05000285/LER-1999-001, :on 990203,shutdown TS Entry Occurred Due to AFW Pump Inoperability.Caused by Inadequate Procedure. Procedure IC-ST-IA-3009 Was Corrected to Properly Perform Surveillance Test.With

:on 990203,shutdown TS Entry Occurred Due to AFW Pump Inoperability.Caused by Inadequate Procedure. Procedure IC-ST-IA-3009 Was Corrected to Properly Perform Surveillance Test.With

ML20207E038
Person / Time
Site:	Fort Calhoun
Issue date:	03/05/1999
From:	Gambhir S, Matzke E OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-99-001, LER-99-1, LIC-99-0019, LIC-99-19, NUDOCS 9903100133
Download: ML20207E038 (6)

LER-1999-001, on 990203,shutdown TS Entry Occurred Due to AFW Pump Inoperability.Caused by Inadequate Procedure. Procedure IC-ST-IA-3009 Was Corrected to Properly Perform Surveillance Test.With

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(1) 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2851999001R00 - NRC Website

text

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March 5, 1999 LIC-99-0019 U.

S.

Nuclear Regulatory Commission Attn: Document Control Desk j

Mail Station F1-137 i

Washington, DC 20555

References:

1.

Docket No. 50-285

)

2.

Letter from OPPD (S. K. Gambhir) to NRC (Document Control Desk) dated June 29, 1998 (LIC-98-0083) i

Subject:

Licensee Event Report 1999-001 Revision 0 for the Fort Calhoun Station Please find attached Licensee Event Report 1999-001 Revision 0 dated

)

March 5, 1999. This report is being submitted pursuant to 10CFR50.73 (a) (2) (1) (B). If you should have any questions, please contact me.

Sincerely,

/~b'df

'S.

K. Gambhir Division Manager Nuclear Operations Division EPM / epm Attachment c:

E. W. Merschoff, NRC Regional Administrator, Region IV L.

R. Wharton, NRC Project Manager W.

C. Walker, NRC Senior Resident Inspector INPO Records Center Winston and Strawn 9903100133 990305 Y

DR g

ADOCK 05000285 ii PDR ;,

45 5121 Employment uth Equal Opgrtunny

(

NxC FORJ 366 U.S. NUCLEAR RE&ULATORY COMalSSION APPRGVED BY C7f,4 N 3. 3150-0104 Hes).

EXPIRES 4/30/98 E"Tiow"Snooutsi seIoHRs RE LE C EA t LICENSEE EVENT REPORT (LER) e,a,'ollf,a'," ora,Tgg,ge,"Eil3,*,a,,o,c==ET,,,WTEyg =

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L'i""'T"o."v i"?JfL"""ECY"^"so41o4), OFFICE OF MAI4A "E*l!^c"M.*PXa"rMll

. (See reverse for required number of PAPEmwonx nEcucTioN Pab sumET,wAsHINomoC assos.

digits / characters for each block).

FACluTY NAM (1)

DoCNET NUMBER (2)

PAGE(3) j Fort Calhoun Station Unit No.1 05000285 10F 5 Tmmp)

Shutdown Technical Specification Entry Leue to Auxiliary Feedwster Inoperability EVENT DATE(5)

LER NUMBER (6)

REPORT DATE(7)

OTHER FACILITIES INVOLVED (8)

~

MONTH DAY YEAR YEAR MONTH DAY YEAR 05000 NUMBE FACIUTY NAME DOCKET NUMBER 02 03 1999 1999 - u01 - 00 03 05 1999 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFit (Check one or more)(11)

+

MODE (9) l' 20.2201(b) 20.2203(aX2Xv)

X 50.73(aX2)(l) 50.73(a)(2Xviii)

POWER 20.2203(aX1) 20.2203(aX3XI) 50.73(aX2)(ii) 50.73(aX2)(x)

