05000244/LER-2009-002, Regarding Plant Trip Due to Loss of Electro-Hydraulic Control System Pressure
| ML100630416 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/26/2010 |
| From: | John Carlin Constellation Energy Nuclear Group, Ginna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 09-002-00 | |
| Download: ML100630416 (8) | |
| 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 |
| 2442009002R00 - NRC Website | |
text
John Carlin Site Vice President R.E. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, New York 14519-9364 585.771.5200 585.771.3943 Fax John.Carlin@cengllc.com CENG
'a joint vehture of Energy-U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 February 26th, 2010 ATTENTION:
SUBJECT:
Document Control Desk R.E. Ginna Nuclear Power Plant Docket No. 50-244 LER 2009-002, Plant trip due to loss of Electro-Hydraulic Control system pressure The attached Licensee Event Report (LER) 2009-002 is submitted under the provisions of NUREG 1022, Event Reporting Guidelines. There are no new commitments contained in this submittal. Should you have any questions regarding the information in this letter, please contact Mr. Thomas Harding at (585) 771-5219.
Attachments: (1) LER 2009-002 cc:
S.J. Collins, NRC D.V. Pickett, NRC Resident Inspector, NRC (Ginna) 1k)PLIQ RC L- ??(0
ATTACHMENT 1 LER 2009-002
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 (9-2007)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. FACILITY NAME R.E. Ginna Nuclear Power Plant
- 2. DOCKET NUMBER
- 13. PAGE 05000 244 1 OF 6
- 4. TITLE Plant Trip Due to Loss of Electro-Hydraulic Control System Pressure
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FMONTH DAY YEAR FI NAME DOCKET NUMBER MONTH DAY YEAR Y
SEQUENTIAL REV O 05000 YER NUMBER NO.
MNH DY YA FACILITY NAME DOCKET NUMBER 12 30 2009 2009
- - 002 -
0 02 26 2010 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
Ei III2701 (b)
[I 2(
)
0.2203(a)(3)(i) l 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) 10 20.2201(d)
E 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
[I 20.2203(a)(1)
El 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[I 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
[I 50.73(a)(2)(iii)
[I 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL E] 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
Z] 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
C3 20.2203(a)(2)(iii)
[]50.36(c)(2)
El 50.73(a)(2)(v)(A)
[:1 73.71 (a)(4)
El 20.2203(a)(2)(iv)
[]50.46(a)(3)(ii)
[]50.73(a)(2)(v)(B)
E] 73.71 (a)(5) 100%
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
[]50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
[I 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)
Thomas Harding, Licensing Director (585) 771-5219CAUSA COMONNT NU-REPORTABLE MANU-REPORTABLE FACTURER TO EPIX
CAUSE
COMPONENT FACTURER TO EPIX B
TG P
Y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION El YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE On December 30, 2009, the R.E. Ginna Nuclear Plant experienced an automatic plant trip from full power.
The trip was initiated upon a signal of two out of two turbine stop valves being fully closed. The two stop valves closed when the electro-hydraulic control system (EHC) could not maintain adequate EHC system pressure to keep the stop valves in the fully open position. The control room operators performed the appropriate actions of procedures E-0, Reactor Trip or Safety Injection and ES-0. 1, Reactor Trip Response. Following the reactor trip, all safety systems operated as designed. The reactor was stabilized in Mode 3.
The cause of the EHC pressure loss was a common mode failure of both EHC pumps caused by the incorrect seal material being used during vendor refurbishment of the pumps that were installed during the 2009 refueling outage (RFO). The specified material was Viton, however evaluation identified that Buna-N was provided which is incompatible with the EH fluid.
Corrective actions to prevent recurrence are summarized in Section IV.B.
NRC FORM 366 (9-2007)U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE SEQUENTIAL REV R.E. Ginna Nuclear Power Plant YEAR NUMBER NOQ..
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- 1.
DESCRIPTION OF EVENT
A.
PRE-EVENT PLANT CONDITIONS:
The reactor was in Operational Mode 1 at 100% power, 2235 psig and 574 degrees F.
Operations was responding to indications of high EH oil temperature.
B.
