05000461/LER-2010-003-01, For Clinton Power Station, Unit 1, Regarding Unexpected Component Actuations Due to Self Test System Design Deficiencies
| ML110620101 | |
| Person / Time | |
|---|---|
| Site: | Clinton  (NPF-062) |
| Issue date: | 02/21/2011 |
| From: | Kearney F Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| U-604005 LER 10-003-01 | |
| Download: ML110620101 (6) | |
| 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 10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation |
| 4612010003R01 - NRC Website | |
text
Exelon.
Nuclear Clinton Power Station 8401 Power Road Clinton, IL 61727 U-604005 SRRS 5A.108 February 21, 2011 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461
Subject:
Licensee Event Report 2010-003-01 Enclosed is Licensee Event Report (LER) No. 2010-003-01: Unexpected Component Actuations Due To Self Test System Design Deficiencies. This report is being submitted in accordance with 10 CFR 50.73.
The enclosed report has been revised to correct an editorial error.
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this report, please contact Mr. A. Khanifar at (217) 937-3800.
Respectfully, F.. Kear jt/_
Site Vice President Clinton Power Station RSF/awh
Enclosures:
Licensee Event Report 2010-003-01 cc:
Regional Administrator - NRC Region III NRC Senior Resident Inspector - Clinton Power Station Office of Nuclear Facility Safety - IEMA Division of Nuclear Safety
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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (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.
- 3. PAGE Clinton Power Station, Unit 1 05000461 1 OF 5
- 4. TITLE Unexpected Component Actuations Due To Self Test System Design Deficiencies
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER SEQUENTIAL REV MONTH DAY YEAR 05000 MONTH DAY YEAR YEAR NUMBER NO. I FACILITY NAME DOCKET NUMBER 03 15 2010 2010 003 -
01 02 21 2011 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
El 20.2203(a)(3)(i) 0l 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
El 20.2201(d)
El 20.2203(a)(3)(ii)
[I 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 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)
El 50.73(a)(2)(iii)
[I 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
[I 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
[E 50.36(c)(2)
[E 50.73(a)(2)(v)(A)
El 73.71(a)(4) 96.9
[ 20.2203(a)(2)(iv)
C3 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
[E 73.71(a)(5)
El 20.2203(a)(2)(v)
[I 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
[E OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in On 8/26/10 at 1223 hours0.0142 days <br />0.34 hours <br />0.00202 weeks <br />4.653515e-4 months <br />, the alarm for RCIC Pump Suction Pressure High, activated in the MCR. RCIC Pump suction pressure was observed to be at 68 pounds per square inch gage (psig) and steady, whereas normal pressure is 55 to 60 psig. After approximately 5 minutes the pressure lowered to 55 psig and the alarm reset. This change in pressure was caused by the RCIC minimum flow valve 1 E51 F01 9 cycling open and back to close.
On 8/26/2010 at 1225 hours0.0142 days <br />0.34 hours <br />0.00203 weeks <br />4.661125e-4 months <br />, Operations identified that RCIC pump suction pressure was high at the same time the Division 1 Direct Current (DC) Bus [BU] Voltage was spiking low. This anomaly was caused by the RCIC minimum flow valve 1 E51 F01 9 cycling open and back to close.
On 8/26/10 at 1550 hours0.0179 days <br />0.431 hours <br />0.00256 weeks <br />5.89775e-4 months <br />, Operators turned off the Division 1 STS power supply and no further unexpected valve operations occurred.
On 8/27/10 at 0251 hours0.00291 days <br />0.0697 hours <br />4.150132e-4 weeks <br />9.55055e-5 months <br />, while performing the Safety System Status Report, operators discovered that the Low Pressure Coolant Injection (LPCI) [BO] A valve 1E12F042A (a PCIV) was unexpectedly in the open position. A computer point shows that the 11E12F042A valve automatically opened on 8/26/10 at 1144 hours0.0132 days <br />0.318 hours <br />0.00189 weeks <br />4.35292e-4 months <br />.