LEVEL (10) 100 20.2203(ax2xi) 20.2203(ax3x:1) 50.73(ax2xiii) 73.71 s,r 3

20.2203(aX2Xii) 20.2203(aX4) 50 73(aX2)(iv)

OTHER

%@$m "gg ch.,4 d'

Mi 20.2203(aX2Xill) 50.36(cX1) 50.73(aX2XV)

Specifyin Abstract bolow g+

20.2203(aX2Xiv) 50.36(cX2) 50.73(aX2Xvil) rin NRC Form 366A

- Ll:Eh SEE CONTACT FOR THIS Ll ER (12)

NAME TELEPHONE NUMBER (include Aree Code)

Erick Matzke, Station Licensing Engineer 402-533-6855 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

RE M T^

CAUSE

SYSTEM COMPONENT MANUFACTURER

CAUSE

SYSTEM COMPONENT MANUFACTURER EP TO EPW i

i t

SUPPLEMENTAL REPORT EXPECTED I14)

EXPECTED MONTH DAY' YEAR YEs SUBMISSION (W yes. complete EXPECTED SUBMIS$10N DATE)

X NO DATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)(16)

On February 2, 1999, at 1235, the steam driven auxiliary f.:.edwater pump (FW-10) was logged as inoperable per Technical Specification (TS) 2.5 in preparation for the planned performance of quarterly surveillance test IC-ST-IA-3009, "Or'rability Test of IA-YCV-1045-C and Close Stroke Test of YCV-1045." This te e manually isolates steam to the FW-10 turbine to allow functional testing

. o f 1.t.-YCV-10 4 5-C, the FW-10 Steam Inlet Instrument Air Accumulator Check Valve.

'During the performance of IC-ST-IA-3009, IA-YCV-1045-C failed to meet the acceptance criteria specified in the surveillance test. On February 3, 1999, at 1202, Fort Calhoun Station commenced a TS 2.0.1 required shutdown. Testing of IA-YCV-1045-C was conducted in parallel with the plant shutdown. The revised surveillance test and FW-10 operability test were successfully completed and FW-10 was restored to an operable condition at 1424. The plant shutdown was subsequently terminated, with plant power at 59 percent.

The root cause of this event is that the initial markup of procedure IC-ST-IA-3009 did not place valve YCV-1045-20-1, Solenoid Isolation to Steam i

Supply to FW-10, in the vented position during the performance of the IA-YCV-1045-C check valve leakage test as it would be for an actual overpressure condition at iW-10.

Procedure IC-ST-IA-3009 was corrected to properly perform the surveillance i

test.

NRC FORM 306 (4-95)

N~C FORM 366A U.S. NUCLEAR RE;ULATORY CoalSSION (4-M)

LICENSEE EVENT REPORT (LER)

FAcli ITY BLAME (1)

DOCKET I FE NUMRFR Mi P A G F (*ll SEQUENTIAL

' R'EVISION g

NUMBER NUMBER Fort Calhoun Station Unit No. 1 05000285 2 OF 5 1999 - 001 -

00 ruicemenna a msma vu umaw wen en cram um tm BACKGROUND Fort Calhoun Station (FCS) Technical Specification (Ts) section 2.5 reads, in part, as follows:

"The reactor coolant shall not be heated above 300 degrees Fahrenheit (F) unless the following conditions are met:

(1) The motor driven auxiliary feedwater pump is operable. The reactor shall not be made critical unless the steam driven auxiliary,feedwater pump is operable. During modes 1 and 2, one auxiliary feedwater pump may be inoperable for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, provided that the redundant component shall be tested to demonstrate operability."

TS 2.0. l(1) reads in part:

"In the event a Limiting Condition for Operation (LCO) and/or associated action requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in at least suberitical and less than 300 degrees F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, unless corrective measures are completed that permit operation under the permissible action requirements for the specified time interval as measured from initial discovery or until the reactor is placed in an operating mode in which the specification is not applicable. Exceptions to these requirements shall be stated in the individual specifications."

