05000446/LER-2006-002, Reactor Trip Due to a Secondary Transient Initiated During Load Rejection Testing
| ML063610127 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 12/18/2006 |
| From: | Madden F TXU Generation Co, LP, TXU Power |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CPSES-200602238, TXX-06184 LER 06-002-00 | |
| Download: ML063610127 (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)(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)(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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 4462006002R00 - NRC Website | |
text
wTXU Power I"XU Power Cxmanche Peak Steam Efechic Station P. 0. 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-200602238 Log # TXX-06184 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 446/06-002-00 Gentlemen:
Enclosed is Licensee Event Report (LER) 06-002-00 for Comanche Peak Steam Electric Station Unit 2, "Reactor Trip Due to a Secondary Transient Initiated During Load Rejection Testing."
This communication contains the following new licensing basis commitments regarding CPSES Units I and 2:
Commitment No.
27416 27417 27418
Description
Operating procedures will be reviewed related to the sequencing of secondary pumps to ensure the MFW pump steam control valve remains in an effective throttling range.
Training will be developed on low power events to ensure that lessons learned from this event are shared.
Secondary system controller responses for Main Steam indicated flow and changes in the dampening for control inputs in the secondary system will be evaluated.
The commitment number is used by TXU Generation Company LP for the internal tracking of CPSES commitments.
A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
- Comanche Peak ° Diablo Canyon
- Palo Verde
- South Texas Project
- Wolf Creek
TXX-06184 Page 2 of 2 Sincerely, TXU Generation Company LP By:
TXU Generation Management Company LLC Its General Partner Mike Blevins By:
Fred W. Madden Director, Oversight and Regulatory Affairs GLM Attachment c -
B. S. Mallett, Region IV M. C. Thadani, NRR Resident Inspectors, CPSES
Enclosure to TXX-06184 NRC FORM 366 U.S. NU(CLEAR REGUILAIORY COMMISSION APPROVEI) BY OMIB NO. 3150-0104 EX IIIRES 06/30/2007 (6-2004)
Estimated burden per response to comply swilh this mandators 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 tIre licensing process and fed back to industry. Send conmments regarding burden estimate to the Records and FOIA1Privacy Service Branch fT-5 F52). U.S. Nuclear Regulatory.Commission. Washington. DC 20555-0001, or by intertet e-mail to infocollectsirtrc.gov, and to the Desk Officer. Office of LIC N S E E EN RE O R (Information ard Regulatory Affairs. NEO13-10202 13150-0104)., ffice ofManag.e.i..
LICEN SEE EVENT REPO RT and Budget, Washington. DC 20503. Ifa means used to irlpose art inflimnration collection doles not display a currently valid OMB/ control number, tire NRCt may not conduct or sponsor, and a person is not required to respond to. the information collection.
Facility Narie (1)
Docker Number (21 Page (3)
COMANCHE PEAK STEAM ELECTRIC STATION UNIT 2 05000446 1 OF 5 Title (4)
Reactor Trip Due to a Secondary Transient Initiated During Load Rejection Testing Esveit D)ate (5)
LER Number (6f Re ort Date (7)
Otier Facilities Involved (81 Month Day v Year Year Sequential Rerision Month Day Year Facility Name Docket Numbers ium N,,
N/A 05000 10 27 2006 2006 002 00 12 18 06 05000 Operatingd This report is submtitted pursuant to the requirements of 10 CFR : (Check all that apply) (II) 20.2201 (b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii)
Forer 20.220( I(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A)
Lete 8
L0).
28%
/
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) 73.71 (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 AbstracR below or in ilforoe space is required, use additional copies of (If more space is required. use additional copies of NRC Fott 366A) (17)
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.
I1.
ANALYSIS OF THE EVENT
A.
SAFETY SYSTEM RESPONSES THAT OCCURRED Both Motor Driven AFW pumps and the Turbine Driven AFW pump automatically started as designed.
B.
