05000446/LER-2006-003, Regarding Reactor Trip Due to Feedwater Regulating Valve Malfunction
| ML063610121 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 12/18/2006 |
| From: | Madden F TXU Power |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CPSES-200602239, TXX-06186 LER 06-003-00 | |
| Download: ML063610121 (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) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 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)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 |
| 4462006003R00 - NRC Website | |
text
I NbCTXU Power "XU Power Comanche Peak Steam Electric Station P O. Box 1002 (E01)
Glen Rose, TX 76043 Tel: 254 897 5209 Fax: 254 897 6652 mike.blevins@txu.com Mike Blevins Senior Vice President &
- - Chief Nuclear Officer Ref: #10CFR50.73(a)(2)(iv)(A)
CPSES-200602239 Log # TXX-06186 December 18, 2006 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NO. 50-446 ACTUATION OF REACTOR PROTECTION SYSTEM LICENSEE EVENT REPORT 446/06-003-00 Gentlemen:
Enclosed is Licensee Event Report (LER) 06-003-00 for Comanche Peak Steam Electric Station Unit 2, "Reactor Trip Due to Feedwater Regulating Valve Malfunction."
This communication contains no new licensing basis commitments regarding CPSES Units I and 2.
A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
- Comanche Peak e Diablo Canyon ° Palo Verde
- South Texas Project @ Wolf Creek
TXX-06186 Page 2 of 2 Sincerely, TXU Generation Company LP By:
TXU Generation Management Company LLC Its General Partner Mike Blevins By:
PAI*
- .]
B:/rei'"d W. Madden Director, Oversight and Regulatory Affairs GLM Attachment c -
B. S. Mallett, Region IV M. C. Thadani, NRR Resident Inspectors, CPSES
I Encloure to TXX-06186 NRC FORM 366 U.S. NUCLFAR REGULATORY COAUMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 06/30/2007 (6-2004)
Estimated burden per response to comply with this rmandators collection request: 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.
Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comttments regarding burden estimate to the Records and FOIAWPrivacy Service Branch iT-5 F52). U.S. Nuclear Regulatoty Comttission. Washington. DC 20555-0001. or by internet e-tttail to infocollectsjinrc.gov. and to the Desk Officer. Office of
~ 1t'LCEN SEE EVTENT REPO RT (LE R)
~
Informationt attd Regulatory Affairs: NEOB-10202 3150-0104). Office of Manage...nt and Budget. WVashington. DC 20503. If1a means used to inmpose an information collection does nor display a
'rertly va lid OMB control nutnber. the NRC nray not cordluct or SporrOr. and a person is not required to respond to. the iformtation collection.
Facility Natte (I)
Docket Number (21 Page (3)
COMANCHE PEAK STEAM ELECTRIC STATION UNIT 2 05000446 1 OF 5
'title (4)
Reactor Trip Due to Feedwater Regulating Valve Malfunction Event Date (5)
ER Number (6)
Retort lDate (7)
Other Facilities Itrso red (8)
Month Day I
Year Year iSequential Reision Mmomthl Day Yea Facility Name Docket Nmnbers Numher Numttbser I
N/A 05000 10 29 2006 2006 003 E
00 12 18 06 05000 Operating 1
This report is submitted pursuant to the requirements of 10 CFR : (Check all that apply) (I I) 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)1vii)
Passer 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.731a)(2)(viii)(A)
Let 2el ll 80%
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)(2)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 20.2203(a)(2)(ii) 50.36(c)(I)(ii)(A)
X 50.73(a)(2)(iv)(A) 50.72(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) 73.7 1(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 SUMMARY OF THE EVENT, INCLUDING DATES AND APPROXIMATE TIMES On October 29, 2006 Comanche Peak Steam Electric Station (CPSES) Unit 2 was in Mode 1 operating at approximately 80% power following the completion of the ninth refueling outage. At 1518 hours0.0176 days <br />0.422 hours <br />0.00251 weeks <br />5.77599e-4 months <br />, while holding for xenon stabilization in preparation for an incore/excore calibration, a "Steam Generator 3 Steam and Feedwater Flow Mismatch" alarm was received. The Unit 2 Balance Of Plant operator (utility, licensed) took manual control of the Steam Generator (SG) 3 Main Feedwater (MFW) flow control valve and raised demand to match feed flow and steam flow.