EVENT:
At approximately 0437 EST on December 30, 2009, the R.E. Ginna Nuclear Plant experienced an automatic reactor trip from full power. The trip was initiated upon a signal of two out of two turbine stop valves being fully closed. The two stop valves closed when the EHC system could not maintain adequate pressure to keep the stop valves in the fully open position. The reduced pressure permitted the stop valve discs to enter the steam flow, which accelerated the closure of the valves.
The function of the EHC high-pressure fluid control system is to provide a motive force which positions the turbine control valves (TCV) in response to electronic commands from the electronic controller, acting through the servo-actuators, and maintain the turbine stop valves (TSV) open. The motive force is provided by two redundant positive displacement vane pumps, which take suction from the fluid reservoir. Immediately downstream of the pumps, two pilot operated unloader valves cycle to regulate system pressure.
During the September 2009 RFO, both EHC pumps were replaced with refurbished pumps.
These pumps were refurbished by a vendor who installed the incorrect seal and O-ring materials. This was not identified by site personnel prior to installation.
Prior to the plant trip, operators were responding to high temperatures in the EHC fluid system and were taking actions to lower temperature. An auxiliary operator identified that the 'B' pump was continuously loading. Operations swapped to the 'A' pump and secured the 'B' pump.
Shortly after pumps were swapped, the auxiliary operator identified rapidly decreasing EH pressure. The trip occurred within a few minutes of this discovery.
Troubleshooting after the event revealed that the pumps were not performing to design, i.e. the pump flow rates were significantly lower during higher system pressure operation. The in-service pumps were removed and examined. The investigation revealed that the pump seals were significantly degraded. The seal material was analyzed by Ginna's Materials Lab and determined to be Buna-N (butadiene), which is not compatible with the EH fluid, Houghto-Safe 1120. Material testing of Buna N in Houghto-Safe fluid has been conducted, and elevated temperatures result in more rapid degradation.U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE I
SEQUENTIAL I
REV R.E. Ginna Nuclear Power Plant YEAR NUMBER NO.
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C.
INOPERABLE STRUCTURES, COMPONENTS OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:
The control room alarm for EH fluid temperature was out of service due to a failed sensor. This equipment was out of service since October 2009 when the sensor was broken during RFO maintenance.
D.
DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES
09/2009 Both EH pumps replaced during RFO 12/30/2009 0415 Operations identifies high temperature condition 12/30/2009 0430 Operations swaps EH operation from "B" to "A" pump 12/30/2009 0437 Turbine Stop Valves Close - Reactor Trip - Entered Mode 3 01/05/2010 0815 Plant entered Mode 1 on restart E.
OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED
The hydraulic actuator for the 'B' turbine stop valve was determined to have failed during the abnormal closure, causing a leak of EHC fluid in the high pressure turbine enclosure. The actuator is not designed to withstand the forces of steam pressure on the valve disc without a coincident turbine trip signal and materials analysis showed this was an abrupt failure of the bolting. In addition, the secondary system transient caused a failure of a 3B feedwater heater relief valve piping connection and subsequent secondary side water hammer events occurred when condensate pumps were secured to stop this leak.
F.
METHOD OF DISCOVERY
Operations was in the process of correcting the high EH fluid temperature condition when the plant trip occurred.
G.
MAJOR OPERATOR ACTION:
Operations entered E-0 and ES-0.1 for a reactor trip and stabilized the plant in Mode 3.
H.
SAFETY SYSTEM RESPONSES:
The reactor protection system operated as expected as a result of the turbine stop valve closure. The Anticipated Transient Without Scram (ATWS) mitigation system actuated, starting the turbine driven auxiliary feedwater pump. The motor driven auxiliary feedwater pumps automatically started prior to this signal based on low-low steam generator level. All systems operated as expected.U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE SEQUENTIAL REV R.E. Ginna Nuclear Power Plant YEAR NUMBER NO.
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- 11.
CAUSE OF EVENT
The events are NUREG-1022 Cause Code B, Design, Manufacturing, Construction/Installation.