Auxiliary Feedwater (AFW) pump FW-10 is one of two safety related AFW pumps at FCS. FW-10 was manufactured by Coffin Turbo Pump, Inc. and is a steam driven pump designed to be independent of Alternating Current (AC) power requirements. FW-6 is the other safety related AFW pump. It is driven by an electric motor supplied from one of the station's 4160 volt vital buses.

The speed (and resultant discharge pressure) of FW-10 is governed by a pneumatic-hydraulic speed control loop which maintains the AFW pump discharge pressure at a fixed differential greater than its steam inlet pressure. This control system is fed from the Instrument Air (IA) system.

In 1990, a commercial grade diesel powered AFW pump (FW-54) was added to the feedwater system. This pump's normal injection path is through the main feedwater train, which is independent of the AFW system. Its water supply is the condensate tank. It can be cross-connected to the AFW piping injection header. FW-54 is now used as the normal source of water during low power plant operation, i.e.,

during plant startups and shutdowns. Prior to the j

installation of FW-54, FW-6 was used.

Due to a problem reported in Reference 2 a modification was installed to close the steam supply valve to FW-10 upon sensing an overpressure condition at the discharge to the pump. (See Figure 1 for a simplified drawing.)

l l

.U.S. NUCLEAR REsULATORY COMMISSION

+

Wes)

LICENSEE EVENT REPORT (LER)

FAcHITYN N lin DOCKET I FIO NUusaFR (g) panF (3)

SEQUENTIAL REVISION

.g NUMBER NUMBER Fort Calhoun Station Unit No. 1 050CO285 3 OF 5 1999 - 001 -

00 TEXT (W more space is,ensnrod, use atmanal cop \\es of NRC Form 366N (11}

EVENT DESCRIPTION

On February 2, 1999, at 1235, FW-10 was logged inoperable per TS 2.5 in preparation for the planned performance of quarterly Surveillance Test IC-ST-IA-3009, " Operability Test of IA-YCV-1045-C and Close Stroke Test of YCV-1045." This test manually isolates steam to the FW-10 turbine (by closing MS-361, the FW-10 Steam Chest Supply Throttle Valve) to allow functional testing of the FW-10 Steam Inlet Instrument Air Accumulator Check Valve (IA-YCV-1045-C), and a timed stroke test of YCV-1045, the Steam to Pump Fw-10 Control Valve, in the close direction. The test is part of the station Inservice Inspection (ISI) program.

i During the performance of IC-ST-IA-3009, the Instrument Air Accumulator Check Valve (IA-YCV-1045-C) failed to meet the acceptance criteria specified in the

)

Surveillance Test. (See Figure 1 for a simplified drawing.) The Surveillance Test requires an accumulator pressure drop of less than or equal to 5.0 pounds per square inch (psi) ir, a'30 minute poried. Test results on February 2,

1999, indicated a leakage rate of 5.5 psi in 30 minutes. A Condition Report was written to document the failed Surveillance Test.

Troubleshooting and retesting of the accumulator check valve was conducted thrcughout the evening and into the morning of February 3, 1999. The repairs and retests conducted during this period were unsuccessful at establishing a leak rate below the specified criteria.

On the morning of February 3, 1999, station personnel detected an error in the test method specified in IC-ST-IA-3009. This error resulted in insufficient differential pressure being applied to the seat of the 3-way solenoid valve YCV-1045-20-2, Solenoid Isolation to Steam Supply to FW-10. A revision to IC-ST-IA-3009 was made. The 24-hour LCO for FW-10 (TS 2.5) was to expire before the revised test could be performed.

On February 3, 1999, at 1202, Fort Calhoun Station commenced a TS 2.0.1 i

(

required shutdown due to approaching the allowed outage time limit for FW-10.

l Testing of IA-YCV-1045-C was conducted in parallel with the plant shutdown.