The LPCI A injection valve is normally closed, and cannot be manually opened with Reactor Pressure Vessel pressure greater than 472 psig. The 1 El 2F042A automatically opens if reactor pressure is less than 472 psig with a LPCI initiation signal present; neither of these conditions was present. Operators restored 1 El 2F042A to its normal closed position at 0251 hours0.00291 days <br />0.0697 hours <br />4.150132e-4 weeks <br />9.55055e-5 months <br /> on 8/27/10.
Troubleshooting was performed to determine the cause of the unexpected valve operations. The troubleshooting consisted of replacing the load driver (LDDR) card for the 1 G33F039 valve, STS power supply voltage monitoring, inspections, data trending, and other methods to eliminate potential causes. When the Division 1 STS power supply was turned off no further unexpected valve operations occurred.
Troubleshooting identified degradation of the Division 1 STS power supply.
During the investigation of these unexpected component actuations an analysis of a spurious automatic isolation of Division 2 PCIVs and shunt trip that occurred on 3/15/10 identified that the 3/15/10 event could be related to the unexpected valve operations on 8/24/10, 8/25/10, and 8/26/10.
Clinton Power Station initiated Event Notification 45901 on 5/5/10 at 1404 Eastern Daylight Time under the provisions of 10 CFR 50.73(a)(1) for an invalid actuation on 3/15/10 that was reportable under the provisions of 10 CFR 50.73 (a)(2)(iv)(A). The notification satisfied the 60-day telephone notification to the NRC Operations Center in lieu of a written licensee event report.
The event notification reported that on 3/15/10 at 0138 hours0.0016 days <br />0.0383 hours <br />2.281746e-4 weeks <br />5.2509e-5 months <br />, with the plant in Mode 1 at about 96.9 percent reactor power, PCIVs automatically closed for isolation Groups 11 and 17, Division 2 Drywell Ventilation (VP)
[VB] and Drywell Cooling (WO) [KM] valves. The valves that automatically closed were inboard PCIVs 1VP005A, 1VP005B, 1VP014A, 1VP014B, 1WO001B, 1WO002B, 1WO551B and 1W0552B. In addition, the shunt trip devices for the breakers [BKR] of the following components tripped: Drywell cooling fan [FAN] 1 B (1 VP01 CB), Drywell cooling fan 1 D (1 VP01 CD), Drywell chiller [CHU] 1 B oil pump [P], DC Motor Control Center [MCC] 1 B ground detection, DC MCC 1 D ground detection, Drywell chiller 1 B (IVP04CB), and Drywell chiller control panel [PL] 1 B. The actuation of primary containment isolation Groups 11 and 17 was complete and the actuation was limited to these two containment isolation valve groups.
Event Notification 45901 reported that the cause of the isolation was a Division 2 load driver card spuriously actuated all of its loads without a valid loss of coolant accident (LOCA) signal or a manual initiation signal
being present. The isolation was reset at 0208 hours0.00241 days <br />0.0578 hours <br />3.439153e-4 weeks <br />7.9144e-5 months <br />, the shunt trips were reset at 0215 hours0.00249 days <br />0.0597 hours <br />3.554894e-4 weeks <br />8.18075e-5 months <br />, and Drywell Cooling was restored by 0259 hours0.003 days <br />0.0719 hours <br />4.282407e-4 weeks <br />9.85495e-5 months <br /> on March 15. The load driver circuit remained in service until a replacement load driver card could be installed on March 19, 2010. An apparent cause evaluation of this issue was unable to identify the cause of the event but concluded that the most probable cause was either an intermittent failure of the card or an intermittent short of the output connectors. Re-evaluation of this cause on 10/12/10 concludes that based on laboratory testing there was no failure of the Load Driver card. The Load Driver card tested satisfactory per original equipment manufacturer and was capable of performing its design function. The re-evaluation further concluded that based on the results of the unexpected valve operations for RT, RCIC, and RHR systems on 8/24/10, 8/25/10, and 8/26/10, the most likely cause of the VP/WO containment isolation valve automatic isolation is the degradation of the Division 2 STS power supply resulting in the misoperation of the NSPS Load Driver card.
