system auto-start on trip of all MFPTs following main feedwater isolation functioned as designed.
V. ASSESSMENT OF SAFETY CONSEQUENCES
The safety significance of this event (Reactor/Turbine Trip) is low because there was no loss of safety function.
This event is categorized in the WBN Updated Final Safety Analysis Report (UFSAR) Chapter 15 Accident Analyses as a Condition II event, a fault of moderate frequency. The event is specifically discussed in UFSAR subsection 15.2.7, Loss of External Electrical Load and/or Turbine Trip, which also encompasses loss of main feedwater flow with subsequent startup of the AFW System.
Although 1B-B AFW pump discharge valve, 1-PCV-3-132, failed to open during this event, the minimum AFW flow requirement of greater than 410 gpm specified in Emergency Operating Instruction, E-0, Reactor Trip or Safety Injection was satisfied. This flow requirement was initially provided by the 1A-A motor driven pump and the available turbine driven pump.
VI. CORRECTIVE ACTIONS
This event was documented within TVAs Corrective Action Program as PER 231080.
The existing AEH circuit cards that were in the signal path that was in use at the time of the trip were replaced and post-maintenance testing was performed to confirm functionality of the electro-hydraulic control system.
WBN maintained the plant below P-9 for an extended period (48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) to provide burn-in time for the new cards to preclude a similar trip due to infant mortality of the new cards.
Failure of the AFW Pump 1B-B discharge valve was documented within TVAs Corrective Action Program as PER 231297. Based on the apparent cause identified in that PER, TVA installed a new I/P converter for 1-PCV-3-132. (9-2007)
LICENSEE EVENT REPORT (LER)
U.S. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE Watts Bar Nuclear Plant 05000390 YEAR SEQUENTIAL NUMBER REV No.
5 OF 5 2010 001 0
VII. ADDITIONAL INFORMATION
A. Failed Components
- 1. Logic card(s) located within Westinghouse Mark IV AEH control system
B. Previous Similar Events
A search of LERs and PERs documenting turbine trips at Watts Bar Unit 1 found no previous failures similar to that which occurred on 5/21/10. Although some circuit card failures within the Analog Electro-Hydraulic (AEH) control system were found, they were not similar in nature nor did they have the same effect on plant operations.
C. Additional Information
None.
D. Safety System Functional Failure This event did not involve a safety system functional failure as defined in NEI 99-02, Revision 5.
E. Loss of Normal Heat Removal Consideration Main Feedwater isolation following reactor/turbine trip is a design feature used to preclude potential RCS overcooling transients. Loss of normal heat removal during this event was expected and was properly mitigated by the automatic startup of the AFW System including operator actions to stabilize the plant at Hot Standby Conditions.
VIII. COMMITMENTS
None.