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LER-2015-002, Regarding Reactor Trip Due to Feed Water Flow Controller Malfunction
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); 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
	4462015002R00 - NRC Website
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Regarding Reactor Trip Due to Feed Water Flow Controller Malfunction
	ML15357A030
Person / Time
Site:	Comanche Peak
Issue date:	12/01/2015
From:	Thomas McCool; Luminant Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	CP-201501014, TXX-15141 LER 15-002-00
	Download: ML15357A030 (5)
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Rafael Flares Luminant Power Senior Vice President P 0 Box 1002 Chief Nuclear Officer 6322 North FM 56 Rafael.Floresl@Luminant~com Glen Rose, TX 76043 Luminant 0 00 C 817 559 0403 F 254 897 6652 CP-201501014 Ref. # 10CFR50.73 TXX-15141 December 1, 2015 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK NUCLEAR POWER PLANT DOCKET NO. 50-446 LICENSEE EVENT REPORT 446/15-002-00 REACTOR TRIP DUE TO FEED WATER FLOW CONTROLLER MALFUNCTION

Dear Sir or Madam:

Enclosed is Licensee Event Report (LER) 446/15-002-00, "Reactor Trip Due To Feedwater Flow Controller Malfunction" for Comanche Peak Nuclear Power Plant (CPNPP) Unit 2.

This communication contains no new licensing basis commitments regarding Comanche Peak Unit 2.

Should you have any questions, please contact Mr. Gary Merka at (254) 897-6613.

Sincerely, Luminant Generation Company LLC Rafael Flores B: Thomas P cCool Vice President, Engineering and Support Enclosure c -

Marc L. Dapas, Region IV Balwant K. Singal, NRR Resident Inspectors, CPNPP

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 0113112017 (02-20141 Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.

Reported lessons learned are incorporated into the licensing process and ted beck to industry.

a_*.*

"*"Send comments regarding berden estimate to the FOIA, Privacy and Information Collections (LER)

Branch (T-S F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by LICErioE E ENT EPO T (ER) internet e-mail to Infocollects.Resource~nrc.gov, and to the Desk Officer, Office of Information sand (See Page 2 for required number of Regulatory Atfairs, NEOB-10202, (3150-0104), Office of Management sand Budget, Washington, DC 20503. Ifsa means useed to impose an information collection does anot display a currently valid 0MB digits/characters fo echbock) control somber, the NRC may sot conduct or sponsor, and a person is not reqeired to respond to, the information collection.

1. FACILITY NAME 2

OKTNME

.PG Comanche Peak Nuclear Power Plant Unit 2 000461O

4. TITLE Unit 2 Manual Reactor Trip Due To Feedwater Flow Controller Malfhnction

5. EVENT DATE

6. LERNUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEOUENTIAL REV MOT A

ER FACILITY NAME DOCKET NUMBER SNUMBER NO.

MOTIAY YA 05000 1 1 1 i i ZJFACILITY NAME DOCKET NUMBER 10 03 2015 2015 002 00 2

01 2015 05000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PU RSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

El 20.2201(b)

[] 20.2203(a)(3)(i)

[] 50.73(a)(2)(i)(C)

[]

50.73(a)(2)(vii) 1 E 20.2201(d)

[]

20.2203(a)(3)(ii)

[]

50.73(a)(2)(ii)(A)

[]

50.73(a)(2)(viii)(A)

El 20.2203(a)(1 )

[] 20.2203(a)(4)

[]

50.73(a)(2)(II)(B)

[]

50.73(a)(2)(viii)(B)

El 20.2203Ca)(2)(i)

[]

50.36(0)(1)(i)(A)

[]

50.73(a)(2)(iii)

[]

50.73(a)(2)(ix)(A)

10. POWER LEVEL El 20.2203(a)(2)(ii)

[]

50.36(c)(1)(ii)(A)

[7] 50.73(a)(2)(iv)(A)

[]

50.73(a)(2)(x)

El 20.2203(a)(2)(iii)

[]

50.36(c)(2)

[] 50.73(a)(2)(v)(A)

El 73.71 (a)(4) 43l[

20.2203(a)(2)(iv)

El 50.46(a)(3)(ii)

[] 50.73(a)(2)(v)(B)

El 73.71 (a)(5)

El 20.2203(a)(2)(v)

El 50.73(a)(2)(i)(A)

[] 50.73(a)(2)(v)(C)

[] OTHER El 20.2203(a)(2)(vi)

[]

50.73(a)(2)(i)(B)

El 50.73(a)(2)(v)(D)Spcf nAsrtbeoorn

I. DESCRIPTION OF THE REPORTABLE EVENT

A. REPORTABLE EVENT CLASSIFICATION This event is reportable under 50.73(a)(2)(iv)(A) 'Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section." The systems that actuated included the Reactor Protection System and the Auxiliary Feedwater System.

This event is not reportable per 10CFR21 because the degraded SG 2-03 feedwater flow control valve positioner upper 0-ring was not designed and manufactured under a quality assurance program complying with 10CFR50, Appendix B.

B. PLANT CONDITION PRIOR TO EVENT At 0958 on October 3, 2015, Comanche Peak Nuclear Power Plant (CPNPP) Unit 2 was in MODE 1 operating at approximately 43%

power shutting down for the 15th refueling outage.

C. STATUS OF STRUCTURES, SYSTEMS, OR COMPONENTS THAT WERE INOPERABLE AT THE START OF THE EVENT AND THAT CONTRIBUTED TO THE EVENT There were no inoperable structures, systems or components that contributed to the event.