The Division 2 STS power supply voltage was measured on 8/27/10 and found to have higher than expected ripples/noise. A decision was then made to turn off the Division 2 STS power supply and replace it with a temporary power supply.
The power supplies for the Divisions 1 and 2 STS have-been replaced with temporary power supplies.
This licensee event report constitutes a revision to Event Notification 45901.
The STS is designed to automatically test Nuclear System Protection System (NSPS) [JG] logic with a primary purpose to improve the availability of the Safety Systems Instrumentation by minimizing the time to detect and determine the failure location. The STS is designed to perform automatic testing of NSPS circuitry which includes logics associated with the Reactor Protection System (RPS) [JC], Emergency Core Cooling System (ECCS), and Containment and Reactor Vessel Isolation Control System [JM]. The STS injects short-duration test signal pulses in the NSPS Logic to verify proper response of the logic to various input combinations. When in the automatic mode, the STS injects test pulses in the logic at 60-minute test intervals. The STS normally operates to automatically test all four NSPS divisions in a continuous, cyclic manner. The STS ceases to operate in automatic mode if failure is detected in any one of the four logic divisions. The STS test sequence can be manually restored to automatic operation for the remaining three divisions. Once the failure is corrected, the STS can then be restored to fully automatic operation for all four divisions. The STS was designed to not influence the safety-related functions of the NSPS cards.
C. CAUSE OF EVENT
The apparent cause of the unexpected valve operations is two design deficiencies which do not allow the STS to meet its design specification. The first design deficiency is the STS 5 VDC power supply does not have low voltage protection to prevent a failure mode that results in the generation of electronic noise. The second design deficiency is the improper coordination timing between the STS coupling capacitor [CAP] and pulse stretcher circuits on the NSPS load driver cards which allowed electronic noise to cause unexpected valve operations. The effects of this combination of power supply low voltage and lack of coordination between the coupling capacitor and pulse stretcher have allowed signals of sufficient duration to reposition valves. When the Division 1 STS power supply was turned off, no further unexpected valve operations occurred.
D. SAFETY CONSEQUENCES
This event is reportable under the provisions of 10 CFR 50.73 (a)(2)(iv)(A) due to the invalid actuation of PCIVs for isolation Groups 11 and 17 on March 15, 2010. This licensee event report constitutes a revision to Event Notification 45901 completed on 5/5/10.
The unexpected closure of PCIVs for isolation Groups 11 and 17 during this event did not compromise their design safety functions which are to automatically close and isolate VP and WO containment penetrations.
By automatically closing to the designed safe position, the valves demonstrated their capability to perform their safety functions to close. Additionally, the closing of these valves did not affect any safety functions.
Actuations of RT system valve 1G33F039, RCIC system valve 1E51F019, and LPCI system valve 1 El 2F042A during this event did not result in a loss of safety function.
E. CORRECTIVE ACTIONS
The power supplies for all four STS Divisions have been replaced with temporary power supplies having low voltage protection.
STS power supplies will be permanently replaced to include low voltage protection as plant conditions allow.
Options for the STS issues will be developed and presented to the Plant Health Committee for approval.
F.
PREVIOUS OCCURRENCES
None
G. COMPONENT FAILURE DATA
Component Description:
5 VDC STC POWER SUPPLY 1C71AK615A POWER, SUPPLY, SWITCH MODE, 10A, 100-130 VAC, 4.5-5.5V, SINGLE Model:
RMT001 -AA-20995 Part Number:
169C8805P004 Manufacturer:
KEPCO, INC