Revised Surveillance Test IC-ST-IA-3009 and FW-10 operability test SE-ST-AFW-3006, " Auxiliary Feedwater Pump FW-10, Steam Isolation Valve, and Check Valve Tests," was successfully completed and FW-10 was restored to an operable condition at 1424. The plant shutdown was subsequently terminated, with plant power at 59 percent. This report is being made pursuant to 10CFR50.73 (a) (2) (i) (B).

h NiC FORJ 366A U.S. NUCLEAR REGULATORY COLMISSION l

l&M)

LICENSEE EVENT REPORT (LER) l l

FACH ITY N AME fin DOCKET I FE NUMRFR (8)

PAGF (3) e SEQUENTIAL

' REVISION mR NUMBER NUMBER Fort Calhoun Station Unit No. 1 05000285 4 OF 5 1999 - 001 -

00 i

l TEKT tt,nore space le,equeed, use additanal copes of NRC form 366A) (17}

SAFETY SIGNIFICANCE

While FW-10 is an important piece of equipment for decay heat removal in a variety of plant accident scenarios, the allowed outage time per the NRC l

approved TS is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In this instance the pump was inoperable for about 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />. The additional 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that the pump was inoperable beyond that allowed l

by the planc's TS did not significantly increase the risk to the plant. Both i

FW-54 and FW-6 were operable during the entire event. FW-54 is diverse from either the FW-6 or FW-10 pump. In addition, the systems required for reactor j

coolant system feed and bleed were operable. Therefore, this condition had minimal impact on the health and safety of the public.

CONCLUSION The root cause of this event is that the initial markup of procedure IC-ST-IA-3009 did not place valve Yev-1045-20-1, Solenoid Isolation to Steam Supply to FW-10, in the vented position during the performance of the IA-YCV-1045-C check valve leakage test as it would be for an actual overpressure condition at FW-10. This configuration is necessary to establish l

sufficient differential pressure across the ports of YCV-1045-20-2 to allow l

proper seating of the valve. Surveillance test IC-ST-IA-3009 has been successfully performed on two previous occasions. During the February 2,

1999, test it was noted that YCV-1045-20-2 exhibited external leakage. There was no l

documented evidence of external leakage from the valve on the two previous test occurrences. On the two previous tests the leak rate was recorded as 1

1 psi in 30 minutes. With the test performed correctly and with the new valve installed, the test resulted in a leak rate of 0 psi in 30 minutes.

CORRECTIVE ACTIONS

i Procedure IC-ST-IA-3009 was corrected to properly perform the surveillance l

test. A root cause analysis has been completed and resulting appropriate corrective actions have been developed to correct the contributing causes to I

this event. These additional corrective actions, while not commitments, will l

be implemented through the condition reporting system.

)

PREVIOUS SIMILAR EVENTS

l LERs 1998-008, 1993-019, 1990-016, and 1989-016 document various problems with l

the auxiliary feedwater system.

i i

l i

t

E FORM 366A U.S. NUCLEAR RE2ULATORY COMMISSION 84s>

LICENSEE EVENT REPORT (LER)

FACILITY NAME (1)

DOCKET t FID NtjMRER (S)

PAGE f 3)

SENUM87R mR N BR Fort Calhoun Station Unit No. 1 05000285 5 OF 5 1999 - 001 -

00 TExi pr more space is requaed. use additenal cupes of NRC Form assA) (17)

Figure 1 4

M ain. Steam lA-YCV-1045-20-FR1 YCV-1045-20-1 YCV-1045-20-2 Inst ument YCV-1045 Exhaust IA-YCV-1045-20-FR2

+

Instrument

/

1 lA-YCV-1045-C FW-10 qm S

3 i't R

v J

YCV-1045-20-1 is shown energized. It deenergizes to open YCV-1045.

~

YCV-1045-20-2 is shown deenergized. It energizes to close YCV-1045. Added in 1998 modification.

YCV-1045 is normally closed and fails open.

E l