After the operator raised the feed flow demand at the SG3 MFW flow control valve, feed flow began to rise and actually exceeded steam flow. The feed flow demand at the SG3 MFW flow control valve was then reduced to lower feed flow to match the steam flow when feed flow dropped off drastically. Demand was once again raised (to 100%) but feed flow continued to lower. The Unit Supervisor (utility, licensed) ordered a reactor trip at 1520 hours0.0176 days <br />0.422 hours <br />0.00251 weeks <br />5.7836e-4 months <br /> due to SG3 level lowering uncontrollably. SG3 level was approximately 40% at the time of the trip and lowering rapidly.
Auxiliary feedwater automatically started as expected due to Lo Lo level in SG3. All systems responded normally during and following the trip and the unit was stabilized in Mode 3.
NRC FORNI 316A t 1-2001)
EncloSure to TXX-06186 NRC FOR-M 366A U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
Facility Name (1)
Docket LER Number (6)
Page(S) sc ehtal 1
- Yea, i
eunta evision COMANCHE PEAK STEAM ELECTRIC STATION UNIT 2 0034 mbet Number 05000446 2006 00 3 OF 5 NARRATIVE (If moe space is required, use additional copies of NRC Fore 366A) (17)
E.
THE METHOD OF DISCOVERY OF EACH COMPONENT OR SYSTEM FAILURE, OR PROCEDURAL OR PERSONNEL ERROR Operators (utility, licensed) in the Unit 2 Control Room received a "Steam Generator 3 Steam and Feedwater Flow Mismatch" alarm.
II.
COMPONENT OR SYSTEM FAILURES A.
FAILURE MODE, MECHANISM, AND EFFECT OF EACH FAILED COMPONENT Not applicable - there were no component failures associated with this event.
B.
CAUSE OF EACH COMPONENT OR SYSTEM FAILURE Not applicable - there were no component failures associated with this event.
C.
SYSTEMS OR SECONDARY FUNCTIONS THAT WERE AFFECTED BY FAILURE OF COMPONENTS WITH MULTIPLE FUNCTIONS Not applicable - there were no component failures associated with this event.
D.
FAILED COMPONENT INFORMATION
Not applicable - there were no component failures associated with this event.
NRC FORM b6FA (1-2001)
Enclosure to TXX-06186 (Ifl nore space is requiicd, use additional copies of (If nore space is icquired, use additional copies of NRC Formi 366A) (17)
IV.
CAUSE OF THE EVENT
The cause of this event was believed to be a loose wire on the SG3 Feedwater regulating valve Weidmuller terminal block. The loose wire in the Feedwater regulating valve control circuit caused a high resistance connection and voltage drop that caused the solenoid valve in the pneumatic control system to vent air while still supplying air from the positioner. This loss of air caused a loss of control and closure of the Feedwater regulating valve. The loose connection was most likely the result of poor workmanship during initial installation of the field cable (believed to be installed during Unit 2 construction).
In 1987, prior to the startup of Units 1 and 2, loose connections in Weidmuller terminals blocks were identified.
As part of the corrective actions for Unit 1, the safety related terminals were inspected and tightened as required. Due to the stage of Unit 2 construction, the Unit 2 corrective actions involved procedure changes to verify tightness. Based on the estimated time of the cable installation, the affected cable's tightness should have been verified under the improved termination procedures. However, it could not be specifically determined how the affected cable became loose.
V.
CORRECTIVE ACTIONS
Corrective actions included tightening the loose wire, checking the connections associated with all four Unit 2 Feedwater regulating valves and all four Unit 2 Feedwater regulating bypass valves for tightness. Although some lack of full tightness conditions were identified, the associated design functions would have been performed for all of these conditions.
As a part of the CPSES Corrective Action Program, the tightness of Weidmuller terminals in Unit I Feedwater regulating and bypass valve tenninal boxes will be verified. Additionally, maintenance procedures will be revised to require a check of tightness of both the instrument and field terminals -whenever either side of a circuit is being worked on a Weidmuller terminal block.
V1.
PREVIOUS SIMILAR EVENTS
There have been no previous similar reportable events at CPSES in the last three years.