05000482/LER-2010-005, Regarding Reactor Trip Due to Low Steam Generator Level from Trip of Main Feedwater Pump

Regarding Reactor Trip Due to Low Steam Generator Level from Trip of Main Feedwater Pump

ML101310391
Person / Time
Site:	Wolf Creek
Issue date:	05/03/2010
From:	Gardner R Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
WO 10-0033 LER 10-005-00
Download: ML101310391 (4)

LER-2010-005, Regarding Reactor Trip Due to Low Steam Generator Level from Trip of Main Feedwater Pump

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)(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)(2)(iv), System Actuation
4822010005R00 - NRC Website

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NUCLEAR OPERATING CORPORATION Rick L. Gardner Plant Manager May 3, 2010 WO 10-0033 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Licensee Event Report 2010-005-00, "Reactor Trip due to Low Steam Generator Level from Trip of Main Feedwater Pump" Gentlemen:

The enclosed Licensee Event Report (LER) 2010-005-00 is being submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A) regarding an Engineered Safety Features Actuation and subsequent reactor trip at Wolf Creek Generating Station.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4156, or Mr. Richard D. Flannigan at (620) 364-4117.

Sincerely, 4R i.*Grne+

RLG/rlt Enclosure cc:

E. E. Collins (NRC), w/e G. B. Miller (NRC), w/e B. K. Singal (NRC), w/e Senior Resident Inspector (NRC), w/e P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET 2iZ7'~

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.

13. PAGE WOLF CREEK GENERATING STATION 05000 482 1 OF 3

4. TITLE Reactor Trip due to Low Steam Generator Level from Trip of Main Feedwater Pump

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.T MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 03 02 2010 2010 005 -

00 05 03 2010 05000

9. OPERATING MODE

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

[o 20.2201(b)

El 20.2203(a)(3)(i)

El 50.73(a)(2)(i)(C)

[I 50.73(a)(2)(vii)

[1 20.2201(d)

El 20.2203(a)(3)(ii)

[I 50.73(a)(2)(ii)(A)

[I 50.73(a)(2)(viii)(A) o 20.2203(a)(1)

El 20.2203(a)(4)

I]

50.73(a)(2)(ii)(B)

[I 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i)

[E 50.36(c)(1)(i)(A)

[1 50.73(a)(2)(iii)

[I 50.73(a)(2)(ix)(A)

10. POWER LEVEL C3 20.2203(a)(2)(ii) 0l 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A) 0l 50.73(a)(2)(x)

[I 20.2203(a)(2)(iii)

[I 50.36(c)(2)

El 50.73(a)(2)(v)(A)

[1 73.71 (a)(4)

El 20.2203(a)(2)(iv)

E] 50.46(a)(3)(ii)

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

El 73.71 (a)(5) 100 [1 20.2203(a)(2)(v)

[I 50.73(a)(2)(i)(A)

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

[I 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)

Richard D. Flannigan, Manager Regulatory Affairs (620) 364-4117CAUSE SYSTEM COMPONENT MANU-REPORTABLE

CAUSE

SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX B

SJ INVT SCI Y

14. SUPPLEMENTAL REPORT EXPECTED 15.SEXPECTED MONTH DAY YEAR SUBMISSION [E YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)

On 3/2/2010 at 1458 CST, while performing procedure SYS PN-200, "Energizing and Deenergizing Inverters PN09 or PN10," Wolf Creek Generating Station (WCGS) experienced a reactor trip due to Steam Generator (SG) Water Level - Low Low actuation signal. The unit received a feedwater isolation and auxiliary feedwater actuation (both motor-driven and turbine-driven) because of the low SG level. All control rods inserted fully and the Reactor Trip System and the Engineered Safety Feature System performed as expected.

The SG Water Level - Low Low actuation signal was initiated due to a trip of the train "A" main feedwater pump as a result of the failed transfer of inverter PN09 to an alternate power supply.

Inverter PN09 did not transfer from the normal to alternate power supply due to the sticking of the reed relay on the static transfer switch circuit board.

The safety significance of this event is low. This event is bounded by analyses as reported in the WCGS Updated Safety Analysis Report (USAR) Section 15.2.7, "Loss of Normal Feedwater Flow." There were no adverse effects on the health and safety of the public.

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 WOLF CREEK GENERATING STATION 05000 482 YEAR NUMBER NO.