DURATION OF SAFETY SYSTEM TRAIN INOPERABILITY Not applicable - there was no safety system train inoperability that resulted from this event.
C.
SAFETY CONSEQUENCES AND IMPLICATIONS OF THE EVENT This event is bounded by the accident analysis in Sections 15.1.2, "Feedwater System Malfunctions That Result in an Increase in Feedwater Flow" and Section 15.2.7, "Loss of Normal Feedwater Flow."
A loss of norimal feedwater resulting from pump failure, valve malfunction, or loss of offsite power leads to a reduction in the capability of the secondary system to remove heat generated in the reactor core. These events are analyzed in section 15.2.7 of the CPSES Updated Final Safety Analysis Report (UFSAR) which uses conservative assumptions in the analysis to minimize the energy removal capability of the AFW system. The October 27, 2006 event occurred with the reactor at approximately 28% power. All systems and components functioned as designed. The event is bounded by the UFSAR accident analysis which assumes an initial power level of 102% and the worst single failure in the AFW system for a loss of feedwater event. The UFSAR analysis shows that a loss of normal feedwater does not adversely affect the core, the reactor coolant systems, or the steam system; therefore, this event posed no threat to the health and safety of the public.
Based on the above, it is concluded that the health and safety of the public was unaffected by this condition and this event has been evaluated to not meet the definition of a safety system functional failure per 10CFR50.73(a)(2)(v).
(If tore space is iequired, use additional copies of NRC Fons 366A) (17)
IV.
CAUSE OF THE EVENT
The cause of this event was the initiation of an oscillation in the Main Steam system while implementing a load rejection test which caused indicated steam flow to oscillate between 0 and 1.4Mlb/HR. The MFW, Heater Drain, and Steam Dump control systems were not able to dampen the oscillations. This caused steam flow and feed flow fluctuations that resulted in a turbine trip, MFW pump trip, and subsequently a manual reactor trip. The collective effect of the tests that were underway and the responsiveness of the secondary control systems for the specific plant conditions at this power level were not fully understood. As a result, the plant's inability to dampen the transient was not anticipated.
Contributing factors for this event included Steam Dump Valve cycling which resulted in indicated steam flow oscillating between 0 and 1.4 Mlb/HR. Forward feed from the Heater Drain pumps resulted in higher MFP suction pressure (less work for the MFW pump) which placed the MFP steam control valve nearer the less stable region so that when Feedwater header pressure began to oscillate, the MFP also began to oscillate.
Plant conditions which established a slightly higher Tave for the third test with a slightly higher Main Steam pressure and larger steam dump demand may have contributed to the Steam Dump valve oscillation.
Implementation of a revised gain setting on the MFP master controller prior to the outage to improve MFP speed control at 100% power may have reduced the capability to dampen oscillations at low power. Due to the Feedwater pump speed oscillations when the master controller was taken to manual, it was apparently at a high peak thus the manual setting was about 600 rpm higher than pre-test levels, increasing feed pressure over steam pressure.
V.
CORRECTIVE ACTIONS
Based on a review of Unit I data, the remaining low power load rejection tests and the four load swings scheduled at high power were cancelled. The review detenrined that no additional infornation was necessary for digital system perfonnance enhancements. Operations perforned a normal startup and brought the Heater Drain system on at a higher power level of about 40%.
The gain on the MFW pump master controller was restored to its original value and the OAT was successfully completed. Operating procedures will be reviewed related to the sequencing of secondary pumps to ensure the MFW pump steam control valve remains in an effective throttling range. Secondary system controller responses for Main Steam indicated flow and changes in the dampening for control inputs in the secondary system will be evaluated. Training will be developed on low power events to ensure that lessons learned from this event are shared, and testing planned subsequent to the twelfth refueling outage on Unit I will be reviewed for application of lessons leamed from this event.
VI.
PREVIOUS SIMILAR EVENTS
There have been no previous similar reportable events at CPSES in the last three years.