This event was entered into the site corrective action program (CR-2010-000084). The turbine stop valves closed due to inadequate EH fluid pressure to keep them open. The EH pumps failed to maintain proper pressure due to degradation of the pump seal caused by chemical attack to the seal material by the EH fluid. The wrong material was used in the seals and O-rings during an offsite vendor refurbishment of the pumps, which were placed in service approximately three months earlier during the 2009 RFO. Since the internal seal materials were not readily accessible, the incorrect materials were not recognized during pump installation. The Preventive Maintenance program allowed for refurbishment as described in the vendor manual instead of replacement as was recommended by Westinghouse who supplied the equipment. In addition, both pumps were scheduled for replacement at the same time, which allowed for a common cause failure to occur.
III
ANALYSIS OF THE EVENT
This analysis is reportable in accordance with 10 CFR50.73, Licensee Events Report System under item (a)(2)(iv) based on the actuation of the Reactor Protection System and Auxiliary Feedwater System.
An assessment was performed considering both the safety consequences and implications of this event with the following conclusions:
Reactor trip breakers opened as required and control rods inserted as designed. Heatup and pressurization of the Reactor Coolant System (RCS) presented no significant challenge to RCS pressure control systems and no Power Operated Relief Valve (PORV) or safety valve actuation occurred. Over-Temperature Delta-Temperature runback criteria were met on three out of four channels and the other channel called for a reactor trip. Maximum steam generator secondary side pressures were well below the atmospheric relief valve pressure setpoint. Automatic actuation of both motor driven auxiliary feedwater pumps occurred when the water level in steam generator 'B' lowered below the low-low level setpoint. The turbine driven auxiliary feedwater pump automatically started due to the ATWS mitigation system on a low feedwater flow signal. All auxiliary feedwater pumps performed as expected and met required flow rates.
The plant transient response is bounded by the Loss of External Electrical Load transient analyzed as part of the licensing basis described in the UFSAR.
A water hammer event occurred on the secondary plant as a result of the condensate pumps being secured in response to a condensate leak from the failed 3B feedwater heater relief valve piping connection. In addition, the 'B' turbine stop valve actuator failed causing an EH leak. While these issues complicated restart of the plant and presented an industrial safety concern, they did not have a significant effect on the plant response to the trip.
VU.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE R.E. Ginna Nuclear Power Plant YEAR NUMBER NO.
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Based on the above considerations, the nuclear safety consequences of this event are very low.
This event impacted NRC performance indicator IE01, Unplanned Scrams per 7000 Critical Hours. This value changed from 0 to 0.8 in the month of December.
IV
CORRECTIVE ACTIONS
A.
ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
Both EH pumps were replaced with new pumps (verified seal materials). The condensate heater relief valve line was repaired and the turbine stop valve actuator cylinder was replaced.
B.
ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE Note: There are no regulatory commitments in this Licensee Event Report Initial Action: A modification was performed to provide a more conservative control room alarm setting for low EH pressure and a repair was performed to the EH temperature sensor.
Long Term Corrective Actions: Updates to the preventive maintenance program are planned such that the pumps will be replaced vs. refurbished, procurement requirements will be augmented, and redundant pump replacement will be staggered in time to minimize common cause failures. A modification is planned to add EHC pressure and temperature indications and alarms to the Plant Process Computer System (PPCS).
Additional corrective and preventive actions are identified in the site Root Cause Analysis Report which addresses contributing organizational and human performance issues.
V.
ADDITIONAL INFORMATION
A.
FAILED COMPONENT EH Pumps 'A' and "B" failed to maintain proper EH pressure due to seal degradation.
These are Denison series T6C vane pumps B.
PREVIOUS LERS ON SIMILAR EVENTS A review of Ginna events over the past five years identified the following similar event:
LER 2007-001 Loss of Electrical Generation Results in Plant Trip
UU.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE I SEQUENTIAL REV R.E. Ginna Nuclear Power Plant YEAR NUMBER NO.
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C.
THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION IDENTIFIER AND SYSTEM NAME OF EACH COMPONENT OR SYSTEM REFERRED TO IN THIS LER:
IEEE 803 FUNCTION IDENTIFIER IEEE 805 SYSTEM INDENTIFICATION COMPONENT PEH01 PEH02 P
P TG TG D.
SPECIAL COMMENTS None