D. NARRATIVE SUMMARY OF THE EVENT, INCLUDING DATES AND APPROXIMATE TIMES On October 3, 2015, CPNPP Unit 2 was operating at approximately 43% while shutting down for the 15th refueling outage. At 0957, Operators (utility, licensed) in the CPNPP Unit 2 Control Room received a Steam Generator (SG) 2-03 level deviation alarm. SG 2-03 level continued rising, and Operators (utility, licensed) took manual control of the SG 2-03 Feedwater flow control valve [EIIS: (S J)

(FCV)], but they were unable to control feed flow. At 0958, a manual reactor trip was initiated in anticipation of an automatic turbine trip at the P-14 setpoint. The Motor Driven Auxiliary Feedwater (AFW) pumps were manually started per Operations procedures to control AFW flow, minimize cool down, and maintain SG levels. All systems responded normally during and following the trip.

E. THE METHOD OF DISCOVERY OF EACH COMPONENT OR SYSTEM FAILURE, OR PROCEDURAL PERSONNEL ERROR Operators (utility, licensed) in the Unit 2 Control Room received a 5G 2-03 level deviation alarm.

II. COMPONENT OR SYSTEM FAILURES A. CAUSE OF EACH COMPONENT OR SYSTEM FAILURE A failure analysis determined that the SG 2-03 Feedwater flow control valve malfunctioned due to a degraded positioner upper 0-ring.

The positioner upper 0-ring failed due to hardening and compression set. This was attributed to the valve manufacturer using an improper material for the valve positioner upper 0-ring.

B. FALURE MODE, MECHANISM, AND EFFECTS OF EACH FAILED COMPONENT The SQ 2-03 feedwater flow control valve malfunctioned due to a degraded positioner upper 0-ring. This 0-ring separates the air supply port from the positioner outlet port. The faulty 0-ring allowed supply air to bypass the upper plunger seat. The failure of this 0-ring allowed supply air to pass uncontrolled through the positioner and provided approximately 45 psig to the actuator diaphragm, which kept the feedwater flow control valve open.

C. SYSTEMS OR SECONDARY FUNCTIONS THAT WERE AFFECTED BY FAILURE OF COMPONENTS WITH MULTIPLE FUNCTIONS The SQ feedwater flow control valves regulate feedwater flow rates to the S~s from approximately 20 percent to full power. These valves also have a safety function to close to provide backup isolation of Main Feedwater flow to the secondary side of the steam generators following a High Energy Line Break. During this event, when operators manually tripped the reactor, the flow control valve solenoids de-energized allowing air to vent from the actuator and the 2-03 SQ feedwater flow control valve performed its safety function to close.

D. FAILED COMPONENT INFORMATION

The SQ 2-03 feedwater flow control valve is a 16 inch, carbon steel, globe valve. The valve is Model Number D-1 00 manufactured by Copes Vulcan Inc. The positioner is a Model 74 SQ manufactured in April 2011 by Siemens.

III. ANALYSIS OF THE EVENT

A. SAFETY SYSTEM RESPONSES THAT OCCURRED The Reactor Protection System actuated as required. The Motor Driven AFW pumps were manually started per Operations procedures to control AFW flow, minimize cool down, and maintain SQ levels.

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 CPNPP Final Safety Analysis Report (FSAR) accident analysis which asstumes conservative initial conditions which bound the plant operating range and other assumptions which could reduce the capability of safety systems to mitigate the consequences of the transient.

Feedwater system malfunctions that result in an increase in feedwater flow are analyzed in section 15.1.2 of the FSAR. The system is analyzed to demonstrate plant behavior in the event that an excessive feedwater addition occurs due to a control system malfunction or operator error. The analysis assumes conservative initial conditions which bound the plant operating range and other assumptions which could reduce the capability of safety systems to mitigate the consequences of the transient. The FSAR analysis shows that the departure from nucleate boiling ratio encountered for an excessive feedwater addition at power is above the limit value and the feedwater malfunction event at no-load is bounded by the feedwater malfunction event at full power. The event of October 3, 2015, occurred at 43% reactor power, and all systems and components functioned as designed. A malfunction of the SQ feedwater flow control valve positioner cannot affect the safety function of the valve to close to provide backup isolation of Main Feedwater flow to the secondary side of the Steam Qenerators following a High Energy Line Break. There were no safety system functional failures associated with this event.

Based on the above, it is concluded that the health and safety of the public were 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).

IV. CAUSE OF THE EVENT

The Unit 2 reactor trip was due to a malfunctioning SG 2-03 feedwater flow control valve. The valve malfunctioned due to a degraded positioner upper 0-ring. This 0-ring separates the air supply port from the positioner outlet port. The degraded 0-ring allowed supply air to bypass the upper plunger seat. This allowed supply air to pass uncontrolled through the positioner and provided approximately 45 psig to the actuator diaphragm, which kept the SG 2-03 feedwater flow control valve open.

V. CORRECTIVE ACTIONS

This condition only applies to the Siemens Model 74SG positioners that are used on the eight feedwater flow control valves (four in Unit 1 and four in Unit 2). Based on review of plant computer data, the demand signals for Unit 1 were normal which indicated this condition does not apply to the Unit 1 positioners. At the time of the failure, two of the Unit 2 positioners were scheduled to be replaced. As a conservative measure, all four of the Unit 2 feedwater flow control valve positioners were replaced.

As a part of the CPNPP Corrective Action Program, periodic monitoring of feedwater flow control valve demand as an early detection of a positioner failure has been established.

VI. PREVIOUS SIMILAR EVENTS

There have been no previous similar reportable events at CPNPP in the last three years.