2 OF 3

2010

-- 005

"° 00 PLANT CONDITIONS PRIOR TO EVENT MODE - 1 Power - 100

EVENT DESCRIPTION

On 3/2/2010 at 1458 CST, Wolf Creek Generating Station (WCGS) experienced an automatic reactor trip due to Steam Generator (SG) Water Level - Low Low actuation signal [EIIS Code: JB]. The SG water level - Low Low actuation signal was caused by the loss of the speed sensor to the train "A" main feedwater (MFW) pump control circuitry [EIIS Code: SJ]. The loss of the speed sensor was caused by the loss of 120 VAC power to the circuit.

The loss of power to the speed sensor occurred when inverter PN09 [EIIS Code: SJ-INVT] was being transferred to its alternate power supply to facilitate a minor maintenance activity to replace a light bulb. The power supply was being transferred to the alternate supply using procedure SYS PN-200, "Energizing and Deenergizing Inverters PN09 or PN10." Inverter PN09 did not transfer from the normal to alternate power supply due to the sticking of the reed relay on the static transfer switch circuit board.

The unit received a feedwater isolation and auxiliary feedwater actuation (both motor-driven and turbine-driven) as a result of the SG water level - Low Low actuation signal. All control rods inserted fully and the Reactor Trip System (RTS) and the Engineered Safety Feature Systems [EIIS Code: JE]

performed as expected. Decay heat removal following the reactor trip was via the atmospheric relief valves and auxiliary feedwater flow to the steam generators. At the time of the reactor trip the steam dump valves were not available as the loss of inverter PN09 caused the loss of the C-9 interlock blocking steam dump operation.

The steam dump valves became available at 1554 CST, on 3/2/2010, when inverter PN09 was reenergized from its alternate power supply.

All systems functioned as designed with the exception of the instrumentation powered by inverter PN09. PN09 does not supply any safety-related equipment. At the time of the reactor trip the train "A" diesel generator [EIIS Code: EK] and the train "A" Class 1 E air conditioning unit [EIIS Code: VI] were out of service for maintenance.

BASIS FOR REPORTABILITY The reactor trip and subsequent actuation of Engineered Safety Feature Actuation System instrumentation described in this event is reportable per 10 CFR 50.73(a)(2)(iv)(A), which requires reporting of "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." Paragraph (B)(1) of 10 CFR 50.73(a)(2)(iv) includes "Reactor protection system (RPS) including: reactor scram or reactor trip." Paragraph (B)(6) of 10 CFR 50.73(a)(2)(iv) includes "PWR auxiliary or emergency feedwater."U.S. NUCLEAR REGULATORY COMMISSION (9-2007)

LICENSEE EVENT REPORT (LER)

1. FACILITY NAME

2. DOCKET

6. LER NUMBER

3. PAGE

--F`SEQUENTrIAL IREV WOLF CREEK GENERATING STATION 05000 482 YEAR NUMBER NO.

3 OF 3

[2010

-- 005 00 ROOT CAUSE During the performance of procedure SYS PN-200, inverter PN09 did not transfer from the normal to alternate power supply due to the sticking of the reed relay on the static transfer switch circuit board.

A contributing cause was the failure to perform timely preventive maintenance (PM) on components exceeding their recommended life expectancy. Life Cycle Management and Preventive Maintenance Optimization plans determined the schedule for implementing PMs for the inverters. The inverters were on a staggered schedule for PM parts replacement during the refueling outages. PM activities on PNO1 0 were completed in Refueling Outage 17 in Fall 2009. PM activities for PNO09 were scheduled for Refueling Outage 18 in Spring 2011.

CORRECTIVE ACTIONS

The static switch control circuit board was replaced along with a number of other circuit boards and fuses. After replacement, the power supplies were transferred successfully several times.

Preventive maintenance has been established to replace the circuit boards in accordance with the manufacturer recommendations.

SAFETY SIGNIFICANCE

The safety significance of this event is low. This event is analyzed as reported in WCGS Updated Safety Analysis Report (USAR) Section 15.2.7, "Loss of Normal Feedwater Flow." Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the reactor coolant system, or the steam system, since the auxiliary feedwater capacity is such that reactor coolant water is not relieved from the pressurizer relief or safety valves.

There were no adverse effects on the health and safety of the public.

OPERATING EXPERIENCE/PREVIOUS EVENTS LER 2009-001-00 described a reactor trip due to a Main Feedwater Regulating Valve (MFRV) closure in response to failures of the primary and secondary fuses for the Westinghouse 7300 control card frame that contained the associated control cards for the MFRV.

LER 2004-002-00 described a reactor trip due to a MFRV closure caused by the valve plug in the MFRV separating from the valve stem.